ML081750035

Entergy'S Initial Statement of Position on New England Coalition Contentions

ML081750035
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	05/13/2008
From:	Travieso-Diaz M Entergy Nuclear Operations, Entergy Nuclear Vermont Yankee, Pillsbury, Winthrop, Shaw, Pittman, LLP
To:	Atomic Safety and Licensing Board Panel
SECY RAS
Shared Package
ML081750042	List: ML081750035 ML081750036 ML081750038 ML081750039
References
50-271-LR, ASLBP 06-849-03-LR, RAS M-60
Download: ML081750035 (191)

Text

12.AS o DOCKETED USNRC ORIGINAL May 13, 2008 (2:41pm)

May 13, 2008 OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF UNITED STATES OF AMERICA I 4UCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

ENTERGY'S INITIAL STATEMENT OF POSITION ON NEW ENGLAND COALITION CONTENTIONS Pursuant to 10 C.F.R. § 2.1207(a)(1) and the Atomic Safety and Licensing Board's

("Board") Revised Scheduling Order dated April 13, 2006 ("Revised Scheduling Order"),'Appli-cants Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (collectively "Entergy") hereby submit their Initial Statement of Position ("Statement") on New England Coa-lition Contentions ("NEC Contentions"). This Statement is supported by the "Joint Declaration of James C. Fitzpatrick and Gary L. Stevens on NEC Contentions 2A and 2B - Environmen-tally Assisted Fatigue" ("Fitzpatrick - Stevens Dir.") and exhibits thereto; the "Joint Declaration of John R. Hoffman and Larry D. Lukens on NEC Contention 3 - Steam Dryer" ("Hoffman -

As directed by the Board, the initial written statements of position by the parties "should be in the nature of a trial brief that provides a precise road map of the party's case, setting out affirmative arguments and applicable legal standards, identifying witnesses and evidence, and specifying the purpose of witnesses and evidence (i.e.,

stating with particularity how the witness, exhibit, or evidence supports a factual or legal position)." Initial Scheduling Order (Nov. 17, 2006) at 10, para. 10C.

>3-SQ)-

Lukens Dir.") and exhibits thereto; and the "Joint Declaration of Jeffrey S. Horowitz and James C. Fitzpatrick on NEC Contention 4 - Flow-Accelerated Corrosion" ("Horowitz - Fitzpatrick Dir.") and exhibits thereto.

I. INTRODUCTION On January 25, 2006, Entergy filed an application pursuant to 10 C.F.R. Part 54 to renew Operating License No. DPR-28 for the Vermont Yankee Station ("VY"). Entergy seeks to ex-tend the license, which expires on March 21, 2012, for an additional twenty years. On March 27, 2006, the Commission published a notice of opportunity to request a hearing on the application.

71 Fed. Reg. 15,220 (Mar. 27, 2006). Among the parties that filed hearing requests challenging Entergy's proposed license renewal was the New England Coalition ("NEC"). The Board granted NEC's hearing request and admitted several of its proposed contentions. Entergy Nu-clear Vermont Yankee, L.L.C., & Entergy Nuclear Operations, Inc. (Vermont Yankee Nuclear Power Station) LBP-06-20, 64 N.R.C. 131, 143 (2006). The Board ruled that the 10 C.F.R. Part 2, Subpart L procedures were appropriate for each of the contentions. Id.

Three of the contentions originally proposed by NEC have been set for hearing: Conten-tion 2A, which criticizes Entergy's analyses of environmentally assisted fatigue of critical reac-tor piping and components;2 Contention 3, which challenges Entergy's plan to monitor and manage aging of the steam dryer during the period of extended operation; 3 and Contention 4, 2 As admitted by the Board, NEC Contention 2A states: ". . [T]he analytical methods employed in Entergy's [en-vironmentally corrected CUF, or] CUFen Reanalysis were flawed by numerous uncertainties, unjustified as-sumptions, and insufficient conservatism, and produced unrealistically optimistic results. Entergy has not, by this flawed reanalysis, demonstrated that the reactor components assessed will not fail due to metal fatigue during the period of extended operation." Memorandum and Order (Ruling on NEC Motions to File and Admit New Con-tention), LBP-07-15, 66 N.R.C. 261, 270 (2007).

Subsequently, the Board admitted new NEC Contention 2B, which challenges the confirmatory analyses per-formed by Entergy in 2008. In admitting the Contention, however, the Board ruled that it is really a subset of Contention 2A and does not merit a separate statement. Order (Granting Motion to Amend NEC Contention 2A)

(Apr. 24, 2008) at 2.

3 As recast by the Board, Contention 3 raises two issues: "1. Whether Entergy has established sound evaluation and implementation procedures to assure that the integrity of the steam dryer is not jeopardized. Specifically, Footnote continued on next page 2

which alleges that Entergy's program for managing flow-accelerated corrosion ("FAC") of pip-ing and components at VY during the license renewal period is inadequate. 4 The facts, testimony and evidence relating to these contentions are described below.

II. APPLICABLE LEGAL STANDARDS 10 C.F.R. § 54.21 (a)(3) requires that a license renewal application demonstrate, for each component within the scope of the license renewal rules, that the effects of aging are being ade-quately managed so that the intended functions will be maintained consistent with the current li-censing basis during the period of extended operation. The standard for this demonstration is one of "reasonable assurance." 10 C.F.R. § 54.29(a). See also Nuclear Power Plant License Renewal Final Rule, 60 Fed. Reg. 22,461, 22,479 (1995) (". .. the [license renewal] process is not in-tended to demonstrate absolute assurance that structures or components will not fail, but rather that there is reasonable assurance that they will perform such that the intended functions ... are maintained consistent with the CLB").

For those components for which time-limited aging analyses ("TLAAs") 5 are provided in accordance with 10 C.F.R. § 54.21(c), the applicant can implement one of three options. The applicant may show that:

Footnote continued from previous page NEC contends that the status of the dryer cracks must be continuously monitored and assessed by a competent engineer. While Entergy has established that it will continuously monitor plant parameters indicative of steam dryer cracking, it has not provided information on its assessment program for the monitoring data or the qualifi-cations of the personnel evaluating this information. 2. Whether a steam dryer aging management program that does not provide a means to estimate and predict stress loads on the dryer during operation for comparison to es-tablished fatigue limits is valid." Memorandum and Order (Ruling on Motion for Summary Disposition of NEC Contention 3) (Sept. 11, 2007), slip op. at 11-12.

Contention 4 states: "Entergy's License Renewal Application does not include an adequate plan to monitor and manage aging of plant piping due to flow-accelerated corrosion during the period of extended operation." LBP-06-20, 64 N.R.C. at 192.

5 TLAAs are defined in 10 C.F.R. § 54.3(a) as those calculations and analyses that "(1) Involve systems, struc-tures, and components within the scope of license renewal, as delineated in § 54.4(a); (2) Consider the effects of aging; (3) Involve time-limited assumptions defined by the current operating term, for example, 40 years; (4)

Were determined to be relevant by the licensee in making a safety determination; (5) Involve conclusions or pro-vide the basis for conclusions related to the capability of the system, structure, and component to perform its in-Footnote continued on next page 3

(i) The analyses remain valid for the period of extended operation; (ii) The analyses have been projected to the end of the period of extended operation; or (iii) The effects of aging on the intended function(s) will be ade-quately managed for the period of extended operation.

10 C.F.R. § 54.21 (c)(1). Of the contentions admitted for litigation in this proceeding, Contention NEC-2 calls for the application of the TLAAs for certain reactor piping and components. The analyses that are the subject of the TLAA here are the ASME Code fatigue analyses required to be performed for the reactor vessel and other key plant components based on thermal and pres-sure transients, e.g., plant heatup and cooldown, expected over the plant over its lifetime. As dis-cussed in the Application (as amended), Section 4.3 ("Metal Fatigue") and Appendix B, Section B. 1.11 ("Fatigue Monitoring"), Entergy intends to demonstrate that its analyses of environmen-tally-assisted fatigue for the components in question show that the effects of aging will be ade-quately managed for the period of extended operation by satisfying the requirements of the ASME Code. The legal issue is whether these analyses satisfy the requirements of 10 C.F.R. § 54.21(c)(1).

The other contentions admitted for adjudication deal with compliance with the aging management requirements with respect to specified components: the steam dryer (Contention NEC-3), and piping and components potentially subject to flow-accelerated corrosion ("FAC")

(Contention NEC-4). License renewal applicants must demonstrate that their programs will be effective in managing the effects of aging on those components during the period of extended operation. 10 C.F.R. § 54.21(a). The applicants need to demonstrate that the components will be Footnote continued from previous page tended functions, as delineated in § 54.4(b); and (6) Are contained or incorporated by reference in the [current li-censing basis] CLB."

4

adequately managed so that the intended function will be maintained consistent with the current licensing basis during the period of extended operation, as required by 10 C.F.R. § 54.21 (a)(3).

Compliance with this regulatory requirement is demonstrated by meeting the general guidance in NUREG-1800, StandardReview Planfor Review ofLicense Renewal Applications for Nuclear Power Plants,Revision 1, U.S. Nuclear Regulatory Commission, September 2005 and the recommendations in NUREG- 1801, GenericAging Lessons Learned (GALL) Report, Revision 1, U.S. Nuclear Regulatory Commission, September 2005. The specific guidance with respect to the steam dryer is found in NUREG-1801,Section XI. M9, BWR Vessel Internals, which calls for a program that includes inspection, flaw evaluation, and repair in conformance with applicable, staff-approved, industry BWR Vessel and Internals Project (BWRVIP) docu-ments. The issue with respect to the steam dryer, therefore, is whether the aging management program for the dryer developed at VY satisfies this guidance.

Likewise, the guidance relative to FAC is found in Section XI.M17, Flow Accelerated Corrosion, of NUREG-1801. The issue then is whether the FAC control program that Entergy intends to implement at VY after license renewal is consistent with the guidance in that section of the GALL Report.

Ill. APPLICANTS' STATEMENT OF POSITION ON FACTUAL ISSUES A. Witnesses and evidence on NEC Contentions 2A and 2B

1. Entergy's Witnesses Entergy's testimony on NEC Contentions 2A and 2B will be presented by a panel of two experts, each with extensive experience in the evaluation of fatigue in boiling water reactor

("BWR") components and first-hand knowledge of how the fatigue evaluations for critical VY reactor components were performed. The first witness on the panel, Mr. James C. Fitzpatrick, has thirty years of experience in design, construction, and modification of nuclear power plant structures, piping systems, pressure vessels, and other equipment. Testimony of James C. Fitz-5

patrick and Gary L. Stevens on NEC Contention 2A / 2B - Environmentally Assisted Fatigue, at-tached to the Joint Declaration of James C. Fitzpatrick and Gary L. Stevens on NEC Contention 2A/ 2B - Environmentally Assisted Fatigue, Entergy Exhibit E2-01 ("Contention 2A/2B Testi-mony") at A3. 6 Mr. Fitzpatrick worked for over twenty years at VY, and in the last six years of his em-ployment he was Senior Lead Engineer, Design Engineering. Id. In that capacity, he was re-sponsible for overseeing the analyses used to predict the long-term performance of critical VY components, including the potential fatigue of metal piping and equipment exposed to the reactor coolant environment. Id. at A13. In particular, he was responsible for the development of En-tergy' s proposed program to manage the effects of fatigue on critical reactor pressure boundary components during the proposed VY license renewal period, and therefore has first-hand knowl-edge of the program and the analyses that support it. Id.

The second Entergy witness, Mr. Gary L. Stevens, is an expert in the application of finite element analysis, fracture mechanics, and structural and fatigue analyses for nuclear components.

Id. at A16. He has extensive experience in the application of American Society of Mechanical Engineers ("ASME") Code Sections III and XI methodology to fatigue analyses of reactor ves-sels and internals components, was the Chairman of former ASME Section XI Task Group on Operating Plant Fatigue Assessments, is the Secretary of the ASME Section XI Working Group on Operating Plant Criteria, is the Secretary of the ASME Section XI Subgroup on Evaluation Standards, and is a member of the ASME Section XI Subcommittee on Nuclear Inservice Inspec-tion. Id. He supervised the Structural Integrity Associates technical staff involved in performing the environmentally assisted fatigue ("EAF") calculations for VY and provided expert technical consultation and review to all aspects of the work. Id. at A18. Mr. Stevens also prepared one of 6 The notation "Exhibit E2-xx" refers to Exhibit xx submitted by Entergy on NEC Contention 2A/2B. A listing of all Entergy exhibits is provided as Attachment 1 to this Statement of Position.

6

the three calculations of the confirmatory analysis, specifically the EAF calculation for the feed-water nozzle, which now represents the calculation of record for that component for VY. Id.

The testimony and opinions of the Entergy witnesses on NEC Contentions 2A and 2B are based on both their technical expertise and experience and their first hand knowledge of the is-sues raised in these contentions.

2. NEC's Witness NEC's witness on these contentions, Dr. Joram Hopenfeld, has provided no indication that he has any experience or expertise whatsoever in the analysis or evaluation of environmen-tally assisted fatigue in reactor piping and components at BWRs. See "Curriculum Vitae for Dr.

Joram (Joe) Hopenfeld," Exhibit NEC-JH_02 to the "Statement of Position, Direct Testimony and Exhibits" ("NEC's Direct") filed by NEC on April 28, 2008.

3. The Evidence The evidence provided by the Entergy witnesses demonstrates that there is no support for the claims made in NEC Contentions 2A and 2B. Entergy's analyses of EAF of the critical VY reactor piping and components demonstrate that EAF is not a credible failure mechanism for such piping and components during the period of extended VY operations after license renewal, and there is no substantial evidence on the record that controverts this finding.

a. EAF Management Program

1. Section 4.3.3 of the License Renewal Application for VY ("Application") presents Entergy's initial assessment of the effects of the reactor coolant environment on fatigue life for nine plant-specific locations of six reactor components at VY selected in accordance NUREG/CR-6260 and theNRC Staff s "GALL Report". Contention 2A/2B Testimony at A19 and Exhib-its E2-04 and E2-05.

2. The initial assessment of environmentally assisted fatigue ("EAF") contained in the VY Ap-plication, which has been refined since the initial filing of the Application, consists of the 7

evaluation of EAF effects for all nine locations in accordance with the provisions of Section X.M1 of the GALL Report by performing environmentally adjusted cumulative usage factor "CUFen" calculations for the nine locations and demonstrating that the total CUFens for 60 years of plant operation remains less than unity. Contention 2A/2B Testimony at A19.

3. The criteria and methodology for performing EAF analyses are specified in Section X.M1 of the GALL Report. The methodology is comprised of three steps: (1) the CUF for a compo-nent is calculated; (2) the environmental multiplier, Fen, is calculated in accordance with the guidance in NUREG/CR-6583 and NUREG/CR-5704; and (3) the CUFen is calculated as the product of the CUF for the component and the corresponding Fen. Id. at A20; Exhibits E2-04 through E2-07.

4. The initial CUFens computed by Entergy for VY are tabulated in Table 4.3.3 of the Applica-tion. As that Table shows, seven of the nine locations had CUFens greater than unity, and therefore greater than the specified criterion of the ASME Code. Contention 2A/2B Testi-mony at A21.

5. To address these results, the Application states (Application, Section 4.3.3 at 4.3-7) that, prior to entering the period of extended operation, for each location that may exceed a CUF of 1.0 when considering environmental effects, VY will implement one of three possible courses of action, including "further refinement of the fatigue analyses to lower the predicted CUFs to less than 1.0."

6. This commitment was modified in Amendment 35 to the Application, which states in part as follows:

At least 2 years prior to entering the period of extended operation, for the locations identified in NUREG/CR-6260 for BWRs of the VY vintage, VY will refine our current fatigue analyses to include the effects of reactor water environment and verify that the cumu-lative usage factors (CUFs) are less than 1. This includes applying the appropriate Fen factors to valid CUFs determined in accor-dance with one of the following:

8

1. For locations, including NUREG/CR-6260 locations, with existing fatigue analysis valid for the period of ex-tended operation, use the existing CUF to determine the environmentally adjusted CUF.

2. More limiting VY-specific locations with a valid CUF may be added in addition to the NUREG/CR-6260 loca-tions.

3. Representative CUF values from other plants, adjusted to or enveloping the VY plant specific external loads may be used if demonstrated applicable to VY.

4. An analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g., NRC-approved code case) may be performed to determine a valid CUF.

Exhibit E2-09, Attachment 3, Commitment 27.

7. Entergy engaged SIA to implement Option 1, that is, perform refined analyses to calculate the CUFs, Fens and CUFens for all nine locations of interest in accordance with the approach described in the GALL Report. Contention 2A/2B Testimony at A23.

8. Final versions of fifteen refined calculations were issued in August and December 2007. Id.

at A25; Exhibits E2-10 through E2-24. The results of the refined calculations show that the CUFens for the nine limiting piping and vessel locations for the sixty years through VY's ex-tended license period are in all cases less than unity, signifying that component failure due to fatigue will not be a concern at VY during the period of extended operation. Contention 2A/2B Testimony at A26.

9. Since performance of the refined analyses has demonstrated that environmentally assisted fatigue will not be a concern during the period of extended operation, no further actions re-garding metal fatigue are currently deemed necessary. Id. at A27. Nonetheless, the condi-tion of piping and components at the locations of interest will continue to be monitored un-der the plant's in-service inspection program through the period of extended plant operation.

9

Id. In addition, the VY Fatigue Monitoring Program will continue to track plant cycles and transients to ensure that the numbers of transient cycles experienced by the plant remain within the analyzed numbers of cycles for all transients. Id.

10. If, at some future time, the results of continued monitoring suggest that the evaluations no longer encompass 60 years of plant operation, submittal of an inspection program for NRC review or replacement of the component in question may become necessary or desirable. Id.

b. Refined and Confirmatory EAF Calculations

11. The methodological approach used in the refined calculations is the same employed in the original analysis referenced in the Application. Id. at A29. However, the refined calcula-tions included additional analyses to determine the CUFs for three of the reactor pressure vessel nozzles, and plant-specific CUF calculations were performed in accordance with ASME Section III for the piping locations in order to develop appropriate plant-specific CUF values for the piping. Id.

12. The refined VY EAF calculations incorporate a number of conservatisms, including:

a. The numbers of transient cycles for 60 years used in the refined calculations is conserva-tive relative to the numbers of transients expected to occur through 60 years of operation.

b. The refined calculations used design basis transient severity definitions, as opposed to the (lesser) actual transient severity.

c. The refined calculations used bounding values for pressure and temperature at EPU condi-tions for the entire 60-year period of plant operation.

d. The refined calculations calculated bounding F,, multipliers using values for temperature, strain rate and sulfur content that were selected to maximize the Fen, multipliers. Id. at A30.

13. SIA Report No. SIR-07-132-NPS, "Summary Report of Plant-Specific Environmental Fa-tigue Analyses for the Vermont Yankee Nuclear Power Station" (Revision 1, dated Decem-ber 2007) provides summaries of the methodology used in the refined analyses and their re-10

sults. Id. at A32; Exhibit E2-24. The results of the analyses, as summarized in Table 3-10 of Report No. SIR-07-132-NPS, demonstrate that the environmentally adjusted fatigue us-age factors for all locations and components analyzed remain within the allowable value of 1.0 through 60 years of VY operation. Contention 2A/2B Testimony at A33; Exhibit E2-24 at 3-18.

14. Upon review of Entergy's refined analyses, the NRC. Staff determined that the overall ap-proach proposed by Entergy was acceptable and found that the performance of the refined calculations would be in conformance with the recommendations in both the Standard Re-view Plan for License Renewal and in GALL Report Section X.M1. Safety Evaluation Re-port ("SER") (February 2008), Section 4.3.3.2 at 4-32 through 4-38.

15. The Staff asked Entergy to explain how the stress intensity for thermal transients (including shear stresses) was calculated for the analyzed components and locations in the refined cal-culations. Entergy explained that, in most cases, shear stresses are negligible for thermal transients for cylindrical components like those used in reactor pressure vessels and piping.

The Staff, however, took the position that shear stresses cannot always be neglected in the calculation of stress intensities used to determine CUFs of all locations, and that while it is appropriate to do so for locations where non-symmetric loadings are not significant, neglect-ing shear stresses for locations with significant geometric discontinuity, such as nozzle cor-ners, could lead to non-conservative results. These Staff concerns were applicable to loca-tions at the blend radius (nozzle comer) regions of three reactor pressure vessel components evaluated in the refined calculations: the feedwater nozzle, the recirculation outlet nozzle, and the core spray nozzle. SER Section 4.3.3.2 at 4-38 through 4-40.

16. The Staff had no concerns with the refined calculations prepared by Entergy for the other six piping and component locations. Id.

11

17. To resolve the Staff's concerns, Entergy proposed, and the NRC Staff accepted, that Entergy perform a confirmatory CUFen analysis of the feedwater nozzle comer using methods that would be acceptable to the NRC. Contention 2A/2B Testimony at A38. The feedwater noz-zle was selected for analysis because (1) it is the limiting nozzle (i.e., has the highest CUFe,)

among the three nozzles regarding which the Staff had questions, (2) it is subjected to more transients and cycles than the other two nozzles, and (3) the transients it experiences are more severe than the transients experienced by the other two nozzles. Id.

18. The Staff agreed to this suggestion and to the choice of the feedwater nozzle comer as the limiting location of interest SER Section 4.3.3.2 at 4 4-41.

19. The confirmatory analysis used the same finite element model, thermal transient definitions, numbers of transient cycles, and water chemistry inputs as the refined analyses performed by SIA for Entergy, but differed from the refined several respects. Contention 2A/2B Testi-mony at A38.

20. The methodology and results of the confirmatory analysis are presented in three calculations (Exhibits E2-25 through E2-27). The feedwater nozzle EAF evaluation was performed for the two controlling locations on the nozzle, the inside surface of the nozzle blend radius (nozzle comer) and at the inside surface of the nozzle safe end. The confirmatory calcula-tions for the nozzle comer yielded a CUF (before application of Fen factors) of 0.089 at the nozzle comer, versus a CUF of 0.064 at the same location using the refined analysis meth-odology (Exhibit E2-27, Section 4.0 and Exhibit E2-24, Table 3-10). The corresponding re-sults for the safe end location showed a reduction in the computed CUF when using the con-firmatory analysis methodology. Contention 2A/2B Testimony at A40.

21. The confirmatory calculation yielded a composite CUFen of 0.353 for the nozzle comer, sig-nificantly below the acceptable limit. The refined calculation, on the other hand, yielded an 12

environmentally adjusted CUF of 0.639 for the nozzle comer, also significantly below the acceptable limit of 1.0. Id.

22. After review, the Staff found that Entergy correctly applied the ANSYS finite element soft-ware; used appropriate input parameters; added the stresses correctly; and applied proper Fen factors for each transient, so that the results of the confirmatory analysis were appropriate and acceptable. Id. at A41; SER Section 4.3.3.2 at 4 - 41 through 4 - 43. The Staff re-quested that the confirmatory analysis be the analysis of record for the feedwater nozzle. Id.

23. The Staff imposed a license condition requiring similar confirmatory analyses for two other nozzles, the recirculation outlet nozzle and the core spray nozzle. Those confirmatory analyses will become the "analyses of record" for those two locations. SER Section 4.3.3.2 at 4 - 41 through 4 - 43. Entergy is to submit these analyses to the Staff no later than two years prior to the start of the period of extended operation, in March 2012. Id. at 4 - 43.

24. It is acceptable to perform these two confirmatory analyses for two reasons. First, there is no requirement to address the effects of coolant environment on component fatigue life prior to license renewal. Contention 2A/2B Testimony at A43; Exhibit E2-03 at 1. On the other hand, the methods for performing the calculations have been defined, and the required per-formance of the calculations two years prior to the license renewal period ensures that the environmental effects are addressed prior to entering the period of extended operation. Con-tention 2A/2B Testimony at A43. Second, because the CUFens at those two nozzle comer locations are very small (0.084 for the recirculation outlet nozzle and 0.167 for the core spray nozzle), it is extremely improbable that the results of the confirmatory analyses of these nozzles would yield CUFenS greater than unity. Id.; Exhibit E2-09, Amendment 35 to Application, Attachment 1.

25. There is no practical difference between the approaches in the refined and confirmatory analyses because they both yield conservatively calculated CUFe,,s for all nine limiting pip-13

ing and vessel locations that are well within the acceptable limit. Thus, regardless of what method one chooses to apply, the conclusion is the same - the critical reactor components will not experience failure due to fatigue during the period of extended operation. Conten-tion 2A/2B Testimony at A44.

c. Response to issues raised in the NEC testimony

26. NEC's consultant Dr. Joram Hopenfeld asserts that Entergy has not supplied to NEC infor-mation necessary to establish the validity of Entergy's CUFen reanalyses, i.e.: "adequate layout drawings of the plant piping" and "a complete description of the methods or models used to determine velocities and temperatures during transients." "Review of Entergy Nu-clear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. ("Entergy") Analyses of the Effects of Reactor Water Environment on Fatigue Life of Risk-significant Components During the Period of Extended Operation" ("NEC Fatigue Report") at 8. However, Entergy supplied to NEC 36 drawings, which showed nozzles and connecting headers for the com-ponents in question. In addition, Entergy supplied to NEC a copy of the Design Information Record ("DIR") that lists all the drawings and other inputs used in the refined calculations.

If NEC or its consultants had required additional drawings they could have identified them through the DIR and requested them in discovery, as they requested other materials. No such a request was ever made. Contention 2A/2B Testimony at A48; Exhibit E2-29.

27. With respect to the alleged lack of"a complete description of the methods or models used to determine velocities and temperatures during transients," NEC asked for such a description through counsel and it was provided to NEC on April 14, 2008. Again, if additional specif-ics on the calculations were needed, NEC could have asked for them. Contention 2A/2B Testimony at A48.

28. NEC alleges in the NEC Fatigue Report the following "errors" in Entergy's Fen calcula-tions: (a) that Entergy used "outdated" statistical equations the Fen parameters, and should 14

have used instead the results in NUREG/CR-6909, ANL-06/08, "Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials," February 2007 ("NUREG/CR-6909," Exhibit E2-30) (NEC Fatigue Report at 10-12) and applied a factor of 17 to correct the CUFs for environmental effects; (b) that Entergy used inappropriate heat transfer equa-tions to calculate the thermal stress for each transient (id. at 12-15); (c) that Entergy has not provided proof that the base metal of the feedwater nozzles is not cracked (id. at 15-16); (c) that the number of plant transients estimated to occur during the operating life of VY is not sufficiently conservative (id. at 16); and (d) that Entergy's calculation of the Fen parameters does not appropriately account for oxygen concentrations and resulting changes in water chemistry (d. at 16-17). In addition, NEC criticizes the refined analyses performed by En-tergy because they used a simplified Green's Function methodology, which allegedly re-sulted in "the underestimation of CUF values by approximately 40%." Id. at 17-18.

29. None of these claims has merit. The NRC has not accepted the use of NUREG/CR-6909 in license renewal fatigue analyses, and the conditions that would cause a multiplier of 17 to exist are primarily associated with high temperature and high dissolved oxygen content for carbon and low alloy steels; those conditions do not exist at VY because the plant is operat-ing using hydrogen water chemistry. 2A/2B Testimony at A50. All parameters that are sig-nificant to the determination of Fen were. considered in the VY EAF analyses. Id. at A5 1.

Surface roughness (finish) effects are incorporated into the Entergy fatigue evaluations via use of the ASME Code design fatigue curves in the CUF calculations. Id. at A52; Exhibit E2-32 at 3.

30. VY periodically inspects the feedwater nozzle for potential cracks in the base metal and has not identified any since the current thermal sleeves were installed. The most recent inspec-tion was conducted during the 2007 refueling outage and showed no evidence of cracks in the base metal of the nozzle. Contention 2A/2B Testimony at A53; Exhibit E2-33.

15

31. All of the challenges to the heat transfer equations utilized in Entergy's CUF calculations are technically unsound. Contention 2A/2B Testimony at A54.

32. The numbers of transient cycles used in Entergy's CUF are conservative projections of the numbers of cycles actually experienced by the plant over its operating history and the mar-gin on the numbers of cycles analyzed over a linear extrapolation of those experienced sig-nificantly exceeds the factor of 1.2 suggested by NEC. Id. at A55.

33. With respect to the environmental adjustment factor Fen, the NEC Fatigue Report (at 16-17) claims that, in calculating the Fen parameters, Entergy did not properly account for unantici-pated oxygen excursions during the extended operations period and did not explain how the chemistry data for the feedwater line or the electrochemical potential measurements relate to the oxygen concentration at the component surface during transients. Those claims are un-founded. The Fen expressions used by Entergy, and documented in NUREG/CR-6583 and NUREG/CR-5704, are not dependent on electrochemical potential. Contention 2A/2B Tes-timony at A56. Also, the VY calculations accounted for variability in oxygen values, in-cluding potential excursions. As discussed in Attachment 2 to Amendment 35 to the Appli-cation (Exhibit E2-09), Entergy investigated the variability in dissolved oxygen from plant data, including several water chemistry excursions, and determined that the oxygen values used in the CUF analyses bounded a mean plus one sigma variation on oxygen. Id.

34. NEC further claims (Fatigue Report at 17) that VY uses the EPRI-BWVRIA computer code to assess oxygen concentrations and has not described how that code was benchmarked against plant data. However, the BWRVIA model was beiichmarked by fitting it to chemis-try data for one plant and then comparing the model with chemistry data from six other plants. Contention 2A/2B Testimony at A57.

35. The Fatigue Report (at 5-6 and 17-18) claims that use of the Green's Function methodology in the refined analysis results in a 40% underestimation of the CUF values. However, the 16

actual difference is small and the difference in CUF due to changes in the Green's Function is also very small. Therefore, use of the Green's Function methodology would not result in a substantial underestimation of the CUF. Also, the confirmatory analysis CUF for the safe end of the same nozzle was 60% lower than that calculated in the refined analysis. Conten-tion 2A/2B Testimony at A58.

36. NEC criticizes Entergy for the failure to perform an error analysis to show the admissible range for each variable included in the analysis (NEC Fatigue Report at 18). In reality, per-forming an error analysis on the stress results is not traditional practice in analyses of this type, and is unnecessary given that bounding input parameters (such as temperature, pres-sure, and heat transfer coefficients) were selected so as to maximize stresses. Contention 2A/2B Testimony at A59.

37. The NEC Fatigue Report (at 18) asserts that the confirmatory calculation for the feedwater nozzle does not bound on the other two nozzles (the recirculation outlet nozzle and the core spray nozzle). However, the feedwater nozzle is the controlling nozzle because it experi-ences the most severe design transients and because it is the location where the relatively colder feedwater returns to the hot reactor vessel, thereby causing the most severe thermal stresses. It is extremely improbable that either of the other two nozzles would experience CUFens in excess of that experienced by the feedwater nozzle. Contention 2A/2B Testimony at A61. Experts from the NRC Staff and the Advisory Committee on Reactor Safeguards

("ACRS") agree that the feedwater nozzle is the limiting component for purposes of CUFen computations. Id.

38. NEC witness Dr. Joram Hopenfeld has performed a recalculation of the CUFens for the nine locations of interest at VY (NEC Fatigue Report at 19-20). Dr. Hopenfeld calculates irrele-vant CUFs by using generic Fens that do not apply to VY and applying methodology that is 17

not sanctioned by the NRC for existing plants. As a result, the results of his recalculation are meaningless and no credit should be given to it. Contention 2A/2B Testimony at A62.

39. Substantial credit, however, should be given to the judgment of the ACRS which, after re-ceiving considerable amount of testimony on the evaluation of environmentally assisted fa-tigue at VY, concluded that the analyses performed by Entergy had demonstrated that the CUFens for the analyzed locations would be below the limit set in the ASME Code. Id. at A63; Exhibit A2-37.

4. Conclusions to be drawn from the evidence

40. In summary, Entergy has performed more refined and confirmatory EAF analyses at limiting piping and vessel locations for VY. These analyses, performed using conservative method-ologies and input parameters, have demonstrated that the CUFens are less than 1.0 for the sixty years of plant operation encompassed by the renewed VY operating license. These analyses collectively demonstrate that the critical VY components will not experience fail-ure due to fatigue during the period of extended operation. For that reason, there is no sup-port for the claims made in NEC Contentions 2A and 2B, which should be rejected.

B. Witnesses and Evidence on NEC Contention 3

1. Entergy's Witnesses Entergy's testimony on NEC Contention 3 will be presented by a panel of two experts, each with extensive experience in monitoring the performance of the VY steam dryer and devel-oping the plan for continued management of the steam dryer after renewal of the VY operating license. The first witness on the panel, Mr. John R. Hoffman, was until his retirement in Sep-tember 2006, employed by Entergy as the Project Manager for the License Renewal Project at VY. Testimony of John R. Hoffman and Larry D. Lukens on NEC Contention 3 - Steam Dryer, attached to the Joint Declaration of John R. Hoffman and Larry D. Lukens on NEC Contention 3

- Steam Dryer, Entergy Exhibit E3-01 ("Contention 3 Testimony") at A2. He has over 37 years 18

of nuclear power engineering experience, and has been associated with VY since 1971. Id. at A2, A3. As Project Manager for the License Renewal Project at VY, he had the responsibility to ensure that all aspects of the license renewal application, including the steam dryer aging man-agement program, were properly developed and were reviewed by the respective subject matter experts at VY. Id. at A5. Mr. Hoffman has a B.E. Degree in Mechanical Engineering from the Cooper Union for the Advancement of Science and Art in 1967, an M.S. Degree in Nuclear En-gineering from the University of Lowell in 1977, and an M.S. Degree in Applied Management from Lesley College in 1985. Id. at A3.

The second witness in Entergy's panel is Mr. Larry D. Lukens. Prior to his retirement in July 2007, Mr. Lukens was employed by Entergy and had, among other responsibilities, that of Supervisor, Code Programs at VY. Contention 3 Testimony at A7. In that position, his respon-sibilities entailed ensuring that the activities required by industry codes, particularly those issued by the American Society of Mechanical Engineers ("ASME"), that are applicable to VY and are the responsibility of Engineering were completed, evaluated, dispositioned, and documented.

The required activities included, for example, those described by the ASME Operation and Maintenance Code for testing pumps and valves; the ASME Boiler & Pressure Vessel ("BPV")

Code for inservice inspection ("ISI"), including containment inservice inspections; the primary containment integrity monitoring program described by 10 C.F.R.50, Appendix J; and the reactor vessel and internals management and monitoring program under the Electric Power Research In-stitute ("EPRI") BWR Vessel & Internals Program ("BWRVIP"), an industry initiative imple-mented with'the concurrence and participation of the NRC. Id. He was directly involved with the License Renewal audits and inspections of Code Programs activities including the inservice testing ("IST"), ISI, Containment ISI, Appendix J, and BWRVIP, and with the Fire Protection programs, and approved the VY License Renewal commitments relating to these programs. Id.

With respect to the steam dryer, Mr. Lukens was responsible for ensuring the proper completion 19

and evaluation of the steam dryer inspections conducted during the 2005 and 2007 refueling out-ages. He was also responsible for overseeing the license renewal aging management program as it applied to the steam dryer. Id. at Al0. Mr. Lukens received a B.S. Degree in Nuclear Engi-neering from the University of Wisconsin, Madison, in 1978. He has over 38 years of nuclear power work experience, including being a qualified reactor operator in the U.S. Navy and an NRC licensed operator at the University of Wisconsin, and nearly 10 years of service as Program Manager for ASME Section XI inservice testing, inservice pressure testing, and containment leak rate testing at an operating nuclear power plant. Id. at A8.

The testimony and opinions of the Entergy witnesses on NEC Contention 3 are based on both their technical expertise and experience and their first hand knowledge of the issues raised in the contention.

2. NEC's Witness NEC's witness on this Contention, Dr. Joram Hopenfeld, has provided no indication that he has any experience or expertise whatsoever in the analysis or evaluation of BWR steam dryer performance or the issues associated with the structural integrity of steam dryers during normal plant operations. See "Curriculum Vitae for Dr. Joram (Joe) Hopenfeld," Exhibit NEC-JH_02 to the "Statement of Position, Direct Testimony and Exhibits" ("NEC's Direct") filed by NEC on April 28, 2008.

3. The Evidence The evidence provided by the Entergy witnesses demonstrates that there is no support for the claims made in NEC Contention 3 with respect to the performance of the VY steam dryer during the period of extended VY operation after renewal of its operating license. Entergy has instituted a program, currently in effect and to be continued after renewal of the VY license, to continuously monitor plant parameters indicative of potential cracking of the steam dryer and properly evaluate and respond to any significant departures of those parameters from their nor-20

mal range. That program is in accordance with industry guidelines and has been accepted by the NRC Staff for implementation during the current period of plant operations at uprated power level.

Entergy has also instituted a program, currently in effect and to be continued after VY li-cense renewal, to perform during each refueling outage thorough visual inspections of the areas of the steam dryer potentially susceptible to crack formation. These inspections are conducted in accordance with industry guidelines and their methodology has been accepted by the NRC Staff for implementation during the current period of plant operations at uprated power level.

Accordingly, there is no factual support for the claims made in NEC Contention 3.

a. Background

41. The steam dryer is a BWR stainless steel component, installed in the reactor vessel above the steam separator assembly and is supported by brackets welded to the inside of the vessel wall below the steam outlet nozzles, whose function is to remove moisture from the steam before it leaves the reactor. Contention 3 Testimony at Al 1. During plant operations, wet steam flows upward and outward through the dryer. Moisture is removed by impinging on the dryer vanes and flows down through drains to the reactor water in the downcomer annu-lus below the steam separators. Id. The VY steam dryer is a non-safety-related, non-Seismic Category I component. Although the steam dryer is not a safety-related component, the assembly is designed to withstand design basis events without the generation of loose parts and the dryer is designed to maintain its structural integrity through all plant operating conditions. Id.

42. In 2002, steam dryer cracking and damage to components and supports for the main steam and feedwater lines were observed at the Quad Cities Unit 2 nuclear power plant, and loose parts shed by the dryer due to flow-induced vibration that caused metal fatigue failure of the dryer had damaged the supports. Id. at A 12.

21

43. The Quad Cities 2 experience raised a concern that a loss of physical integrity of the dryer such that loose dryer sections or parts are released to the reactor steam space (that is, the space in the reactor where steam is confined above the water) could potentially migrate to other components and could have adverse impact on safety-related equipment. Id. at A13-A14.

44. While the formation of cracks on the surface of a steam dryer is not in itself cause for con-cern, the existence of those cracks needs to be identified and evaluated before the cracks progress to the point where they could cause a loss of physical integrity of the dryer, result-ing in loose parts. Id. at A15.

b. VY's response to industry experience

45. In response to the Quad Cities 2 event, Entergy substantially modified the steam dryer at VY during the Spring 2004 refueling outage to improve its capability to withstand the higher flow induced vibration loadings that could result from operation of the plant at EPU levels.

Id. at A16; Exhibit E3-04. In addition, VY instituted a program of dryer monitoring and in-spections to provide assurance that the flow-induced loadings under normal operation at EPU levels did not result in the formation or propagation of cracks on the dryer. Contention 3 Testimony at A16; Exhibit E3-05. The program was reviewed and approved by the NRC and included as a license condition as part of the power uprate license amendment issued on March 2, 2006. Contention 3 Testimony at Al 6.

46. As power was increased in 2006 from the original licensed power level to full extended power uprate ("EPU") conditions, there was continuous monitoring of plant parameters in-dicative of dryer performance, including measurement at least once per week of moisture carryover and periodic measurement of main steam line pressure. Id. at A 17. Following completion of EPU power ascension testing, moisture carryover measurements continue to be made periodically, and other plant operational parameters that would be symptomatic of 22

loss of steam dryer structural integrity (main steam line flow, reactor vessel water level, steam dome pressure) continue to be monitored and their values trended. Id. This monitor-ing program will continue to be implemented during the period of extended operation after renewal of the VY license. Id.

47. In addition, the VY steam dryer was inspected during plant refueling outages in the Fall of 2005 (before completion of the EPU) and Spring of 2007 (after one year of operation at EPU power levels). Id. at A17. The dryer is scheduled to be inspected again during the refueling outages in the Fall of 2008 and the Spring of 2010, with a partial inspection scheduled for the Fall of 2011. Id. Inspections will continue in the license renewal period starting with the first refueling after March 2012. Id.

c. VY's Steam Dryer Aging Management Program

48. In its License Renewal Application, Entergy addresses aging management of the VY steam dryer as follows:

Cracking due to flow-induced vibration in the stainless steel steam dryers is managed by the BWR Vessel In-ternals Program. The BWR Vessel Internals Program currently incorporates the guidance of GE-SIL-644, Revision 1. VYNPS will evaluate BWRVIP-139 once it is approved by the staff and either include its rec-ommendations in the VYNPS BWR Vessel Internals Program or inform the staff of VYNPS's exceptions to that document.

License Renewal Application, § 3.1.2.2.11 "Cracking due to Flow-Induced Vibration."

49. GE-SIL-644 recommends that BWR licensees institute a program for the long term monitor-ing and inspection of their steam dryers. It provides detailed inspection and monitoring guidelines (Exhibit E3-06, Appendices C and D).

50. The proposed steam dryer management program conforms to the guidance in the NRC GALL Report (NUREG-1 801), which was confirmed by the NRC Staff in its Safety Evalua-23

tion Report ("SER") for the VY license renewal. Contention 3 Testimony at A23; SER at 3-175.

51. The monitoring component of the proposed VY steam dryer management program consists of assessing the status of the steam dryer continuously by the plant operators and VY's tech-nical staff through the monitoring of certain plant parameters. Contention 3 Testimony at A24; VY Off-Normal Procedure ON-3178, Exhibit E3-08.

52. VY Off-Normal Procedure ON-3178 alerts the operators that any of the following events could be indicative of significant dryer damage: (a) sudden drop in main steam line flow

>5%; (b) >3 inch difference in reactor vessel water level instruments; and (c) sudden drop in steam dome pressure >2 psig. In addition, periodic measurements of moisture carryover are performed, and changes in moisture carryover are evaluated in accordance with the re-quirements of GE-SIL-644 to determine whether significant cracking has occurred. Conten-tion 3 Testimony at A24.

53.. Abnormal values of the monitored plant parameters would indicate that the steam leaving the reactor has a high moisture content, which in turn could indicate that steam is escaping through a crack in the dryer. Such escape would be symptomatic of a significant crack that might result in loss of physical integrity of the dryer. Contention 3 Testimony at A25; Ex-hibit E3-06, Appendix D.

54. Moisture carryover is measured by plant chemistry personnel using procedure OP-0631 Ap-pendix F (Exhibit E3-10). If moisture carryover is determined to be greater than the limit stated in the procedure (currently 0.19%), the procedure requires that a Condition Report

("CR") be written, the Shift Manager notified, and actions taken in accordance with Off-Normal Procedure ON-3178. Contention 3 Testimony at A27.

55. If the moisture carryover is in the range of 0.19% to 0.35%, Off- Normal Procedure ON-3178 requires that plant management and engineering be informed. Additionally, the Opera-24

tional Decision Making Initiative process would be initiated, to provide an especially me-thodical, systematic, conservative decision making process affecting station operation. In addition, an engineering evaluation in accordance with EN-OP- 104 "Operability Determina-tions" would be performed. Contention 3 Testimony at A28.

56. If moisture carryover exceeds 0.35% station management is notified and operability evalua-tion is requested. If the results of the evaluation do not support continued plant operation, the reactor is brought to hot shutdown. In either case, experienced qualified engineering personnel will determine the significance of the abnormal moisture carryover measurement.

Id.

57. According to Procedure ON-3178, if the engineering evaluation of plant data confirms that steam dryer damage may have occurred, a plant shutdown is initiated such that the plant is placed in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Contention 3 Testimony at A32.

58. The personnel involved in determining the significance of the moisture carryover and other measured parameters are required to be qualified in the application of the operability deter-mination procedure EN-OP-0104 (Exhibit E3-11). Contention 3 Testimony at A31. A pre-requisite for procedure qualification is the requirement that the individual(s) be enrolled in the Engineering Support Personnel training program and that their capability to perform in-dependent engineering work be assessed by their supervisor. If an engineer or his supervisor feels the engineer needs additional training to maintain or enhance his level of expertise, that training is incorporated into the performance goals for the year. Id.

59. The purpose of the measurements of moisture carryover, main steam line flow, reactor ves-sel water level, and steam dome pressure is not to enable Entergy to determine whether a dryer crack is about to form, but to provide early warning to the plant personnel that a crack may have developed so that appropriate, timely action may be taken before undesirable ef-fects ensue as a result of the crack. Contention 3 Testimony at A33.

25

60. While there is no technology that will predict when a crack will initiate, the monitoring pro-grams at VY ensure that any steam dryer cracks that develop are detected before they grow to a size that would be of concern. Id. at A34.

61. With respect to dryer inspections during plant refueling outages, the details of the visual in-spection program to be implemented are set forth in the section of GE-SIL-644 devoted to BWR-3 steam dryers, which is Appendix C, pp. 15-16. The dryer inspections are to be per-formed in accordance with the VY BWRVIP Program Plan, VY-RPT-06-00006 (Exhibit E3-12) and GE-SIL-644, Revision 1. Contention 3 Testimony at A35.

62. The dryer examinations consist of VT-I and VT-3 examinations of accessible internal and external welds and plates in the steam dryer potentially susceptible to crack formation.

VT-I and VT-3 examinations are defined by ASME Boiler & Pressure Vessel ("BPV")

Code Section XI, and the non-destructive examination technicians who perform and review these examinations are qualified in accordance with ASME BPV Section XI. Id. at A35-A36. The inspections are performed by qualified non-destructive examination ("NDE") in-spection personnel, using qualified NDE techniques appropriate for BWR steam dryer in-spections. Because of the large number of individual examinations to be performed during a refueling outage, this work is typically contracted out to qualified vendors, including the re-actor supplier (General Electric). Id. at A38.

63. The inspection data are reviewed by qualified Level III NDE personnel and are subject to fi-nal acceptance by Entergy Level III NDE personnel. VY typically contracts both the Level II and Level III services for reactor vessel and internals examinations, including the steam dryer examinations. As an additional quality step, VY requires that these examinations also be reviewed by an Entergy Level III NDE technician. The Entergy Level III review and ap-proval is required to be completed on 100% of the steam dryer examinations prior to its re-turn to service. Id. at A40.

26

64. All detected indications (imperfections or unintentional discontinuities that may or may not be cracks) are evaluated by qualified structural engineers, who are experienced with BWR steam dryer crack evaluation. Typically, these indications are evaluated by engineers who are on the staff of the reactor vendor, and the evaluations and conclusions are reviewed and accepted by qualified Entergy structural engineers. Id. at A42.

65. An indication is classified as recordableor relevant if it is visible to the resolution of the ex-amination technique. All recordable indications found by the Level II NDE technician who performs the examination are identified and documented in the corrective action program.

All recordable indications that are confirmed by the Level III NDE technician are evaluated by Engineering to determine whether or not they are rejectable. Rejectable indications are those that must be repaired prior to restarting the plant. Repair of rejectable indications is an ASME BPV Code Section XI requirement. Id. at A43.

66. If the characteristics of a particular indication do not rule out fatigue, the indication is typi-cally classified as a potential fatigue indication. Subsequent examinations determine whether the indication is growing. A fatigue crack would be expected to grow; a crack that does not grow would not show the characteristic behavior of fatigue and would not be of concern. Id. at A44. At VY. all recordable indications are reinspected at each refueling out-age until at least two consecutive inspections show no growth. Id.

67. VY has not identified any steam dryer cracks that are consistent with fatigue, and this con-clusion is supported by the fact that the identified indications have not grown during subse-quent operating cycles. Id. at A45.

68. During the Spring 2004 refueling outage, in preparation for EPU, the dryer received a base-line VT-I inspection of all accessible areas deemed potentially susceptible to crack forma-tion. These examinations comprised 287 weld and plate examinations. A total of 20 indica-27

tions were identified, of which 2 were weld-repaired, and 18 were determined acceptable to use as-is. Contention 3 Testimony at A46; Exhibit E3-13.

69. The steam dryer inspections performed during the Fall 2005 outage examined all high-stress areas, as identified in GE-SIL-644. In addition, all areas that had been repaired or modified in the Spring 2004 outage were reinspected, as well as those indications that were found and evaluated to be acceptable for use as-is during the Spring 2004 outage. These examinations comprised 113 internal and external weld examinations. A total of 66 indications were iden-tified, including 20 previously identified indications and repaired areas from 2004, all of which were found acceptable for use as-is. The increase in identified VT-I indications was due to increased resolution of the VT-I examinations. Contention 3 Testimony at A46; Ex-hibit E3-14.

70. During the Spring 2007 outage, all accessible susceptible areas of the steam dryer were in-spected, consistent with the guidance in GE-SIL-644, Revision 1. The previously repaired areas, the identified high stress areas, as well as those indications that were previously found and evaluated to be acceptable for use as-is were also examined. A total of approximately 448 individually identified steam dryer examinations were performed. A total of 66 indica-tions were recorded, including 47 of those identified in 2005 and 19 previously unidentified indications. These 19 previously unidentified indications were the result of the increased examination scope in 2007 compared to that in 2005 (448 in 2007 and 94 in 2005), and the fact that all accessible susceptible areas of the steam dryer had been subjected to the im-proved resolution VT-1. All the indications identified in 2007 were accepted for use as-is.

Contention 3 Testimony at A46; Exhibit E3-15.

71. The steam dryer inspections conducted in the Spring of 2007 followed approximately one year of full power operation at the EPU condition. The examinations were conducted using enhanced examination resolution, which provides improved detection levels over those 28

achievable by using the prescribed VT- I examination process. A total of 66 dryer indica-tions. Each of these 66 indications was evaluated by qualified structural engineers, experi-enced in evaluating indications in BWR steam dryers. Each of the indications was accepted to "use as-is." No growth was noted in the previously identified indications. None of the cracks were determined to be associated with fatigue. Contention 3 Testimony at A49.

72. NEC's consultant Dr. Joram Hopenfeld raises two claims regarding the VY steam dryer management program: (1) the monitoring of plant parameters indicative of potential dryer cracks is insufficient to prevent fatigue cracks from forming and propagating in the period between dryer inspections; and (2) a dryer management program must include estimating the stress loadings on the dryer and ensuring that they remain within the stress limits of the dryer material. Dr. Joram Hopenfeld, "Assessment of Proposed Program to Manage Aging of the Vermont Yankee Steam Dryer Due to Flow-Induced Vibrations" (April 25, 2008),

NEC Exhibit NEC-JH_54 ("NEC Dryer Report").

73. With respect to Dr. Hopenfeld statement that "[m]oisture monitoring only indicates that a failure has occurred; it does not prevent the failure from occurring" (NEC Dryer Report at 5), monitoring of plant parameters will not predict the incipient formation of dryer cracks, but it will identify the existence of a crack sufficiently large to adversely affect dryer per-formance and flag the risk of structural failure of the dryer. Contention 3 Testimony at A53.

Since the steam dryer has completed two years of EPU operations without the detection of large cracks, this provides a high degree of assurance that the steam dryer is not subject to rapid flaw growth due to high cycle fatigue. Thus, the monitoring program will be sufficient to provide an "early warning" of potential dryer failure so that action can be taken prior to the occurrence of such failure. Id.

74. Dr. Hopenfeld asserts that most parameter monitoring may indicate the formation of only those steam dryer cracks that increase moisture carryover, but those cracks that do not lead 29

to significant moisture carryover may continue to grow undetected (NEC Dryer Report at 4).

However, since all of the reactor steam flows through the steam dryer, it is very unlikely that any damage to the dryer would not also result in a decrease in efficiency of the steam dryer and an increase in moisture carry-over that would cause a change in one or more of the monitored parameters (steam flow rate, reactor vessel water level and/or steam dome pres-sure). Contention 3 Testimony at A54.

75. Dr. Hopenfeld further asserts (NEC Dryer Report at 1-2) that Entergy's program to date of visual inspection and moisture monitoring have been ineffective in identifying cracking at the time it occurs, when it occurs in between inspections. However, there is no need to iden-tify a crack the moment it occurs, because the intent of VY's program is to monitor material conditions on a frequency sufficient to identify and mitigate any flaws before they can grow to a size that would be detrimental to the integrity of the component. The overwhelming majority of visual indications at VY have not grown since they were first identified, and those few indications that were determined to need repair had not reached critical size (that is, they had not had a negative effect on steam dryer integrity) prior to repair. Contention 3 Testimony at A55.

76. Dr. Hopenfeld goes on to state (NEC Dryer Report at 3) that, once fatigue cracks initiate, they propagate very fast when exposed to alternating stresses of sufficient magnitude and frequency, so that even if one does not find cracks during an inspection, there is absolutely no reason why such cracks would not start propagating once the plant is restarted. However, that postulation assumes that there will be alternating stresses of sufficient magnitude and frequency to cause cracks to propagate rapidly. VY's operating experience after the EPU (exemplified by the data collected during the 2007 inspection and the subsequent year of monitoring of plant operating parameters) demonstrates that the stresses experienced by the 30

dryer are insufficient to initiate and propagate fatigue cracks. Contention 3 Testimony at A56.

77. Dr. Joram Hopenfeld asserts that the aging management program for the VY steam dryer should include "some means of estimating and predicting stress loads on the dryer, estab-lishing load fatigue margins, and establishing that stresses on the dryer will fall below ASME fatigue limits" (Pre-Filed Direct Testimony of Dr. Joram Hopenfeld Regarding NEC Contentions 2A, 2B, 3 and 4, NEC Exhibit NEC-JH_01 at A16). Stress load estimation and prediction, however, are unnecessary because confirmation that stresses on the VY steam dryer remain within its fatigue limits is provided daily by the fact that the dryer has been able to withstand without damage the increased loads imparted on it during power ascension and for the two years of operation since the EPU was implemented. Thus, d ryer perform-ance to date demonstrates that none of the stresses on the dryer has exceeded the endurance limit for the component. Contention 3 Testimony at A61. There will be no change in dryer loads or stresses during the license renewal period of operation; hence, there is no reason to expect that the dryer will be subjected to increased stresses in the future. Id.

78. Moreover, industry experience shows that BWR steam dryers in use during uprated power operations that have inadequate fatigue resistance will most likely exhibit this inadequacy during the first fuel cycle. In other words, operation of a steam dryer for a year or two is sufficient to accumulate enough fatigue cycles to cause significant cracking in susceptible areas of the dryer. Conversely, good performance (such as exhibited by the VY steam dryer) during the first operating cycle after the uprate strongly suggests that the dryer will not experience a fatigue-induced failure. Id. at A62.

79. Dr. Hopenfeld also expresses the view (NECDryer Report at 6) that Entergy should have in-troduced additional analytical tools for predicting the loads on the dryer. However, the ana-lytical tools that were used during the uprate proceeding to demonstrate that loads on the 31

dryer will be below its endurance limits were utilized as part of the design validation process that demonstrated the adequacy of the design and established the current licensing basis.

Because the predicted loads on the dryer were shown to be below the endurance limit, the design analysis was not time limited and thus does not need to be revisited at the license re-newal stage, where only time limited aging analyses need to be evaluated. Contention 3 Testimony at A63. Further, the loadings on the dryer derive from plant geometries (pipe lengths, diameters, flows, pipe connections, etc.) that have not changed since the uprate was implemented, so there has been no change to the loadings on the dryer and the resulting stresses. Therefore, there is no reason to provide continued instrumentation to measure loadings or further analytical efforts. Id.

4. Conclusions to be drawn from the evidence

80. In summary, Entergy has instituted a program, currently in effect and to be continued after renewal of the \fY license, to continuously monitor plant parameters indicative of potential cracking of the steam dryer and properly evaluate and respond to any significant departures of those parameters from their normal range, and to perform during each refueling outage thorough visual inspections, conducted in accordance with industry guidelines, of the areas of the steam dryer potentially susceptible to crack formation. The fact that the VY steam dryer has shown no evidence of fatigue induced cracks after two years of EPU operation strongly indicates that routine inspection of the steam dryer during the period of extended operation will be sufficient to provide reasonable assurance of continued steam dryer integ-rity. In all, the steam dryer inspection and monitoring plan that Entergy will implement dur-ing the period of extended operation after license renewal will assure that the aging effects on the steam dryer will be adequately managed. For that reason, there is no support for the claims made in NEC Contention 3, which should be rejected.

32

C. Witnesses and evidence on NEC Contention 4

1. Entergy's Witnesses Entergy's testimony on NEC Contention 4 will be presented by a panel of two experts, each with extensive experience in the management of flow accelerated corrosion ("FAC") in boiling water reactor ("BWR") components. The first witness on the panel, Dr. Jeffrey S.

Horowitz, has more than 36 years of experience in the field of nuclear energy and related disci-plines and 22 years of experience specializing in FAC and nuclear safety analysis. Testimony of Jeffrey S. Horowitz and James C. Fitzpatrick on NEC Contention 4 - Flow-Accelerated Corro-sion, attached to the Joint Declaration of Jeffrey S. Horowitz and James C. Fitzpatrick on NEC Contention 4 - Flow-Accelerated Corrosion, Entergy Exhibit E4-01 ("Contention 4 Testimony")

at A3. Dr. Horowitz designed and implemented a computer program to assist utilities in deter-mining the most likely places for FAC wear to occur, and thus the key locations to inspect for component wall thinning. He developed the computer programs CHEC (Chexal-Horowitz Ero-sion Corrosion) in 1987, CHECMATE (Chexal-Horowitz Methodology for Analyzing Two-Phase Environments) in 1989, and CHECWORKS (Chexal-Horowitz Engineering Corrosion Workstation) in 1993. Contention 4 Testimony at A6. He has performed, by himself or with an-other engineer, audits of the FAC programs at over fifty nuclear units in the United States and Canada, including a FAC program audit at VY, in April 2007. Contention 4 Testimony at A7.

Dr. Horowitz played a significant role in drafting NSAC-202L, entitled 'Recommendations for an Effective Flow-Accelerated Corrosion Program," and each of its three revisions, which has become the most important standard-setting document for the conduct of FAC control programs in the United States. Contention 4 Testimony at A8. Dr. Horowitz has authored numerous arti-cles and given numerous presentations regarding FAC. Contention 4 Testimony at A9-A1O.

The second witness on the panel, Mr. James C. Fitzpatrick, has thirty years of experience in design, construction, and modification of nuclear power plant structures, piping systems, pres-sure vessels, and other equipment. Contention 4 Testimony at A13, A15 and A16. Mr. Fitz-33

patrick was the Cognizant Engineer for the VY FAC Program through June 2007, and was re-sponsible for developing the scope of refueling outage inspections, providing on-site engineering support, screening and evaluating piping and components, determining if the sample of piping locations designated for inspection during a refueling outage needed to be expanded, coordinat-ing piping and component repairs and replacements, updating the CHECWORKS models of plant piping systems, and maintaining the FAC Program Manual supporting documents. Conten-tion 4 Testimony at A 16.

The testimony and opinions of the Entergy witnesses on NEC Contention 4 are based on both their technical expertise and experience and their first hand knowledge of the issues raised in this Contention.

2. NEC's Witness NEC's witnesses on this Contention, Dr. Joram Hopenfeld, Dr. Rudolf Hausler, and Mr.

Ulrich Witte, have provided no indication that they have any experience or expertise whatsoever in the analysis or evaluation of FAC in reactor piping and components at BWRs. See "Curricu-lum Vitae for Dr. Joram (Joe) Hopenfeld," Exhibit NEC-JH_02, "Curriculum Vitae of Dr. Ru-dolf Hausler," Exhibit NEC-RH_02, and "Curriculum Vitae of Ulrich Witte," Exhibit NEC-JH_02 to the "Statement of Position, Direct Testimony and Exhibits" ("NEC's Direct") filed by NEC on April 28, 2008.

3. The Evidence The evidence provided by the Entergy witnesses demonstrates that there is no support for the claims made in NEC Contention 4. The current FAC Program, which will be used during the license renewal period, meets industry practice as reflected in NSAC-202L and has been re-viewed, audited and inspected with only minor, mostly administrative, issues identified. The FAC Program that will be used during the period of extended operation after license renewal will assure that the aging effects of FAC will be adequately managed.

34

a. VY FAC Program

81. Section B. 1.13 of the License Renewal Application for VY ("Application") (Exhibit E4-04),

sets forth the VY program for addressing FAC. It is consistent with the program described in the NRC guidance document "Generic Aging Lessons Learned (GALL) Report -- Tabulation of Results," NUREG-1801, Vol. 2, Rev. 1 (Sep. 2005) ("NUREG-1801" or "GALL Report"),

Section XI.M1 7, Flow Accelerated Corrosion (Exhibit E4-05). Contention 4 Testimony at A17 and Exhibit E4-04 at B-47. There are no exceptions in the Application to the guidance in NUREG-1801 with respect to FAC.

82. The VY FAC Program includes, as recommended in the GALL Report and the NSAC-202L guidelines, "procedures or administrative controls to assure that the structural integrity of all carbon steel lines containing high-energy fluids (two-phase as well as single-phase) is main-tained." Contention 4 Testimony at A17 and Exhibit E4-05 at XI.M-61. A program imple-mented in accordance with the EPRI guidelines predicts, detects, and monitors FAC in plant piping and other components, such as piping elbows and reducers, as recommended in the GALL Report. Id. The FAC Program is implemented in accordance with EPRI guidelines.

Contention 4 Testimony at A19-A20.

83. The FAC Program includes the following activities: (a) conducting an analysis to determine critical locations; (b) performing baseline inspections to determine the extent of thinning at these locations; and (c) performing follow-up inspections to confirm the predictions, or repair-ing or replacing components as necessary. Contention 4 Testimony at Al 8.

84. To ensure that all of the aging effects caused by FAC are properly managed, NRC guidance recommends that the FAC Program make use of, among other tools, a predictive computer program, such as CHECWORKS, that implements the guidance in NSAC-202L to satisfy the criteria specified in 10 C.F.R. Part 50, Appendix B, "criteria for development of procedures and control of special processes." Contention 4 Testimony at Al 8 and Exhibit E4-05 at 35

XI.M-61. The FAC Program uses CHECWORKS for that purpose. Contention 4 Testimony atA19-A21.

85. VY is a relatively small and simple plant. There are fewer FAC-susceptible systems and piping components than at a typical plant, and many of those were either originally con-structed of FAC-resistant materials or have been replaced with FAC-resistant materials since their initial installation. The original plant design and the component replacements result in a significantly smaller amount of FAC-susceptible piping at Vermont Yankee as compared to the typical nuclear power plant of similar size. Contention 4 Testimony at A22-A23.

86. At VY, the addition of oxygen into the condensate/feedwater stream mitigates the effects of FAC on piping exposed to single phase flow. Contention 4 Testimony at A24.

b. CHECWORKS

87. CHECWORKS is a multi-purpose computer program designed to assist FAC engineers in identifying potential locations of FAC vulnerability. Contention 4 Testimony at A26.

CHECWORKS utilizes plant-specific user inputs defining (1) the oxygen concentration in the feedwater and at the reactor steam effluent (".g., main steam nozzle); (2) thermodynamic con-ditions; and (3) flow rates, to calculate the water chemistry at each location in the model.

These inputs are applied, together with user-defined component geometry, to an EPRI-proprietary algorithm (the Chexal-Horowitz correlation) to provide an estimate of the rate of FAC for each modeled component. Contention 4 Testimony at A27.

88. CHECWORKS is used to model a particular nuclear unit by specifying global plant data, in-cluding a schematic representation of the plant heat balance diagram ("HBD") (i.e., major lines and connectivity of the power producing portion of the nuclear plant). The HBD model constructed in CHECWORKS is then populated with the thermodynamic conditions representative of each power level at which the plant has operated at or is contemplated to operate. The user then inputs the oxygen concentration conditions that have been used or 36

are anticipated. These inputs define the operational history of the plant in terms of what power levels have been used with what water chemistry for how long. A "Pass 1 Analysis" is conducted in CHECWORKS to report predicted wear rates. The results of the Pass 1 Analysis, together with other information including operating experience at similar units, are normally used by the FAC engineer to generate a list of components for inspection. Conten-tion 4 Testimony at A28.

89. Inspection data may be input into CHECWORKS in the form of a matrix of thickness read-ings covering the components, typically taken from ultrasonic measurements of the wall thickness at local points (i.e., grid points) or from scanning the component and recording the minimum thickness at grid points. Inspection data are not required for a Pass 1 Analysis.

Contention 4 Testimony at A28.

90. When inspection data are available, a "Pass 2 Analysis" can be run. A Pass 2 Analysis com-pares the measured inspection results to the calculated wear rates and adjusts the FAC rate calculations to account for the inspection results. Contention 4 Testimony at A28.

91. CHECWORKS calculations have been subject to verification. The correlations in one of the predecessor programs to CHECWORKS, CHEC, were initially based on FAC laboratory testing data from France and the United Kingdom and a combination of laboratory and plant operational data from Germany. Contention 4 Testimony at A30. When CHECMATE was written and when CHECWORKS was revised in the mid-1990s, a large amount of plant in-spection data was used to refine the accuracy of the CHECWORKS' predictions. These data sources included assembled data from a variety of U.S. nuclear units as well as available laboratory data from England, France and Germany. Contention 4 Testimony at A30 and Exhibit E4-08 at 7 7-33.

92. The use of the program does not change on account of a power uprate (or any other change in operating parameters), and remains essentially as outlined above. All that needs to be 37

done is to update plant-specific inputs into the CHECWORKS program. When a power uprate is implemented, a user simply does what he would normally do as part of any Pass 2 Analysis - update the relevant variables ( thermodynamic conditions, temperature, oxy-gen concentration, etc.), and let the program calculate the predicted FAC wear. The Pass 2 Analysis can be used as a planning tool by performing it in advance of the uprate to deter-mine if, under uprate conditions, systems and sub-systems would experience significantly greater FAC rates than those predicted before the uprate. CHECWORKS was specifically designed to accommodate power uprates and is routinely used throughout the U.S. nuclear industry for this purpose. Contention 4 Testimony at A33 and Exhibit E4-09 at ¶¶ 19, 20.

93. At VY, the only CHECWORKS inputs that affect FAC wear rate that need to be changed to model uprate conditions were the flow rate and the temperature. Contention 4 Testimony at A33. These were updated at VY upon implementation of the EPU. Contention 4 Testimony at A62 and Exhibit E4-32.

c. Adequacy of Data Collection

94. The VY FAC Program uses five criteria for selecting which components and locations will be inspected for potential FAC effects during a plant refueling outage. Those factors, which are consistent with the guidance in NSAC-202L, are: (1) pipe wall thickness measurements from past outages; (2) predictive evaluations performed using the CHECWORKS computer code; (3) industry experience related to FAC; (4) results from other plant inspection pro-grams; and (5) engineering judgment. CHECWORKS assists power plant engineers in de-termining the most likely places for FAC to occur, and thus, the key locations to inspect for pipe wall thinning, but is only one of the tools that Entergy will use for that purpose. Con-tention 4 Testimony at A40.

95. It is not necessary to "re-calibrate" or "benchmark" CHECWORKS when plants have changed their water chemistry, the power output has been increased, or other operational 38

changes have taken place. Contention 4 Testimony at A34, A41. At VY, the new values for flow rate and temperature are used as inputs into CHECWORKS, and CHECWORKS pro-vides FAC rate calculations for the modeled components under the uprated conditions. Con-tention 4 Testimony at A4 1.

96. Only the flow rate and temperature changed at VY due to the power uprate. Because these are the only changes, the FAC rates established after the uprate will be constant and the ef-fect of the uprate on FAC will, therefore, be apparent with the first inspection after the uprate. Contention 4 Testimony at A41.

97. The first post-uprate inspection at VY was performed in the Spring of 2007. Contention 4 Testimony at A41 and VY-RPT-08-0002, Rev. 0 (Exhibit E4-10). The results of that in-spection demonstrate that data from inspections before and after the uprate of large bore components in the feedwater system, which experience continuous flow, show that essen-tially no wear has occurred since the implementation of the EPU in March, 2006. Conten-tion 4 Testimony at A41 and Exhibit E4-10, Section 8. Nonetheless, as an added measure of conservatism, Entergy will increase the inspection scope by at least 50% for the first three outages following the EPU. Contention 4 Testimony at A41. In 2005, in RFO 25, the last refueling outage prior to the EPU, there were a total of 35 inspections performed, including 27 large bore inspections. Exhibit E4-38. In 2007, in RFO 26, the first outage since the EPU, the inspection scope was increased by more than 50%, as there were a total of 63 in-spections performed, including 49 large bore inspections. Contention 4 Testimony at A41 and Exhibit E4-10. While these additional inspections are not needed to "calibrate" CHECWORKS, they will provide additional, confirmatory data points for the use of the FAC Program. Contention 4 Testimony at A41.

39

d. Response to issues raised in the NEC testimony

98. NEC witness Dr. Hopenfeld alleges that, in order to establish the rate of FAC, data are needed from either: (1) inspection of all risk-significant susceptible components "operat[ing]

at [a] minimum [of] three inspection periods before a trend can be established," with "[f]ive inspection periods [being] the time interval between component inspection and the estab-lishment of a corrosion rate for a given component at a given location" (NEC-JH_36 at 15);

or (2) a "look at historic plant data in terms of the time scale for the occurrence of large, risk-significant wall thinning events," which Dr. Hopenfeld asserts is 16 years. Id. at 16.

Neither approach is supported by Dr. Hopenfeld's testimony and exhibits or industry prac-tice.

99. The operational experience cited by Dr. Hopenfeld does not indicate any problems in the proper use of CHECWORKS as part of a FAC Program nor does it support either of the ap-proaches he proposes. The plants referred to by Dr. Hopenfeld where FAC-related events occurred had no FAC program before the accident (i.e., Surry, Trojan) (IN 86-106, Exhibit E4-1 1 and IN 87-36, Exhibit E4-12), or their FAC program was not applied to the compo-nent that experienced a FAC-induced failure (i.e., San Onofre) (NEC-JH_46) or had a FAC program that did not follow the guidance in NSAC-202L (i.e., Clinton, Fort Calhoun and Mihama) (NEC-JH_51 at 1-2; NEC-JH_53, Section 6). Contention 4 Testimony at 42.

100.NEC witness Mr. Witte states that "[s]eparate industry guidance supports five to ten years of data trending. Trending to the high end of the range is appropriate where variables affecting wear rate, such as flow velocity, have significantly changed, as at VYNPS following the 120% power up-rate." NEC-UC_03 at 22. This statement, however, is not supported by the document cited, which is not "industry guidance," but a report produced at the behest of the Petroleum Safety Authority of Norway regarding aging management and life extension in the U.S. nuclear power industry. NEC-UW_13 at iii. The document discusses starting a 40

condition assessment program where no aging management program has been in place pre-viously. The reference is not applicable to an established program and has nothing to do with CHECWORKS. Contention 4 Testimony at A43.

101.Dr. Hopenfeld (NEC-JH_36 at 15) asserts that one of the reasons a 10-15 year period of data collection is needed to benchmark CHECWORKS for use at VY is that there was a reduc-tion in the oxygen content of the plant, further increasing the potential for FAC. Dr.

Hopenfeld cites in support of this statement page 3.2 of the summary report of the evalua-tion performed by SIA on environmentally assisted fatigue, provided by NEC as NEC ex-hibit NEC-JH_18 at 3.2.

102.This assertion, however, is not supported by the document he cites and is directly contra-dicted by uncontradicted data provided in the License Renewal Application. Measured plant data demonstrate no reduction in the oxygen concentrations in the feedwater system. Ex-hibit E4-18.

103.NEC's consultants assert that the EPU of 20% represents a situation where the predictive ef-ficacy of CHECWORKS will be diminished and that data from existing plant experience cannot be used to predict the effect of post-uprate conditions. Dr. Hopenfeld (NEC-JH_36 at 15) ("... without specifying how each variable separately effects corrosion, does not ad-dress the issue of how the corrosion rate at a given location would be affected when the ve-locity changes by 20% at a given plant."); Mr. Witte (NEC-UW_03 at 22-23) ("... VYNPS is unique in its approach of Constant Pressure Power Up-rate to 120%").

104.These claims do not have any merit. The algorithms in CHECWORKS used to predict the FAC wear rate are based on extensive laboratory and plant data, including data on FAC wear rates where the flow rate and the temperatures exceed those present at VY after the uprate. This assures that the FAC wear rates predicted by CHECWORKS are accurate.

41

Multiple plants, in addition to VY, have used CHECWORKS as part of an EPU. Contention 4 Testimony at A45 and Exhibit E4-09 at ¶¶ 19-21.

105.Mr. Witte's assertion that "... 50% of those [plants] have experienced FAC related prob-lems" (NEC-UW_03 at 23) is also not supported by any data. The plants referred to by Mr.

Witte did not have FAC-related problems. Contention 4 Testimony at A46 and Exhibits E4-13 through E4-17.

106. Dr. Hopenfeld asserts that CHECWORKS and its predecessor codes are not acceptable for predicting FAC because FAC is non-linear and local, and "the required correct inputs that account for local turbulence are not included in CHECWORKS." NEC-JH_36 at 6-7; see also id. at 4, 11, 12, 15. Laboratory tests, however, demonstrate that FAC, unlike erosion, causes damage in a manner that is linear with time. Contention 4 Testimony at A47 and Ex-hibit E4-19; Exhibit E4-08 at 7-6 and Figures 3-6 and 3-7.

107.The normal feature of FAC wear - widespread wear over an extended area - is what causes significant problems (e.g., the need for pipe replacements or the occurrence of pipe rup-tures). The global nature (i.e., widespread effect) of the FAC damage is consistent with the experience of FAC-induced ruptures. The photographs of failures at Surry, Fort Calhoun, and at a Czech nuclear unit, for example, clearly show the large area of thinned material.

Contention 4 Testimony at A47 and Exhibit E4-08 at Figures 1-3, 4-31 and 4-41.

108.The NSAC-202L guidance, which is also the practice at VY, calls for the inspection of the entirety of each component and the attached piping or the piping section selected for inspec-tion. The inspection is conducted in sufficient detail to identify any FAC-caused degrada-tion anywhere in the entire location being inspected. In virtually all cases, the degradation caused by FAC occurs over a fairly wide area (comparable to the size of the fitting). There-fore, pinpointing the "exact" location in a component or piping section where FAC will oc-cur is infeasible and unnecessary. Contention 4 Testimony at A47.

42

109.Dr. Hopenfeld claims (NEC-JH_36 at 7) that "it is the local flow velocity that directly con-trols the local turbulence and not the average velocity" and that CHECWORKS is flawed because it is "based on average flow velocities." This claim lacks merit because CHECWORKS uses "geometric factors" to relate the maximum degradation occurring in a component, such as an elbow, to the degradation predicted to occur in a straight pipe. This approach was developed by Keller in the 1970s and is similar to the approach taken in other FAC computer programs. Contention 4 Testimony at A48 and Exhibit E4-08, Table 3-1 at 3-11, Table 7-1 at 7-3, and 7 7-8.

1 10.Dr. Hopenfeld's claim (NEC-JH_36 at 4) that "the mass transfer coefficient varies with the 0.8th power of the velocity for straight pipes and the square of the velocity for curved pipes" is not supported. It is contradicted by a large body of mass transfer data and correlations published in the technical literature for the geometries encountered in piping systems. For all known geometries including straight pipes, bends, and flow restrictions, the dependence of mass transfer coefficient on velocity is less than unity. Contention 4 Testimony at A49, Exhibit E4-22 at Figure 8 and Exhibit E4-23 at Figures 5, 6, 8 and 9 (showing the exponent on the Reynolds number (i.e., a dimensionless number directly proportional to the velocity) is between 0.5 and unity). This results in a dependence of FAC rate on velocity that is slightly less than linear (i.e., doubling the velocity will not quite double the rate of FAC).

Id.

111 .Dr. Hopenfeld observes (NEC-JH_36 at 12) that "CHECWORKS is not a mechanistic model ... " and claims that "the correlation of CHECWORKS was performed in an unscien-tific manner." This claim is without merit. No existing FAC model is mechanistic. The model used in CHECWORKS and other FAC programs takes a broader approach that does not deal with the microscopic processes involved, but relates physical and chemical parame-ters with the entirety of the corrosion process. Contention 4 Testimony at A50. The predic-43

tive algorithms in CHECWORKS were developed using all available plant and laboratory data and approximate, without directly reproducing, the mechanistic details of the corrosion process, such as the rate of mass transport within the porous oxide. Id.

112.Dr. Hopenfeld (NEC-JH_36 at 9-11) and Mr. Witte (NEC-UW_03 at 9-10) refer to several reactors and fossil units where FAC has allegedly not been detected in components, in some instances leading to pipe ruptures. These examples, however, do not involve any situation in which the proper use of CHECWORKS or its predecessor programs was ineffective in pre-venting a FAC failure. Contention 4 Testimony at A52.

1 13.Dr. Hopenfeld and Mr. Witte raise concerns about the implementation of the FAC Program over the last several years. Dr. Hopenfeld questions the implementation of the FAC Pro-gram as it relies on engineering judgment as one of the factors in selecting the locations where inspections will be made during refueling outages. Dr. Hopenfeld questions the im-portance VY has given to higher flow velocities during EPU operation in selecting the com-ponent locations, stating that it is the turbulence of the flow that determines FAC suscepti-bility. NEC-JH_36 at 12. He also challenges the selection of the highest length of piping as candidates for performing inspections. NEC-JH_36 at 11. Dr. Hopenfeld further states that "the selection of the correct grid size for UT measurements is one of the most critical in-spection tasks" and criticizes the CHECWORKS guidelines for selection of grid size as they are applied at VY. NEC-JH_36 at 7, 14-16. Dr. Hausler (NEC-RH_03 at 9) also states, "It would be erroneous for the utility to continue to rely on grids established prior to EPU since these grids may not specifically capture the FAC phenomena observed at lesser velocities."

114.None of these concerns have merit. Entergy's FAC program takes risk significance and component susceptibility to failure into account. Contention 4 Testimony at A56 and Ex-hibit E4-06, Sections 5.2 and 5.3. Dr. Hopenfeld's (NEC-JH_36 at 11) claim that Entergy believes "that length and the highest velocities control corrosion," is based on a mistran-44

scribed quote in the transcript of the November 30, 2005 Advisory Committee on Reactor Safeguards meeting. The discussion in the transcript around the misquoted testimony con-cerned the inspections to be performed during three upcoming refueling outages. In the discussion, Entergy was stating that, given the low wear rates that had been measured, En-tergy would be inspecting the locations that had the highest length of time since the last in-spection and the locations with the highest velocities. Contention 4 Testimony at A56.

1 15.At VY, the grid size for inspections is specified by an Engineering Standard, "Flow Accel-erated Corrosion Component Scanning and Gridding Standard." Contention 4 Testimony at A56 and Exhibit E4-25. There are two aspects to grid size: (1) when degradation is found, the grid size is normally made smaller in that area to more accurately define the wear area; and (2) in inspecting a component, the larger the pipe, the larger amount of material that may be lost before the component fails, allowing for a "larger" grid (i.e., the defect size that would cause failure varies directly with the size of the pipe). Both of these approaches are consistent with NSAC-202L, Rev. 2, Section 4.5.3. Contention 4 Testimony at A56. At VY, an additional step is taken in performing the inspections. Rather than recording the thickness reading at particular grid points, the components inspected at VY are scanned in their entirety. This is done by moving an ultrasonic transducer over the entire surface within a grid "square." The data logger automatically records the minimum reading anywhere within the grid square and the qualified inspector verifies that reading. This ensures that the thinnest readings in the component are found. Contention 4 Testimony at A56, A57.

116.Mr. Witte asserts that data from previous FAC inspections (prior to the EPU) were not en-tered into the CHECWORKS database (NEC-UW_03 at 2, 3, 6, 7-8, 15, 16, 17); that CHECWORKS was not updated with the uprate parameters (id. at 5, 23); that, for the period 2000-2006 VY failed to use a current version of CHECWORKS (id. at 6, 17); that four com-ponents were predicted in 2004 to have wall thinning beyond operability limits (id. at 17-18, 45

22); that open corrective actions identified in condition reports may not have been com-pleted (id. at 3-4, 18-19); that ranking of small bore piping was not done (id. at 8, 20); that the number of inspection points were reduced after the 2005 outage (id. at 7, 8, 20); and that the 2006 refueling outage inspection "scope, planning, documentation, and procedural analysis appear to have been performed under a superseded program document" (id. at 5, 7, 20-21).

117.None of these claims have merit. There is no factual support for Mr. Witte's assertions (NEC-UW_03 at 15) Entergy was "aware of the problematic state of the program for many years" or that (NEC-UW_03 at 2, 18) the FAC Program was "unsatisfactory." The quality assurance audit cited by Mr. Witte, QA-8-2004-VY-1, categorically states that "[n]one of the findings or areas for improvement, individually or in the aggregate, were indicative of significant programmatic weaknesses which would impact the overall effectiveness of the Engineering Programs assessed." Exhibit E4-26 at 2. Mr. Witte does not explain what sig-nificance, if any, that quality assurance audit has with respect to the adequacy of the FAC Program. Contention 4 Testimony at A60. The CHECWORKS models were updated with all applicable inspection data during the Summer and Fall of 2000. Contention 4 Testimony at A61, A62 and Exhibit E4-28. Additional updates were performed for the feedwater sys-tem in 2003. Contention 4 Testimony at A61, A62 and Exhibit E4-29. Another CHECWORKS update was performed in 2006. Contention 4 Testimony at A61, A62 and Exhibit E4-30. The CHECWORKS update performed in 2006 confirmed again that the pre-viously predicted wear rates were conservative. Contention 4 Testimony at A62 and Exhib-its NEC-UW_10 and E4-31, E4-35 through E-38. As inspection data were obtained and in-corporated into the models, Pass 2 Analyses were performed and the predicted wear rates were correlated to the measured data. In all cases, the inclusion of the inspection data re-duced the predicted wear rates and increased the times to minimum wall thickness. Thus, 46

not entering data from a particular inspection into CHECWORKS would not suggest that "susceptible locations may not have been inspected." Id.

118.Mr. Witte states (NEC-UW_03 at 19) that "the VY FAC program was primafacie in non-compliance with its CLB" because "in 2005 a sixth CR was written, CR-VTY-2005-02239, stating 'CHECWORKS predictive model for Piping FAC inspection program was not up-dated per appendix D of PP 7028."' There is no basis for Mr. Witte's assertion that not en-tering the most recent inspection data into CHECWORKS rendered the FAC Program in noncompliance with VY's current licensing basis ("CLB"). VY's CLB incorporated the recommendations in EPRI NSAC-202L, Rev.2 (Exhibit E4-33) by reference in FAC Pro-gram Procedures PP7028 (Exhibit E4-34) and, later, ENN-DC-315, Rev. 1 (NEC-UWI 2),

which were in effect during the period from 1999 to 2006. Section 4 of NSAC-202L, Rev.

2, does not specify a specific interval for model updates. It states: "It is recommended that whenever possible, the Predictive Plant Model utilize the results of wall thickness inspec-tions to enhance the FAC predictions. In CHECWORKS this is called Pass 2 analysis."

Because there is no specific interval required for entering additional inspection data into CHECWORKS, no departure from the CLB took place. Contention'4 Testimony at A63.

119.VY updated the version of CHECWORKS it used from CHECWORKS FAC 1.OD to CHECWORKS FAC 1 .OF in 2000. Contention 4 Testimony at A65 and Exhibit E4-28.

Version 1 .OF was used for the 2003 and 2006 model updates. CHECWORKS FAC 1.OG was installed in 2006. There were no differences in versions 1.OD, 1 .OF, and 1.OG with re-spect to water chemistry and wear rate predictions for BWRs (Exhibit E4-39). Nothing re-garding the version in use at any particular point in time had any effect on the use of CHECWORKS as a tool as part of the FAC Program. Contention 4 Testimony at A65.

120.Mr. Witte states (NEC-UW_03 at 17) that, "[iun 2004, at least four VYNPS components, in-cluding the condensate system and the extraction steam systems, were determined to have 47

'negative time to Tmin,' meaning that wall thinning was being predicted as beyond operabil-ity limits and should be considered unsafe with potential rupture at anytime." These state-ments demonstrate a misunderstanding of how CHECWORKS is used at VY. The "deter-mination" that the four components had "negative times to Tmin" is a theoretical conclusion based on the results of CHECWORKS, and is not based on actual inspection data. As such, there would be no need to "write condition reports for this condition," (NEC-UW_03 at 18) as Mr. Witte states. CRs are written when inspection data indicate there is an actual prob-lem and additional inspections are then performed as corrective actions. If a planning tool, like CHECWORKS, indicates an area of potential concern, inspections of that area are scheduled. The only FAC susceptible component identified in the 2004 Scoping Worksheets (CD30TE02DS) was scheduled for inspection. The actual inspection data show that the en-tire component meets design code with significant margin. Contention 4 Testimony at A66 and Exhibit E4-37 at 12.

121. Mr. Witte states (NEC-UW_03 at 19): "The 2006 cornerstone report shows a number of in-dicators as yellow, with lists of open CR corrective actions, and a new CR written in August 30, 2006. The report lists six corrective actions and four CRs that were written as early as 2003 that remain open." The FAC Program Health Report, "Cornerstone Rollup" shows the overall FAC Program status as Green. Exhibit NEC-UW_07 at 1. The report rates twenty-seven different areas. Of these, two were rated as "Yellow": (1) Owner Availability and (2)

Open Actions Items. Id. at 4, 6. A yellow indicator for Open Action Items is triggered if any action item, regardless of its importance, is more than one year old. Six LO-VTYLO action items are listed. These are not Condition Reports, nor are they corrective actions from condition reports. They are commitments. The Corrective Action Program is used to track all commitments. There is no safety significance to these commitments. The items listed are for completion of program administrative tasks. Contention 4 Testimony at A67.

48

122.Mr. Witte states (NEC-UW_03 at 20): "Ranking of small bore piping was not done. With no ranking, the basis for selection of high susceptible points for small bore piping is not evi-dent." However, at VY, the initial scoping and inspection selection of small bore piping was performed in 1993 and 1995. The scope and criteria for determining the inspection locations is documented in FAC Program documents (Exhibits E4-41 and E4-42). The small bore in-spections were initiated prior to the inclusion of small bore guidance provided in NSAC-202L. Contention 4 Testimony at A68.

123.Mr. Witte (NEC-UW_03 at 20) states: "A flow-accelerated corrosion related pipe break as-sociated with a 1" elbow, SSH (WO 06-6880), appears to have occurred in 3 quarter 2006." However, this was not a FAC-related pipe break. A pinhole leak was identified dur-ing operator rounds on an elbow on the 1" drain line from the steam seal header to the con-denser. No "pipe break" had occurred. The elbow was replaced in RFO 26. The damage found was due to droplet impingement, not FAC. Contention 4 Testimony at A69.

124.Mr. Witte states (NEC-UW_03 at 20) that "Entergy apparently reduced the number of FAC inspection data points between the 2005 and 2006 refueling outage, in violation of its com-mitment to increase inspection data point by 50%. The 2005 refueling outage inspection called for 137 large-bore inspection points. The 2006 refueling outage inspection presented to the ACRS on June 5, 2007, covered only 63 points."

125.This claim is not supported. There was no refueling outage in 2006. In RFO 25 in the Fall of 2005, a total of 35 inspections were performed. Of those, 27 were large bore UT inspec-tions. Exhibit E4-38. In RFO 26 (conducted in the Spring of 2007), a total of 63 inspections were performed. Of those, 41 were large bore UT inspections. Exhibit E4-38. The increase in the number large bore UT inspections from RFO 25 to RFO 26 was more than 50%, in accordance with Entergy's plans. Contention 4 Testimony at A70.

49

126.Mr. Witte states (NEC-UW_03 at 20): "The 2006 Refueling outage FAC Inspection scope, planning, documentation, and procedural analysis all appear to have been performed under a superseded program document. ENN-DC-315 Rev. 1 was effective March 15, 2006." This claim lacks merit. The guidance used to perform the scoping had not been superseded. The scoping process for RFO 26 started before RFO 25 was complete and well before the March 15, 2006 effective date of ENN-DC-315, Rev.1 (NEC-UW_12). The RFO 26 scoping was performed using the same criteria as contained in Section 5.3 of ENN-DC-315, Rev. 1. The scoping criteria in ENN-DC-315, Rev. 1, is the same as under the superseded VY procedure PP 7028 (Exhibit E4-34). Contention 4 Testimony at A71.

4. Conclusions to be drawn from the evidence 127.In summary, Entergy has had an effective FAC Program in place at VY for over twenty years. The program has detected wear as designed and components have been replaced prior to thinning below minimum design thickness. The FAC Program at VY uses CHECWORKS as a tool in planning inspections, evaluating inspection data, and managing the UT data col-lected. While an effective tool, it is only one of several used in the VY FAC Program to identify the locations to be inspected during refueling outages of the plant. CHECWORKS has a well-established track record of use in FAC management programs, including BWRs which have undergone uprates. The input of new values for the plant-specific variables af-fected by the uprate at VY - flow rate and temperature - is all that is required for CHECWORKS to continue to be used as part of the FAC Program. The current FAC Pro-gram, which will be the FAC Program used during the license renewal period, meets indus-try practice as reflected in NSAC-202L and has been reviewed, audited and inspected with only minor, mostly administrative, issues identified. The FAC Program that will be used during the period of extended operation after license renewal will assure that the aging ef-50

fects of FAC will be adequately managed. For that reason, there is no support for the claims made in NEC Contention 4 and it should be rejected.

Respectfully Submitly.d, #

David R. Lewis Matias F. Travieso-Diaz Blake J. Nelson PILLSBURY WINTHROP SHAW PITTMAN LLP 2300 N Street, N.W.

Washington, DC 20037-1128 Tel. (202) 663-8000 Counsel for Entergy Nuclear Vermont Yankee, LLC, and Entergy Nuclear Operations, Inc.

Dated: May 13, 2008 51

ATTACHMENT 1 INDEX OF EXHIBITS TO ENTERGY'S DIRECT TESTIMONY EXHIBITS ON NEC CONTENTION 2 (ENVIRONMENTALLY ASSISTED FATIGUE)

Exhibit E2 Joint Declaration of James C. Fitzpatrick and Gary L. Stevens Exhibit E2 James C. Fitzpatrick Resume Exhibit E2 Thadani December 26, 1999 Memo Exhibit E2 NUREG CR-6260 (excerpt)

Exhibit E2 Section X.M1 of GALL Vol 2 Exhibit E2 NUREG CR-6583 Exhibit E2 NUREG CR-5704 Exhibit E2 Gary L. Stevens Resume Exhibit E2 VY LRA Amendment 35 Exhibit E2 VY- 16Q-301 R0 Exhibit E2 VY-16Q-302R0 Exhibit E2 VY- 16Q-303r0 REDACTED Exhibit E2 VY-16Q-304R0 Exhibit E2 VY-16Q-305r0 Exhibit E2 VY-16Q-306R0 Exhibit E2 VY-16Q-307r0 Exhibit E2 VY-16Q-308R0 Exhibit E2 VY-16Q-309R1 Exhibit E2 VY- 16Q-31 OR1 Exhibit E2 VY-16Q-31 IRO Exhibit E2 VY- 16Q-401 rO REDACTED Exhibit E2 VY-16Q-402r0 2

Exhibit E2 VY-16Q-403rl Exhibit E2 VY-16Q-404rl REDACTED Exhibit E2 VY-19Q-301r0.

Exhibit E2 VY-19Q-302r0 Exhibit E2 VY-19Q-303r0 Exhibit E2 VY LRA Amendment 34 Exhibit E2 List of Drawings Provided to NEC Exhibit E2 NUREG CR-6909 Exhibit E2 ACRS 02072008 Meeting Transcript(excerpt)

Exhibit E2 WRC Bulletin 487 (excerpt)

Exhibit E2 Excerpts from 2007 VY Feedwater Nozzle Inspection Report Exhibit E2 Excerpt from EPRI Report BWVRVIP 120 Exhibit E2 VY LRA Amendment 33 Exhibit E2 Transcript of ACRS Meeting 030608 (excerpt)

Exhibit E2 VYACRS Letter 3-20-2008 EXHIBITS ON NEC CONTENTION 3 (STEAM DRYER)

Exhibit E3 Joint Declaration of John R. Hoffman and Larry D. Lukens re NEC Contention 3 Exhibit E3 John R. Hoffman's Resume Exhibit E3 Larry D. Lukens' Resume Exhibit E3 Attachment 2 to EPU Application Exhibit E3 Attachment 6 to Supplement 33 to EPU Application Exhibit E3 GE-SIL 644 Rev. 1 Exhibit E3 Off-Normal Procedure ON 3178 Exhibit E3 Excerpt from GALL Report, Vol 2 Exhibit E3 GE-SIL 644 Rev. 2 3

Exhibit E3 VY Operating Procedure OP 0631 Exhibit E3 VY Procedure EN-OP-0104 Exhibit E3 VY-RPT-06-00006 Exhibit E3 VY 2004 Dryer Inspection Results Summary Exhibit E3 VY 2005 Dryer Inspection Results Summary Exhibit E3 VY 2007 Dryer Inspection Results Summary Exhibit E3 2007 GE Dryer Flaw Evaluation Report (Excerpt)

EXHIBITS ON NEC CONTENTION 4 (FLOW-ACCELERATED CORROSION)

Exhibit E4 Joint Declaration of Horowitz and Fitzpatrick re NEC Contention 4 Exhibit E4 Horowitz CV Exhibit E4 Fitzpatrick Exhibit E4 Section B.1.13 of LRA Exhibit E4 Section XI.M17 of GALL Report Exhibit E4 EN-DC-315 Exhibit E4 NSAC_202LR3_050206 (non-proprietary)

Exhibit E4 FAC Book selected pages Exhibit E4 Declaration of Neil Wilmshurst Exhibit E4 VY-RPT-08-0022 Exhibit E4 IN 86-106 (Surry)

Exhibit E4 IN 87-36 (Trojan)

Exhibit E4 OE- 17412 (Clinton)

Exhibit E4 OE- 18478 (Clinton)

Exhibit E4 OE-20246 (Clinton)

Exhibit E4 OE-17654 (Clinton)

Exhibit E4 OE-21421 4

Exhibit E4 LRA A3 FDW 02 at LPH PPB Exhibit E4 Effects of Chemistry on Corrosion-Erosion Exhibit E4 OE-15860 Exhibit E4 OE-20127 (Calvert Cliffs)

Exhibit E4 Heat and Mass Transfer Downstream Exhibit E4 The Use of a Corrosion Process Exhibit E4 Mihama Final Exhibit E4 ENN-EP-S-005 Exhibit E4 QA-8-2004-VY-1 Exhibit E4 EPRI Letter 2-28-2000 Exhibit E4 EPRI CHECWORKS Wear Rate Cycles 20&21 Exhibit E4 EPRI CHECWORKS WEAR RATE Cycle22B (2003)

Exhibit E4 EPRI CHECWORKS WEAR RATE Cycle25 (2006)

Exhibit E4 CR-VTY-2006-02699 Exhibit E4 VY-RPT-05-00012-R0 Exhibit E4 NSAC-202L-R2 Exhibit E4 PP 7028 Exhibit E4 2001 RFO Inspection Report Exhibit E4 2002 RFO Inspection Report Exhibit E4 2004 RFO Inspection Report Exhibit E4 2005 RFO Inspection Report Exhibit E4 v1.OGEnhancements Exhibit E4 Inspection Location 2004 RFO Exhibit E4 FAC Small Bore Database RevO Exhibit E4 FAC Small Bore Database Revl 5

ORIGINAL May 12, 2008 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

JOINT DECLARATION OF JAMES C. FITZPATRICK AND GARY L.

STEVENS ON NEC CONTENTION 2A/ 2B -

ENVIRONMENTALLY ASSISTED FATIGUE James C. Fitzpatrick and Gary L. Stevens state as follows under penalty of perjury:

1. We have prepared the attached "Testimony of James C. Fitzpatrick and Gary L. Stevens on NEC Contention 2A/2B - Environmentally Assisted Fatigue" in the above captioned proceeding.

2. The factual statements and opinions we express in the cited testimony are true and correct to the best of our personal knowledge and belief.

3. We declare under penalty of perjury that the foregoing is true and correct.

Executed on May 12, 2008 J s C4 itzpatrick

A~aryL.

Stevens Executed on May 12, 2008

May 12, 2008 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

TESTIMONY OF JAMES C. FITZPATRICK AND GARY L. STEVENS ON NEC CONTENTION 2A / 2B - ENVIRONMENTALLY ASSISTED FATIGUE I. WITNESS BACKGROUND James C. Fitzpatrick ("JCF")

Q1. Please state your full name.

Al. (JCF) My name is James C. Fitzpatrick.

Q2. By whom are you employed and what is your position?

A2. (JCF) I am currently employed by AREVA, NP as an Engineering Supervisor. Until March 7, 2008, 1 was employed by Entergy Nuclear Operations, Inc. ("Entergy") as a Senior Lead Engineer in Design Engineering at the Vermont Yankee Nuclear Power Station ("VY").

Q3. Please summarize your educational and professional qualifications.

A3. (JCF) My professional and educational experience is described in the curriculum vitae attached to this testimony as Exhibit E2-02.

Briefly summarized, I have thirty years experience in design, construction, and modification of nuclear power plant structures, piping systems, pressure vessels, and in the seismic evaluation of

mechanical and electrical equipment. Twenty-two of those years are in operating plant engineering support in both the mechanical and structural areas. I have been responsible for the development and implementation of plant design changes, inspection programs, equipment specifications, installation support, outage support, and operability evaluations of degraded components.

Q4. What is the purpose of your testimony?

A4. (JCF) The purpose of my testimony is to address Contentions 2A and 2B submitted by the New England Coalition ("NEC") in this proceeding. As admitted by the Atomic Safety and Licensing Board ("Board"), and as subsequently modified by the Board on November 7, 2007, NEC Contention 2A (which encompasses Contention 2B) reads:

... [T]he analytical methods employed in Entergy's [environmentally corrected CUF, or] CUFen Reanalysis were flawed by numerous uncertainties, unjustified assumptions, and insufficient conservatism, and produced unrealistically optimistic results. Entergy has not, by this flawed reanalysis, demonstrated that the reactor components assessed will not fail due to metal fatigue during the period of extended operation.

Memorandum and Order (Ruling on NEC Motions to File and Admit New Contention), LBP-07-15, 66 N.R.C. 261, 270 (2007).

Q5. What is metal fatigue?

A5. (JCF) In general, metal fatigue is an age-related degradation mechanism caused by cyclic mechanical and thermal stresses at a location on a metallic component. The results of fatigue can be observed in the cracking of components subjected to cyclic stresses of sufficient magnitude and duration.

2

There are two general categories of fatigue which are important in nuclear power plant design and analysis. These are low cycle fatigue ("LCF") and high cycle fatigue ("HCF"). There are substantive differences between the two in terms of the major factors controlling fatigue life. High cycle fatigue is generally associated with machinery and rotating equipment with small fluctuations in stress and strain and a high number of applied cycles. The design of the American Society of Mechanical Engineers ("ASME") Boiler and Pressure Vessel Code ("Code")

Class 1 (pressure-retaining boundary) vessels and piping is controlled by LCF and is associated with a relatively small number of large, localized fluctuations in stress and strain typically caused by thermal and pressure transients.

Q6. How are the fatigue limits on nuclear power plant components established?

A6. (JCF) The design specifications for a given safety-related component specify the number of mechanical and thermal cycles that the component is expected to experience during its design life, and define the safety limits and applicable codes that must be satisfied. For components exposed to the primary reactor coolant pressure boundary, the specified requirements for evaluation of cyclic loading and thermal conditions are contained in Section III of the ASME Code for Class 1 components.

Q7. What are the cumulative usage factors or "CUFs" for nuclear power plant components?

A7. (JCF) For a Class 1 component, stress cycles from the loadings specified in the governing design specification will produce total stresses of several different magnitudes. The number of times these stress magnitudes occur also varies. The cumulative effect of the various stress magnitudes and associated number of cycles is evaluated by means of a linear damage relationship known as 3

Miner's Rule. The allowable number of cycles for a given alternating stress range is determined from the ASME design fatigue curve for the material being evaluated. The fatigue usage for that stress cycle is the ratio of the number of applied stress cycles (n) to the allowable number of stress cycles (N) from the ASME design fatigue curve. The cumulative usage factor

("CUF") for the component is the sum of the individual usage factors for all of the various stress magnitudes.

At any point in time, the CUF for a component represents the fraction of the allowable fatigue cycles that the component has experienced up to that time.

Q8. Is there a limit on the acceptable CUF factors?

A8. (JCF) Yes. An ASME Code Section III criterion requires that the CUF for a Class 1 component not exceed unity. Stated another way, the total number of applied stress cycles is not to exceed the allowable number of stress cycles. Because of safety factors included in the development of the ASME Code design fatigue curves, the use of a 1.0 limit on the CUFs is intended to provide an additional safety factor in the design against component failure. Therefore, exceeding the 1.0 limit does not imply that the component will fail.

Q9. What is environmentally assisted fatigue or "EAF"?

A9. (JCF) For components (equipment and piping) exposed to reactor coolant water, the fatigue life, as measured by the allowable number of stress cycles, is reduced compared to the components' fatigue life when exposed to an air environment. The ASME design fatigue curves were developed based on laboratory testing of specimens in an air environment, with safety factors incorporated into the curves to account for several factors, 4

including environment. Laboratory testing of specimens in water under reactor operating conditions indicates that, under certain situations additional environmental factors may need to be included in the calculated CUF to fully accommodate reactor coolant operating conditions. Accounting for the effects of operating in a reactor coolant environment in the fatigue analysis is called environmentally assisted fatigue ("EAF") analysis.

EAF analysis was addressed in NRC Generic Safety Issue ("GSI")

190, "Fatigue Evaluation of Metal Components for 60-year Plant Life."

Q10. How was GSI 190 resolved?

A10. (JCF) GSI 190 was closed out by the NRC Staff in December 1999 without the imposition of additional requirements on operating reactors for their initial license term (40 years), based on a determination that the potential for EAF posed no significant increase in the risk of component failure (see Exhibit E2-03).

However, the NRC Staff concluded that, due to the increase in the probability of leakage at plants operating beyond their original 40-year license term, licensees should address the effects of coolant environment on component fatigue life as aging management programs are formulated in support of license renewal. Id.

Qll. Where are the standards to be used in EAF analyses for license renewal applications set?

All. (JCF) As stated in NUREG- 1800, Section 4.3.2.2, one method acceptable to the Staff for performing EAF analyses is to assess the impact of the reactor coolant environment on those critical components identified in NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, U.S. Nuclear Regulatory Commission, February 1995" ("NUREG/CR-6260," Exhibit E2-04) for the 5

appropriate vendor/vintage plant. These critical components may be evaluated by applying environmental correction factors to the existing ASME Code fatigue analyses. Criteria and methodology for performing EAF analyses are also specified in Chapter X, "Time Limited Aging Analyses Evaluation of Aging Management Programs Under 10CFR54.21(c)(1)(iii),"Section X.M1 "Metal Fatigue of Reactor Coolant Pressure Boundary," of the Generic Aging Lessons Learned ("GALL") Report, NUREG-1801 (Rev.

1) (Exhibit E2-05). The GALL Report recommends the use of NUREG/CR-6583, "Effects of LWR Coolant Environments on Fatigue Design Curves of Carbon and Low-Alloy Steels"

("NUREG/CR-6583", Exhibit E2-06) to evaluate EAF effects for carbon and low alloy steel components, and NUREG/CR-5704, "Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels" ("NUREG/CR-5704")

(Exhibit E2-07) to evaluate EAF effects for stainless steel components.

Q12. How is the effect of the reactor coolant environment on component fatigue quantified?

A12. (JCF) As defined in NUREG/CR-6583 and NUREG/CR-5704, the CUF for a component exposed to reactor coolant may multiplied by an adjustment factor or "EAF multiplier", when appropriate environmental conditions exist. This results in an environmentally adjusted CUF or "CUFen". The resulting CUFen must still not exceed unity.

Q13. What has been your role in addressing environmentally assisted fatigue issues at VY?

A13. (JCF) As Senior Lead Engineer at VY, I was responsible for overseeing the analyses used to predict the long-term performance of critical VY components, including the potential fatigue of 6

metal piping and equipment exposed to the reactor coolant environment. I was also responsible for the development of Entergy's proposed program to manage the effects of fatigue on critical reactor pressure boundary components during the proposed VY license renewal period.

Gary L. Stevens ("GLS")

Q14. Please state your full name.

A14. (GLS) My name is Gary Lance Stevens.

Q15. By whom are you employed and what is your position?

A15. (GLS) I am a Senior Associate at Structural Integrity Associates, Inc. ("SIA").

Q16. Please summarize your educational and professional qualifications.

A16. (GLS) My professional and educational experience is summarized in the curriculum vitae attached to this testimony as Exhibit E2-08. Briefly summarized, I specialize in the application of finite element analysis, fracture mechanics, and structural and fatigue analyses to nuclear components. I have extensive experience in the application of ASME Code Sections III and XI methodology to fatigue and fracture analyses of reactor vessels and internals components. I was the Chairman of the former ASME Section XI Task Group on Operating Plant Fatigue Assessments, am the Secretary of the ASME Section XI Working Group on Operating Plant Criteria, the Secretary of the ASME Section XI Subgroup on Evaluation Standards, and a member of the ASME Section XI Subcommittee on Nuclear Inservice Inspection.

7

Q17. What is the purpose of your testimony?

A17. (GLS) The purpose of my testimony is to address those aspects of NEC Contention 2A/2B that relate to calculations of EAF of components at VY performed by SIA under my supervision.

Q18. What has been your role with respect to the EAF calculations for VY?

A18. (GLS) My role with respect to the EAF calculations for VY was to supervise the SIA technical staff involved in performing the EAF calculations, and to provide expert technical consultation and review to all aspects of the work. I also prepared one of the three calculations of the confirmatory analysis, specifically the EAF calculation for the feedwater nozzle, which now represents the calculation of record for that component for VY. Finally, I was the Project Manager for the fatigue analysis of the VY feedwater nozzle completed in 2004, portions of which were utilized in the VY confirmatory analysis.

II. DESCRIPTION OF VY'S PROPOSED EAF MANAGEMENT PROGRAM Q19. Would you please describe the program that VY proposes to implement to address EAF issues during the period following license renewal?

A19. (JCF) Section 4.3.3 of the License Renewal Application for VY

("Application") presents Entergy's initial assessment of the effects of the reactor coolant environment on fatigue life for the plant-specific locations at VY equivalent to those identified in NUREG/CR-6260 for the older vintage General Electric plants.

The component locations identified in NUREG/CR-6260 and endorsed by the GALL Report are: (1) the reactor vessel shell and lower head; (2) the reactor vessel feedwater nozzle; (3) the reactor recirculation piping (including the reactor inlet and outlet nozzles); (4) the core spray line reactor vessel nozzle and 8

associated Class 1 piping; (5) the residual heat removal ("RHR")

return line Class 1 piping; and (6) the feedwater line Class 1 piping. Due to the inclusion of both piping and nozzles, as well as the different materials for the nozzle forgings and nozzle safe ends, a total of nine locations for the six components identified in the NUREG/CR-6260 list above were evaluated for EAF at Vermont Yankee.

The initial assessment of EAF contained in the VY Application, which has been refined since the initial filing of the Application, consists of the evaluation of EAF effects for all nine locations in accordance with the provisions of Section X.M1 of the GALL Report by performing CUFen calculations and demonstrating that the total CUFen for 60 years of plant operation remains less than unity. These calculations make use of projected occurrences of plant transients for 60 years of operation.

VY's proposed program to address EAF issues during the period of extended operation is to continue to monitor all relevant plant transients and ensure that the calculations remain valid throughout the period of extended operation.

Q20. How is the analysis of EAF for those components and locations performed?

A20. (JCF) The criteria and methodology for performing EAF analyses are specified in Section X.M1 of the GALL Report. The methodology is comprised of three steps: (1) the CUF for a component is calculated; (2) the environmental multiplier, Fen, is calculated in accordance with NTUREG/CR-6583 and NUREG/CR-5704; and (3) the CUFen is calculated as the product of the CUF for the component and the corresponding Fen.

9

Q21. Does the Application contain calculated CUFen values for the limiting piping and vessel locations specified in NUREG/CR-6260?

A21. (JCF) Yes. Section 4.3 of the Application documents the initial evaluation of metal fatigue for Class 1 and selected non-Class 1 components for the period of extended operation. Piping components in limiting locations did not have specific CUFs because they were designed under American National Standards Institute ("ANSI") Code B3 1.1, which does not require explicit fatigue evaluation. For these piping components, CUFs were reported in the Application based on generic values provided in NUREG/CR-6260. The CUF values reported in NUREG/CR-6260 that are applicable to VY are based on fatigue analyses performed for a representative BWR plant that also originally used ANSI Code B31.1.

Table 4.3-1 of the Application shows the CUFs for the various VY Class 1.components based on the number of transients projected to occur over a 60-year operating life. The number of transient cycles projected to be experienced by each component through the period of extended operation is shown in Table 4.3-2 of the Application. The numbers of transients provided in Table 4.3.2 were conservatively calculated by projecting the numbers of cycles that had occurred as of May 2004 through the period of extended operation of the plant.

The Fen for each of the nine limiting piping and vessel locations was calculated based on the methodology from NUREG/CR-6583 (Exhibit E2-06) for the carbon and low alloy steel components, and NUREG/CR-5704 (Exhibit E2-07) for the stainless steel components, as set forth in Section X.M1 of the GALL Report.

10

Finally, the CUFs for each location were multiplied by the Fen values to obtain the resulting CUFens.

The initial CUFenS computed by Entergy for VY are tabulated in Table 4.3.3 of the Application. As that table shows, seven of the nine locations had CUFens greater than unity, and therefore greater than the specified criterion of the ASME Code. However, some of those results were not plant-specific, and have been superseded by subsequent refined, VY-specific analyses.

Q22. What actions are proposed in the Application to address these results?

A22. (JCF) The Application states (Application, Section 4.3.3 at 4.3-7):

Prior to entering the period of extended operation, for each location that may exceed a CUF of 1.0 when considering environmental effects, VYNPS will implement one or more of the following:

(1) further refinement of the fatigue analyses to lower the predicted CUFs to less than 1.0; (2) management of fatigue at the affected locations by an inspection program that has been reviewed and approved by the NRC (e.g., periodic non-destructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC);

(3) repair or replacement of the affected locations.

Should VY select the option to manage environmental-assisted fatigue during the period of extended operation, details of the aging management program such as scope, qualification, method, and frequency will be provided to the NRC prior to the period of extended operation.

Q23. Is that Entergy's current licensing commitment with respect to environmentally assisted fatigue during the license renewal period?

11

A23. (JCF) No. This commitment was modified in Amendment 35 to the Application, which states as follows (Exhibit E2-09 hereto, Attachment 3, Commitment 27):

At least 2 years prior to entering the period of extended operation, for the locations identified in NUREG/CR-6260 for BWRs of the VY vintage, VY will refine our current fatigue analyses to include the effects of reactor water environment and verify that the cumulative usage factors (CUFs) are less than 1. This includes applying the appropriate Fen factors to valid CUFs determined in accordance with one of the following:

1. For locations, including NUREG/CR-6260 locations, with existing fatigue analysis valid for the period of extended operation, use the existing CUF to determine the environmentally adjusted CUF.

2. More limiting VY-specific locations with a valid CUF may be added in addition to the NUREG/CR-6260 locations.

3. Representative CUF values from other plants, adjusted to or enveloping the VY plant specific external loads may be used if demonstrated applicable to VY.

4. An analysis using an NRC-approved version of the ASME code or NRC-approved alternative (e.g.,

NRC-approved code case) may be performed to determine a valid CUF.

During the period of extended operation, VY may also use one of the following options for fatigue management if ongoing 12

monitoring indicates a potential for a condition outside the analysis bounds noted above:

1) Update and/or refine the affected analyses described above.

2) Implement an inspection program that has been reviewed and approved by the NRC (e.g., periodic nondestructive examination of the affected locations at inspection intervals to be determined by a method acceptable to the NRC).

3) Repair or replace the affected locations before exceeding a CUF of 1.0.

Q24. Has Entergy implemented any of these options at the present time?

A24. (JCF) Yes. Entergy engaged SIA to perform refined analyses to calculate the CUFs, Fens and CUFens for all nine locations of interest in accordance with the approach described in the GALL Report.

Q25. Did SIA perform those refined calculations?

A25. (GLS) Yes. Final versions of fifteen refined calculations and associated reports were issued in early August 2007.

Subsequently, in December 2007, four of the fifteen calculations and reports were slightly modified to correct an input data error.

The methodology and all but one of the results were unaffected by the correction. (The calculation whose results were affected showed an increase in CUFen from 0.0432 to 0.1698, a result still much less than unity.) Exhibits E2-10 through E2-24 are copies of the final version of the refined calculations and reports.

Q26. What were the results of the refined calculations?

13

A26. (GLS) The results of the refined calculations show that the CUFens for the nine limiting piping and vessel locations for the sixty years through VY's extended license period are in all cases less than unity, signifying that component failure due to fatigue will not be a concern at VY during the period of extended operation. These results were provided to the NRC in Amendment 35 to the Application (Exhibit E2-09).

Q27. What does Entergy intend to do with respect to the other options specified in the Application?

A27. (JCF) Performance of the refined analyses has demonstrated that environmentally assisted fatigue will not be a concern during the period of extended operation. Therefore, no further actions regarding metal fatigue, including implementation of the other options, are currently deemed necessary.

The condition of piping and components at the locations of interest will continue to be monitored under the plant's in-service inspection program through the period of extended plant operation. In addition, the VY Fatigue Monitoring Program will continue to track plant cycles and transients to ensure that the numbers of transient cycles experienced by the plant remain within the analyzed numbers of cycles for all transients. If, at some future time, the results of continued monitoring suggest that the evaluations no longer encompass 60 years of plant operation, implementation of the options of submitting an inspection program for NRC review or replacing the component in question may become necessary or desirable.

III. DESCRIPTION OF THE REFINED FATIGUE ANALYSIS Q28. Please describe the methodology used to perform VY's refined fatigue analysis.

14

A28. (GLS) The first step in the analysis was to re-calculate CUFs for the locations and components of interest to reflect actual VY operating experience, and to compute CUFs for the B3 1.1 piping locations. The numbers of cycles for sixty years were calculated based on the design specifications and the numbers of cycles actually experienced by the plant as of June 2007, projected out to 60 years of operation. In addition, VY has implemented an extended power uprate ("EPU"). The effects of the uprate were incorporated into the CUF computations.

The second step was to collect relevant plant operating parameters, primarily dissolved oxygen, for both pre-uprate and post-uprate operating conditions, as well as pre- and post-hydrogen water chemistry ("HWC") implementation. The Fen factors were re-calculated for each component and location using the Fen methodology published in NUREG/CR-6583 for carbon and low alloy steels, and NUREG/CR-5704 for stainless steels.

The third step was to calculate CUFens for the various components and locations as the product of the re-calculated CUFs and the re-calculated Fen factors for each component and location.

Q29. Does the refined methodology you just discussed differ from the previous approach to computing CUFen?

A29. (GLS) No. The methodological approach is the same. However, the refined calculations included additional analyses to determine the CUFs for three of the reactor pressure vessel nozzles, and plant-specific CUF calculations were performed in accordance with ASME Section III for the piping locations in order to develop appropriate plant-specific CUF values for the piping.

15

Q30. What conservatisms are incorporated into the refined CUFen calculations?

A30. (GLS) The primary conservatisms incorporated into the refined VY EAF calculations are as follows:

a. The numbers of transient cycles for 60 years used in the refined calculations is conservative relative to the numbers of transients expected to occur through 60 years of operation.

b. The refined calculations used design basis transient severity definitions, as opposed to the (lesser) actual transient severity.

c. The refined calculations used bounding values for pressure and temperature at EPU conditions for the entire 60-year period of plant operation.

d. The refined calculations calculated bounding Fen multipliers using values for temperature, strain rate and sulfur content that were selected to maximize the Fen multipliers.

Q31. What was Entergy's role in the preparation of these calculations?

A31. (JCF) Entergy developed a Design Information Record which validates and documents all of the plant specific information used to perform the EAF calculations. Also, Entergy reviewed and approved the refined calculations performed by SIA.

Q32. Are the results of these refined analyses summarized in a report?

A32. (GLS) Yes. SIA Report No. SIR-07-132-NPS, "Summary Report of Plant-Specific Environmental Fatigue Analyses for the Vermont Yankee Nuclear Power Station" (Revision 1, dated December 2007) (Exhibit E2-24 hereto), provides summaries of the methodology used in the refined analyses and their results.

Q33. What do the results show in terms of CUFens?

A33. (GLS) The results of the analyses, as summarized in Table 3-10 of Report No. SIR-07-132-NPS, demonstrate that the 16

environmentally adjusted fatigue usage factors for all locations and components analyzed remain within the allowable value of 1.0 through 60 years of VY operation.

IV. NRC STAFF REVIEW OF REFINED ANALYSES Q34. Did Entergy advise the NRC Staff of its intention to implement Option 1 of the three specified options in Section 4.3.3 of the Application?

A34. (JCF) Yes. Entergy advised the NRC Staff of its intent to implement Option 1 (performance of refined analyses) by letter BVY-07-054 dated July 30, 2007, ADAMS Accession No. ML072140847.

Q35. What was the NRC Staff's response to the approach presented by Entergy?

A35. (JCF) The NRC Staff determined that the overall approach proposed by Entergy was acceptable. The Staff found that the refined EAF calculations proposed by Entergy would be based on the recommendations for performing such calculations in NUREG/CR-6583 for carbon and low alloy steel components, and in NUREG/CR-5704 for stainless steel components, and that the methods for determination of stresses and CUFs would be in accordance with an NRC-endorsed edition of the ASME Code Section III, Division 1 Subsection NB, Subarticles NB-3200 or NB-3600, as applicable to the specific component. The Staff accordingly concluded that performance of the refined calculations would be in conformance with the recommendations in both the Standard Review Plan for License Renewal and in GALL Report Section X.M1. These conclusions are documented in the Safety Evaluation Report ("SER") issued by the NRC Staff in February 2008 (SER Section 4.3.3.2 at 4-32 through 4-38).

Q36. Did Entergy provide the refined analysis results to the NRC Staff?

17

A36. (JCF) Yes. Entergy provided the NRC Staff with a summary of the refined analysis results in letter BVY-07-066 dated September 17, 2007, ADAMS Accession No. ML072670135.

Q37. Did the NRC Staff review the refined analyses?

A37. (JCF) Yes. The NRC Staff reviewed the refined analyses during an audit on October 9-10, 2007. During the audit, the Staff asked Entergy to explain how the stress intensity for thermal transients (including shear stresses) was calculated for the analyzed components and locations. Entergy explained that, in most cases, shear stresses are negligible for thermal transients for cylindrical components like those used in reactor pressure vessels and piping.

The Staff, however, took the position that shear stresses cannot always be neglected in the calculation of stress intensities used to determine CUFs of all locations, and that while it is appropriate to do so for locations where non-symmetric loadings are not significant, neglecting shear stresses for locations with significant geometric discontinuity, such as nozzle comers, could lead to non-conservative results. These Staff concerns were applicable to locations at the blend radius (nozzle comer) regions of three reactor pressure vessel components evaluated in the refined calculations: the feedwater nozzle, the recirculation outlet nozzle, and the core spray nozzle. It is important to note that the Staff had no concerns with the stress calculations for all other six piping and component locations. This sequence of events is discussed in the VY SER (SER Section 4.3.3.2 at 4-38 through 4-40).

Q38. How was this technical difference of opinion resolved?

A38. (GLS) At a public meeting between Entergy representatives and the NRC Staff on January 8, 2008, Entergy suggested performing a confirmatory CUFen analysis of the feedwater nozzle comer 18

using methods that would be acceptable to the NRC. The feedwater nozzle was selected for analysis because (1) it is the limiting nozzle (i.e., has the highest CUFen) among the three nozzles regarding which the Staff had questions; (2) it is subjected to more transients and cycles than the other two nozzles; and (3) the transients it experiences are more severe than the transients experienced by the other two nozzles. The Staff agreed to this suggestion and to the choice of the feedwater nozzle comer as the limiting location of interest (SER Section 4.3.3.2 at 4 4-41).

Q39. How was the confirmatory analysis performed?

A39. (GLS) The confirmatory analysis used the same finite element model, thermal transient definitions, numbers of transient cycles, and water chemistry inputs as the refined analyses performed by SIA for Entergy. The confirmatory analysis differs from the refined analysis in three main respects:

(1) While both methods use a detailed finite element model of the feedwater nozzle, when the thermal transient stress histories were determined, the confirmatory analysis computed six component stress histories for each transient using the ANSYS finite element computer code, whereas the refined analysis used a Green's Function approach based on a simplified single stress component difference, computed from the results of ANSYS, to obtain the stress time history for all of the transients. The use of Green's Functions, which is a well-documented mathematical technique, allows for significantly less computational effort compared to evaluating each of the same transients individually via the finite element software.

(2) In the confirmatory analysis, each of the thermal transients produces three orthogonal components and three shear stress 19

components, as is typical from structural mechanics theory. These six components are combined to obtain a maximum stress intensity history for all evaluated transients. Shear stresses are included in the computation. In the refined analysis, on the other hand, only the maximum stress difference, which is essentially equal to the stress intensity computed from the finite element program, is used. In using the maximum single stress component difference in the Green's Function approach, the analyst assesses the magnitude of individual shear stresses and determines whether the resulting stress intensity history closely approximates the stress intensity history that would result from a stress intensity evaluation and combination process that uses all six stress components.

(3) In the confirmatory calculation, a maximum Fen is computed for each incremental stress load-pair, which constitute the paired transient stress state points into which the applied loading history is separated for calculating the CUF. Each Fen value is based on the maximum transient temperature unique to each load pair, and the contributions of all load pairs are added to produce a composite CUFen. In the refined analysis, on the other hand, a single, maximum Fen is applied to the total CUF resulting from all load pairs, and is based on the maximum transient temperature for all load pairs. This Fen selection technique is more conservative than the approach used in the confirmatory calculation, as it results in a single bounding Fn value.

Q40. What were the results of the confirmatory analysis?

A40. (GLS) The methodology and results of the confirmatory analysis are presented in three calculations (Exhibits E2-25 through E2-27). The feedwater nozzle EAF evaluation was performed for the two controlling locations on the nozzle, the inside surface of the 20

nozzle blend radius (nozzle comer), and at the inside surface of the nozzle safe end. The confirmatory calculations for the nozzle comer yielded a CUF (before application of Fen factors) of 0.089 at the nozzle comer, versus a CUF of 0.064 at the same location using the refined analysis methodology (Exhibit E2-27, Section 4.0 and Exhibit E2-24, Table 3-10). The corresponding results for the safe end location showed a reduction in the computed CUF when using the confirmatory analysis methodology.

As stated earlier, the confirmatory calculation used specific Fens for each transient load pair in the CUF calculation, applied them to the corresponding CUFs for each load pair, and added the results for all load pairs, yielding a composite CUFen of 0.353 for the nozzle comer, again significantly below the acceptable limit.

The refined calculation, on the other hand, applied a single, bounding Fen factor of 10.05, yielding an environmentally adjusted CUF of 0.639 for the nozzle comer, also significantly below the acceptable limit of 1.0. See id.

Q41. Did the NRC Staff review the confirmatory analysis?

A41. (GLS) Yes. The results of the confirmatory analyses were submitted for review by the NRC Staff via Amendment 34 to the Application on January 30, 2008 (Exhibit E2-28). Also, the NRC Staff performed an audit of the calculations on February 14, 2008.

After review, the Staff found that Entergy correctly applied the ANSYS finite element software; used appropriate input parameters; added the stresses correctly; and applied proper Fen factors for each transient, so that the results of the confirmatory analysis were appropriate and acceptable. The Staff asked what the CUFen for the nozzle comer would be for the confirmatory analysis if a single, bounding Fen factor were used as was done in 21

the refined analysis. Applying the previously-determined single bounding Fe,, value (10.05), the confirmatory analysis would yield a CUFCn of 0.893 for the nozzle cormer, higher than that obtained in the refined analysis, but still below the acceptable limit of 1.0.

The Staff concluded from this result that utilization of the analytical tools and simplifications used in the refined analyses could underestimate the CUFe, for the feedwater nozzle comer.

Accordingly, the Staff requested that the confirmatory analysis be the analysis of record for the feedwater nozzle. The Staff also imposed a license condition requiring similar confirmatory analyses for two other nozzles, the recirculation outlet nozzle and the core spray nozzle. Those confirmatory analyses will become the "analyses of record" for those two locations (SER Section 4.3.3.2 at 4 - 41 through 4 - 43).

Q42. When are these two additional confirmatory analyses to be performed?

A42. (JCF) No later than two years prior to the start of the period of extended operation, in March 2012.

Q43. Why is it acceptable not to perform the two additional confirmatory analyses prior to approving the renewal of the VY license?

A43. (JCF) It is acceptable for two reasons. First, GSI 190 was closed out by the NRC Staff in December 1999 without the imposition of additional requirements for a plant's initial license term (40-year)

(see Exhibit E2-03). The NRC Staff concluded that licensees are to address the effects of coolant environment on component fatigue life only as part of the aging management programs formulated in support of license renewal. The methods are

• defined, and the required performance of the calculations two years prior to the license renewal period ensures that the environmental effects are addressed prior to entering the period of extended operation. Second, because the CUFens at those two 22

nozzle comer locations are very small (0.084 for the recirculation outlet nozzle and 0.167 for the core spray nozzle), it is extremely improbable that the results of the confirmatory analyses of these nozzles would yield CUFns greater than unity. This is particularly true given that the confirmatory analysis of the feedwater nozzle comer which, as discussed earlier, is the limiting case among the three nozzles, yielded acceptable results, and the results for the other two nozzles are expected to fall well below the feedwater nozzle results. (Exhibit E2-09, Amendment 35 to Application, Attachment 1).

Q44. What is the significance of the different approaches taken by Entergy for estimating environmentally adjusted CUFs?

A44. (GLS) There is no practical difference between the two approaches because they both yield conservatively calculated CUFens for all nine limiting piping and vessel locations that are well within the acceptable limit. Thus, regardless of what method one chooses to apply, the conclusion is the same - the critical reactor components will not experience failure due to fatigue during the period of extended operation.

V. RESPONSE TO ISSUES RAISED IN THE NEC TESTIMONY ON CONTENTION 2A!2B Q45. Have you had the opportunity to review the testimony and exhibits submitted by NEC in this proceeding on April 28, 2008 relating to NEC Contention 2A/2B?

A45. (JCF, GLS) Yes, we have.

Q46. What testimony did you review?

A46. (JCF, GLS) We reviewed the direct testimony of Joram Hopenfeld, NEC Exhibit NEC-JH_01; Dr. Hopenfeld's report entitled "Review of Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. ("Entergy") Analyses of the 23

Effects of Reactor Water Environment on Fatigue Life of Risk-significant Components During the Period of Extended Operation" ("NEC Fatigue Report"), NEC Exhibit NEC-JH_03; and NEC Exhibits NEC-JH_ 04 through NEC-JH_35 and NEC-JH_62.

Q47. What are the main technical issues raised in NEC's testimony and exhibits on Contention 2A/2B?

A47. (JCF, GLS) NEC's consultant Dr. Joram Hopenfeld asserts that Entergy has not supplied to NEC information necessary to establish the validity of Entergy's CUFen reanalyses, i.e.:

"adequate layout drawings of the plant piping" and "a complete description of the methods or models used to determine velocities and temperatures during transients." See NEC Fatigue Report at

8. He also finds fault with Entergy's refined fatigue analysis, with the criticisms generally falling into two broad categories: a critique of Entergy's calculations of the Fen factors, and a critique of the calculations of 60-year CUFs. NEC Fatigue Report at 10-

17. With respect to the confirmatory analysis, Dr. Hopenfeld asserts that, even though that analysis addresses the inaccuracies in CUF values used in Entergy's refined analysis resulting from the use of Green's Functions, the confirmatory analysis does not address the other "errors" in the refined analysis, id. at 18, and that none of the analyses includes an "error analysis to show the admissible range for each variable." Id. He also asserts that the confirmatory analysis of the feedwater nozzles does not bound the analysis for other components. Id. at 18-19.

A. Failure to Supply Pertinent Information Q48. How do you respond to NEC's complaint that information necessary to establish the validity of Entergy's CUFen refined and confirmatory analyses was not provided?

24

A48. (JCF) Early in the discovery process in this proceeding, between December 2006 and March 2007, Entergy supplied to NEC 36 drawings, which showed nozzles and connecting headers for the components in question. A listing of the drawings provided is included in Exhibit E2-29. In addition, Entergy supplied to NEC a copy of the Design Information Record ("DIR") that lists all the drawings and other inputs used in the refined calculations. If NEC or its consultants had required additional drawings, they could have identified them through the DIR and requested them in discovery, as they requested other materials. No such a request was ever made.

With respect to the alleged lack of"a complete description of the methods or models used to determine velocities and temperatures during transients," NEC asked for such a description through counsel and I provided the following information, which was conveyed to NEC on April 14, 2008: "that information is provided in the calculations that constitute the fatigue analyses, and is documented in each of the calculations. For example, temperatures are given in Figures 4 through 20 of Calculation VY-16Q-302-RO. In addition, flow rates are also presented in Section 3.2 (Thermal Loads) of calculation VY-16Q-301-RO.

Given the geometries described in the calculations, an experienced analyst can calculate velocities from flow rates using the continuity equation." Again, if additional specifics on the calculations were needed, NEC could have asked for them.

B. Calculation of Environmental Correction Factors Q49. What criticisms does NEC raise with respect to Entergy's calculation of the Fen environmental fatigue correction factors?

25

A49. (GLS) NEC alleges in the NEC Fatigue Report the following "errors" in Entergy's Fen calculations: (a) that Entergy used "outdated" statistical equations from NUREG/CR-6853 and NUREG/CR-5704 to calculate the Fen parameters, and should have used instead the results in NUREG/CR-6909, ANL-06/08, "Effect of LWR Coolant Environments on the Fatigue Life of Reactor Materials," February 2007 ("NUREG/CR-6909," Exhibit E2-30) (NEC Fatigue Report at 10-12) and applied a factor of 17 to correct the CUFs for environmental effects; (b) that Entergy used inappropriate heat transfer equations to calculate the thermal stress for each transient (d. at 12-15); (c) that Entergy has not provided proof that the base metal of the feedwater nozzles is not cracked (id. at 15-16); (c) that the number of plant transients estimated to occur during the operating life of VY is not sufficiently conservative (id. at 16); and (d) that Entergy's calculation of the Fen parameters does not appropriately account for oxygen concentrations and resulting changes in water chemistry (id. at 16-17). In addition, NEC criticizes the refined analyses performed by Entergy because they used a simplified Green's Function methodology, which allegedly resulted in "the underestimation of CUF values by approximately 40%." Id. at 17-18.

Q50. Is NEC correct in claiming that Entergy should have used the NUREG/CR-6909 guidance?

A50. (GLS) No. The NRC has not accepted the use of NUREG/CR-6909 in license renewal fatigue analyses.

The standards to be used for license renewal purposes are identified in Section X.M1 of the GALL Report, where it is clearly indicated that NUREG/CR-6583 and NUREG/CR-5704 contain acceptable methodology for computing Fen multipliers.

26

Moreover, an NRC Staff expert has explained at an Advisory Committee on Reactor Safeguards ("ACRS") meeting on the VY license renewal that, whereas the methodology in NUREG/CR-6909 is more recent, it is intended for use for new plant construction, and that the methods given in NUREG/CR-6583 and NUREG/CR-5704 generally yield higher Fen multipliers for currently operating plants compared to those that result from the NUREG/CR-6909 methodology (Exhibit E2-3 1).

Indeed, for carbon and low alloy steels, NUREG/CR-6909 acknowledges (Exhibit E2-30 at 38) that: "Relative to the earlier expressions, correction factors determined... for carbon steels are 48% higher, and ... for low-alloy steels are z18% lower...

NUREG/CR-6909 further states: "...The discussions.. .indicate that the current Code requirement of a factor of 20 on cycles... is conservative by at least a factor of 1.7. Thus, to reduce this conservatism, fatigue design curves based on the ANL model for carbon and low-alloy steels have been developed using factors of 12 on life and 2 on stress..." (Exhibit 30 at 19). In addition, for stainless steel, NUREG/CR-6909 states: "Relative to the earlier expressions, correction factors determined... are 45-60%

lower." Id. at 63. Therefore, the use of NUREG/CR-6583 and NUREG/CR-5704 is appropriately conservative and there is no reason for license renewal applicants to use the methodology contained in NUREG/CR-6909. Finally, the assertion (NEC Fatigue Report at 12) that a correction factor of 17 should be applied to the CUF values to correct for environmental effects is applicable only to "certain environmental and loading conditions..." (NUREG/CR-6909 at 3) and is inappropriate for the VY environment and loading conditions. The conditions that would cause a multiplier of 17 to exist are primarily associated 27

with high temperature and high dissolved oxygen content for carbon and low alloy steels. Those conditions do not exist at VY.

Q51. Is there any validity to the claim in the NEC Fatigue Report (at 11) that, in calculating the Fen parameters, Entergy failed to consider many factors deemed important in NUREG/CR-6909, including mean stress, surface finish, size and geometry, and loading history?

AS1. (GLS) No. All parameters that are significant to the determination of Fen were considered in the VY EAF analyses. In particular, NUREG/CR-6583 determines the Fen values for carbon and low alloy steel as a function of sulfur content, temperature, dissolved oxygen, and strain rate (NUREG/CR-6583, Exhibit E2-06 at 69); and NUREG/CR-5704 determines the Fen for stainless steel as a function of temperature, dissolved oxygen, and strain rate (NUREG/CR-5704, Exhibit E2-07, at Section 5). Indeed, most of the additional factors cited by Dr. Hopenfeld are taken into account in the ASME fatigue curves, which reduce the allowable cycles by a factor of 20 or the stress by a factor of 2, whichever is more limiting, to account for the effects of material variability and data scatter, as well as size effects, surface finish, and atmosphere, and adjusted to account for maximum mean stress effects (e. Exhibit E2-32 at 3-5). Loading history is very conservatively accounted for in the CUF calculations by using the ASME Code Section III range-pair cycle counting methodology, which assumes all transients occur in the most severe order possible.

Q52. The NEC Fatigue Report (at 11) also claims that Entergy did not provide any data on the surface roughness of the components it evaluated. Is this a legitimate concern?

A52. (JCF) No. Surface roughness (finish) effects are incorporated into the Entergy fatigue evaluations via use of the ASME Code design fatigue curves in the CUF calculations. The ASME Code design 28

fatigue curves include a factor of safety of 20 of allowable cycles, which includes a factor of 4.0 to account for "surface finish, atmosphere, etc." (Exhibit E2-32 at 3).

Q53. NEC also raises the possibility that surface cracks may exist at the blend radius and the base metal of the feedwater nozzle (NEC Fatigue Report at 11-12, 15-16).

Do such cracks exist?

A53. (JCF). No. VY periodically inspects the feedwater nozzle for potential cracks in the base metal and has not identified any since the current thermal sleeves were installed. The most recent inspection was conducted during the 2007 refueling outage and showed no evidence of cracks in the base metal of the nozzle. See Exhibit E2-33, which includes summaries of the 2007 inspection results and a picture showing where the inspections were made.

Q54. NEC challenges the validity of the three heat transfer equations used in Entergy's analysis to calculate the thermal stress for each transient. What is your response to NEC's challenges?

A54. (GLS) (1) One of the concerns expressed by NEC is that, if the pipe run is short, the assumption of fully developed turbulent flow in Equation (1) would not apply. Each of the feedwater nozzles has at least 48 inches of horizontal pipe on the upstream end.

This length is more than sufficient for fully developed turbulent flow to occur through the nozzle region.

(2) Another criticism by NEC is that no correction is made in Equation (1) for the ratio of viscosities evaluated at the bulk and wall temperatures. However, such a correction would only be necessary if there were wide temperature differences in the flow.

Due to the high flow velocity for most of the analyzed conditions, as well as the presence of a vertical run of pipe prior to the horizontal inlet pipe to the nozzle, sufficient mixing is present to minimize flow temperature differences. In addition, Equation (1) is valid for moderate temperature differences between wall and 29

fluid conditions, which would account for any minor variations in flow temperature differences.

(3) NEC claims that the connecting pipe "is probably at some angle to the nozzle." However, this claim is groundless. The connecting pipes are horizontal and normal to the nozzle centerline. This configuration was confirmed during Entergy walkdowns in response to NRC IE Bulletin 79-14, "Seismic Analysis for As-Built Safety Related Piping Systems." Having the piping normal to the nozzle is necessary as a practical matter in order to obtain sufficient fit-up for welding the pipe to the nozzle during construction.

(4) NEC states that use of Equation (2) is inappropriate because it does not adequately represent the physical considerations of flow that changes from forced to natural convection, nor does it describe the variation in heat transfer coefficient along the pipe.

The VY analyses, however, assumed that an abrupt change occurs between forced and natural convection flow states, which conservatively introduces abrupt changes in heat transfer compared to a mixed forced/free convection region. In addition, the heat transfer coefficient determined using Equation (2) is dependent upon the diameter of the flow path, thereby allowing this equation to properly describe the variation in heat transfer along the nozzle. For the VY analyses, the bounding nozzle diameter in each region of the nozzle was used in Equation (2) so as to produce the most conservative heat transfer coefficients for all locations along the length of the nozzle.

(5) NEC alleges that Equation (3) is only applicable to laminar flow and it defines an average heat transfer coefficient instead of a local heat transfer coefficient. However, Equation (3) is applicable to steam condensation under a low vapor Reynolds number, which does not necessarily imply laminar conditions. A 30

low Reynolds number (which generally equates to a low velocity) is appropriate for the single transient evaluated for VY where steam condensation conditions exist at the nozzle location due to the low water level specified during the transient. In addition, the heat transfer coefficient determined using Equation (3) is dependent upon the diameter of the flow path, which allows this equation to properly describe the local variation in heat transfer along the nozzle. For the VY analyses, the bounding diameter was used in Equation (3) so as to produce the most conservative heat transfer coefficient for all locations along the length of the nozzle.

Q55. The NEC Fatigue Report (at 16) claims that Entergy made the incorrect assumption that the number of transients that the plant will experience throughout its life varies linearly with time, whereas the number of transients should "at a minimum be multiplied by a factor of 1.2 to account for the probability of an increase in unanticipated failures due to the 20% power uprate." Are NEC's assertions correct?

A55. (JCF) No. Entergy did not assume that the number of transients that the plant will experience varies linearly with time. The transients used in the EAF analyses are a combination of the original VY design transients and additional, more detailed design conditions from a later BWR 4 design specification. (The later BWR plants have more detailed thermal transient definitions based on the operating experience from earlier BWRs.) Then VY projections for 60 years were made based on all available sources, including the numbers of cycles for 40 years in the VY reactor pressure vessel Design Specification, the numbers of cycles actually analyzed in the VY Design Stress Report, and the numbers of cycles experienced by VY after approximately 35 years of operation (July 2007).

31

The numbers of cycles used in the EAF analyses are conservative projections of the numbers of cycles actually experienced by the plant over its operating history. For example, 300 startup -

shutdown cycles were used in the VY EAF analyses. The original 40-year Design Specification specified 200 cycles. As of May 2008, VY had experienced 95 startup - shutdown cycles after 36 years of operation. A linear projection of this value would yield approximately 160 cycles after 60 years of operation, which is

,significantly less than the 300 cycles Entergy used in the VY EAF analyses. Therefore, the margin on the numbers of cycles analyzed significantly exceeds the factor of 1.2 suggested by NEC.

Further, Entergy has committed to monitor the transient count throughout the period of extended operation to verify that these assumptions remain valid, and will take appropriate corrective action if these assumptions do not remain valid. This Fatigue Monitoring Program is described in Section A.2.11 of the Application.

Q56. The NEC Fatigue Report (at 16-17) claims that, in calculating the Fen parameters, Entergy did not properly account for unanticipated oxygen excursions during the extended operations period and did not explain how the chemistry data for the feedwater line or the electrochemical potential measurements relate to the oxygen concentration at the component surface during transients. Are these criticisms correct?

A56. (GLS) No. The Fen expressions documented in NUREG/CR-6583 and NUREG/CR-5704 are not dependent on electrochemical potential.

The expressions are dependent on dissolved oxygen level, which is the parameter used by Entergy in the VY CUFen analyses. The VY CUF calculations accounted for variability in oxygen values, including potential excursions. As discussed in Attachment 2 to 32

Amendment 35 to the Application (Exhibit E2-09), Entergy investigated the variability in dissolved oxygen from plant data, including several water chemistry excursions, and determined that the oxygen values used in the CUF analyses bounded a mean plus one sigma variation on oxygen.

For carbon and low alloy steel, the Fen multipliers are reduced in value for low oxygen levels. All of the VY analyses used the bounding value of oxygen, including at least a one-sigma uncertainty, so as to maximize the Fen multiplier.

For stainless steel, the Fen multipliers are increased for oxygen levels below 50 parts per billion ("ppb"). Such potential increases were considered in the VY EAF analyses by using the BWRVIA results based on 40 ppb oxygen condition in the feedwater line.

The BWRVIA oxygen concentrations throughout the vessel are relatively insensitive to a 10 ppb variation in the feedwater dissolved oxygen.

Q57. NEC further claims (Fatigue Report at 17) that VY uses the EPRI-BWVRIA computer code to assess oxygen concentrations and has not described how that code was benchmarked against plant data, so "one must assume the oxygen concentrations that were used by Entergy to calculate the Fens contain unknown errors." Is this a valid concern?

A57. (JCF, GLS) No. The BWRVIA model was developed to give reasonable comparisons with plant chemistry recirculation and steam oxidant data. This was achieved by adjusting a number of input parameters in the model within their levels of uncertainty.

EPRI benchmarked the model by fitting it to chemistry data for one plant and then comparing the revised model with chemistry data from six other plants. The comparison of the model with steam and recirculation oxidant data was adequate and all chemistry parameters were defensible. See Exhibit E2-34.

33

Q58. The Fatigue Report (at 5-6 and 17-18) claims that use of the Green's Function methodology in the refined analysis results in a 40% underestimation of the CUF values, a result that NEC obtains by comparing the CUFs for the refined analysis and the confirmatory analysis for the feedwater nozzle. Is NEC's analysis correct?

A58. (GLS) No. Although the confirmatory analysis for the feedwater nozzle calculated a CUF for the nozzle comer that was 40%

higher than the values yielded by the refined analysis, that difference was the result of approximately 20 different inputs and assumptions used in the confirmatory analysis compared to those used in the refined analysis. The reason for this difference was not specifically investigated, as it was not the intent of the confirmatory analysis to duplicate the results of the refined analysis. However, the results documented by Entergy in Exhibit E2-35 (Amendment 33 to Application, Attachment 1 at 4) show that the difference in CUF due to changes in the Green's Function is very small. Therefore, use of the Green's Function methodology would not result in a substantial underestimation of the CUF. Also, it is important to note that the confirmatory analysis CUF for the safe end of the same nozzle was 60% lower than that calculated in the refined analysis.

C. Lack of Error Analysis Q59. NEC criticizes Entergy for the failure to perform an error analysis to show the admissible range for each variable included in the analysis (NEC Fatigue Report at 18). Is this a valid criticism?

A59. (GLS) No. Performing an error analysis on the stress results is not traditional practice in analyses of this type, and is unnecessary given that bounding input parameters (such as temperature, pressure, and heat transfer coefficients) were selected so as to maximize stresses. Use of nominal or mean input parameters and establishment of an error band on the final results would lead to reduced stress (and therefore lower CUF) results.

34

D. Confirmatory Calculations Q60. What criticisms does NEC raise regarding the confirmatory calculation of CUFen for the feedwater nozzle?

A60. (GLS) The NEC Fatigue Report (at 18) asserts that the confirmatory calculation for the feedwater nozzle does not account for the various alleged deficiencies in the refined calculation other than those relating to the use of Green's Functions. In addition, Dr. Hopenfeld does not agree that the confirmatory analysis results are bounding on the other two nozzles (the recirculation outlet nozzle and the core spray nozzle).

He asserts that "[t]here are considerable differences in geometry, heat transfer characteristics and loadings between the feedwater and the other two nozzles. These differences could result in different stress distributions which would affect the CUFs." (Id. at 18-19).

Q61. Is there any validity to NEC's claim that the feedwater nozzle CUFen results are not bounding on the other two nozzles?

A61. (JCF) No. The feedwater nozzle is the controlling nozzle because it experiences the most severe design transients and because it is the location where the relatively colder feedwater returns to the hot reactor vessel, thereby causing the most severe thermal stresses. The feedwater nozzle has the highest fatigue usage factors of the three nozzles in question. Repeated industry analytical experience for 40 years by many vendors and consultants has shown that the fatigue usage is typically higher for the feedwater nozzle than it is for any other reactor pressure vessel nozzle. For those reasons, it is extremely improbable that either of the other two nozzles would experience CUFens in excess of that experienced by the feedwater nozzle.

35

I would also note that the experts at the ACRS and the NRC Staff agree that the feedwater nozzle is the bounding component in terms of CUFen value. This is demonstrated by the discussion in the SER (Section 4.3.3.2 at 4 - 40) and the statement by the ACRS Vice Chairman Mario Bonaca at the March 6, 2008 meeting in which the VY fatigue analyses were discussed (Exhibit E2-36 at 91).

E. Dr. Hopenfeld's CUFen Recalculation Q62. The NEC Fatigue Report includes (at 19-20) Dr. Hopenfeld's recalculation of the CUFens for the nine locations of interest at VY. Do you have any comments on Dr. Hopenfeld's calculation?

A62. (GLS) Yes. Dr. Hopenfeld calculates irrelevant CUFs by using generic Fens that do not apply to VY and applying methodology that is not sanctioned by the NRC for existing plants. As a result, the results of his recalculation are meaningless.

Dr. Hopenfeld provides no details as to specifically how he calculated his Fen multipliers. However, Dr. Hopenfeld does reference the CUF values from Table 4.3-3 of the Application and cites NUREG/CR-6909 as his source for computing his Fen multipliers. Even using NUREG/CR-6909 as his source (which is inappropriate for the reasons discussed earlier), Dr. Hopenfeld's approach is unduly conservative in that it uses generic CUF values taken from NLUREG/CR-6260 for B.3 1.1 components that are not VY-specific. Therefore, Dr. Hopenfeld is not using CUFs that reflect the actual plant conditions and transients (which were used when Entergy subsequently calculated the CUFs for these components in accordance with ASME Section III). Further, the equations for calculating Fe, in NUREG/CR-6909 are intended to apply to updated fatigue curves proposed in that document, and those fatigue curves are not the same as the fatigue curves from 36

the ASME Code upon which the CUFs in NUREG/CR-6260 are based. Therefore, no credit should be given to Dr. Hopenfeld's CUFen recalculation.

VI. ACRS ENDORSEMENT OF ENTERGY'S PROGRAM TO MANAGE ENVIRONMENTALLY ASSISTED FATIGUE Q63. Has the ACRS examined the program proposed by Entergy for managing environmentally assisted fatigue at VY during the period of extended operations after license renewal?

A63. (JCF, GLS) Yes. There was extensive discussion by Entergy and the NRC Staff on the environmentally assisted fatigue issue before the ACRS at their meetings on February 8 and March 6, 2008. See Exhibit E2-36. Both of us presented at those meetings.

In its letter to the NRC Chairman following review of the proposed license renewal for VY, the ACRS concluded:

The applicant has chosen to address environmentally assisted fatigue by demonstrating that the cumulative usage factor (CUF) at the most sensitive locations will remain below 1.0 throughout the period of extended operation, considering both mechanical and environmental effects. Analyses were performed by the applicant using assumptions to be monitored and verified during the period of extended operation. These analyses showed that the CUF at all analyzed locations will remain below 1.0 throughout the period of extended operation. However, for those locations with geometric discontinuities or non-symmetric loads such as the feedwater nozzle, the reactor recirculation outlet nozzle, and the core spray line nozzle, the staff challenged the methodology used by the applicant because this methodology neglects shear stresses on the component. At the request of the staff, the applicant performed an additional analysis of the expected limiting location, the feedwater nozzle, using an approved methodology that accounts for all stress components. This analysis confirmed that the CUF will not exceed 1.0 during the period of extended 37

operation. Since this analysis showed that the original methodology could underestimate the CUF, the staff has concluded that additional analyses are needed for the reactor recirculation outlet and the core spray line nozzles. These three analyses will be the analyses-of-record for these components.

Performance of the remaining analyses at least two years before entering the period of extended operation will be a license condition. We agree with the staff's conclusion.

We agree with the staff that there are no issues related to the matters described in 10 CFR 54.29(a)(1) and (a)(2) that preclude renewal of the operating license for VYNPS. The programs established and committed to by ENO provide reasonable assurance that VYNPS can be operated in accordance with its current licensing basis for the period of extended operation without undue risk to the health and safety of the public. The ENO application for renewal of the operating license for VYNPS should be approved.

(Exhibit E2-37, Letter from ACRS Chairman William J. Shack to NRC Chairman Dale E. Klein, dated March 20, 2008 at .3-4).

VII.

SUMMARY

AND CONCLUSIONS Q64. Please summarize your testimony.

A64. (JCF, GLS) Our testimony can be summarized as follows:

Entergy has performed more refined EAF analyses at nine limiting piping and vessel locations for VY. These analyses, performed using conservative methodologies and input parameters, have demonstrated that the CUFens are less than 1.0 for the sixty years of plant operation encompassed by the renewed VY operating license.

The acceptability of the CUFens for the nine limiting piping and vessel locations at VY has been further established by the performance of a confirmatory calculation for a limiting component, the feedwater nozzle, using a methodology endorsed by the NRC Staff. This confirmatory analysis 38

again yielded CUFens well within the required ASME Code limit.

These analyses collectively demonstrate that the critical VY components will not experience failure due to fatigue during the period of extended operation.

Q65. What are your conclusions regarding the assertions in NEC Contentions 2A and 2B?

A65. (JCF, GLS) We conclude that there is no support for the claims made in NEC Contentions 2A and 2B.

Q66. Does thatconclude your testimony?

A66. (JCF, GLS) Yes, it does.

39

ORIGINAL May 9, 2008 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

JOINT DECLARATION OF JOHN R. HOFFMAN AND LARRY D. LUKENS ON NEC CONTENTION 3 - STEAM DRYER John R. Hoffman and Larry D. Lukens state as follows under penalty of perjury:

1. We have prepared the attached "Testimony of John R. Hoffman and Larry D.

Lukens on NEC Contention 3 - Steam Dryer" in the above captioned proceeding.

2. The factual statements and opinions we express in the cited testimony are true and correct to the best of our personal knowledge and belief.

3. We declare under penalty of perjury that the foregoing is true and correct.

Executed on May 9, 2008 Hoffman-JýY--

LaEeueD. Lukens Executed on May 9, 2008

May 9, 2008 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

TESTIMONY OF JOHN R. HOFFMAN AND LARRY D. LUKENS ON NEC CONTENTION 3 - STEAM DRYER I. WITNESS BACKGROUND John R. Hoffman ("JRH")

Q1. Please state your full name.

Al. (JRH) My name is John R. Hoffman.

Q2. By whom are you employed and what is your position?

A2. (JRH) I am an independent consultant. Prior to September 2006, I was employed by Entergy Nuclear Operations, Inc. ("ENO") and had, among other responsibilities, that of Project Manager for the License Renewal Project at the Vermont Yankee Nuclear Power Station ("VY"). I retired from ENO's employment in September 2006.

Q3. Please summarize your educational and professional qualifications.

A3. (JRH) My professional and educational experience is described in the curriculum vitae attached to this testimony as Exhibit E3-02.

Briefly summarized, I have over 37 years of nuclear power engineering experience. I received a B.E. Degree in Mechanical Engineering from the Cooper Union for the Advancement of Science and Art in 1967, an M.S. Degree in Nuclear Engineering from the University of Lowell in 1977, and an M.S. Degree in Applied Management from Lesley College in 1985. I am a Registered Professional Engineer in the States of Massachusetts and Vermont. I have been associated with VY since 1971.

Q4. What is the purpose of your testimony?

A4. (JRH) The purpose of my testimony is to address, on behalf of Entergy Nuclear Vermont Yankee, LLC and ENO (collectively "Entergy"), Contention 3 submitted by the New England Coalition

("NEC") in this proceeding. As admitted by the Atomic Safety and Licensing Board ("Board"), NEC Contention 3 reads:

Entergy's License Renewal Application does not include an adequate plan to monitor and manage aging of the steam dryer during the period of extended operation.

Memorandum and Order (Ruling on Standing, Contentions, Hearing Procedures, State Statutory Claim, and Contention Adoption), LBP-06-20, 64 N.R.C. 131, 187 (2006). The scope of NEC Contention 3 has been subsequently narrowed by the Board, which has ruled that two issues remain to be adjudicated with respect to the contention:

1. Whether Entergy has established sound evaluation and implementation procedures to assure that the integrity of the steam dryer is not jeopardized. Specifically, NEC.

contends that the status of the dryer cracks must be continuously monitored and assessed by a competent engineer. While Entergy has established that it will continuously monitor plant parameters indicative of steam dryer cracking, it has not 2

provided information on its assessment program for the monitoring data or the qualifications of the personnel evaluating this information.

2. Whether a steam dryer aging management program that does not provide a means to estimate and predict stress loads on the dryer during operation for comparison to established fatigue limits is valid.

Memorandum and Order (Ruling on Motion for Summary Disposition of NEC Contention 3) (September 11, 2007), slip op.

at 11-12.

Q5. What has been your role at VY as it relates to NEC Contention 3?

A5. (JRH) In the License Renewal Project at VY, my team was responsible for the development of the proposed program to manage the aging of the VY steam dryer during the renewed license operating period. As Project Manager, I had the responsibility to ensure that all aspects of the license renewal application, including the steam dryer aging management program, were properly developed and were reviewed by the respective subject matter experts at VY.

LARRY D. LUKENS ("LDL")

Q6. Please state your full name.

A6. (LDL) My name is Larry D. Lukens.

Q7. By whom are you employed and what is your position?

A7. (LDL) I am an independent consultant. Prior to July 2007, I was employed by ENO and had, among other responsibilities, that of Supervisor, Code Programs at VY. In that position, my responsibilities entailed ensuring that the activities required by industry codes, particularly those issued by the American Society 3

of Mechanical Engineers ("ASME"), that are applicable to VY and are the responsibility of Engineering are completed, evaluated, dispositioned, and documented. The required activities included, for example, those described by the ASME Operation and Maintenance Code for testing pumps and valves; the ASME Boiler & Pressure Vessel ("BPV") Code for inservice inspection

("ISI"), including containment inservice inspections; the primary containment integrity monitoring program described by 10 C.F.R.50, Appendix J; and the reactor vessel and internals management and monitoring program under the Electric Power Research Institute ("EPRI") BWR Vessel & Internals Program (BWRVIP), an industry initiative implemented with the concurrence and participation of the NRC. I was directly involved with the license renewal audits and inspections of Code Programs activities including the inservice testing ("IST"), ISI, Containment ISI, Appendix J, and BWRVIP, and with the Fire Protection programs, and I approved the VY license renewal commitments relating to these programs. I retired from ENO's employment in July 2007.

Q8. Please summarize your educational and professional qualifications.

A8. (LDL) My professional and educational experience is summarized in the curriculum vitae attached to this testimony as Exhibit E3-03. Briefly summarized, I received a B.S. Degree in Nuclear Engineering from the University of Wisconsin, Madison, in 1978. I have over 38 years of nuclear power work experience.

My relevant experience includes being a qualified reactor operator in the U.S. Navy and an NRC licensed operator at the University of Wisconsin, and nearly 10 years of service as Program Manager for ASME Section XI inservice testing, inservice pressure testing, and 4

containment leak rate testing at an operating nuclear power plant. I have been associated with VY since 2002.

Q9. What is the purpose of your testimony?

A9. (LDL) The purpose of my testimony is to address those aspects of NEC Contention 3 that relate to the steam dryer inspections that have been conducted to date and those to be performed during the period of plant operations after renewal of the VY license.

Q10. What has been your role at VY as it relates to NEC Contention 3?

A10. (LDL) As Supervisor of Code Programs, I was responsible for ensuring the proper completion and evaluation of the steam dryer inspections conducted during the 2005 and 2007 refueling outages.

I was also responsible for overseeing the license renewal aging management program as it applied to the steam dryer.

II. OVERVIEW A. Background Qll. Would you please describe briefly the VY steam dryer?

All. (JRH) In a boiling water reactor ("BWR"), the steam dryer is a stainless steel component whose function is to remove moisture from the steam before it leaves the reactor. The dryer is installed in the reactor vessel above the steam separator assembly and is supported by brackets welded to the inside of the vessel wall below the steam outlet nozzles.

During plant operations, wet steam flows upward and outward through the dryer. Moisture is removed by impinging on the dryer vanes and flows down through drains to the reactor water in the downcomer annulus below the steam separators. The VY steam dryer is a non-safety-related, non-Seismic Category I component.

Although the steam dryer is not a safety-related component, the 5

assembly is designed to withstand design basis events without the generation of loose parts and the dryer is designed to maintain its structural integrity through all plant operating conditions.

Q12. Have concerns about steam dryer performance arisen in the nuclear industry?

A12. (JRH) Yes. In 2002, steam dryer cracking and damage to components and supports for the main steam and feedwater lines were observed at the Quad Cities Unit 2 nuclear power plant. It was determined that loose parts shed by the dryer due to flow-induced vibration that caused metal fatigue failure of the dryer had damaged the supports. The Quad Cities experience raised concern in the industry about the need to assure the physical integrity of steam dryers.

Q13. Did those concerns relate to the performance of the dryer or to the potential effects of a steam dryer failure?

A13. (JRH) The latter. The steam dryer does not perform any safety functions and is not required to prevent or mitigate the consequences of accidents. Therefore, the condition of the steam dryer is only of concern to the extent that a failure of the dryer could have adverse impact on safety-related equipment.

Q14. What do you mean by failure?

A14. (JRH) A loss of physical integrity of the dryer such that loose dryer sections or parts are released to the reactor steam space (that is, the space in the reactor where steam is confined above the water) and potentially migrate to other components.

Q15. Does the formation of cracks in the dryer's surface constitute a failure of the dryer?

A15. (JRH) No. However, the existence of those cracks needs to be identified and evaluated before the cracks progress to the point 6

where they could cause a loss of physical integrity of the dryer, resulting in loose parts.

Q16. Were actions taken at VY in response to the Quad Cities 2 event?

A16. (JRH) Yes. Quad Cities Unit 2 had implemented a power uprate analogous to the extended power uprate ("EPU") that Entergy was planning to implement at VY. Accordingly, Entergy substantially modified the steam dryer at VY during the Spring 2004 refueling outage to improve its capability to withstand the higher flow induced vibration loadings that could result from operation of the plant at EPU levels. The modifications, intended to increase the structural strength of the dryer, are described in Attachment 2 to Supplement 8 (dated July 2, 2004) to the EPU Application, ADAMS Accession No. ML042090103, Exhibit E3-04 hereto. In addition, VY instituted a program of dryer monitoring and inspections to provide assurance that the flow-induced loadings under normal operation at EPU levels did not result in the formation or propagation of cracks on the dryer. The program is described in Attachment 6 to Supplement 33 (dated September 14, 2005) to the EPU Application, ADAMS Accession No. ML052650122, Exhibit E3-05. The program was reviewed and approved by the NRC and included as a license condition as part of the power uprate license amendment issued on March 2, 2006.

Q17. Has a dryer monitoring and inspection program been implemented at VY since.

the plant uprate was accomplished?

A17. (JRH) Yes. As power was increased from the original licensed power level to full EPU conditions, there was continuous monitoring of plant parameters indicative of dryer performance.

The program included measurement at least once per week of moisture carryover and periodic measurement of main steam line pressure. Following completion of EPU power ascension testing, 7

moisture carryover measurements continue to be made periodically, and other plant operational parameters that would be symptomatic of loss of steam dryer structural integrity (main steam line flow, reactor vessel water level, steam dome pressure) continue to be monitored and their values trended. This monitoring program will continue to be implemented during the period of extended operation after renewal of the VY license.

In addition, the steam dryer was inspected during plant refueling outages in the Fall of 2005 (before completion of the EPU) and Spring of 2007 (after one year of operation at EPU power levels).

The dryer is scheduled to be inspected again during the refueling outages in the Fall of 2008 and the Spring of 2010, with a partial inspection scheduled for the Fall of 2011. Inspections will continue in the license renewal period starting with the first refueling after March 2012. The inspections are conducted in accordance with the recommendations of General Electric's Service Information Letter ("SIL") No. 644, Revision 2 (August 30, 2006), ("GE-SIL-644"), Exhibit E3-09 hereto, which is an updated version of the GE-SIL-644 Revision 1 document referenced in the License Renewal Application (Exhibit E3-06). Again, this inspection program will continue during the period of extended operation.

B. Issues Raised By Contention Q18. What is your understanding of the technical issues raised by NEC Contention 3?

A18. (JRH) NEC Contention 3 raises two issues regarding the adequacy of the program, in VY's License Renewal Application, to monitor and manage aging of the steam dryer during the period of extended operation. Those issues are: (1) Whether Entergy has established sound inspection and evaluation procedures to assure that the physical integrity of the steam dryer is not jeopardized; and (2) 8

Whether a steam dryer aging management program that does not provide a means to estimate and predict stress loads on the dryer during operation for comparison to established fatigue limits is valid.

Q19. Do you agree with the assertion in NEC Contention 3 that the License Renewal Application is deficient in these two respects?

A19. (JRH, LDL) No.

Q20. Will you please summarize the basis for your disagreement?

A20. (JRH, LDL) NEC's Contention 3 claims that the status of the dryer cracks must be continuously monitored and assessed by a competent engineer and suggests that Entergy will not carry out these activities during the license renewal period. However, Entergy's proposed steam dryer aging management program includes continuous monitoring of the previously mentioned plant parameters for which a departure from normal range of values would be symptomatic of significant steam dryer cracking, assessment of the monitoring data, and evaluation of the significance of changes in the data by several levels of qualified personnel. The on-line monitoring is augmented by the refueling outage inspections discussed above. Thus, there is no deficiency in the proposed program to monitor and manage dryer performance during the license renewal period.

With respect to the alleged need to provide a means to estimate and predict stress loads on the dryer during operation for comparison to established fatigue limits, no such a need exists because the monitoring program, supplemented by the periodic dryer inspections during refueling outages, is sufficient to diagnose whether significant dryer cracking has occurred before such cracking results in dryer failure. The same approach is used in 9

other aspects of fatigue monitoring in the reactor system components. Those fatigue monitoring programs do not require the estimation of actual loads on the components.

Thus, the inspection and monitoring program for the steam dryer is consistent with the methodology and rigor applicable to safety-related components in other inspection programs. Real time estimation of stresses imparted on the dryer during plant operation is unnecessary to assure its integrity.

III. DESCRIPTION OF VY'S STEAM DRYER AGING MANAGEMENT PROGRAM A. Programmatic Basis for the Proposed Aging Management Program for the VY Steam Dryer Q21. Please describe the aging management program for the steam dryer included in the VY License Renewal Application.

A21. (JRH) In its License Renewal Application, Entergy addresses aging management of the VY steam dryer as follows:

Cracking due to flow-induced vibration in the stainless steel steam dryers is managed by the BWR Vessel Internals Program. The BWR Vessel Internals Program currently incorporates the guidance of GE-SIL-644, Revision 1. VYNPS will evaluate BWRVIP-139 once it is approved by the staff and either include its recommendations in the VYNPS BWR Vessel Internals Program or inform the staff of VYNPS's exceptions to that document.

License Renewal Application, § 3.1.2.2.11 "Cracking due to Flow-Induced Vibration." GE-SIL-644 recommends that BWR licensees institute a program for the long term monitoring and inspection of their steam dryers. It provides detailed inspection and monitoring guidelines (see GE-SIL-644 Rev. 1, Exhibit E3-06 hereto, Appendices C and D).

10

Q22. Does the proposed monitoring and inspection program for the VY steam dryer conform to the recommendations in SIL-644?

A22. (JRH, LDL) Yes.

Q23. Does the proposed monitoring and inspection program for the VY steam dryer conform to the recommendations in the GALL Report (NUREG-1801)?

A23. (JRH, LDL) Yes. The GALL Report calls for a plant-specific aging management program to be evaluated. See Exhibit E3-

08. In its Safety Evaluation Report, the NRC Staff concluded that Entergy's commitment to implement BWRVIP- 139, if approved by the NRC Staff prior to the period of extended operation, will result in the aging of the steam dryer to be adequately managed, as recommended by the GALL Report.

SER at 3-175.

B. Dryer Monitoring Program Q24. Please describe the proposed steam dryer monitoring program to be implemented during the period of extended plant operation following renewal of the VY license.

A24. (JRH) The status of the steam dryer is assessed continuously by the plant operators and VY's technical staff through the monitoring of certain plant parameters. VY Off-Normal Procedure ON-3178 (Exhibit E3-07 hereto) alerts the operators that any of the following events could be indicative of significant dryer damage:

(a) sudden drop in main steam line flow >5%; (b) >3 inch difference in reactor vessel water level instruments; and (c) sudden drop in steam dome pressure >2 psig. In addition, periodic measurements of moisture carryover are performed, and changes in moisture carryover are evaluated in accordance with the requirements of GE-SIL-644 to determine whether significant cracking has occurred. The monitoring program now in place will continue for the entire license renewal period.

11

In addition to implementing its docketed commitment to GE-SIL-644, Revision 1, VY has adopted the additional recommendations in GE-SIL-644 Revision 2 (Exhibit E3-09 hereto) for its BWRVIP Program. The purpose of Revision 2 to GE-SIL-644 is to update the monitoring guidance in Appendix D. These updates reinforce the need for continuous monitoring of plant parameters.

Q25. How would this monitoring detect cracking?

A25. (JRH) Abnormal values of the monitored plant parameters would indicate that the steam leaving the reactor has a high moisture content, which in turn could indicate that steam is escaping through a crack in the dryer. See Exhibit E3-06, Appendix D. Such escape would be symptomatic of a significant crack that might result in loss of physical integrity of the dryer.

Q26. What additional recommendations are contained in Revision 2 to GE-SIL-644?

A26. (LDL) Appendix D of GE-SIL-644, "Monitoring Guidelines,"

which provides guidelines for monitoring moisture carryover and other parameters that may be indicative of dryer damage, has been expanded from three pages in Rev. 1 to six pages in Rev. 2, and the guidance it contains is commensurately more specific and precise.

The changes to Appendix D provide much more detailed guidance for plant personnel in effectively monitoring moisture carryover as an indicator of dryer cracking.

Q27. How is the monitoring of moisture carryover to be performed?

A27. (JRH) Moisture carryover is measured by plant chemistry personnel using procedure OP-0631 Appendix F (Exhibit E3-10).

If moisture carryover is determined to be greater than the limit stated in the procedure (currently 0.19%), the procedure requires that a Condition Report ("CR") be written, the Shift Manager 12

notified, and actions taken in accordance with Off-Normal Procedure ON-3178.

Q28. How and by whom is the significance of the measurements of moisture carryover to be assessed?

A28. (JRH) The Shift Manager performs the initial assessment of the data. If the moisture carryover is in the range of 0.19% to 0.35%,

Off- Normal Procedure ON-3178 requires that plant management and engineering (the General Manager, Operations Management, Systems Engineering, Reactor Engineering, and Design Engineering Mechanical) be informed. Additionally, the Operational Decision Making Initiative ("ODMI") process is to be initiated. The ODMI process is used whenever an emergent plant issue comes up that requires an especially methodical, systematic, conservative decision making process affecting station operation.

The ODMI procedure provides the structure and process for these decisions and helps to ensure that senior management oversight, vendor expertise, fleet expertise, industry experience and any additional resources are applied in an effective, efficient manner.

In addition, an engineering evaluation in accordance with EN-OP-104 "Operability Determinations" is to be performed. If moisture carryover exceeds 0.35% station management is notified and operability evaluation is requested. If the results of the evaluation do not support continued plant operation, the reactor is brought to hot shutdown. In either case, experienced qualified engineering personnel will determine the significance of the abnormal moisture carryover measurement.

Q29. How is the monitoring of main steam line flow, reactor vessel water level, and steam dome pressure performed?

13

A29. (JRH) Data on these parameters are available continuously in the control room, and the control room operators monitor and record these data.

Q30. What happens if abnormal values of main steam line flow, reactor vessel water level, and steam dome pressure are measured?

A30. (JRH) If any of the action levels for main steam line flow, reactor vessel water level, and steam dome pressure are reached (as described in Procedure ON-3178), the procedure requires that a moisture carryover measurement be performed by the plant's Chemistry Department. The results of that measurement dictate the required response, as described above.

Q31. What are the qualifications of the personnel assessing the significance of the moisture carryover and other measured parameters?

A31. (JRH) The personnel involved in determining the significance of the moisture carryover and other measured parameters are required to be qualified in the application of the operability determination procedure EN-OP-0104 (Exhibit E3-11 hereto). A prerequisite for procedure qualification is the requirement that the individual(s) be enrolled in the Engineering Support Personnel training program and that their capability to perform independent engineering work be assessed by their supervisor. This is part of Entergy's Engineering Support Personnel training program, which includes an annual assessment of individual training needs by the engineer and his supervisor. Thus, if an engineer or his supervisor feels the engineer needs additional training to maintain or enhance his level of expertise, that training is incorporated into the performance goals for the year. Thus, by virtue of their training, the members of the VY technical staff are qualified to make a decision as to the significance of moisture carryover measurements.

14

Q32. Will the plant continue to operate if values of moisture carryover or the other parameters indicate that steam dryer cracking may have occurred?

A32. (JRH) No. As stated in Procedure ON-3178, "IF the engineering evaluation of plant data confirms that steam dryer damage may have occurred, THEN: initiate a plant shutdown per OP 0105, place the plant in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, evaluate reportability per AP 0156."

Q33. Will the measurements of moisture carryover, main steam line flow, reactor vessel water level, and steam dome pressure enable Entergy to determine whether a dryer crack is about to form?

A33. (JRH) No. The purpose of these measurements is to provide early warning to the plant personnel that a crack may have developed so that appropriate, timely action may be taken before undesirable effects ensue as a result of the crack.

Q34. Why is it acceptable for the plant personnel not to know of the likely formation of steam dryer cracks in advance of their actual occurrence?

A34. (JRH) There is no technology that will predict when a crack will initiate. Crack formation can be caused by one of three mechanisms: stress relief, fatigue, or intergranular stress corrosion. These mechanisms are not unique to the steam dryer.

Stress relief cracks typically form early in a component life and tend to arrest once the locked-in fabrication stresses are relieved through crack formation. Fatigue cracks can be experienced by components subjected to cyclic stresses of sufficient magnitude and duration. Stress corrosion cracks can occur when a component of a given material is subjected to a particular combination of stresses and environmental conditions.

Inspection and monitoring programs are developed and implemented to ensure that any cracks developing as a result of one 15

of these mechanisms are detected before they develop to a size that would be of concern.

C. Dryer Inspection Program Q35. Please describe the guidance that will be used in performing the steam dryer monitoring program during the period of extended plant operation following renewal of the VY license.

A35. (LDL) Because VY has a BWR-3 steam dryer, the details of the visual inspection program to be implemented are set forth in the section of GE-SIL-644 devoted to such dryers, which is Appendix C, pp. 15-16. The dryer inspections are to be performed in accordance with the VY BWRVIP Program Plan, VY-RPT-06-00006 (Exhibit E3-12) and GE-SIL-644, Revision 1.

The dryer examinations consist of VT-I and VT-3 examinations of accessible internal and external welds and plates in the steam dryer potentially susceptible to crack formation. VT-I and VT-3 examinations are defined by ASME Boiler & Pressure Vessel

("BPV") Code Section XI, and the non-destructive examination technicians who perform and review these examinations are qualified in accordance with ASME BPV Section XI.

Q36. Please describe the VT-I and VT-3 examinations.

A36. (LDL) Briefly, these examinations can be described as follows:

A VT-I visual examination under BWRVIP standards (such as the steam dryer inspections) is one capable of achieving a resolution to discern a 0.044 inch (slightly over 1/32 inch) high lower case character with no ascender or descender strokes (e.g., an a, c, e, or o) on an 18% neutral gray card. A VT-1 examination determines the condition of a part, component or surface, including cracks, wear, corrosion, erosion, or physical damage on the surfaces of the part or component.

16

A VT-3 visual examination is intended to determine the general mechanical and structural condition of components, such as the verification of clearances, settings, physical displacements, loose or missing parts, debris, corrosion, wear, erosion, or the loss of integrity at bolted or welded connections. A VT-3 examination under the BWRVIP standards requires character recognition of 0.105 inch characters on an 18% neutral gray card, similar to the VT-I demonstration.

The VT-I and VT-3 steam dryer examinations at VY also include the steam dryer hold-down bolts, the tie bars, the manway cover, the lifting eye assemblies, and the lifting eye attachment welds.

In addition, other examinations have been added to the scope of dryer inspections as a result of industry experience. For example, the Spring 2007 examination scope included VT-1 examination of all four steam dryer support lugs, which are reactor vessel wall integral attachments. Although not specifically identified as-being within the required scope of steam dryer inspections, certain U.S.

BWRs have experienced accelerated wear on these lugs.

Therefore, VY chose to inspect the support lugs as a good practice addition to the inspection program. No accelerated wear was noted on theselugs at VY.

Steam dryer examinations at VY are performed using high resolution color cameras and are recorded directly to digital video disks ("DVDs") for review and evaluation. A "resolution demonstration" is performed by aiming the camera at a "Sensitivity Resolution Contrast Standard" and verifying that the lighting and the equipment at actual test conditions meets the 0.044 inch resolution requirement for the examination being performed.

17

The resolution demonstration is also recorded on DVDs for future review. Therefore, the technicians and structural engineers reviewing the inspection results see exactly the same surface conditions that the inspecting technician saw during the examination.

Q37. How often are the steam dryer inspections conducted?

A37. (LDL) The steam dryer is inspected during each scheduled refueling outage, approximately every 18 months.

Q38. Who performs the steam dryer inspections?

A38. (LDL) The inspections are performed by qualified non-destructive examination ("NDE") inspection personnel, using qualified NDE techniques appropriate for BWR steam dryer inspections. Because of the large number of individual examinations to be performed during a refueling outage, this work is typically contracted out to qualified vendors, including the reactor supplier (General Electric).

Q39. What are the training and qualification requirements applicable to Level II NDE personnel?

A39. (LDL) A Level II NDE technician is qualified in accordance with the requirements for Level II NDE technicians in the ASME BPV Code Section XI. This qualification includes documented training that must be given by a Level III NDE technician who is qualified in the examination of interest; a visual acuity and color acuity test; and documented NDE experience under the direct oversight and supervision of a qualified NDE Level III technician.

Qualifications to perform VT-1 and VT-3 examinations are achieved separately and are subject to separate, documented qualification processes. The NDE technician must renew these qualifications; must re-take the vision tests; and must document examination experience as part of ongoing proficiency 18

maintenance. The specific requirements for this qualification and requalification process are found in ASME BPV Code Section XI.

Q40. Who evaluates the dryer inspection data?

A40. (LDL) The inspection data are reviewed by qualified Level III NDE personnel and are subject to final acceptance by Entergy Level III NDE personnel. The ASME BPV Code requires that the examinations be performed by a qualified NDE technician and that the examinations and their results be reviewed and approved by a qualified NDE technician. VY typically contracts both the Level II and Level III services for reactor vessel and internals examinations, including the steam dryer examinations. As an additional quality step, Vermont Yankee requires that these examinations also be reviewed by an Entergy Level III NDE technician. The Entergy Level III review and approval is required to be completed on 100% of the steam dryer examinations prior to its return to service.

Q41. What are the training and qualification requirements applicable to Level III NDE personnel?

A41. (LDL) Similar to the Level II NDE technician, a Level III NDE technician is qualified in accordance with the requirements for Level III NDE technicians in the ASME BPV Code Section XI.

This qualification includes documented training that must be conducted by a Level III NDE technician who is qualified in the examination of interest; a visual acuity and color acuity test; and documented NDE experience under the direct oversight and supervision of a qualified NDE Level III. The principal difference between a Level II and a Level III technician is the amount of formal training and the number of hours of documented examination experience. The NDE technician (both Level II and Level III) must renew these qualifications; must re-take the vision 19

tests; and must document examination experience as part of ongoing proficiency maintenance. The specific requirements for this qualification and requalification process are found in ASME BPV Section XI.

Q42. How does Entergy assess the significance of any detected evidence of potential dryer cracks?

A42. (LDL) All detected indications are evaluated by qualified structural engineers, who are experienced with BWR steam dryer crack evaluation. Typically, these indications are evaluated by engineers who are on the staff of the reactor vendor, and the evaluations and conclusions are reviewed and accepted by qualified Entergy structural engineers.

Q43. What are indications?

A43. (LDL) Indications are surface discontinuities. An indication is an imperfection or unintentional discontinuity that is detected by nondestructive examination. Not all indications are cracks.

Sometimes an indication is a shadow that results from a peculiarity of the surface condition or illumination. An indication is classified as recordableor relevant if it is visible to the resolution of the examination technique. For example, any apparent surface imperfection identified during a VT-I would be considered recordable or relevant, and any surface imperfection that is 0.044 inch or greater would be visible. All recordable indications found by the Level II NDE technician who performs the examination are identified and documented in the corrective action program. All recordable indications that are confirmed by the Level III NDE technician are evaluated by Engineering to determine whether or not they are rejectable. Rejectable indications are those that must be repaired prior to restarting the plant. Repair of rejectable indications is an ASME BPV Code Section XI requirement.

20

Q44. How are recordable steam dryer indications evaluated?

A44. (LDL) Recordable indications in the steam dryer are evaluated to determine whether the indication represents a potential crack or just a surface imperfection. As stated previously, cracks in BWR steam dryers are one of the following three types: fatigue cracks; intergranular stress corrosion cracks ("IGSCC"), and stress relief cracks. Cracks in the steam dryer are typically stress relief cracks and self-arrest when the stress is relieved.

In evaluating a VT- examination, it is possible to determine from the video recording of the examination whether an indication has the characteristics of a fatigue crack, an IGSCC crack, or a stress relief crack, based on its location and physical characteristics.

IGSCC cracks arrest when the initiating condition (a combination of stress and environment) is mitigated. IGSCC cracks typically appear in the heat-affected zone ("HAZ") of a weld. Stress relief cracks may occur at any location with sufficient residual manufacturing stress.

If the characteristics of a particular indication do not rule out fatigue, the indication is typically classified as a potential fatigue indication. Subsequent examinations determine whether the indication is growing. A fatigue crack would be expected to grow; a crack that does not grow would not show the characteristic behavior of fatigue and would not be of concern.

An important aspect of visual examinations is that all recordable (relevant) indications are examined in subsequent cycles to determine whether any of the indications is an active crack. All recordable indications are reinspected at each refueling outage until at least two consecutive inspections show no growth.

21

Q45. What provides assurance that indications will not become cracks that propagate to the point of causing a dryer failure during the period between inspections?

A45. (LDL) Part of the indication evaluation is the determination of the potential for growth during the following eighteen month operating cycle until the next refueling inspection. IGSCC cracks are typically short and tight (on the order of 1 inch long and show no measurable width) and typically do not grow in subsequent cycles, since protection by hydrogen injection and noble metal application produces an environment that resists IGSCC propagation.

Stress relief cracks may be larger and may show measurable width.

After the residual stress is relieved, a stress relief crack is arrested.

Vermont Yankee has not identified any steam dryer cracks that are consistent with fatigue, and this conclusion is supported by the fact that the identified indications have not grown during subsequent operating cycles.

Q46. How many steam dryer inspections have been conducted at VY and what have been their results?

A46. (LDL) The VY steam dryer has been inspected as part of the BWRVIP inspection program since VY's adoption of the BWRVIP guidelines. The first dryer inspections were conducted in the Spring 1998 refueling outage. VT-3 examinations of the dryer and steam separator were performed, and five indications were identified on tack welds. The Fall 1999 refueling outage reexamined the five tack welds where indications had been identified in the previous outage. The Fall 2002 outage included VT-I and VT-3 examinations of steam dryer cover plates and welds.

Most recently, during the Spring 2004 refueling outage, in preparation for EPU, the dryer received a baseline VT-I inspection of all accessible areas deemed potentially susceptible to crack 22

formation. These examinations comprised 287 weld and plate examinations. See Exhibit E3 -13, a summary that I prepared of the 2004 inspections and their results based on my review of the dryer inspection results reported by General Electric, which conducted the inspections. A total of 20 indications were identified, of which 2 were weld-repaired, and 18 were determined acceptable to use as-is. Physical modifications to the VY steam dryer were made in 2004 to increase the dryer's resistance to vibration loadings.

The steam dryer inspections performed during the Fall 2005 outage examined all high-stress areas, as identified in GE-SIL-644. In addition, all areas that had been repaired or modified in the Spring 2004 outage were reinspected, as well as those indications that were found and evaluated to be acceptable for use as-is during the Spring 2004 outage. See Exhibit E3-14, a summary that I prepared of the 2005 inspections and their results based on my review of the dryer inspection results reported by ARE VA, which conducted the inspections. These examinations comprised 113 internal and external weld examinations. A total of 66 indications were identified, including 20 previously identified indications and repaired areas from 2004, all of which were found acceptable for use as-is. The increase in identified VT-i indications was due to increased resolution of the VT- I examinations. Therefore, several indications not previously visible were identified in the 2005 examinations, even though the number of examinations was less than in 2004.

During the Spring 2007 outage, all accessible susceptible areas of the steam dryer were inspected, consistent with the guidance in SIL-644, Revision 1. See Exhibit E3-15, a summary that I prepared of the 2007 inspections and their results based on my review of the dryer inspection results reported by General Electric, which conducted the inspections. The previously repaired areas, the 23

identified high stress areas, as well as those indications that were previously found and evaluated to be acceptable for use as-is were also examined. The examination included susceptible accessible internal and external welds and plates. A total of approximately 448 individually identified steam dryer examinations were performed. VY specified that the examination quality in 2007 must be sufficient to identify all indications found in the 2005 examinations. The 0.044 inch character resolution was required and documented on the examination DVDs.

A total of 66 indications were recorded, including 47 of those identified in 2005 and 19 previously unidentified indications.

These 19 previously unidentified indications were the result of the increased examination scope in 2007 compared to that in 2005 (448 in 2007 and 94 in 2005), and the fact that all accessible susceptible areas of the steam dryer had been subjected to the improved resolution VT-1 as a result of the 2005 experience. All 75 of the indications identified in 2007 were accepted for use as-is.

In the years between 2004 and 2007, certain additional areas potentially susceptible to crack formation were identified beyond those identified as susceptible as of 2004. Therefore, the total scope of the 2007 dryer inspection (463 examinations) was larger than that of the 2004 inspection (287 examinations) by 176 examinations.

The number of examinations performed in the dryer inspections conducted since 2004 and the number of indications found in each can be summarized as follows (See Exhibits E3-13, 14 and 15):

24

Year of Total Recorded Rejected Indications Previously Dryer Dryer Indications Indications Dispositioned Recorded Exam Exams (Repaired) Use As-Is Indications Re-Identified*

2004 287 20 2 18 N/A 2005 113 66* 0 66 20 (100%)

2007 463 66 0 66 47**(71.2%)

• The 66 recorded indications in 2005 included 18 relevant indications reported in 2004 plus the repaired areas on the two rejected indications from 2004. Therefore, there were 46 newly identified relevant indications.

• • In 2007, 47 previous indications were re-identified; 19 new relevant indications were identified in 2007 because of increased scope and improved VT resolution compared to 2004; and 18 previous indications were not re-identified in 2007, either due to surface conditions or because the previous indication was determined to be non-relevant.

Q47. What was your role in those inspections?

A47. (LDL) Having become acting Supervisor of Code Programs, my role in the Fall 2005 outage was to ensure the proper completion and evaluation of the steam dryer inspections, as well as other Code-required tests and inspections.

In 2007, I was again directly involved in the oversight of all Code Programs inspection and testing activities. As related to the steam dryer inspections, I had multiple daily contacts with NDE Level II, Level III, and supervisory personnel to monitor the progress and appropriateness of steam dryer inspections and their timely evaluation and resolution. I was directly involved in the conclusions and decisions regarding the evaluation and resolution.

of steam dryer indications, and participated in the telephone conversation with the NRC staff in which we reported our preliminary examination results prior to startup.

Q48. How were the steam dryer inspections and evaluations conducted?

A48. (LDL) The inspections were performed using the VT-1 visual examination technique. The examinations were performed by qualified Level II NDE personnel, using approved industry 25

standard techniques, and the results were evaluated by qualified Level III NDE personnel. As previously discussed, as an additional quality check, 100% of the steam dryer examinations were reviewed and accepted by qualified Entergy Level III NDE personnel. The examination video, which was directly recorded in color, was transmitted to the reactor vendor structural engineers, who had access to exactly the same information that the NDE technicians used to make the initial assessment that 75 indications were recordable. These evaluations were reviewed, discussed, and accepted by Entergy Level III NDE personnel, structural engineering and site management. Both Entergy and the reactor vendor's structural engineers agreed that none of the indications required repair.

Q49. What have been the results of the steam dryer inspections conducted while the EPU was being implemented and thereafter?

A49. (LDL) As noted earlier, the Spring 2004 steam dryer baseline inspections found 20 indications, of which 2 were repaired and the remainder (18) were deemed acceptable for "use as-is." All of the indications were evaluated, and none were determined to be fatigue cracks.

The Fall 2005 steam dryer inspection found 66 indications, including the 20 identified in 2004, all of which were evaluated and dispositioned "use as-is." None of the previously identified indications showed growth, which suggests that fatigue is not occurring. Further, the fact that none of the previously identified indications showed growth in the 2005 inspection is evidence that the additional indications were in all likelihood not new, but simply became visible due to the enhanced examination capability.

Again, all of the indications were carefully evaluated, and in no instance was any indication of fatigue evident.

26

The steam dryer inspections conducted in the Spring of 2007 followed approximately one year of full power operation at the EPU condition. The examinations were again conducted using enhanced examination resolution, which provides improved detection levels over those achievable by using the prescribed VT-1 examination process. Due to differences in surface conditions and illumination, some conditions that were identified as indications in 2005 were no longer identified as such in 2007. As a result, 47 of the 66 indications identified in 2005 were again identified in 2007.

In addition, 19 new indications were identified for a total of 66 dryer indications. Each of these 66 indications was evaluated by qualified structural engineers, experienced in evaluating indications in BWR steam dryers. Each of the indications was accepted to "use as-is." No growth was noted in the previously identified indications. None of the cracks were determined to be associated with fatigue.

IV. RESPONSE TO CLAIMS IN NEC'S TESTIMONY AND EXHIBITS ON NEC CONTENTION 3 Q50. Have you had the opportunity to review the testimony and exhibits submitted by NEC in this proceeding on April 28, 2008 relating to NEC Contention 3?

A50. (JRH, LDL) Yes, we have.

Q51. What testimony did you review?

A51. (JRH, LDL) We reviewed the direct testimony of Joram Hopenfeld, NEC Exhibit NEC-JH_0 1; Dr. Hopenfeld's report entitled "Assessment of Proposed Program to Manage Aging of the Vermont Yankee Steam Dryer Due to Flow-Induced Vibrations" (April,25, 2008), NEC Exhibit NEC-JH 54 ("NEC 27

Dryer Report"); and NEC Exhibits NEC-JH_ 55 through NEC-JH_61.

Q52. What are the main claims raised in NEC's testimony and exhibits on Contention 3?

A52. (JRH, LDL) NEC's consultant Dr. Joram Hopenfeld makes two main claims regarding the VY steam dryer management program:

(1) that monitoring of plant parameters indicative of potential dryer cracks is insufficient to prevent fatigue cracks from forming and propagating in the period between dryer inspections; and (2) that a dryer management program must include estimating the stress loadings on the dryer and ensuring that they remain within the stress limits of the dryer material.

A. Alleged Limitations in Steam Dryer Monitoring Program Q53. In the NEC Dryer Report, Dr. Hopenfeld states that "[m]oisture monitoring only indicates that a failure has occurred; it does not prevent the failure from occurring." NEC Dryer Report at 5. Is monitoring subject to the limitation described by Dr. Hopenfeld?

A53. (JRH) No. Monitoring of plant parameters will not predict the incipient formation of dryer cracks, but it will identify the existence of a crack sufficiently large to adversely affect dryer performance and flag the risk of structural failure of the dryer.

Since the steam dryer has completed two years of EPU operations and will have completed eight years of operation at EPU conditions prior to entering the period of extended operation in 2012, such extended operation without the detection of large cracks provides a high degree of assurance that the steam dryer is not subject to rapid flaw growth due to high cycle fatigue. Thus, the monitoring program is sufficient to provide an "early warning" of potential dryer failure so that action can be taken prior to the occurrence of such failure.

28

Q54. Dr. Hopenfeld also states: "Most parameter monitoring (moisture, steam flow, water level, dome pressure) may indicate the formation of only those steam dryer cracks that increase moisture carryover; those cracks that do not lead to significant moisture carryover may continue to grow undetected." NEC Dryer Report at 4. Is this a valid concern?

A54. (JRH) No. The steam dryer is a direct part of the steam flow path in the reactor. Unexplained changes in steam flow rate, reactor vessel water level and/or steam dome pressure can each be indicative of a change in the overall steam path pressure drop, and therefore a loss in dryer efficiency that may be caused by steam dryer damage. That is why these parameters are continuously monitored. Since all the reactor steam flows through the steam dryer, it is very unlikely that any damage to the dryer would not also result in a decrease in efficiency of the steam dryer (and thus result in an increase in moisture carry-over and a change in one or more of the monitored parameters).

Q55. Dr. Hopenfeld further asserts (NEC Dryer Report at 1-2) that "the history of steam dryer cracking at the VY plant indicates that Entergy's program to date of visual inspection and moisture monitoring have been ineffective in identifying cracking at the time it occurs, when it occurs in between inspections." What is your response to this assessment?

A55. (LDL) Dr. Hopenfeld misunderstands the purpose of the current and proposed steam dryer management program at VY. Entergy's program of visual inspections and moisture carryover and other parameter monitoring uses the latest approved examination techniques and is consistent with the current industry best practice.

All monitoring programs-including the inservice inspection program, which forms the basis for assuring the continued integrity of the reactor coolant system-are based on the principle that periodic monitoring and inspection, informed by the knowledge of plant materials and the physics of stress, strain, flaw initiation and crack propagation, will monitor material conditions on a frequency that is sufficient to identify and mitigate any flaws 29

before they can grow to a size that would be detrimental to the integrity of the component under consideration.

The results of the steam dryer inspection and monitoring program have been quite effective to date and will continue to be effective after license renewal. The program has demonstrated that steam dryer integrity can be assured by periodic inspection and monitoring. The overwhelming majority of visual indications have not grown since they were first identified, and those few indications that* were determined to need repair had not reached critical size (that is, they had not had a negative effect on steam dryer integrity) prior to repair. They also have not shown evidence of new growth since they were repaired.

Q56. Dr. Hopenfeld goes on to state (NEC Dryer Report at 3) that, once fatigue cracks initiate, they propagate very fast when exposed to alternating stresses of sufficient magnitude and frequency, so that even if one does not find cracks during an inspection, there is absolutely no reason why such cracks would not start propagating once the plant is restarted. Is this concern well founded?

A56. (LDL) Dr. Hopenfeld's concern is based on the assumption that there will be alternating stresses of sufficient magnitude and frequency to cause cracks to propagate rapidly. However, VY's operating experience after the EPU (exemplified by the data collected during the 2007 inspection and the subsequent year of monitoring of plant operating parameters) demonstrates that the stresses experienced by the dryer are insufficient to initiate and propagate fatigue cracks. Therefore, as the examinations in 2007 show, there is reasonable assurance that there are no stresses sufficient to initiate and propagate a fatigue crack in the Vermont

• Yankee steam dryer.

Q57. Dr. Hopenfeld further notes (NEC Dryer Report at 3) that, in its evaluation of the indications found during the 2007 refueling outage, General Electric determined 30

that the indications were IGSCC cracks but "GE did not rule out the possibility of continued crack growth by fatigue." Is his comment correct?

A57. (LDL) The reference cited by Dr. Hopenfeld and included with his testimony (Exhibit NEC-JH_59) contains no such statement. What General Electric actually stated was: "It is recommended that the visual indications reported in References 1 and 2 be accepted as-is for continued operation for at least one additional operating cycle.

Repair is not recommended at this refueling outage. These indications are most likely IGSCC and therefore they will propagate very slowly if at all. These indications have little or no structural impact on the steam dryer assembly and do not pose a risk of creating lost parts during the next operating cycle. These indications should be visually inspected during the next refueling outage to confirm there has been little or no growth." See Exhibit E3-16 at 4. There is no doubt, based on the history of the indications inspected during the 2007 refueling outage, that they are IGSCC cracks and not fatigue cracks.

Q58. Do the results of the most recent dryer inspections shed any light on the long term outlook for the physical integrity of the VY steam dryer?

A58. (LDL) Yes. The most recent steam dryer inspections show that the VY steam dryer has a modest number of IGSCC and stress relief indications typical of its age and service. These inspections show that none of the indications identified to date are active; that is, they exhibit no discernible growth from one inspection to the next.

B. Alleged need to estimate dryer stress loads Q59. What are the stress loadings to which steam dryers are subjected during plant operations?

A59. (JRH) There are two types of stress loadings imposed on the steam dryer (as well as other plant components) during nuclear power plant operations. There are the normal operating loads that are 31

experienced day-in and day-out over the life of the plant. These loads are generally lower than the design basis (accident) loads, but because of their long duration or frequency they can induce fatigue damage. The design basis loads, on the other hand, are one-time loads. It is the normal operating loads that can cause the eventual failure of a steam dryer from vibration-induced fatigue.

Q60. NEC consultant Dr. Joram Hopenfeld asserts that the aging management program for the VY steam dryer should include "some means of estimating and predicting stress loads on the dryer, establishing load fatigue margins, and establishing that stresses on the dryer will fall below ASME fatigue limits" (Pre-Filed Direct Testimony of Dr. Joram Hopenfeld Regarding NEC Contentions 2A, 2B, 3 and 4, NEC Exhibit NEC-JH_01 at A16). Do you agree with Dr. Hopenfeld's assessment of the necessity of "estimating and predicting" stress loads on the steam dryer as part of the steam dryer aging management program?

A60. (JRH) No. While it is appropriate to estimate and predict stress loads on the steam dryer during the plant design and for the design validation performed for the EPU, it is not necessary to do so as part of the steam dryer aging management program, and there is no regulatory requirement or industry guidance that calls for ongoing estimation of steam dryer stresses.

Q61. Why is it not necessary to estimate and predict dryer stresses?

A61. (JRH) Confirmation that stresses on the VY steam dryer remain within its fatigue limits is provided daily by the fact that the dryer has been able to withstand without damage the increased loads imparted on it during power ascension and for the two years of operation since the EPU was implemented. Dryer performance to date demonstrates that none of the stresses on the dryer has exceeded the endurance limit for the component. It is important to note that there will be no change in dryer loads or stresses during the license renewal period of operation; hence, there is no reason to expect that the dryer will be subjected to increased stresses in the future.

32

Also, as I indicated earlier, the dryer monitoring program, supplemented by the periodic dryer inspections during refueling outages, is sufficient to diagnose whether significant dryer cracking has occurred before such cracking results in dryer failure. The same approach is used in other aspects of fatigue monitoring in the reactor system components, whose fatigue monitoring program does not require the estimation or prediction of actual loads on the components.

C. Effect of Uprate on Steam Dryer Performance Q62. In the NEC Dryer Report (at 3-4) Dr. Hopenfeld asserts that, as a result of the EPU implemented by Entergy in 2006, there has been a 20% increase in steam flow velocity, which in turn has increased the potential for fluctuating local pressure loadings on the dryer that may approach the natural frequency of the dryer. Is this a valid concern?

A62. (JRH) No. Industry experience shows that BWR steam dryers in use during uprated power operations that have inadequate fatigue resistance will most likely exhibit this inadequacy during the first fuel cycle. In other words, operation of a steam dryer for a year or two is sufficient to accumulate enough fatigue cycles to cause significant cracking in susceptible areas of the dryer. Conversely, good performance (such as exhibited by the VY steam dryer) during the first operating cycle after the uprate strongly suggests that the dryer will not experience a fatigue-induced failure.

Q63. Dr. Hopenfeld also expresses the view (NEC Dryer Report at 6) that instead of removing the instrumentation used during the power ascension phase of implementing the extended power uprate to estimate the loadings on the steam dryer, Entergy should have improved the analytical tools for predicting the loads on the dryer, perhaps by conducting additional scaling test at GE at the San Jose facility. Do you agree with this view?

A63. (JRH) No. The analytical tools that were used during the uprate proceeding to demonstrate that loads on the dryer will be below its 33

endurance limits were performed as part of the design validation process that demonstrated the adequacy of the design and established the current licensing basis. Because the predicted loads on the dryer were shown to be below the endurance limit, the design analysis was not time limited and thus does not need to be revisited at the license renewal stage, where only time limited aging analyses need to be evaluated. Further, the loadings on the dryer derive from plant geometries (pipe lengths, diameters, flows, pipe connections, etc.). Those have not changed since the uprate was implemented, so there has been no change to the loadings on the dryer and the resulting stresses. Therefore, there is no reason to provide continued instrumentation to measure loadings or further analytical efforts.

Q64. Are the stresses on other plant components, such as piping, measured or monitored during normal plant operations?

A64. (JRH) No, they are not. Nor is there need for such measurements.

For those components, as for the steam dryer, a "defense in depth" approach is implemented, involving: 1. conservative prediction of loads; 2. conservative structural analyses to ensure stress limits are satisfied; 3. confirmation of design during start-up testing; and 4.

periodic inspections to confirm satisfactory performance or provide early warning of unexpected performance. The first three of these actions occur as part of the design process, or during the design validation for an uprate. Thereafter, as the facility operates, the inservice inspection program, which is conducted to the requirements of 10 C.F.R. § 50.55a and the ASME Boiler &

Pressure Vessel Code,Section XI examines feedwater nozzles, core spray nozzles, reactor recirculation nozzles, main steam nozzles and other components that have been determined potentially susceptible to fatigue. All of these ISI inspections are conducted to monitor safety-related piping and components for 34

aging effects due to fatigue and other environmental factors.

Routine inservice inspections have proven effective for monitoring the condition of piping and components and detecting early indications of aging, well before those components degraded to the point that their integrity might be compromised. These methods of component integrity monitoring do not rely on stress measurements.

The ISI program performs periodic visual and surface examinations, as well as volumetric examinations using ultrasonic tests ("UT") of the thickness of a pipe wall or weld on certain safety related welds and nozzle connections. By monitoring the condition of the metal through the pipe wall from the inside diameter to the outside diameter, any change in the metal subject to this UT examination is noted and evaluated in the corrective action program. The ISI program is a monitoring and trending program that does not rely on detailed stress analysis or direct stress measurements.

V.

SUMMARY

AND CONCLUSIONS Q65. Please summarize your testimony.

A65. (JRH, LDL) Our testimony can be summarized as follows:

Entergy has instituted a program, currently in effect and to be continued after renewal of the VY license, to continuously monitor plant parameters indicative of potential cracking of the steam dryer and properly evaluate and respond to any significant departures of those parameters from their normal range. That program is in accordance with industry guidelines and has been accepted by the NRC Staff for implementation during the current period of plant operations at uprated power level.

Entergy has instituted a program, currently in effect and to be continued after VY license renewal, to perform during each refueling outage thorough visual inspections of the 35

areas of the steam dryer potentially susceptible to crack formation. These inspections are conducted in accordance with industry guidelines and their methodology has been accepted by the NRC Staff for implementation during the current period of plant operations at uprated power level.

• The most recent steam dryer inspections show that the VY steam dryer has a modest number of service induced stress relief indications typical of its age and service. These inspections show that none of the indications identified to date are active; that is, they exhibit no discernible growth from one inspection to the next.

The fact that the VY steam dryer has shown no evidence of fatigue induced cracks after two years of EPU operation strongly indicates that routine inspection of the steam dryer during the period of extended operation will be sufficient to provide reasonable assurance of continued steam dryer integrity.

The steam dryer inspection and monitoring plan that Entergy will implement during the period of extended operation after license renewal will assure that the aging effects on the steam dryer will be adequately managed.

Q66. What are your conclusions regarding the assertions in NEC Contention 3?

A66. (JRH, LDL) We conclude that there is no factual support for the claims made in NEC Contention 3.

Q67. Does that conclude your testimony?

A67. (JRH, LDL) Yes, it does.

36

ORIGINAL May 12, 2008 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of ))

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

JOINT DECLARATION OF JEFFREY S. HOROWITZ AND JAMES C. FITZPATRICK ON NEC CONTENTION 4 -

FLOW-ACCELERATED CORROSION Jeffrey S. Horowitz and James C. Fitzpatrick state as follows under penalty of perjury:

1. We have prepared the attached "Testimony of Jeffrey S. Horowitz and James C. Fitzpatrick on NEC Contention 4 - Flow-Accelerated Corrosion" in the above captioned proceeding.

2. The factual statements and opinions we express in the cited testimony are true and correct to the best of our personal knowledge and belief

3. We declare under penalty of perjury that the foregoing is true and correct.

Executed on May 12, 2008

mC.z am- C. Fitzpatrick Executed on May 12, 2008

May 12, 2008 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

TESTIMONY OF JEFFREY S. HOROWITZ AND JAMES C. FITZPATRICK ON NEC CONTENTION 4 - FLOW-ACCELERATED CORROSION I. WITNESS BACKGROUND Jeffrey S. Horowitz ("JSH")

Q1. Please state your full name.

Al. (JSH) My name is Jeffrey S. Horowitz.

Q2. By whom are you employed and what is your position?

A2. (JSH) I am an independent consultant.

Q3. Please summarize your educational and professional qualifications.

A3. (JSH) My professional and educational experience is de-scribed in the curriculum vitae attached to this testimony as Exhibit E4-02. Briefly summarized, I have more than 36 years of experience in the field of nuclear energy and related disciplines. For the last 22 years, I have specialized in flow-accelerated corrosion ("FAC") and nuclear safety analysis.

My main client during this time has been the Electric Power

Research Institute ("EPRI"). I have also consulted for utilities that operate nuclear power plants, including Arizona Public Service, Exelon Nuclear, Pacific Gas & Electric, and Southern California Edison. In Canada, I have consulted for the CANDU Owners Group and Ontario Power Generation. I hold four degrees in mechanical engineering. Three of these degrees, including a doctor of science degree, are from the Massachusetts Institute of Technology.

Q4. What is the purpose of your testimony?

A4. (JSH) The purpose of my testimony is to address, on behalf of Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc. (collectively "Entergy"), Contention 4 sub-mitted by the New England Coalition ("NEC") in this pro-ceeding. As admitted by the Atomic Safety and Licensing Board ("Board"), NEC Contention 4 reads:

Entergy's License Renewal Application does not include an adequate plan to monitor and manage aging of plant piping due to flow-accelerated corrosion during the period of extended operation.

Memorandum and Order (Ruling on Standing, Contentions, Hearing Procedures, State Statutory Claim, and Contention Adoption), LBP-06-20, 64 N.R.C. 131, 192 (2006).

Q5. What is FAC?

A5. (JSH) FAC is a degradation mechanism that attacks carbon steel piping and vessels exposed to moving water or wet steam. It is important to understand that FAC is only one of several mechanisms that can affect the physical integrity of piping and components. The term "FAC" was coined to avoid the ambiguities present in the previously used term - "ero-2

sion-corrosion." Specifically, FAC is a corrosion mechanism, not a mechanical damage mechanism (i.e., erosion). Erosive damage also occurs in nuclear piping, but such damage is normally confined to small leaks.

FAC occurs because the protective oxide layer that builds on the surface of carbon steel components dissolves into the flow stream. This attack occurs under specific water chemistry conditions. If FAC is not detected, the piping or vessel walls will become progressively thinner, normally globally (i.e.,

over a broad area of the component), until the material in the affected area can no longer withstand internal pressure and other applied loads and a rupture (rather than a leak) eventu-ally occurs. It is the global nature of FAC wear that causes a pipe rupture, whereas localized damage due to other mecha-nisms (e.g., erosion only) causes leaks and does not impact the structural integrity of the piping.

As defined, FAC only attacks carbon steel components in the presence of purified flowing water or wet steam. It does not attack steels containing other fluids, such as oil. Steels con-taining appreciable amounts of chromium have been found immune to FAC.

Q6. What has been your professional involvement with FAC issues?

A6. (JSH) My involvement with FAC dates back to December 1986, when an elbow in the condensate system at the Surry Unit 2 nuclear plant failed catastrophically. This failure caus-ed steam and hot water to be released into the turbine build-ing, resulting in the deaths of four workers and severe injuries to others. Post-accident investigations revealed that FAC was 3

the cause of the degradation to the elbow. At that time, the U.S. nuclear fleet did not have programs in place to deal with single-phase (i.e., water only) piping degradation caused by FAC. Some programs were in place to deal with two-phase (i.e., water and steam) piping degradation, but in general, these programs were very limited in their scope.

In response to the Surry accident, EPRI became committed to developing a computer program that would assist utilities in determining the most likely places for FAC wear to occur, and thus the key locations to inspect for pipe wall thinning.

The late Bindi Chexal, the EPRI Program Manager, gave me the job of designing and implementing such a program. I de-veloped the computer program CHEC (Chexal-Horowitz Ero-sion Corrosion) and demonstrated and released it to U.S. utili-ties in 1987. CHEC was replaced by CHECMATE (Chexal-Horowitz Methodology for Analyzing Two-Phase Environ-ments) in 1989. CHECMATE expanded on the capabilities of CHEC by adding algorithms to calculate FAC under two-phase conditions. CHECMATE was the first program to ac-curately predict two-phase FAC. CHECMATE was later re-placed by the current program, CHECWORKS (Chexal-Horowitz Engineering Corrosion Workstation), in 1993. Each new version built on the success of the previous program and incorporated user feedback, improvements in software tech-nology, and available laboratory and plant data into the mod-eling used in the programs. I remained the technical lead per-son in the development of these new and revised versions.

Q7. Have you been asked to review the FAC programs for nuclear power plants?

A7. (JSH) Yes. I have performed, by myself or with another engi-neer, audits of the FAC programs at over fifty nuclear units in 4

the United States and Canada. The most recent FAC program audit I conducted was at VY, in April 2007.

Q8. Have you been involved in the development of industry standards governing FAC programs?

A8. (JSH) Yes. After the first several audits I performed, the need became apparent for a guidance document that would help utilities improve and standardize their FAC programs.,

NSAC-202L, entitled "Recommendations for an Effective Flow-Accelerated Corrosion Program," was the document cre-ated to meet this need. I played a key role in drafting the original version of NSAC-202L and resolving numerous util-ity and U.S. Nuclear Regulatory Commission ("NRC") com-ments on it. Since that time, I have played a significant role in each of the three subsequent revisions to NSAC-202L, which has become the most important standard-setting document for the conduct of FAC control programs in the United States.

NSAC-202L also has been accepted as a valuable guidance tool by the Institute of Nuclear Power Operations ("INPO")

and the NRC.

Q9. Have you written books or technical papers on FAC?

A9. (JSH) After developing CHECWORKS, I co-authored three books on FAC and related issues. One book is a compendium of FAC science and experience; it is the most complete refer-ence available on the subject of FAC. The other two books deal with thermal-hydraulic issues. I have also authored or co-authored more than 30 EPRI reports related to FAC and nuclear safety issues. I was the principal investigator and sole author of 16 of them. Among the most important reports with which I have been involved are a study of weld attack in nu-5

clear piping and preliminary guidance for the protection of piping against damage from erosive forms of attack.

I have also written a number of technical papers on FAC, in-cluding papers presented at the International Conference on FAC, "FAC2008" held in Lyons, France in March 2008, at "Water Chemistry of Nuclear Reactors - Chimie 2002" held in Avignon, France (a meeting attended by over 300 interna-tional scientists and engineers), at ASME Pressure Vessel and Piping Conferences, at a Nuclear Regulatory Commission Water Reactor Safety Meeting, and at other technical meet-ings.

Q10. Have you given lectures or technical presentations on FAC?

A10. (JSH) I have made technical presentations at each of the semi-annual CHUG (CHECWORKS Users Group) meetings.

CHUG meetings typically attract between 50 and 100 utility engineers and station managers. I have made presentations at every one of the 38 CHUG meetings. My presentations typi-cally cover the results of research I have performed or are technical presentations regarding FAC. In addition to making presentations, I have served as session chair and moderated various discussion groups.

I have also conducted more than two dozen two or three-day training sessions covering FAC and the use of the EPRI com-puter programs (CHEC, CHECMATE and CHECWORKS).

These training sessions have been held in the United States and in foreign countries, including Belgium, Canada, the Czech Republic, Japan, South Korea and Taiwan. The train-ing sessions have been attended by utility engineers, utility managers, engineers from the 1NPO, and the NRC Staff. I 6

also was an invited participant in the NRC Erosion-Corrosion Workshop in February 1993.

I have developed for EPRI two computer-based training mod-ules. One of these modules covers FAC and the other covers erosive attack on piping in power plants. These modules have been distributed to EPRI member utilities. I also continue to be actively involved in training people to use the latest ver-sions of CHECWORKS.

James C. Fitzpatrick ("JCF")

Qll. Please state your full name.

All. (JCF) My name is James C. Fitzpatrick.

Q12. By whom are you employed and what is your position?

A12. (JCF) I am employed by AREVA, NP as an Engineering Su-pervisor. Until March 2008, I was employed at Entergy Nu-clear Operations, Inc. ("Entergy") as a Senior Lead Engineer in Design Engineering at VY.

Q13. Please summarize your educational and professional qualifications.

A13. (JCF) My professional and educational experience is de-scribed in the curriculum vitae attached to this testimony as Exhibit E4-03. Briefly summarized, I have thirty years ex-perience in the design, construction, and modification of nu-clear power plant structures, piping systems, pressure vessels, and in the seismic evaluation of mechanical and electrical equipment. Twenty-two of those years are in operating plant engineering support in both the mechanical and structural ar-eas. I have been responsible for the development and imple-mentation of plant design changes, inspection programs, 7

equipment specifications, installation support, outage support, and operability evaluations of degraded components.

Q14. What is the purpose of your testimony?

A14. (JCF) The purpose of my testimony is to address those aspects of NEC Contention 4 that relate to Entergy's activities to ad-dress FAC at VY, and particularly the FAC Inspection Pro-gram in place at VY, which is to be continued during the pe-riod of plant operations after renewal of the VY license.

Q15. What has been your role with respect to FAC control activities at VY?

A15. (JCF) My involvement with FAC dates back to 1987. While employed at Yankee Atomic Electric Company, I assisted in the preparation of VY's response to NRC Bulletin 87-01, "Thinning of Pipe Walls in Nuclear Power Plants," issued as a result of the December 1986 Surry accident. Later, I per-formed the first modeling of plant piping systems at VY using the-EPRI CHEC code to help select the single-phase piping component inspection locations for the 1989 refueling outage.

I was responsible for the development of a long term "Piping Erosion-Corrosion Inspection Program," ("FAC Program")

for VY in 1990. Development of this FAC Program involved determining the scope of piping potentially affected by FAC, modeling the plant systems using the CHECMATE code, de-veloping the criteria and procedures for performing the in-spections, and evaluating inspection data. I have either pro-vided engineering support or have been responsible for im-plementing the FAC Program at VY for the thirteen refueling outages from 1989 thorough 2007.

8

My responsibilities with respect to FAC included reviewing industry experience with FAC and assessing the impact of that experience on the FAC Program at VY. As such, I have developed contacts with other plant FAC Program Engineers by attending many of the EPRI-sponsored CHUG meetings since 1987. I also participated in the NRC Erosion-Corrosion Workshop in February 1993. 1 have participated either as a team member or a technical specialist in audits and assess-ments of FAC programs at six other nuclear plants.

Q16. What were your most recent duties with respect to FAC at VY?

A16. (JCF) I was the Cognizant Engineer for the VY FAC Program through June 2007. I was responsible for developing the scope of refueling outage inspections, providing on-site engi-neering support, screening and evaluating piping and compo-nents, determining if the sample of piping locations desig-nated for inspection during a refueling outage needed to be expanded, coordinating piping and component repairs and re-placements, updating the CHECWORKS models of plant pip-ing systems, and maintaining the FAC Program Manual and supporting documents.

II. DESCRIPTION OF VY'S PROPOSED FAC PROGRAM Q17. Would you please describe the program that VY proposes to implement to control FAC during the period following license renewal?

A17. (JCF) As stated in Section B. 1.13 of the License Renewal Ap-plication for VY ("Application") (Exhibit E4-04), the VY program for addressing FAC is consistent with the program described in the NRC guidance document "Generic Aging Lessons Learned (GALL) Report -- Tabulation of Results,"

NUREG-1801, Vol. 2, Rev. 1 (Sep. 2005) ("NUREG-1801" or "GALL Report"),Section XI.M17, Flow Accelerated Cor-9

rosion (Exhibit E4-05). Exhibit E4-04 at B-47. There are no exceptions in the Application to the guidance in NUREG-1801 with respect to FAC.

The original VY FAC Program was instituted prior to the is-suance of EPRI's guidance document NSAC-202L. How-ever, the FAC Program's documents have been revised as necessary over time to conform to the recommendations in the various revisions to NSAC-202L. The FAC Program cur-rently in effect (set forth in Entergy Procedure EN-DC-315, Rev. 0, Exhibit E4-06) substantially follows the current ver-sion of NSAC-202L, NSAC-202L-R3 (Exhibit E4-07).

The VY FAC Program includes, as recommended in the GALL Report and the NSAC-202L guidelines, "procedures or administrative controls to assure that the structural integrity of all carbon steel lines containing high-energy fluids (two-phase as well as single-phase) is maintained." Exhibit E4-05 at XI.M-61. A program implemented in accordance with the EPRI guidelines predicts, detects, and monitors FAC in plant piping and other components, such as piping elbows and re-ducers, as recommended in the GALL Report. Id.

Q18. Can you explain how the FAC Program is used as an aging management tool?

A18. (JCF) The FAC Program includes the following activities: (a) conducting an analysis to determine critical locations; (b) per-forming baseline inspections to determine the extent of thin-ning at these locations; and (c) performing follow-up inspec-tions to confirm the predictions, or repairing or replacing components as necessary. Id. NSAC-202L (Exhibit E4-07) provides the general guidelines that are implemented in the FAC Program.

10

To ensure that all the aging effects caused by FAC are prop-erly managed, NRC guidance recommends that the FAC Pro-gram make use of, among other tools, a predictive computer program, such as CHECWORKS, that implements the guid-ance in NSAC-202L to satisfy the criteria specified in 10 C.F.R. Part 50, Appendix B, "criteria for development of pro-cedures and control of special processes." Exhibit E4-05 at XI.M-61.

Q19. How does the FAC Program proposed for the license renewal period at VY com-pare to the program currently being implemented?

A19. (JCF) The FAC Program during the license renewal period will be identical to the existing program. The program will conform to the EPRI guidelines contained in NSAC-202L. It will include "procedures or administrative controls to assure that the structural integrity of all carbon steel lines containing high-energy fluids (two-phase as well as single-phase) is maintained." It will also provide detailed instructions on:

" how to conduct the inspections;

• how to evaluate the inspection data;

" the acceptance criteria for inspected components;

" the disposition of components failing to meet accep-tance criteria;

" the expansion of the sample to other components simi-lar to those failing to meet acceptance criteria; and

• the updating of CHECWORKS models to incorporate inspection data.

II

Exhibit E4-06, Section 5.0.

Q20. What inspections are performed under VY's FAC Program?

A20. (JCF) The VY FAC Program conforms to the inspection rec-ommendations contained in NSAC-202L. See Exhibits E4-06, Section 5.0, and E4-07. The FAC Program calls for piping and component inspections to be conducted at each refueling outage, with the items to be inspected being selected based on:

" required re-inspections and recommendations from previous outages.

• CHECWORKS susceptibility rankings or to calibrate the CHECWORKS models.

• industry/ utility/ station experience including items identified through work orders and condition reports.

• the susceptible non-modeled large bore and small bore program piping.

• engineering judgment.

See Exhibit E4-06, Section 5.3.

Q21. Could you explain in more detail how CHECWORKS is used in the FAC Pro-gram?

A21. (JCF) The FAC Program at VY primarily uses CHECWORKS' FAC wear rate analysis. VY uses CHECWORKS as a tool in planning inspections, evaluating inspection data, and managing the ultrasonic thickness ("UT") data compiled over the past thirteen refueling outages at Vermont Yankee.

Q22. Are there features of the VY design that result in a reduction of the amount of pip-ing and components at a typical plant that are potentially susceptible to FAC?

12

A22. (JCF) Yes. Compared'to the majority of nuclear power plants in operation, VY is a relatively small and simple plant. There are fewer FAC-susceptible systems and piping components than at a typical plant, and many of those were either originally constructed of FAC-resistant materials or have been replaced with FAC-resistant materials since their initial installation.

VY has vane-type moisture separators with no reheat steam system. This eliminates a large amount of FAC-susceptible piping and a number of components known to be susceptible to FAC found in a typical nuclear power plant.

The extraction steam system piping, which contains a signifi-cant portion of the two-phase piping in a power plant, was originally constructed from FAC-resistant materials. A number of other components and associated piping subject to two-phase flow (wet steam) have been replaced with FAC-resistant materials.

The original plant design and the component replacements re-sult in a significantly smaller amount of FAC-susceptible pip-ing at Vermont Yankee as compared to the typical nuclear power plant of similar size.

Q23. Please describe the use of FAC-resistant material at VY.

A23. (JCF) The most effective action in a Boiling Water Reactor

("BWR") to minimize potential FAC effects is to use piping materials that are resistant to FAC. As previously stated, the original design of VY already incorporated FAC-resistant pip-ing for the entire extraction steam system. In addition, since the plant went into operation, carbon steel piping and equip-13

ment in a number of systems has been progressively replaced with FAC-resistant materials. These include:

" All 10 of its feedwater heaters.

• Both low pressure turbine casings, including the attached extraction steam nozzles and piping.

" All of the two-phase flow piping in the moisture separator drains system.

" The majority of the two-phase flow piping in the heater drains system except at the lowest pres-sure feedwater heaters.

• The majority of the turbine cross around piping.

" Small bore steam drain lines to the condenser for the high pressure cooling injection system, the reactor core isolation cooling system, and the advanced off-gas system.

" Small bore shell vent lines for all four of the high pressure feedwater heaters.

Nearly all of the large bore piping at VY which is exposed to two-phase flow was either originally constructed with, or re-placed with, FAC-resistant material.

The fluid environments in the remaining FAC susceptible large bore piping systems are either high quality (dry) steam or sin-gle-phase flow.

Q24. What other actions have been undertaken at VY to limit the effects of FAC?

A24. (JCF) The addition of the oxygen injection system in 1980 im-proved the water chemistry with respect to minimizing FAC.

Q25. Can you explain the role of water chemistry at VY with respect to the limiting of FAC?

14

A25. (JCF) At VY, the addition of oxygen into the conden-sate/feedwater stream mitigates the effects of FAC on piping exposed to single phase flow.

Oxygen is injected into the condensate and feedwater trains just downstream of the condensate pumps. This results in about 40 parts per billion ("ppb") dissolved oxygen in the condensate and feedwater trains. This level of dissolved oxy-gen serves to reduce the rate of FAC because, as mentioned previously, FAC is a dissolution process in which the oxide layer on the carbon steel pipe components dissolves into a flowing stream of water or water-steam mixture. By main-taining this concentration of dissolved oxygen in the conden-sate and feedwater lines, the stability of the oxide film is en-hanced and the rate of dissolution is reduced; hence, the po-tential for corrosion is decreased. This reduction of rates is clearly shown in Section 5.3.2.1, Table 5-2 of Exhibit E4-07.

III. DESCRIPTION OF CHECWORKS Q26. Can you briefly explain how CHECWORKS works as an analytic tool as part of a FAC Program?

A26. (JSH) CHECWORKS is a multi-purpose computer program designed to assist FAC engineers in identifying potential loca-tions of FAC vulnerability. CHECWORKS is designed to be used by plant engineers as a tool in identifying piping loca-tions susceptible to FAC, predicting FAC wear rates, planning inspections, evaluating inspection data, and managing inspec-tion data.

Q27. Would you please describe in general terms how CHECWORKS calculates ex-pected FAC rates?

15

A27. (JSH) The rate of FAC is a function of a number of variables that define: (1) the water chemistry; (2) the flow rate; (3) the geometry of the components; (4) the material properties of the components; (5) temperature; and (6) steam quality. Exhibit E4-07, Section 1.1.

CHECWORKS utilizes plant-specific user inputs defining; (1) the oxygen concentration in the feedwater and at the reac-tor steam effluent ("g. main steam nozzle); (2) thermody-namic conditions; and (3) flow rates, to calculate the water chemistry at each location in the model. These inputs are ap-plied, together with user-defined component geometry, to an EPRI-proprietary algorithm (the Chexal-Horowitz correlation) to provide an estimate of the rate of FAC for each modeled component.

Q28. How is the CHECWORKS evaluation performed for a particular plant?

A28. (JSH) The modeling of a nuclear unit starts with specification of global data. This process begins with the plant heat balance diagram ("HBD"). The HBD is a schematic representation of the major lines and connectivity of the power producing por-tion of the nuclear plant. The HBD model constructed in CHECWORKS is then populated with the thermodynamic conditions representative of each power level at which the plant has operated at or is contemplated to operate. The user then inputs the oxygen concentration conditions that have been used or are anticipated. These inputs define the opera-tional history of the plant in terms of what power levels have been used with what water chemistry for how long.

As discussed above, the user also inputs information concern-16

ing the piping systems to be analyzed. Most of this informa-tion is at the component level and deals with geometry, wall thickness, operating conditions, and pipe material. CHEC-WORKS includes over fifty geometry models to represent various component geometries. In cases where the component geometry does not match any of the models, the CHEC-WORKS user is instructed to either use a conservative model or schedule the component for inspection. Likewise, CHEC-WORKS conservatively assumes that steel components con-tain the lowest amount of alloying elements allowed by the specification (typically, zero). Such an assumption disregards the beneficial effects of some alloying elements (e.g., chro-mium) in retarding the onset of FAC.

Based on these user inputs, a "Pass 1 Analysis" is conducted to report predicted wear rates. The results of the Pass 1 Analysis, together with other information including operating experience at similar units, are normally used by the FAC en-gineer to generate a list of components for inspection.

Once this information is specified in the plant database, the plant engineers are able to conduct wear rate analyses of any or all of the piping defined in the database.

Inspection data may also be input into CHECWORKS. In-spection data may be input in the form of a matrix of thickness readings covering the component. Typically, these data sets are from ultrasonic measurements of the wall thickness at lo-cal points (i.e., grid points) or from scanning the component and recording the minimum thickness at grid points. Inspec-tion data are not required for a Pass 1 Analysis.

17

When inspection data are available, a "Pass 2 Analysis" can be run. A Pass 2 Analysis compares the measured inspection results to the calculated wear rates and adjusts the FAC rate calculations to account for the inspection results. The pro-gram does this by comparing the predicted amount of degra-dation with the measured degradation for each of the in-spected components. Using statistical methods, a correction factor is determined which is applied to all components in a given pipe line - whether or not they were inspected.

In addition to refining the Pass I Analysis, Pass 2 Analyses provide feedback to the analyst with respect to the goodness of fit of the model to actual results, the location of any out-liers, and the possibility of modeling improvements.

Q29. How is the power level history of the plant taken into account in performing the CHECWORKS evaluation?

A29. (JSH) In using CHECWORKS, the engineer breaks the oper-ating time of the plant into a number of periods with a nomi-nally constant power level and reasonably constant water chemistry. For each of these periods, the program calculates a corrosion rate for each component considered in the analysis.

The product of the corrosion rate and operating time (i.e., the predicted degree of corrosion) is added up for all the operating periods. Thus, the program predicts the "lifetime" corrosion for each component considered.

Q30. Have the results of CHECWORKS calculations been subject to verification?

A30. (JSH) Yes. The correlations in one of the predecessor pro-grams to CHECWORKS, CHEC, were initially based on FAC laboratory testing data from France and the United Kingdom and a combination of laboratory and plant operational data 18

from Germany.

When CHECMATE was written, and again when CHEC-WORKS was revised in the mid-1 990s, a large amount of plant inspection data were used to refine the accuracy of the program's predictions. These data sources included assem-bled data from a variety of U.S. nuclear units as well as avail-able laboratory data from England, France and Germany (Ex-hibit E4-08 at 7 7-33).

CHUG, the users' group associated with the program, has met twice a year since 1989 and has been the major source of feedback on the adequacy of the program. As Mr. Neil Wilmshurst, EPRI's Director of Nuclear Plant Technology, has stated in this proceeding in the "Declaration of Neil Wilmshurst in Support of EPRI's Opposition to Motion to Compel," dated April 18, 2008 (Exhibit E4-09), "[n]uclear

[p]ower plants properly using CHECWORKS have never re-ported a failure in a steam and feed water system pipe or com-ponent of 2" in diameter or greater." Exhibit E4-09 at ¶ 11.

Q31. How is CHECWORKS "updated" if there is a change in conditions at a plant us-ing CHECWORKS?

A31. (JSH) The CHECWORKS Pass 2 Analysis is performed with updated user input data that include inspection results, infor-mation on material replacements, operating regimes, and changes to operating conditions, such as flow rate or tempera-ture and the water chemistry used. None of the algorithms are modified by plant-specific data. Exhibit E4-09 at ¶ 23.

Q32. How can a user be assured that changes made to the plant's operating parameters remain within the modeling capabilities of CHECWORKS?

19

A32. (JSH, JCF) This is done in two ways. The range of input vari-ables is defined in the users' manual and checked by the pro-gram while the data are being input. Further, there is a data checking feature which ensures that the data are within the al-lowable range for program operation. At VY, Entergy con-firmed that the temperature and flow velocities resulting from the uprate are within the range of the correlations built into CHECWORKS.

Q33. How is CHECWORKS updated for a power uprate?

A33. (JSH) The use of the program does not change on account of a power uprate (or any other change in operating parameters),

and remains essentially as outlined above. All that needs to be done is to update plant-specific inputs into the CHECWORKS program. When a power uprate is implemented, a user simply does what he would normally do as part of any Pass 2 Analy-sis - update the relevant variables (.g., thermodynamic con-ditions, temperature, oxygen concentration, etc.), and let the program calculate the predicted FAC wear. The Pass 2 Analysis can be used as a planning tool by performing it in advance of the uprate to determine if, under uprate conditions, systems and sub-systems would experience significantly greater FAC rates than those predicted before the uprate.

CHECWORKS was specifically designed to accommodate power uprates and is routinely used throughout the U.S. nu-clear industry for this purpose. Exhibit E4-09 at ¶¶ 19, 20. It is important to emphasize that with the implementation of the power uprate at VY the only CHECWORKS inputs which af-fect wear rates that changed were the flow rate and the tem-perature.

20

Q34. Is it necessary to "recalibrate" or "benchmark" CHECWORKS when operating conditions change after a power uprate?

A34. (JSH) No. Power uprates are no different from other opera-tional changes. In fact, the differences in rates experienced in a power uprate are generally smaller than those experienced by units when their water chemistry changes. It has never been necessary to "re-calibrate," "re-baseline" or "bench-mark" CHECWORKS when plants have changed their water chemistry, the power output has been increased, or other op-erational changes have taken place.

IV. ISSUES RAISED IN NEC CONTENTION 4 Q35. Have you had the opportunity to review the testimony of NEC's consultants filed in this proceeding on April 28, 2008?

A35. (JSH, JCF) Yes, we have.

Q36. What testimony did you review?

A36. (JSH, JCF) We reviewed the direct testimony of Joram Hopenfeld, Exhibit NEC-JH_01; Dr. Hopenfeld's report, titled "Review of Entergy License Renewal Application for Ver-mont Yankee Nuclear Power Station: Program for Manage-ment of Flow-Accelerated Corrosion," Exhibit NEC JH_36; Exhibits NEC-JH 37 - NEC-JH_53; the direct testimony of Dr. Rudolf Hausler, Exhibit NEC-RH_01; Dr. Hausler's re-port, titled "Discussion of the Empirical Modeling of Flow-Induced Localized Corrosion of Steel Under High Shear Stress," Exhibit NEC RH_03; the direct testimony of Ulrich Witte, Exhibit NEC-UW_01; Mr. Witte's report, titled "Evaluation of Vermont Yankee Nuclear Power Station Li-cense Extension: Proposed Aging Management Program for 21

Flow-Accelerated Corrosion," Exhibit NEC UW_03; and Ex-hibits NEC-UW_04 - NEC-UW_22.

Q37. What are the issues raised by NEC in its testimony with respect to NEC Conten-tion 4?

A37. (JSH, JCF) NEC raises two categories of issues: (1) the al-leged insufficiency of the data that will be collected between the implementation of the extended power uprate ("EPU") in March 2006 and the start of extended operations in March 2012 to properly "benchmark" the CHECWORKS program so that it can give accurate FAC wear rate calculations; and (2) whether the implementation of the FAC Program to date has been adequate and will support an adequate FAC management program after license renewal.

Q38. Does NEC challenge any element of the proposed FAC Program as set forth in the License Renewal Application?

A38. (JSH, JCF) No. NEC's concerns are not with the proposed FAC Program (except for the "benchmarking" of CHEC-WORKS), but with Entergy's ability to properly implement.

the FAC Program based on NEC's concerns about its past his-tory.

A. ADEQUACY OF DATA COLLECTION Q39. What is the issue in controversy with respect to the adequacy of the data used to predict FAC effects on VY components?

A39. (JSH, JCF) NEC's consultants claim that CHECWORKS can-not be "calibrated" to model the operating conditions at VY since the EPU before the expiration of the current VY operat-ing license. Dr. Hopenfeld (Exhibit NEC-JH_36 at 25) con-tends that 10-15 years of data would be needed to calibrate the CHECWORKS model. Dr. Hausler (NEC-RH_01 at 3) 22

claims that it is his opinion "that 12-15 years is a reasonable estimate of the time necessary to calibrate the CHECWORKS model." Mr. Witte (Exhibit NEC-UW_03 at 21) contends that "separate industry guidance supports five to ten years of data trending" for CHECWORKS.

Q40. Do you agree with these assessments?

A40. (JCF) No. As indicated earlier, VY uses five criteria for se-lecting which components and locations will be inspected for potential FAC effects during a plant refueling outage. Those factors, which are consistent with the guidance in NSAC-202L, are: (1) pipe wall thickness measurements from past outages; (2) predictive evaluations performed using the CHECWORKS computer code; (3) industry experience re-lated to FAC; (4) results from other plant inspection pro-grams; and (5) engineering judgment.

CHECWORKS assists power plant engineers in determining the most likely places for FAC to occur, and thus, the key lo-cations to inspect for pipe wall thinning. However, it is only one of the tools that Entergy will use for that purpose.

With respect to the need to "calibrate" CHECWORKS, En-tergy will be able to use the CHECWORKS program effec-tively to assist in identifying the locations where piping in-spections should be performed based on data collected at VY since 1989 and in the three sets of inspections that will be conducted during refueling outages between the implementa-tion of the EPU and the expiration of the current license.

Those inspections will yield data for four and a half years of operation at the EPU levels.

23

Q41. Does the uprate have any effect on the ability of CHECWORKS to be used as an effective tool as part of the VY FAC Program?

A41. (JSH, JCF) No. As discussed above, there is no need to "re-calibrate" CHECWORKS when the operating conditions are due to an uprate change. The new values for flow rate and temperature are simply used as inputs into CHECWORKS and CHECWORKS provides FAC rate calculations for the mod-eled components under the uprated conditions. Because only the flow rate and temperature are changed due to the power uprate, any FAC rates established after the uprate will be con-stant and the effect of the uprate on FAC will, therefore, be apparent with the first inspection after the uprate. This first post-uprate inspection, VY-RPT-08-0002, Rev. 0 (Exhibit E4-

10) was performed in the Spring of 2007. The results of that inspection demonstrate that data from repeat inspections (be-fore and after the uprate) of large bore components in the fe-edwater system, which experience continuous flow, show that essentially no wear has occurred since the commencement of the EPU in March, 2006. Exhibit E4-10, Section 8.

As an added measure of conservatism, Entergy will increase the inspection scope by at least 50% for the first three outages following the EPU. In 2005, in RFO 25, the last refueling outage prior to the EPU, there were a total of 35 inspections performed, including 27 large bore inspections. Exhibit E4-

38. In 2007, in RFO 26, the first outage since the EPU, the in-spection scope was increased by more than 50%, as there were a total of 63 inspections performed, including 49 large bore inspections. Exhibit E4-10. While these additional inspec-tions are not needed to "calibrate" CHECWORKS, they will 24

provide additional, confirmatory data points for the use of the FAC Program.

Q42. Dr. Hopenfeld asserts that, in order to establish the rate of FAC, data are needed from either: (1) inspection of all risk-significant susceptible components "op-erat[ing] at [a] minimum [of] three inspection periods before a trend can be estab-lished," with "[f]ive inspection periods [being] the time interval between compo-nent inspection and the establishment of a corrosion rate for a given component at a given location" (NEC-JH_36 at 15); or (2) a "look at historic plant data in terms of the time scale for the occurrence of large, risk-significant wall thinning events," which Dr. Hopenfeld asserts is 16 years. Id. at 16. Do you agree with these assertions?

A42. (JSH, JCF) No. Neither approach is appropriate for determin-ing the rate of FAC through CHECWORKS. During the past 30 years, there has been a great deal of research performed to understand the features and the parametric dependencies of FAC. The approaches suggested by Dr. Hopenfeld, which es-sentially call for inspection of every potentially susceptible run of piping three times over five inspection periods, discard all analytical work done by the industry and substitute a brute-force unscientific approach. The combination of CHEC-WORKS and the EPRI guidance have eliminated the need for such an approach.

Dr. Hopenfeld's second approach reduces to the first approach and is equally unworkable. Dr. Hopenfeld appears to assert that in order to determine the rate of FAC, one has to deter-mine how long it takes particular components to fail and gather operating data for every component over that time pe-riod, which he claims is 16 years. However, such an approach is also equivalent to not having a FAC Program. Dr.

Hopenfeld does not explain how the second approach differs from his first approach other than by taking longer. Presuma-bly, inspections of essentially every potentially susceptible run 25

of piping would be required to assure that excessive FAC wear did not occur.

The operational experience cited by Dr. Hopenfeld does not indicate any problems in the proper use of CHECWORKS as part of a FAC Program nor does it support either of the ap-proaches he proposes. The plants referred to by Dr.

Hopenfeld where FAC-related events occurred had no FAC program before the accident (i.e., Surry, Trojan) (see IN 86-106, Exhibit E4-11 and IN 87-36, Exhibit E4-12), or their FAC program was not applied to the component that experi-enced a FAC-induced failure (i.e., San Onofre) (.see NEC-JH_46) or had a FAC program that did not follow the guid-ance in NSAC-202L (i.e., Clinton, Fort Calhoun and Mihama)

(see, e.g., NEC-JH_51 at 1-2; NEC-JH_53, Section 6). More-over, the logic of Dr. Hopenfeld's assertions regarding his be-lief that 16 years of operational data are needed is belied by units that have had failures with fewer than 16 years of opera-tion. Dr. Hopenfeld provides no basis for excluding these from consideration as historic plant data. For example, just considering some of the events referenced by Dr. Hopenfeld, the Millstone Unit 3 failure occurred after approximately 4 years of operation, the Sequoyah J-tube experience occurred after approximately 6 years of operation, and the feedring failure at San Onofre Unit 3 occurred after approximately 6 years of operation. Dr. Hopenfeld does not consider these his-toric plant data or explain why they should not be considered in evaluating the validity of his proposed approaches.

Q43. Mr. Witte states that "[s]eparate industry guidance supports five to ten years of data trending. Trending to the high end of the range is appropriate where vari-ables affecting wear rate, such as flow velocity, have significantly changed, as at 26

VYNPS following the 120% power up-rate.", NEC-UC_03 at 22. Do you agree with this assertion?

A43. (JCF) No. The statement Mr. Witte quotes is taken out of context. The document to which Mr. Witte refers, "Aging Management and Life Extension in the U.S. Nuclear Power Industry," published by the Chockie Group International, Inc.,

is not "industry guidance." It is a report produced at the be-hest of the Petroleum Safety Authority of Norway. NEC-UWI 3 at iii. The section from which Mr. Witte quotes in-volves a description of the EPRI Preventive Maintenance Ba-sis Program and describes the use of historical data to assess the performance and reliability of plant equipment, generi-cally, by reviewing 5 to 10 years of the most recent operating history, typically from the plant's work order database. This generic procedure applies to starting a condition assessment program where no aging management program has been in place previously. The reference is simply not applicable to an established program and has nothing to do with CHEC-WORKS. Nor would it apply to VY, which has had a formal aging management program for FAC in place since 1990. At the end of the current license, the program will have been in place for over 21 years, with at least 20 years of piping in-spection data having been accumulated.

Q44. Dr. Hopenfeld (NEC-JH_36 at 15) asserts that one of the reasons a 10-15 year period of data collection is needed to benchmark CHECWORKS for use at VY is that there was a reduction in the oxygen content of the plant, further increasing the potential for FAC. Dr. Hopenfeld cites in support of this statement page 3.2 of the summary report of the evaluations performed by SIA on environmentally assisted fatigue, provided by NEC as NEC exhibit NEC-JH_1 8 at 3.2. Do you agree with this assertion?

A44. (JCF) No. First of all, the cited report page does not state that the oxygen content of the plant has been reduced. It simply 27

states that the plant water chemistry was switched to Hydro-gen Water Chemistry ("HWC") in 2003. The change to HWC did not change, nor was it expected to change, the oxygen concentrations in the feedwater system, as demonstrated by measured plant data. Exhibit E4-18.

B. CHECWORKS MODELING Q45. NEC's consultants assert that the EPU of 20% represents a situation where the predictive efficacy of CHECWORKS will be diminished and that data from exist-ing plant experience cannot be used to predict the effect of post uprate conditions.

Dr. Hopenfeld (NEC-JH_36 at 15) ("... without specifying how each variable separately effects corrosion, does not address the issue of how the corrosion rate at a given location would be affected when the velocity changes by 20% at a given plant."); Mr. Witte (NEC-UW_03 at 22-23) ("... VYNPS is unique in its approach of Constant Pressure Power Up-rate to 120%"). Do you agree with these assertions?

A45. (JSH) No. Dr. Hopenfeld's and Mr. Witte's statements do not accurately reflect how CHECWORKS incorporates data from other nuclear units. As discussed above, the correlations built into CHECWORKS are based on laboratory experiments on modeled geometries, published correlations, and operating data from many nuclear units.

Mr. Witte's assertion that the use by CHECWORKS of in-spection data from other plants is not helpful because the VY conditions after the extended power uprate are different from those at other units denotes a lack of understanding of how CHECWORKS operates. As discussed above, the data used to develop the predictive algorithms in CHECWORKS en-compass the conditions at VY after the uprate. The algo-rithms used to predict the FAC wear rate are based on exten-sive laboratory and plant data, including data on FAC wear rates where the flow rate and the temperatures exceed those 28

present at VY after the uprate. This assures that the FAC wear rates predicted by CHECWORKS are accurate.

VY is certainly not unique in using CHECWORKS as part of an EPU. Exhibit E4-09 at ¶¶ 19-21.

Q46. Is Mr. Witte's assertion that "... 50% of those [plants] have experienced FAC re-lated problems" (NEC-UW_03 at 23) relevant to the prediction of FAC at VY?

A46. (JSH) No. Mr. Witte's citation does not support his state-ment. The operating experiences of the plants cited by Mr.

Witte are either inapplicable to VY or altogether irrelevant.

At Clinton (one of the plants cited by Mr. Witte), three sepa-rate instances of piping degradation are reported. The first in-stance refers to INPO operating experience reports OE 17412 and OEl 8478 (Exhibits E4-13 and E4-14). These reports re-fer to impingement degradation found in the feedwater heater vent lines. Impingement damage is a form of erosive attack, unrelated to FAC.

The second instance refers to INPO operating experience re-port OE20246 (Exhibit E4-15). This report refers to FAC wear found in extraction lines within the condenser. The re-port clearly states that the damaged lines were assumed to carry superheated steam (and thus were immune from FAC at-tack) and were mistakenly excluded from the CHECWORKS model. However, there was enough condensation in the con-denser to remove the superheat from these lines, resulting in wet steam which can and did cause FAC wear in these lines.

The incident, however, is not relevant to VY because (1) the similar lines at VY are composed of FAC resistant material; and (2) VY does not have a reheater, so there are no super-heated lines that might be mistakenly excluded from modeling 29

in CHECWORKS.

The third instance refers to INPO Operating Experience OE17654 (Exhibit E4-16). OE17654 deals with degradation downstream of orifices at the condenser in several systems.

The damage mechanism was found to be cavitation erosion or liquid droplet impingement erosion, not FAC.

At Dresden, the second plant on whose experience Mr. Witte relies, two instances are cited. The first was a loss of con-denser vacuum. The experience report, OE21421 (Exhibit E4-17), does not indicate that the cause of the leakage was deter-mined and certainly does not ascribe the cause to FAC. The other instance seems to concern erosion on the exterior sur-

• face of a vent line within the condenser. This erosion was not degradation due to FAC. In neither case is there any basis for linking the issues to FAC wear or to the implementation of the power uprate at Dresden.

Mr. Witte does not provide a reference for his assertion about Quad Cities. We have made inquiries and have not identified any FAC-related incidents at Quad Cities after its power uprate.

Q47. Dr. Hopenfeld asserts that CHECWORKS and its predecessor codes are not ac-ceptable for predicting FAC because FAC is non-linear and local, and "the re-quired correct inputs that account for local turbulence are not included in CHECWORKS." NEC-JH_36 at 6-7; see also id. at 4, 11, 12, 15. Is this state-ment correct?

A47. (JSH) No. Dr. Hopenfeld is in error in claiming that FAC is a non-linear phenomenon. Unlike erosion mechanisms, FAC causes damage in a manner that is linear with time (i.e., there is a constant corrosion rate). This has been demonstrated in 30

numerous laboratory tests and by the fact that field measure-ments match predictions using a linear model. For that rea-son, laboratory tests designed to measure the impact of, for example, water chemistry on FAC rates are often run for a pe-riod ofjust hours - enough time to establish a trend. In fact, when operating conditions are intentionally changed, the rate of FAC responds almost immediately to the new conditions.

See, e.g., Exhibit E4-19; Exhibit E4-08 at 7-6 and Figures 3-6 and 3-7.

With respect to the allegedly local nature of FAC wear, al-

• though local FAC wear is occasionally seen - normally near a geometric discontinuity - such local wear usually results in only minor effects (e.g*, leaks). The normal feature of FAC wear - widespread wear over an extended area - is what causes significant problems (L.& the need for pipe replace-ments or the occurrence of pipe ruptures).

This distinction can be seen by comparing the catastrophic FAC-induced failures, such as the one at Surry, with other in-stances, such as those described at Calvert Cliffs in OE 15860 (Exhibit E4-20) and OE 20127 (Exhibit E4-21) where local-ized FAC caused leaks, but not pipe ruptures. By contrast, at Surry, there was not "localized" wall thinning (as Dr.

Hopenfeld mischaracterizes the event). NEC-JH_36 at 2.

Rather, there was widespread loss of material typical of attack by FAC (Exhibit E4-08 at 1-5). The global nature (i.e., wide-spread effect) of the FAC damage is consistent with the ex-perience of FAC-induced ruptures. The photographs of fail-ures at Surry, Fort Calhoun, and at a Czech nuclear unit, for example, clearly show the large area of thinned material. (Ex-31

hibit E4-08 at Figures 1-3, 4-31 and 4-41).

Dr. Hopenfeld also states that "one must know the exact spot on a given component where conditions are most favorable to FAC." NEC-JH_01 at 12. This statement is clearly incorrect.

The NSAC-202L guidance, which is also the practice at VY, calls for the inspection of the entirety of each component and the attached piping or the piping section selected for inspec-tion. The inspection is conducted in sufficient detail to iden-tify any FAC-caused degradation anywhere in the entire loca-tion being inspected. In virtually all cases, the degradation caused by FAC occurs over a fairly wide area (comparable to the size of the fitting). Therefore, pinpointing the "exact" lo-cation in a component or piping section where FAC will occur is infeasible and unnecessary.

Q48. Do you agree with Dr. Hopenfeld's claim that "it is the local flow velocity that directly controls the local turbulence and not the average velocity" and that CHECWORKS is flawed because it is "based on average flow velocities"? NEC-JH_36 at 7.

A48. (JSH) No. Dr. Hopenfeld is incorrect in claiming that the use of average flow velocities is flawed. Engineering calculations are often premised on the use of average flow velocities. Both Dr. Hopenfeld and Dr. Hausler refer to such an example in their reports (NEC-JH_36 at 3, NEC-RH_01 at 5) - calculat-ing pressure drop in turbulent flow.

When calculating the pressure drop across a component, such as an elbow, an engineer would normally use well-established values to relate the pressure drop in the elbow to the pressure drop in a straight pipe. An engineer would not usually resort to computational fluid dynamics, as Dr. Hopenfeld suggests.

32

To do so would require the engineer to calculate the flow field relating the average inlet velocity to the local velocity, and then integrate the local shear stress field to obtain the pressure drop.

Similar to the engineer using well-established values of pres-sure drop and flow velocity in a straight pipe and comparing it to the pressure drop in the elbow, CHECWORKS uses "geo-metric factors" to relate the maximum degradation occurring in a component, such as an elbow, to the degradation pre-dicted to occur in a straight pipe. This approach was devel-oped by Keller in the 1970s and is similar to the approach taken in other FAC computer programs. The use of geometric factors in the three most prominent FAC programs are pre-sented in Exhibit E4-08 in Table 3-1 at 3-11, Table 7-1 at 7-3, and at 7 7-8.

Q49. Dr. Hopenfeld claims that "the mass transfer coefficient varies with the 0 .8 th power of the velocity for straight pipes and the square of the velocity for curved pipes," NEC-JH_36 at 4. Do you agree with this claim?

A49. (JSH) No. The portion of this statement concerning elbows (and by extension other geometries) is incorrect and is contra-dicted by a large body of mass transfer data and correlations published in the technical literature for the geometries encoun-tered in piping systems. The portion of this statement con-cerning elbows (and by extension geometries other than straight pipes) is incorrect. For all known geometries includ-ing straight pipes, bends, and flow restrictions, the depend-ence of mass transfer coefficient on velocity is less than unity.

For example, data from Tagg, et al. (Exhibit E4-22 at Figure

8) and Poulson and Robinson (Exhibit E4-23 at Figures 5, 6, 8 and 9) shows the exponent on the Reynolds number (i.e., a 33

dimensionless number directly proportional to the velocity) is between 0.5 and unity. This results in a dependence of FAC rate on velocity that is slightly less than linear (i.e., doubling the velocity will not quite double the rate of FAC).

The mass transfer correlations built into CHECWORKS are based on laboratory experiments on modeled geometries, pub-lished correlations and plant data from many nuclear units, all of which have shown a less than linear relationship exists be-tween velocity and the rate of FAC wear, including velocities higher than those present at VY after the uprate. The FAC wear rates vary roughly with velocity and do not increase with velocity in the non-linear manner claimed by Dr. Hopenfeld.

Q50. What is your response to Dr. Hopenfeld's observation that "CHECWORKS is not a mechanistic model.. ." and that "the correlation of CHECWORKS was per-formed in an unscientific manner"? NEC-JH_36 at 12.

A50. (JSH) No existing FAC model is mechanistic. A "mechanistic model" would be based directly on the physical processes in-volved. The model used in CHECWORKS and other FAC programs takes a broader approach that does not deal with the microscopic processes involved. Instead, it relates physical and chemical parameters with the entirety of the corrosion process. However, the fact that CHECWORKS is not mecha-nistic does not mean that it is not an effective predictive pro-gram. The predictive algorithms in CHECWORKS were de-veloped using all available plant and laboratory data and, therefore approximate, without directly reproducing, the mechanistic details of the corrosion process, such as the rate of mass transport within the porous oxide. The successful use of CHECWORKS and its predecessor programs for more than 34

20 years provides additional support for the claim that CHECWORKS is an effective tool for inspection planning.

Q51. Does the non-mechanistic nature of the CHECWORKS model affect its ability to be used as part of the VY FAC Program?

A51. (JSH) No. As discussed earlier, CHECWORKS was specifi-cally designed to predict the effects on plants when their oper-ating conditions change. The only relevant parameters that change as a result of the power uprate at VY are flow rate and temperature, both of which are accounted for within the CHECWORKS model.

Q52. Dr. Hopenfeld (NEC-JH_36 at 9-11) and Mr. Witte (NEC-UW_03 at 9-10) refer to several reactors and fossil units where FAC has allegedly not been detected in components, in some instances leading to pipe ruptures. Would you please dis-cuss the operating experience cited and its relevance, if any, to the FAC Program at VY?

A52. (JSH) Generally, Dr. Hopenfeld characterizes these events as:

(1) being a failure of CHECWORKS or predecessor pro-grams; or (2) evidencing a failure of the EPRI guidelines. In the cited instances, the events Dr. Hopenfeld cites do not in-volve one or either of these.

San Onofre. "May 1990 - Erosion and corrosion was dis-covered in the feed distribution piping of units 2 and 3 at San Onofre - IN 91-019" and "June 1993 - Through wall FAC of two J-tubes in Unit 2 Steam Generator at San Onofre. IN 93-06." NEC-JH_36 at 9-10. The failed components - feed distribution piping and J-tubes - were within the pressure boundary of the steam generator, and as such are not nor-mally, nor were they in this case, modeled using CHEC-WORKS (or its predecessor programs). The experience at San Onofre was one of the first reported instances of FAC 35

degradation in the feed ring of a steam generator. With re-spect to the feed distribution piping, NEC's own exhibit, NEC-JH_46, shows that the NRC reported that "[t]he licensee determined that the root cause contributing to the degradation of the feedwater distribution system piping to be inadequate design of the feedring and feedring supports." Id. at 2. More-over, the EPRI guidelines (NSAC-202L) were not issued until November 1993, after the two events occurred.

Fort Calhoun. "April 1997- 6 square foot rupture of a 12-inch elbow at Ft. Calhoun. IN 97-84." NEC-JH_36 at 9-10.

While CHECWORKS was being used at the station, a post-accident investigation by the NRC indicated that the utility failed to use CHECWORKS properly. NEC-JH_51 at 2. The NRC found that, with respect to this event, the FAC engineer should have been aware from the CHECWORKS output that the area should have been investigated, as field measurement results did not match the program's predictions due to model-ing errors. Id. at 2.

Mihama. "August 9, 2004 - A secondary pipe ruptured, 5 workers were killed and 6 more were injured at the MIHAMA plant in Japan." NEC-JH_36 at 10. As Dr. Hopenfeld con-cedes, the operators of Mihama did not use CHECWORKS or other predictive programs; nor did they use an inspection pro-gram consistent with NSAC-202L. Rather, they used a Japa-nese procedure that relied on inspections and trending alone (Exhibit E4-24). There are considerable differences between the Japanese approach prior to Mihama and the U.S. approach, including not using predictive tools like CHECWORKS and not having an industry-wide inspection standard.

36

Millstone 3. "December 1990 - Two six inch pipes were damaged .as a result of wall thinning at Millstone 3. The first pipe completely sheared off while the second was sheared by 1/2 to "4". The pipe eroded from its original thickness of 0.28 inches to 0.11 inches." NEC-JH_36 at 9. This piping was not analyzed by CHEC or CHECMATE, as NEC-JH_47 makes clear: "Although the licensee had identified the MSD system as one of the systems to be analyzed for erosion/corrosion susceptibility, that analysis was not performed because of a communication error." Id. at 2. Moreover, the EPRI single-phase guidance was not issued until November 1993, after the event occurred.

Susquehanna Unit 1. "May 1992 - Unexpected high ero-sion rates in the feedwater piping at Susquehanna Unit 1 (BWR) in a section of piping that could not be isolated from the reactor vessel. IN 92-35." NEC-JH_36 at 10. This area was not analyzed with CHEC or CHECMATE. Likewise, the EPRI single-phase inspection guidelines were not in place.

North Anna. "Wear in the feedwater nozzle at North Anna in the safety-related area of the plant." NEC-JH_36 at 10. Dr.

Hopenfeld provides no citation for this event.

Sequoyah. "November 1994 - 180-degree crack in a 14" condensate piping at Sequoyah. IN 95-11." NEC-JH_36 at 10.

Although the line was analyzed with CHECWORKS, there was an error in modeling the component that resulted in inac-curate results. NEC-JH_50 at 2.

37

Callaway. "August 1999 - Double ended pipe break in a moisture separator at Callaway. IN 36015." NEC-JH_36 at

10. The event report shows that the event actually occurred in a reheater drain line, not a moisture separator drain line.

NEC-JH_45. Although it was analyzed with CHECWORKS and the area in question was inspected, it was not inspected in a manner consistent with NSAC-202L.

Kewaunee. "April 2004 - A work order to inspect the el-bow for wall-thinning at Kewaunee was cancelled after wall thickness in a nearby elbow was evaluated by the licensee and deemed acceptable. The extrapolation of inspection results from one elbow to the other elbow was inappropriate." NEC-JH_36 at 10. The line in question is not FAC-susceptible '(raw water system). It was not analyzed with CHECWORKS and is not covered by NSAC-202L.

In short, none of Dr. Hopenfeld's examples involve a situation in which proper use of CHECWORKS or its predecessor pro-grams was ineffective in preventing a FAC failure.

Q53. Is there any merit to Mr. Witte's (NEC-UW_03 at 8-9) argument that, because failures have occurred at Surry, Pleasant Prairie [fossil] and Mihama after less than ten years of operation, an extended period of "baselining" would be re-quired?

A53. (JSH) No. Even though accidents did occur after about 10 years of operation at each of these plants, there were no com-puter programs in use at any of the plants to predict the risk of FAC or to protect piping and components against FAC. There were no inspections of the potentially affected areas or any sort of FAC program at Surry, San Onofre and Pleasant Prai-rie; nor was there a program in place that would have identi-fied the potential failure location before it occurred. Thus, the 38

operating time before failure had nothing to do with the effec-tive use or accuracy of CHECWORKS. At Mihama, there was a program in place that did not use a predictive method-ology, and the accident was caused by a programmatic, not a technical, failure.

Thus, the experiences at the plants cited by NEC do not sup-port NEC's claim that CHECWORKS is unsuitable for its in-tended purpose, or that an extended period is necessary to calibrate the program after the EPU.

C. ADEQUACY OF VY'S FAC PROGRAM

1. Programmatic Issues Q54. What is the issue in controversy with respect to the adequacy of VY's FAC Pro-gram?

A54. (JSH, JCF) Dr. Hopenfeld and Mr. Witte raise concerns about the implementation of the FAC Program over the last several years.

Dr. Hopenfeld questions the implementation of the FAC Pro-gram as it relies on engineering judgment as one of the factors in selecting the locations where inspections will be made dur-ing refueling outages. Dr. Hopenfeld questions the impor-tance VY has given to higher flow velocities during EPU op-eration in selecting the component locations, stating that it is the turbulence of the flow that determines FAC susceptibility.

NEC-JH_36 at 12. He also challenges the selection of the highest length of piping as candidates for performing inspec-tions. NEC-JH_36 at 11.

Dr. Hopenfeld further states that "the selection of the correct grid size for UT measurements is one of the most critical in-39

spection tasks" and criticizes the CHECWORKS guidelines for selection of grid size as they are applied at VY. NEC-JH_36 at 7, 14-16.

Q55. Do you agree with Dr. Hopenfeld's assertion that the VY FAC Program at VY is deficient in these respects?

A55. (JSH, JCF) No.

Q56. What is the basis for your disagreement with Dr. Hopenfeld?

A56. (JSH, JCF) There are four areas of criticism by Dr. Hopenfeld, all invalid: use of engineering judgment; velocity versus tur-bulence; use of pipe length as a basis for selection; and selec-tion of grid size.

The use of engineering judgment as one way of selecting in-spection locations is specifically recommended by NSAC-202L. Dr. Hopenfeld states that, "[t]here is no indication that components to be included in the FAC program are selected on the basis of [risk significance and component susceptibility to failure]; instead, component selection is left to the judgment of plant operators." NEC-JH_36 at 8. Entergy's FAC pro-gram, however, does take risk significance and component susceptibility to failure into account. Exhibit E4-06, Sections 5.2 and 5.3.

Dr. Hopenfeld's argument that there should be a greater than linear dependence on velocity (NEC-JH_36 at 4) is unsup-ported and, as discussed earlier, is contradicted by an exten-sive body of laboratory results (see Exhibits E4-22 at Figure 8 and E4-23 at Figures 5, 6, 8 and 9).

40

Dr. Hopenfeld (NEC-JH_36 at 11) asserts that Entergy be-lieves "that length and the highest velocities control corro-sion." Dr. Hopenfeld bases this assertion on a quote taken from the transcript of the November 30, 2005 Advisory Committee on Reactor Safeguards ("ACRS") meeting. The quoted language contains a transcription error. The discussion around the quotation concerns the inspections to be performed during three upcoming refueling outages. In the discussion, Entergy was stating that, given the low wear rates that had been measured, Entergy would be inspecting the locations that had the highest length of time since the last inspection and the locations with the highest velocities. Dr. Hopenfeld's inter-pretation of the text as evidencing that Entergy uses the long-est length of pipe as a criterion for inspections is incorrect.

Dr. Hopenfeld's assertions criticizing CHECWORKS guide-lines for the selection of grid size (NEC-JH_36 at 7, 14-16) are just wrong. The grid size is normally specified in the util-ity's procedure governing inspections and is not related to CHECWORKS. Usually, this recommendation will follow the guidelines of NSAC-202L. In the case of VY, the grid size is specified by an Engineering Standard, "Flow Acceler-ated Corrosion Component Scanning and Gridding Standard" (Exhibit E4-25). Historically, grid size is related to the physi-cal size of the component being inspected for degradation.

There are two aspects to grid size: (1) when degradation is found, the grid size is normally made smaller in that area to more accurately define the wear area; and (2) in inspecting a component, the larger the pipe, the larger amount of material that may be lost before the component fails, allowing for a "larger" grid (i.e., the defect size that would cause failure var-ies directly with the size of the pipe). Both of these ap-41

proaches are consistent with NSAC-202L, Rev. 2, Section 4.5.3.

At VY, an additional step is taken in performing the inspec-tions. Rather than recording the thickness reading at particu-lar grid points, the components inspected at VY are scanned in their entirety. This is done by moving an ultrasonic trans-ducer over the entire surface within a grid "square." The data logger automatically records the minimum reading anywhere within the grid square and the qualified inspector verifies that reading. This ensures that the thinnest readings in the compo-nent are found.

Q57. Dr. Hausler (NEC-RH_03 at 9) states: "It would be erroneous for the utility to continue to rely on grids established prior to EPU since these grids may not spe-cifically capture the FAC phenomena observed at lesser velocities." Is his state-ment correct?

A57. (JSH) No. This statement is not correct. As stated previously, the FAC Program at VY provides for the inspection of the en-tirety of each component and the attached piping or the piping section selected for inspection. Therefore, the size of the grid is the same both before and after the uprate. The inspection is conducted in enough detail to identify any FAC-caused degra-dation anywhere in the entire location being inspected. In virtually all cases, the degradation caused by FAC occurs over a fairly wide area (comparable to the size of the fitting).

2. Alleged Program Implementation Deficiencies Q58. What historical deficiencies does Mr. Witte allege in the VY FAC Inspection Program?

A58. (JCF) Mr. Witte asserts that data from previous FAC inspec-tions (prior to the EPU) were not entered into the CHEC-WORKS database (NEC-UW_03 at 2, 3, 6, 7-8, 15, 16, 17);

42

that CHECWORKS was not updated with the uprate parame-ters (id. at 5, 23); that, for the period 2000-2006 VY failed to use a current version of CHECWORKS (id. at 6, 17); that four components were predicted in 2004 to have wall thinning be-yond operability limits (id. at 17-18, 22); that open corrective actions identified in condition reports may not have been com-pleted (id. at 3-4, 18-19); that ranking of small bore piping was not done (id. at 8, 20); that the number of inspection points were reduced after the 2005 outage (id. at 7, 8, 20); and that the 2006 refueling outage inspection "scope, planning, documentation, and procedural analysis appear to have been performed under a superseded program document" (id. at 5, 7, 20-21).

Q59. Do you agree with these criticisms?

A59. (JCF) No. Each of these criticisms either demonstrates a mis-understanding of the FAC Program on the part of Mr. Witte or is erroneous.

Q60. Mr. Witte states (NEC-UW 03 at 15) that Entergy was "aware of the problematic state of the program ,for many years" and describes (NEC-UW_03 at 2, 18) the FAC Program as "unsatisfactory." Do you agree with these statements?

A60. (JCF) No. Mr. Witte is making an assertion regarding the FAC Program being "problematic" that is not supported by the document he cites. Mr. Witte cites NEC-UW_09 as the basis for Entergy being "aware of the problematic state of the program" (NEC-UW_03 at 15), but that document is a quality assurance audit, No. QA-8-2004-VY-1 (Exhibit E4-26), which categorically states that "[n]one of the findings or areas for improvement, individually or in the aggregate, were indicative of significant programmatic weaknesses which would impact the overall effectiveness of the Engineering Programs as-43

sessed." Id. at 2. Nor does Mr. Witte try to explain what sig-nificance, if any, this quality assurance audit has with respect to the adequacy of the FAC Program.

Q61. Mr. Witte (NEC-UW_03 at 15) states that VY was, "...notified by EPRI as early as 2000 that it had not been fully updating the CHECWORKS model in use at VYNPS with plant inspection data collected or plant modifications performed during previous inspections. Entergy apparently ignored the warning." Is Mr.

Witte's statement factually accurate?

A61. (JCF) No. The statement is inaccurate. The EPRI evaluation report (Exhibit E4-27) was received in February 2000. The CHECWORKS models were updated with all applicable in-spection data during the Summer and Fall of 2000. Exhibit E4-28. Additional updates were performed for the feedwater system in 2003. Exhibit E4-29. Another CHECWORKS up-date was performed in 2006. Exhibit E4-30.

Q62. Mr. Witte makes several other statements to the effect that CHECWORKS was not properly updated with data from previous FAC inspections (prior to the EPU)

(NEC-UW_03 at 2, 3, 6, 7-8, 15, 16, 17). Do you agree with these statements?

A62. (JCF) No. Each of these statements is essentially a repetition of the assertion addressed in my previous answer.

Mr. Witte makes other inaccurate statements regarding VY's input of data into CHECWORKS. He asserts that the "model was not kept current during a seven-year period and suggests that susceptible locations may not have been inspected during this time period." NEC-UW_03 at 16. In fact, all applicable inspection data were updated during the Summer and Fall of 2000. Additional updates were performed for the feedwater system in 2003. In addition, inspections performed in 2001, 2002, 2004, and in 2005 showed that the wear rates predicted by the CHECWORKS model were consistently conservative; 44

thus the alleged failure to "update" the CHECWORKS model would not have resulted in under-prediction of the FAC risks because the inspection planning and component selections made during those outages were based in part on the conserva-tively high wear rates predicted by CHECWORKS. The CHECWORKS update performed in 2006 confirmed again that the previously predicted wear rates were conservative.

Exhibits NEC-UW_10 and E4-3 1.

Mr. Witte's conclusion that the purported failure to update CHECWORKS "suggests that susceptible locations may not have been inspected during this period" is only unfounded speculation. In fact, all susceptible piping was identified inde-pendently from the CHECWORKS results. Exhibit E4-32.

In implementing the FAC Program at VY, the original CHECMATE and later CHECWORKS Pass 1 Analyses (per-formed with the initial use of CHECMATE and CHEC-WORKS) calculated FAC wear rates and predicted time to minimum wall thickness based on plant-specific variables.

Under the FAC Program, inspections were then performed for the components with the highest wear rates and lowest time to minimum wall thickness. As inspection data were obtained and incorporated into the models, Pass 2 Analyses were per-formed and the predicted wear rates were correlated to the measured data. In all cases, the inclusion of the inspection data reduced the predicted wear rates and increased the times to minimum wall thickness. Thus, not entering data from a particular inspection into CHECWORKS would not suggest that "susceptible locations may not have been inspected," as Mr. Witte asserts.

45

The data and inspection reports clearly indicate that Mr.

Witte's statement has no basis. Exhibits E4-35 through E4-

38. The CHECWORKS update performed in 2006 confirmed that the previously predicted wear rates (before the 2006 up-date) were conservative (see Exhibits NEC-UW_10 and E4-

31) and the results of the updated model did not identify any instance where recommended inspections were not performed.

Exhibit E4-3 1.

Contrary to Mr. Witte's assertion, there was no lapse related to the FAC Program. The susceptibility analysis was updated in October 2005 to include changes in the relevant input variables - flow rate and temperature - associated with the power uprate. Exhibit E4-32.

Q63. Mr. Witte states (NEC-UW_03 at 19) that "the VY FAC program was primafa-cie in noncompliance with its CLB" because "in 2005 a sixth CR was written, CR-VTY-2005-02239, stating 'CHECWORKS predictive model for Piping FAC inspection program was not updated per appendix D of PP 7028"'. Is he right?

A63. (JCF) No. Comparison of the CHECWORKS predictions with subsequent inspection data showed that the CHEC-WORKS predictions were conservative (i.e., predicted higher wear rates than observed during the inspection). For that rea-son, even if the most recent inspection data had not been en-tered into the CHECWORKS program, the result would have been over-estimation of FAC wear. The CR cited by Mr.

Witte correctly concludes, therefore, that not updating the CHECWORKS database with the most recent inspection data was not necessary in order to determine the appropriate scope of the RFO 25 inspection.

46

There is no basis for Mr. Witte's assertion that not entering the most recent inspection data into CHECWORKS rendered the FAC Program in noncompliance with VY's current licens-ing basis ("CLB"). VY's CLB incorporated the recommenda-tions in EPRI NSAC-202L, Rev.2 (Exhibit E4-33) by refer-ence in FAC Program Procedure(s) PP7028 (Exhibit E4-34) and, later, ENN-DC-315, Rev. 1 (NEC-UWI 2), which were in effect during the period from 1999 to 2006. Section 4 of NSAC-202L, Rev. 2, does not specify a specific interval for model updates. It merely states: "It is recommended that whenever possible, the Predictive Plant Model utilize the re-sults of wall thickness inspections to enhance the FAC predic-tions. In CHECWORKS this is called Pass 2 analysis."

Because there is no specific interval required for entering ad-ditional inspection data into CHECWORKS, no departure from the CLB took place.

Q64. Mr. Witte also states (NEC-UW_03 at 19): "From 1999-2006, the plant was es-sentially operating in a state in which component wear was improperly trended and pipe conditions were actually unknown. Reliance on CHECWORKS for this time period for predicting grid points, ranking susceptible components, and in-specting new points was therefore virtually without technical or empirical value."

Is there any validity to his conclusions?

A64. (JCF) No. Mr. Witte's assertions assume (1) that CHEC-WORKS was the only tool used for choosing inspection points, and (2) that trending of component wear is a necessary part of CHECWORKS use. Neither assumption is true. Dur-ing the time period that Mr. Witte references, inspections were conducted, data were evaluated, and component wear rate was trended, all in accordance with the FAC Program. The results of these inspections, data evaluations and trending have been provided to NEC (Exhibits E4-35 through E4-38). Signifi-47

cantly, Mr. Witte does not point to any deficiency in those re-cords.

Q65. Mr. Witte states (NEC-UW_03 at 17): "During the years 2000-2006, the VYNPS FAC program apparently used an outdated version of the CHECWORKS soft-ware. As far back as 2000, EPRI recommended that VYNPS update to the current version of the software, but the recommendation was not implemented until 2006." Is this a correct statement?

A65. (JCF) No. As noted earlier, VY updated the version of CHECWORKS it used from CHECWORKS FAC 1.OD to CHECWORKS FAC 1.OF in 2000. Exhibit E4-28. Version 1.OF was used for the 2003 and 2006 model updates.

CHECWORKS FAC 1.OG was installed in 2006.

More fundamentally, Mr. Witte does not state what conse-quences would follow from the failure to use the most current version of CHECWORKS. In fact, there were no differences in versions 1.OD, 1.OF, and 1.OG with respect to water chemis-try and wear rate predictions for BWRs (see Exhibit E4-39).

Nothing regarding the version in use at any particular point in time had any effect on the use of CHECWORKS as a tool as part of the FAC Program. Nor would it have any effect on the implementation of the FAC Program during the license re-newal period.

Q66. Mr. Witte states (NEC-UW_03 at 17) that "[i]n 2004, at least four VYNPS com-ponents, including the condensate system and the extraction steam systems, were determined to have 'negative time to Tmin,' meaning that wall thinning was being predicted as beyond operability limits and should be considered unsafe with po-tential rupture at anytime." What is the significance of the statement quoted by Mr. Witte?

A66. (JCF) Mr. Witte's interpretation of this assertion demonstrates his misunderstanding of how CHECWORKS is used at VY.

The document to which Mr. Witte refers for support for these 48

statements is the VY Scoping Worksheets (Exhibit E4-40 at 5) developed in preparation for the 2004 refueling outage. These statements refer to the predicted wear based on the CHEC-WORKS wear rates which, as discussed above, had previously been determined to be conservative.

Mr. Witte fails to clarify that the "determination" that the four components had "negative times to Tmin" is a theoretical conclusion based on the results of CHECWORKS, and is not based on actual inspection data. As such, there would be no need to "write condition reports for this condition," (NEC-UW_03 at 18) as Mr. Witte states. CRs are written when in-spection data indicate there is an actual problem and addi-tional inspections are then performed as corrective actions. If a planning tool, like CHECWORKS, indicates an area of po-tential concern, inspections of that area are scheduled.

The only FAC susceptible component identified in the 2004 Scoping Worksheets (CD30TE02DS) was scheduled for in-spection. The actual inspection data show that the entire com-ponent meets design code with significant margin. See Ex-hibit E4-37 at 12.

Q67. Mr. Witte states (NEC-UW_03 at 19): "The 2006 cornerstone report shows a number of indicators as yellow, with lists of open CR corrective actions, and a new CR written in August 30, 2006. The report lists six corrective actions and four CRs that were written as early as 2003 that remain open." Is this an accurate characterization?

A67. (JCF) The statement is inaccurate. The FAC Program Health Report, "Cornerstone Rollup" shows the overall FAC Program status as Green. Exhibit NEC-UW_07 at 1. The report rates twenty-seven different areas. Of these, two were rated as "Yellow": (1) Owner Availability and (2) Open Actions 49

Items. Id. at 4, 6. A yellow indicator for Open Action Items is triggered if any action item, regardless of its importance, is more than one year old. Six LO-VTYLO action items are listed. These are not Condition Reports, nor are they correc-tive actions from condition reports. They are commitments.

The Corrective Action Program is used to track all commit-ments. There is no safety significance to these commitments.

The items listed are for completion of program administrative tasks.

Q68. Mr. Witte states (NEC-UW_03 at 20): "Ranking of small bore piping was not done. With no ranking, the basis for selection of high susceptible points for small bore piping is not evident." Is this statement accurate?

A68. (JCF) This statement is not accurate. At VY, the initial scop-ing and inspection selection of small bore piping was per-formed in 1993 and 1995. The scope and criteria for deter-mining the inspection locations is documented in FAC Pro-gram documents (Exhibits E4-41 and E4-42). The small bore inspections were initiated prior to the inclusion of small bore guidance provided in NSAC-202L.

Q69. Is Mr. Witte (NEC-UW_03 at 20) correct when he states: "A flow-accelerated corrosion related pipe break associated with a 1" elbow, SSH (WO 06-6880), ap-pears to have occurred in 3rd quarter 2006"?

A69. (JCF) No. A pinhole leak was identified during operator rounds on an elbow on the 1" drain line from the steam seal header to the condenser. No "pipe break" had occurred. The elbow was replaced in RFO 26. The damage found was due to droplet impingement, not FAC.

Q70. Mr. Witte states (NEC-UW_03 at 20): "Entergy apparently reduced the number of FAC inspection data points between the 2005 and 2006 refueling outage, in viola-tion of its commitment to increase inspection data point by 50% The 2005 refuel-ing outage inspection called for 137 large-bore inspection points. The 2006 refuel-50

ing outage inspection presented to the ACRS on June 5, 2007, covered only 63 points." Do you agree with Mr. Witte's statement?

A70. (JCF) No, the statement is inaccurate. First, there was no refu-eling outage in 2006. Second, in RFO 25 in the Fall of 2005, a total of 35 inspections were performed. Of those, 27 were large bore UT inspections. Exhibit E4-38. In RFO 26 (con-ducted in the Spring of 2007), a total of 63 inspections were performed. Of those, 41 were large bore UT inspections. Ex-hibit E4-38. The increase in the number large bore UT in-spections from RFO 25 to RFO 26 was more than 50%, in ac-cordance with Entergy's plans.

Q71. Mr. Witte states (NEC-UW_03 at 20): "The 2006 Refueling outage FAC Inspec-tion scope, planning, documentation, and procedural analysis all appear to have been performed under a superseded program document. ENN-DC-315 Rev. 1 was effective March 15, 2006." Is Mr. Witte correct?

A71. (JCF) No, the statement is inaccurate. There was no 2006 re-fueling outage. Additionally, the guidance used to perform the scoping had not been superseded. The scoping process for RFO 26 started before RFO 25 was complete and well before the March 15, 2006 effective date of ENN-DC-315, Rev. 1 (NEC-UW_12). The RFO 26 scoping was performed using the same criteria as contained in Section 5.3 of ENN-DC-315, Rev. 1. The scoping criteria in ENN-DC-315, Rev. 1, is the same as under the superseded VY procedure PP 7028 (Exhibit E4-34, Appendix E, Section E.2).

The most recent FAC inspections, performed in RFO 26, were performed under the appropriate versions of the Entergy pro-cedures EN-DC-315, Rev.0 and ENN-NDE-9.05.

51

V.

SUMMARY

AND CONCLUSIONS Q72. Please summarize your testimony.

A72. (JSH, JCF) Our testimony can be summarized as follows:

Entergy has had an effective FAC Program in place at VY for over twenty years. The program has detected wear as designed and components have been replaced prior to thin-ning below minimum design thickness.

VY is a relatively small and simple plant that has fewer FAC-susceptible piping components than a typical plant of comparable size. The piping and components in many of the two phase flow systems were either originally con-structed of FAC-resistant materials or have been replaced with FAC-resistant piping and components since their ini-tial installation.

The FAC Program at VY uses CHECWORKS as a tool in planning inspections, evaluating inspection data, and man-aging the UT data collected. While an effective tool, it is only one of several used in the VY FAC Program to iden-tify the locations to be inspected during refueling outages of the plant.

• No nuclear power plant properly using CHECWORKS has reported a large bore piping failure.

• CHECWORKS has a well-established track record of use in FAC management programs, including BWRs which have undergone uprates. The input of new values for the plant-specific variables affected by the uprate at VY - flow rate and temperature - is all that is required for CHEC-WORKS to continue to be used as part of the FAC Pro-gram.

• The current FAC Program, which will be the FAC Program used during the license renewal period, meets industry practice as reflected in NSAC-202L and has been reviewed, audited and inspected with only minor, mostly administra-tive, issues identified.

• The FAC Program that will be used during the period of extended operation after license renewal will assure that the aging effects of FAC will be adequately managed.

52

Q73. What are your conclusions regarding the assertions in NEC Contention 4?

A73. (JSH, JCF) We conclude that there is no factual support for the claims made in NEC Contention 4.

Q74. Does that conclude your testimony?

A74. (JSH, JCF) Yes, it does.

53

ORIGINAL UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )

)

Entergy Nuclear Vermont Yankee, LLC ) Docket No. 50-271-LR and Entergy Nuclear Operations, Inc. ) ASLBP No. 06-849-03-LR

)

(Vermont Yankee Nuclear Power Station) )

CERTIFICATE OF SERVICE I hereby certify that copies of "Entergy's Statement of Position on New England Coalition Contentions, .... Joint Declaration of James C. Fitzpatrick and Gary L. Stevens on NEC Contention 2 (Environmentally-Assisted Fatigue)," "Joint Declaration of John R. Hoffman and Larry D. Lukens on NEC Contention 3 (Steam Dryer)," "Joint Declaration of Jeffrey S. Horowitz and James C. Fitzpatrick on NEC Contention 4 (Flow Accelerated Corrosion)," and Exhibits E2-02 through E2-37, E3-02 through E3-16, and E4-02 through E4-42 were served on the persons listed below by deposit in the U.S. Mail, first class, postage prepaid, and where indicated by an asterisk by electronic mail, this 1 3 th day of May, 2008.

• Administrative Judge *Administrative Judge Alex S. Karlin, Esq., Chairman Dr. Richard E. Wardwell Atomic Safety and Licensing Board Atomic Safety and Licensing Board Mail Stop T-3 F23 Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Washington, D.C. 20555-0001 ask2(@nrc.gov rew Anrc. gov

• Administrative Judge *Secretary William H. Reed Att'n: Rulemakings and Adjudications Staff 1819 Edgewood Lane Mail Stop 0-16 C1 Charlottesville, VA 22902 U.S. Nuclear Regulatory Commission whrcville(iembarqmail.com Washington, D.C. 20555-0001 secy@nrc.gov, hearingdocket(anrc. gov

• Office of Commission Appellate Adjudication Atomic Safety and Licensing Board Mail Stop 0-16 C I Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Washington, D.C. 20555-0001 OCAAmail(dnrc.gov

• Lloyd Subin, Esq. *Sarah Hofmnann, Esq.

• Mary Baty, Esq. Director of Public Advocacy Office of the General Counsel Department of Public Service Mail Stop 0- 15 D21 112 State Street - Drawer 20 U.S. Nuclear Regulatory Commission Montpelier, VT 05620-2601 Washington, D.C. 20555-0001 Sarah.hofmann(distate.vt.us LBS3(@nrc.gov; mcb 1@nrc.gov

• Anthony Z. Roisman, Esq. *Ronald A. Shems, Esq.

National Legal Scholars Law Firm *Karen Tyler, Esq.

84 East Thetford Road Shems, Dunkiel, Kassel & Saunders, PLLC Lyme, NH 03768 9 College Street aroismangnationallegalscholars.comn Burlington, VT 05401 rshems(asdkslaw.com ktyler@sdkslaw.com

• Peter L. Roth, Esq. *Marcia Carpenter, Esq.

Office of the New Hampshire Attorney General Atomic Safety and Licensing Board Panel 33 Capitol Street Mail Stop T-3 F23 Concord, NH 03301 U.S. Nuclear Regulatory Commission Peter.roth@doj.nh.gov Washington, D.C. 20555-0001 mxc7(@i)nrc.gov

• Lauren Bregman, Law Clerk *Diane Curran, Esq.

Atomic Safety and Licensing Board Harmon, Curran, Spielberg, & Eisenberg, U.S. Nuclear Regulatory Commission L.L.P.

Mail Stop: T-3 F23 1726 M Street N.W., Suite 600 Washington, D.C. 20555-0001 Washington, D.C. 20036 Lauren. Breaman(dnrc.g1ov dcurran(&harmoncurran.com

• James R. Milkey, Esq.

Assistant Attorney General, Chief Environmental Protection Division Office of the Attorney General One Ashburton Place, 18th Floor Boston, MA 02108 j im.milkey(Rstate.ma.us K

Matias F. rr2ul pc~A-Thi 2