ML081650169

New England Coalition, Incs Motion to Late-File Rebuttal Testimony of Ulrich Witte

ML081650169
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	06/06/2008
From:	Tyler K New England Coalition, Shems, Dunkiel, Kassel, & Saunders, PLLC
To:	Karlin A, Wendy Reed, Richard Wardwell Atomic Safety and Licensing Board Panel
SECY RAS
References
50-271-LR, ASLBP 06-849-03-LR, RAS M-82
Download: ML081650169 (43)

Text

P OAS Hk-DOCKETED USNRC June i1, 2008 (8:00am)

OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF UNITED STATES NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD Before Administrative Judges:

Alex S. Karlin, Chairman Dr. Richard E. Wardwell Dr. William H. Reed In the Matter of

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ENTERGY NUCLEAR VERMONT YANKEE, LLC

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Docket No. 50-271-LR and ENTERGY NUCLEAR OPERATIONS, INC.

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ASLBP No. 06-849-03-LR

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(Vermont Yankee Nuclear Power Station)

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NEW ENGLAND COALITION, INC's MOTION TO LATE-FILE REBUTTALTESTIMONY OF-ULRICH WITTE Pursuant to 10 CFR § 2.323 and the Initial Scheduling Order ¶ 9, New England Coalition, Inc. (NEC) requests authorization to late-file the testimony of Ulrich Witte in support of NEC's Rebuttal Statement of Positioný attached hereto as Exhibit 3.

NEC timely filed its Rebuttal Statement of Position, on June 2, 2008. As Mr.

Witte explains in his Declaration attached-hereto-as Exhibit 1, he was unable to complete his testimony in support of NEC's Rebuttal Statement of Position byNEC's June 2,: 2008 deadline.because he was ill from Friday, May.30t1 through Sunday, June 1st NEC's Counsel: was&unaware of Mr. Witte's illness or inability to complete, his testimony until Monday, June 2, 2008, andý therefore could not meet the deadline for a request to extend the proceeding schedule pursuant to.the Initial Scheduling Order ¶9. See, EXhibit 2, Declaration'of Karen Tyler.

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The Board could readily avoid any prejudice to other parties that may result from the late submission of Mr. Witte's testimony by extending the deadline for submission of portions of each party's proposed direct examination plan that concern Mr. Witte's rebuttal testimony by four (4) days. No other modifications to the proceeding schedule would be necessary. Thus, admission of Mr. Witte's late-filed rebuttal testimony will serve the "NRC's interest in providing a fair and expeditious resolution of the issues sought to be adjudicated by the parties in the proceeding." 10 CFR § 2.332(b).

NEC has consulted all other parties to this proceeding concerning this motion.

The States of Vermont, New Hampshire and Massachusetts are not opposed. Entergy and the NRC Staff! are opposed.

June 6, 2008 New England Coalition, Inc.

by:

Andrew Rauobvo SHEMS DUNKIELIKASSEL & SAUNDERS PLLC

-For the: firm Attorneys f6r NEC

'The NRC Staff is opposed even.though it late-filed.substantive revisions to:the direct testimony, of Kenneth C. Chang Conce.*ng NECContentions 2A & 2B (Metal Fatigue) more than a week after the StafPs deadline for filing that testimony. See, Chang Correction Letter with Enclosures (May 22, 2008).

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EXHIBIT I UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No. 50-271 -LR ENTERGY NUCLEAR VERMONT YANKEE, LLC, and ASLBP No. 06-849-03-LR ENTERGY NUCLEAR OPERATIONS, INC.

June 6, 2008 (Vermont Yankee Nuclear Power Station)

DECLARATION OF ULRICH WITTE

1.

My name is Ulrich Witte.

2.

I provided direct testimony in support of New England Coalition, Inc.'s (NEC)

Initial Statement of Position in the above-captioned proceeding, filed April 28, 2008.

3, I have also drafted testimony in'support of NEC's Rebuttal Statement of Position, filed June 2, 2008. I was unable to complete my rebuttal testimony in time to meet NEC's June 2, 2008 filing deadline because I was ill from Friday, May 30 through Sunday, June 1.

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

Ulrich Witte At i-4,'Pk* Connecticut, this L46 day of June, 2008 personally appeared Ulrich Witte, and having subscribed his name acknowledges his signature to be his free act and deed.

Before me:

Notary Public My Commission Expires

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EXHIBIT 2 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of ENTERGY NUCLEAR VERMONT YANKEE, LLC, and ENTERGY NUCLEAR OPERATIONS, INC.

(Vermont Yankee Nuclear Power Station)

Docket No. 50-271-LR ASLBP No. 06-849-03-LR June 5, 2008 DECLARATION OF KAREN TYLER

1.

My name is Karen Tyler.

2.

Ulrich Witte first informed me on June 2, 2008 that he was unable, due to illness, to complete his testimony in support of New England Coalition, Inc.'s (NEC) Rebuttal Statement of Position in time to meet NEC's filing deadline.

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

Karen Tyler

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At Burlington, Vermont, this 6h day of June, 2008 personally appeared Karen Tyler, and having subscribed her name acknowledges her 'signature to be her free act and deed.

Before me:

Notary Public My Commissioni Expires

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Exhibit 3

UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of

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ENTERGY NUCLEAR VERMONT YANKEE, LLC

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Docket No. 50-271-LR and ENTERGY NUCLEAR OPERATIONS, INC.

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ASLB No. 06-849-03-LR

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Vermont Yankee Nuclear Power Station

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PRE-FILED REBUTTAL TESTIMONY OF ULRICH WITTE REGARDING NEW ENGLAND COALITION, INC.'S CONTENTIONS 2A, 2B AND 4 Q1.

Please state your name.

Al.

My name is Ulrich Witte.

Q2.

Have you previously provided testimony in this proceeding?

A2.

Yes. I provided direct testimony in support of New England Coalition, Inc.'s (NEC) Initial Statement of Position, filed April 28, 2008.

Q3.

Have you reviewed the initial statements of position, direct testimony and exhibits concerning NEC's Contentions 2A and 2B filed by Entergy and the NRC Staff?

A3.

Yes. I have reviewed Entergy's Initial Statement of Position on New England Coalition Contentions (May 13, 2008), and the Joint Declaration of James C. Fitzpatrick and Gary L. Stevens on NEC Contention 2A/2B - Environmentally-Assisted Fatigue (May 12, 2008) and exhibits thereto. I have also reviewed the NRC Staff Initial Statement of Position on NEC Contentions 2A, 2B, 3, and 4, the Affidavit of John R. Fair I

Concerning NEC Contentions 2A & 2B (Metal Fatigue) (May 13, 2008) and exhibits thereto, the Affidavit of Kenneth Chang Concerning NEC Contentions 2A & 2B (Metal Fatigue) (May 12, 2008) and exhibits thereto, and the revised Affidavit of Dr. Chang provided on May 22, 2008.

I. NEC's Contentions 2A and 2B - Environmental Assisted Metal Fatigue Analysis Q4.

Please describe your qualifications to provide testimony concerning NEC's Contentions 2A and 2B.

A4.

I have extensive experience in original stress analysis in qualifying Class 1 and Class 2 pipe and components, and applicable ASME codes as well as ANSI B3 1.1 codes, in particular in the design, analysis, construction, and qualification of Class 1 and 2 systems within the domestic nuclear industry. This experience includes, for example, original stress analysis for McGuire, Catawba, and V.C. Summers Power Plants. In addition, I have performed non-linear finite element analysis for a number of components and I am familiar with Swanson's computer algorithms such as ANSYS., RELAP, and other commercial analytical computer programs. Under contract to EPRI, I conducted detailed correlation studies of non-linear finite element analysis code predictions against actual in situ testing of piping, and components at the Indian Point 1 Nuclear facility after the plant was closed. The results are, published in EPRI Report Number 8480, -

Seismic Piping Test and Analysis, 1980:

Q5. Do you agree that Entergy's "confirmatory" CUFen analysis of the feedwater nozzle fully incorporates thermal fatigue history for the feedwater nozzles?

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A5. No. The NRC questioned the Applicant's "simplified analysis" with respect to the Feedwater nozzle as part of Request for Additional Information (RAI) dated October 9, 2007, during NRC LR Audit. The Staff was unsatisfied with the responses by Entergy, dated October 19, 2007 and November 14, 2007. During a meeting with Staff on January 8, 2008, the Applicant committedto performing refined analysis on the Feedwater nozzle including the use of actual operational thermal fatigue histories, as opposed to derived histories from the GE Specification. Incorporation of operational histories of the Feedwater nozzle was made a formal commitment in BVY 08-008, dated February 5, 2008.

An operational event that results in an unanalyzed thermal transient to the reactor vessel is relevant and cannot simply be set aside as licensees did for some period of time.

The event at Vermont Yankee (VY)was no exception. The causal relationship between the event as found in historical records and the consequences in terms of thermal shock is key. During the early years of plant start-up and operation there where many unplanned forced shutdowns..I found 42 for VY. Not exactly asilky'smooth running reactor. Three were downright dangerous.

GE and the Licensee did not fully predict all of the events in their shutdown estimates. Hence, those that were outliers needed detailed analysis. During the mid-1980s and into the 1990s this fact came to light starting with NUREG 0599 and others.

Operational events led to the need for careful and refined transient analysis. The simplified method was shown to be overly dependent on skillful and experienced engineering. New methods removed the uncertainties and doubts of accuracy in CUF and 3

CUFen. Not just cycle counting but examination of derivative temperature changes forced on the reactor vessel, the associated safe end, and on, of course, the feedwater nozzle as well., I know, because I was required immediately to notify the Technical Support Center (the emergency response area assembling management to provide technical support) for just such an event occurred on December 2 6 th, 1986, at 6am, which brought down another plant for many months, placing the plant under its emergency plan. There was a'concem that the plant would never operate again.

Based uponmy examinationof Vermont Yankee's historical records and my own experience of the challenge of maintaining nuclear plant operational history beginning with plant start-up, it appears to me that major thermal transients have likely not been.

incorporated into the operational history, as referenced in the SER. This deficiency is particularly significant where the reactor vessel has experienced an unplanned and unanalyzed transient that was outside the engineered design basis. Occurrence of these events throughout the industry was not as uncommon as one might presume.

Assessment of transient impact. to specific component life is required following such an event toreestablish fidelity with the plant's design basis and is accompanied by additional fatigue analysis. The outcome of the engineering analysis holds one of three possibilities: (1) severe damage has occurred to the nozzle or vessel (less likely), (2) no additional fatigue usage outside the GE Specifications has occurred (also not likely), or (3) some additional usage outsidethe GE Specifications has occurred and therefore the component life is shortened (likely). Assessment and incorporation of the assessment of these impacts into plant operating records is essential to providing a basis for effective

,*aging management programs.

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An example of an historical Vermont Yankee event with the potential to impact the useful life of a number of systems, structures, and components occurred on December 1, 1972. On that date, the reactor automatically scrammed when an internal fault on a startup transformer resulted in a loss of offsite power. The emergency diesel generators automatically started and connected to their electrical buses. The high pressure coolant injection (HPCI) system got an automatic start signal on high drywell pressure, but failed to start. The operators manually started HPCI. Three relief valves opened when reactor pressure increased to 1,130 pounds per square inch gauge. A fourth relief valve should have opened, but failed to do so. One of the three relief valves that opened chattered on its seat about 100 psig below its set point. The transient was significant as reflected by the fact that odds of a core melt from this single event were 1-4E-3. See, Exhibit UW-24.

More significant to the issue of fully recovering the record of all transients and accurately incorporating them in assessing remaining fatigue life is the assessment of wear, damage, and stress on each relevant component during each significant transient event.

• There are other examples of transients that appear to have not been incorporated as input in the refined fatigue analysis. During the period from 1973 through 1977, Vermont Yankee experienced 42 unplanned forced shutdowns. This is a significant number, and expended much of the fatigue life of the reactor vessel and feedwater nozzle. See Exhibit UW-25.

Of these 42 forced shutdowns, in 1976 Vermont Yankee experienced 10 unplanned reactor scrams. Exhibit UW-24. One of these, on July 6, 1976, occurred during surveillance testing when the air operator, plunger on a relief valve did not move when air was applied. Two of the other three relief valves failed. The failures were traced to air 5

operator diaphragms damaged during excessive heating. The damage was attributed to improper insulation in the proximity of the diaphragms and an extended operating cycle.

Core melt frequency for this event was an astoundingly high number 6.25 E-2. Exhibit UW-24. Again, the event stressed a number of systems and impacted the fatigue life of numerous components.

I made a comparison of the Engineering Design Input document, EN-DC-141, Rev. 3 provided to NEC by Entergy, to available records contained in the following documents and as compared to the responses provided to Dr. Chang's questions contained in Exhibit UW-26, "NRC Audit 10/09/07, with responses provided 10/18/07."

It appears that, in Entergy's calculation of 60-year CUFs in its CUFen reanalyses, operational histories were not properly or accurately compiled and that instead of documented transients, estimated thermal transient histories were used to predict the number of Reactor Thermal Cycles for 60 years. Purported added conservatisms remain unqualified and unjustified. The estimates of thermal transients are provided on, Page 1 of 6, EN-DC-141, Rev. 3. See Exhibit UW-27 "Design Input Record, Environmental Fatigue Analysis for Vermont Yankee Nuclear Power Station."

Q6. Why is this of concern in assessing the validity of Entergy's CUFen reanalysis?

A6. Refined fatigue analysis fidelity largely turns on correct design inputs. The simplified Green's Function method challenged by Staff on January 8, 2008 and in other records, was essentially about uncertainty in assumptions and estimates. My observation is that this particular design input is an ungrounded estimate, an assumption, and not an actual historical number; any conclusion stemming from it, therefore, cannot be relied on without,orroboration. Clearly, to proceed with estimates based on a flawed record of all 6

transient events is not appropriate. The rationale provided for not using actual transient operational cycles as found in Exhibit UW-26 at sequential page no. 8 (Bates number NEC069994), is not valid in the event of a thermal transient event that was outside the original design basis. Entergy, has not shown that those events were incorporated.

Second, the estimated transient history - assumption - may or may not be conservative. As noted above, the plant experienced certain transients during its operational life from initial plant start up and testing, commercial operation, then uprate to 120% power beginning in 2004. Actual excursions, in particular those that appear to be outside the GE design specifications, should have been accounted for in the refined analysis. From the analysis provided, at least in the first example, they were not.

Third, considering Extended.Power Uprate contributing factors such as increased flow, component modification, increased vibration, and increased core heat and neutron flux, the transients experienced, by the plant beginning with power escalation to 120%

should be given more weight in forecasting thermal transient cycles. There is no credible basis provided in.the Applicant's analysis that justifies thermal cycle projections to 60 years.

In summary, by using estimated histories as opposed to actual history, specific transients that shorten the component fatigue life appear not to be acknowledged or included in the Applicants fatigue analysis, making the results including CUFen unsubstantiated.

II.

NEC's Contention 4: Flow Accelerated Corrosion Plan 7.

Q7.

Have you reviewed the initial statements of position, direct testimony and exhibits concerning NEC's Contention 4 filed by Entergy and the NRC Staff?

A7.

Yes. I have reviewed Entergy's Initial Statement of Position on New England Coalition Contentions (May 13, 2008), and the Joint Declaration of James C. Fitzpatrick and Dr. Jeffrey Horowitz on NEC Contention 4 - Flow Accelerated Corrosion (May 12, 2008) and exhibits thereto. I have also reviewed the NRC Staff Initial Statement of Position on NEC Contentions 4, and the Affidavit of Kaihwa R. Hsu and Jonathan G.

Rowley Concerning NEC Contention 4 (Flow-Accelerated Corrosion) (May 13, 2008),

and exhibits thereto.

Q8.

Entergy contends that you have no experience or expertise relevant to the testimony you have provided concerning NEC's Contention 4. How do you respond?

A8.

I have extensive experience in development of engineering programs including controls for design change processes, configuration management programs and comprehensive initiatives in affecting operating nuclear power stations. These processes typically involve complex multifunction, and multi-organization challenges. These programs are often mandated under federal regulations, or committed programs for a licensee to re-establish fidelity with its current design basis and license conditions. I have substantial experience in, for example, implementation and validation of NUREG 0737, "Clarification of TMI Action Plan Requirements," rand was a principal manager in the successful restoration of Indian Point 3 from the NRC's Watch *list, as well as Millstone Units 2 and 3. For the Tennessee Valley Authority, specifically the completion of the Watts Bar Nuclea* Plant, I developed a p;ogrmn entitled "Program to Assure Completion and Quality." For Georgia Power's Plant Hatch, I developed and implemented a 8

Configuration Management Program, led in-house Safety System Functional Inspections, and an Electrical Distribution Function Inspection so as to prevent Plant Hatch from going on the NRC's watch list. For Northeast Utilities, I developed a multiple department and multi-function program to reestablish the fidelity of the design basis and licensing basis, including identifying, dispositioning and either eliminating or implementing over 30,000 regulatory commitments. My leadership in establishing and implementing these programs

- successful initiatives was well-received by the Licensee and well-received by the regulator. By their transparency to the community, they were generally accepted as improvements by the Licensee in protecting the health and safety of the public and minimizing risk to public assets.

As a seasoned engineer, manager, and problem solver, my expertise and track record demonstrate successfully implemented solutions to complex organizational, technical, or regulatory challenges in nuclear plant operations.

Applying my expertise in Engineering Design Control Programs, I note that Entergy's proposed Flow Accelerated Corrosion management program is~based on use of a predictive modeling tool derived fromn an empirically based program with heavy reliance on engineering judgment, coupled with experience, oversight, and effective monitoring of FAC-related wear to certain vulnerable plant. systems. My expertise in program management focuses on correct and effective implementation of the program and finding a record that is auditable, defendable. against program requirements and transparent. To quote the NRC Staff s position regarding flow. accelerated corrosion, "Corrosion is not an exact science. Due to epistemic and aleatory uncertainty,, absolute wear rates cannot be determined...." NRC Staff Initial Statement of Position-at 20. Thusthe burdenin 9

constructing and maintaining an effective FAC program must emphasize' reliance on engineering judgment, coupled with experience, oversight, and effective monitoring of FAC-related wear.

While I do not purport to be intimately familiar with the empirically based CHECWORKS algorithm, I can attest to sufficient expertise in evaluating the fidelity of a comprehensive FAC program. I believe that the parties and witnesses are not in dispute that an effective flow accelerated program is highly dependent on sound engineering judgment and precise implementation, including the program goal of effective management of the predictive results, so as to preclude wall thinning beyond acceptance criteria during the license renewal period.

A. Summary Rebuttal Q9.

Do you believe that Entergy's Flow Accelerated Corrosion Management Program as implemented to date will be adequate for purposes of aging management during the period of extended operation, as Entergy and the NRC Staff assert in their initial statements of position and direct testimony?

A9.

No. Entergy asserts on page 34, 35, and 37 of their Intial Statement of Position to New England Coalition Contentions, that their intention to credit the existing program as demonstrated to be adequate with no changes planned. Staff underwrites this assertion as well on page 20 of the NRC Staff's Initial Statement of Position on NewEngland Coalition Contentions. I do not agree, the program as implemented to date is adequate.

NEC raised significant concerns* regarding the Flow-Accelerated Corrosion Program and asserted that the application' for License Rene wal submitted by Entergy for

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Vermont Yankee does not include an adequate plan to monitor and manage aging of plant 10'

equipment due to flow-accelerated corrosion during extended plant operation. The responses provided in summary disposition as well as Entergy's Reply and Staff's Reply do not address NEC's concerns and in fact raise troubling new concerns beyond simply the sufficiency of the Vermont Yankee flow-accelerated corrosion program as presently credited for license renewal.

The Applicant's response summarized during-motion for summary disposition is that it's present FAC program is consistent with industry guidance including EPRI NSAC

• 202L R.3 and that the use of the CHECWORKS model is a central element in the FAC program implementation. The Applicant stated that it is relying on its current program for FAC management for the license renewal period, and "furthermore, the FAC program that will be implemented by Entergy is the same program being carried out today... [and] will meet all regulatory guidance.". See Entergy Reply at 34.

Entergy represents that it will rely on its current FAC management program for purposes of FAC management during, the license renewal period, that no changes to this program are planned, and that this program complies with EPRI guidelines.. See, Entergy's Initial Statement of Position on New England Coalition Contentions at 34 ("The current FAC program, which will be used during the license renewal period, meets industry practice as reflected in NSAC-202L..."). My review provided in pre-filed testimony shows that Entergy's current program is not in compliance with EPRI guidelines.

Q10. Entergy asserts on page 34 of its Initial Statement of Position that "the program has been reviewed, audited,* and inspected with only minor, mostly 11

administrative issues identified," and discounts its own Quality Assurance audit, which declared the, program "unsatisfactory." How doyou respond?

A1O. I believe that.these statements indicate that Entergy may have ignored or misconstrued the fundamental requirements of I OCFR Part 50, Appendix B, "Quality Assurance Requirements for Nuclear Power Plants." It appears that federal requirements for Quality Assurance (QA) are being set aside. Quality Assurance Division Audit No.

QA-8-2004-VY-1 declared the Flow Accelerated Program "unsatisfactory," submitted two Condition Reports, and found five findings and seven areas of improvement. See, Exhibit NEC-UW_09 at 2. Yet Entergy's Initial Statement of Position interprets the 38-page document as containing "only minor, mostly administrative issue[s]." Entergy Initial Statement of Position at 34.

Furthermore, the Entergy asserts this single analytical tool for predicting unacceptable wall thinning should, as policy, be set aside as it was for four components, See Exhibit NEC-UW_20 at 5 of 14. Thus the Entergy provides a second indicator where the Licensee obliquely.waived Appendix B requirements for Quality Assurance. See Entergy Statement of Initial Position at 48:

That again is misapplication of the requirements of Appendix B, which is particular to the Flow Accelerated Program, where the Applicant's only defense to its failure to prepare condition reports associated with unacceptable wall thinning, a prediction derived fromits own analysis, is somehow that this component shown not to be meeting quality standards isdeemed acceptable "as is'until the next outage. Therefore, there are two indications of a troubling and clearly deep-seated failure to properly implement the requirements of a compliant Quality Assurance Program. Appendix B to 12'

10 CFR Part 50 requires among other things,Section III, "Design Control; and Section XVI, "Corrective Action" The latter section of the rule includes the following:

Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the significant condition adverse to quality, the cause of the condition, and the corrective action taken shall be documented and reported to the appropriate levels of management.

Quality Assurance requirements are not apracticethat may or may not be voluntarily implemented.by the Licensee, but are in fact are regulatory requirements promulgated under federal rules. The Applicant incorrectly asserts that a failure theoretically predicted by the CHECWORKS model is somehow treated differently than a failure predicted by actual inspection data.. The Applicant is incorrect in assuming that a failure predicted by CHECWORKS does not meet the threshold for a condition report, with timely follow-up or corrective action, as fundamentally required under Appendix B.

The Licensee has no regulatory grounds to escape a determination of potential failure by reason of its assertion that "if a planning tool such as CHECWORKS..... determines a theoretical conclusion.., as such no condition reports are required." See Entergy Statement of Initial position at 48. This improper rationale is essentially analogous to a Licensee ignoring a Technical Specification requirement calling for declaration of a component or system to be classified as inoperable and a Limiting Condition of Operation started if a surveillance is missed. In theanalogous situation, a component is administratively (theoretically) declared inoperable, although its actual functionality is unknown.

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The consequences of the Licensee's apparent policy regarding Appendix B requirements, for Vermont Yankee's Flow Accelerated Corrosion Program are significant and have broad implications to multiple programs relied upon for renewal. Essentially, following the Licensee's logic every program can be viewed as theoretical when it is intended to be a predictive tool. The implications of Entergy's statements are profound and raise questions regarding credibility of all the Aging Related Management Programs proposed and Entergy's actual intentions for monitoring, and maintaining the plant if the license is extended.

Qll.

Has applicant provided in its response any reasonable assurance that pipe thinning beyond code limits will riot occur in the period between outages?

All.

No. Quite to the contrary, the applicant has stated at page 48 of its Initial Statement of Position, in reference to page 5 of 14 of PP7028 Piping Inspection Program, Exhibit NEC-UW_20, that wear rates predicted to exceed code limits will not be acted upon until the next outage. Based on statements made by the Applicant regarding pipe thinning predictions including negative time to inspect (described as negative Tmin in the document) and predictions of unacceptable wear rates leading to thinning beyond code limits prior to the next outage, coupled with the decision to not prepare condition reports (or an analogous report consistent with requirements of a corrective action program as part of Appendix B), it is my opinion that reasonable assurance is not provided, and that the NRC Staff erroneously concluded that the program is complete, correct and adequate.

Therefore, my opinion is that the staff erroneously concluded that the program is complete, correct and adequate.

V;,

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Q. 12 Does Entergy's Initial Statement of Position resolve the programmatic weaknesses you identified in your direct testimony, including open corrective actions, stale open action items from condition reports, and the negative assessment of the program stated in the 2006 cornerstone roll up report?

A12.

No. Entergy characterizes the issues. I have identified as shortcomings in the documentation paperwork with no substantive implications. I disagree. Any one of the Quality Assurance findings are significant. For example, a classic indictor of a problematic program is age of open corrective actions. A second indicator is number of Condition Reports, and number of extensions planned and then postponed to implement necessary actions to maintain the program current. Data drawn was sometimes more than fifteen years old.:

Entergy expends much discussion, largely on a generic basis, on what ought to constitute a good FAC program. Entergy Statement of Initial Position at 36. However, Entergy does not respond to or take into consideration the VY's actual repeated historical failures to implement the FAC program from 1999 to the.present day, which I have identified in my report, filed in. this proceeding as Exhibit NEC-UW-03. With few exceptions, these numerous programmatic failures go unchallenged by Entergy.

Most significantly, successive implementation of CHECWORKS to current plant design inputs is undisputed as a mandatory element of the program, as required under NSAC 202L rev. 2 and rev. 3. Entergy makes no claim that this was consistently done.

Successive data passes.at appropriate intervals, with scope selection, current operating conditions etc, taken into consideration are a fundamental element to identifying appropriate grid selection points, and trending of wear items. However, this obligation 1,5

was consistently ignored for many years and at best done in fragments for many outages.

See Exhibit NEC-UW_03, "Evaluation of Vermont Yankee Nuclear Power Station License Extension." This approach places the reviewer in the untenable position of having to look a look at wear data for trends with only very limited data points and then speculate as to whether the data set is sufficient. This approach is invalid.

Detailed Review of Entergy and Staff Reply Q13.

Do you take issue with the general merits of the approach to FAC management recommended in NSAC 202L?

A13.

No. My focus is strictly'on the adequacy of the implementation of NSAC 202L at VY.

Q14. On Page 38 of its Initial Statement of Position, Entergy makes thefollowing assertion regarding FAC Susceptibility review: "the only CHECWORKS inputs affecting FAC wear rate that need to be changed to model uprate conditions were the flow rate and the temperature. These were updated at VY upon implementation of the EPU." Do you agree that flow rate and temperature are the only inputs that were necessary to incorporate into the model?

A14. No. I disagree. Identification of the added inputs should be made, incorporating the results of all pertinent susceptibility analyses. Apparently, this has not been done. First, Exhibit E4-32 is a copy of a susceptibility analysis performed by Entergy in 2005. This analysis was performed fully five years after the previous analysis was completed in 2000.

This five year gap is found by examining the dates associated with the 2005 Susceptibility analysis. Numerous changes to the plant occurred between 2000 and 2005. For example, in 2003, the reactor recirculation and residual. heat removal piping was replaced. See, Exhibit NEC'-UW_27 at 6, Attachment 1. Second, operational factors (such as TECH 16

SPEC changes, configuration changes, and material changes) should have triggered a new susceptibility analysis well before the analysis performed in 2005.

In brief, beginning in 2004, substantial plant modifications were performed, including system modifications etc, yet a current Susceptibility Analysis was not performed until 2005. The premise that only flow rate and temperature input changes were needed is not properly supported and incorrect.

It is apparent that Vermont Yankee's FAC program management was broken from February 28, 2000 through October 25, 2005 -based upon lack of Susceptibility Analysis alone. A. comparison of program scope for piping inclusion, exclusion, small bore, large bore, fluid type etc, should have been incorporated into the FAC Program under the station Engineering Design Controls program on an ongoing basis-essentially any time a plant modification, system function change, or operational change was contemplated.

Based upon the Applicant's information provided on page 38 of Entergy's Statement of Initial Position, as well as the Table 2 of Exhibit E4-32, the susceptibility analysis was set aside for more than five years, losing both continuity and assurance that all modifications have been evaluated and taken into consideration.

Proper grid point selection, proper sampling, proper frequency and the consistent integration of new data all. serve to remove speculation and uncertainty in the accuracy of CHECWORKS. This fact by itself provides:the impetus for a "new baseline," especially in light of the fact -that a current baseline is, for all practical purposes, lacking. In conjunction with the relative uniqueness of the CPPU power uprate-chemistry changes, geometry changes, and of course velocity changes, the need for a "new baseline" is compelling. The strength of the CHECWORKS and the NSAC 202L methodology 17

endorsed in the GALL Report, is in its successive passes with tight control of changes in requisite input variables. These core elements have yet to be implemented.

In 2005, Entergy relied on ancient susceptibility data for component selection points, such as small bore piping from data circa 1993. See Exhibit NEC-UW_20 at page 12 of 14. Five small bore points were selected that had never been inspected previously, indicating, loss of control of the program. Entergy's defense of this methodology raises significant doubt as to the effica'cy of the current program, and therefore the FAC program for the license renewal period.

A lack of a timely susceptible review can only serve to skew the results appropriate selection of specific wear points. An updated and inclusive Susceptibility Review should definitely have been required by NRC Staff in their review. It apparently was not.

The Susceptibility review did not appear to address wear points associated with plant modifications5and based upon the descoping of the inspection, even after*

recommending by engineering judgment, to include certain points they were not. See Exhibit E4-38 referenced in Entergy's Statement of Initial Position at page 39.

Q15.-On page 39 of its Initial Statement of Position, Entergy states that in 2007, RFO 26, the first outage since the EPU, the inspection scope was a total of 63 inspections performed, including 49 large bore inspections. Do you believe that Entergy met its commitment to increase the scope of inspection by 50%?

A15. No. It is apparent on reviewing the record that Entergy first reduced the effective inspection scope and then enlarged it, in the process offsetting any "increase." A mirror 18

analogy would be the retail store that raises its prices on certain goods, prior to offering them at a sale discount.

Entergy's commitment to increase the number of inspection points by 50% was made in response to an RAI, acknowledged in Entergy's Statement of Initial Position at 39, but this commitment was tacitly fulfilled by increasing the number of inspection points for RFO 26 only after decreasing the number of inspection points (by descoping) for RFO 25. The Scoping document for RFO 25 contained significantly more inspection points. See, Exhibit NEC-UW_20, "PP7028 Piping FAC Inspection Program FAC INSPECTION PROGRAM RECORDS FOR 2005 REFUELING OUTAGE." On page 20, it states "The planned 2005 RFO inspection scope consists of 0137 large bore components at 16 locations... [a]lso, any industry, or plant events that occur in the interim may necessitate an increase in the planned scope." In addition, criteria for inspection of components outside of CHECWORKS grid selection is articulated to include points simply because of the lengthy intervals since previous inspections.. These include Feedwater piping, and Mainsteam piping. Id. at 3.

However, the number called for in the above scoping document is considerably more than the actual number of large bore components reported to be inspected during RFO 25, as in Exhibit E4-38, where the Applicant notes that it limited its inspection to 27 large bore points. The actual inspection of 63 large bore points.for RFO 26 is about 1/22 of the number of planned inspection. points for RFO 25, not 50% more.

Q16. Entergy, disagrees with your statement in direct testimony that "trending to the high end of the range [for bench marking] is appropriate where variables 19

affecting wear rate,.such as flow velocity, have significantly changed, as at VYNPS following the 120% power up-rate...". How do you respond?

A16.

Entergy questions the relevance of the report brought forward in my direct testimony in support of this statement. The report in question is "Aging Management and Life Extension in the U.S. Nuclear Power Industry," Exhibit NEC-UW_13, or the "Chockie Report." Entergy asserts that this report does not support trending to the high end of the range where variables such as flow velocity etc have significantly changed, because it 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.

The Chockie Report most certainly assimilates industry guidance, including regulatory rules and implementation of those -rules, and compiles aging programs strictly with respect to the United States domestic nuclear power plants. On page 38, it answers exactly what is required if there is no pre-existing baseline, as is the case for Vermont Yankee. The use of the report by the Norway Petroleum Safety Authority has no bearing on its content. The report is on point to Contention 4.

The Chockie Report is applicable to the question of what constitutes an adequate baseline. Entergy assumes that its present baseline is adequate. I believe after examination of the failure to adequately implement the program, that VY does not have an adequate baseline. The Chockie Report is a concise primer on the effective implementation of NSAC 202L, including CHECWORKS, and by inference impeaches Entergy's Application as well as the adequacy of NRC Staff Review.

20

Q17 Do you agree with Entergy's statement contained in a single paragraph on page 45 of Entergy's Initial Statement of Position that the following eight claims you made in your direct testimony have no merit?

a. "that data from previous FAC inspections (prior to the EPU) were not entered into the CHECWORKS database (NEC-UW_03 at 2, 3, 6, 7-8, 15, 16, 17);"

b. "that CHECWORKS was not updated with the uprate parameters (id. at 5, 23);

c. that, for the period 2000-2006, VY failed to use a current version of CHECWORKS (id. at 6, 17);"

d. "that four components were predicted in 2004 to have wall thinning beyond operability limits (id. at 17418, 22);"

e. "that open corrective actions identified in condition reports may not have been completed (id. at 3-4, 18-19);"

f. "that-ranking of small bore piping was not done (id. at 8, 20);"

g. "that the number of inspection. points were reduced after the 2005 outage (id. at 7, 8, 20); and"

h. "that the 20061 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)."

A17. No. I disagree. Entergy states that these claims have no merit but does not actually refute them, or specifically address the majority of the documents I cite in support of my direct testimony. Entergy's reply to my direct testimony consists primarily of conclusory denials.

Q18. Does this conclude your rebuttal testimony?

A18. Yes 21

OG/0S/'2E18 14:40 2033896657 NORTHERN LIGHTS ENGI PAGE 01/01 I declare under penalty of perjiry that the foregoing is true and coiTect.

Ulrich Witte At L/J 4"'**, Connecticut, this L dA day of June, 2008 personally appeared Ulrich Witte, and having subscribed his name acknowledges his signature to be his free act and deed.

Before me:

Notary Public My Commission Expires

NEC-UW_24 Nuclear Near-Misses Odds*

Date Reactor Location 625.0 7

6 1976 Vermont Yankee Vernon VT 1470.6 12 1

1972 Vermont Yankee Vernon VT 3448.3 4 23 1991 Vermont Yankee Vernon VT

• Odds from NRC reports of how likely can lead to reactor meltdown. The value represents how often such an' even't would result in a meltdown. For example, a value of 50 means that, on average, a meltdown will occur once every 50 years.

Friday, May 30, 2008 Page I of I

NEC-UW_25 Year Shutdowns V erm ontY an eelElill:1-IbIl:

1973 1974 1975 1976 1977.

1978 1979 1981 1982 1984 1987 1988 1990 1991 1992 1993 1994 1995 2001 2002 2003 2004 2005 2007 2

16 9

10 5

8 5

1 6

1 1

2 5

4 2

5 3

2 2

2 2

2 2

2 M ii' lllil axnnux~i:

Tuesday, May 27, 2008 Page 63 of 67

NEC-UW_26 NRC LR Audit 10M09/07 EAF Questions from Ken Chang Thursday 10118107 6:30 PM EST Question IQuestion No.

z 0

__.*~he SMECode defines that stress intensity-(Sl) from two temperature transients is calculated from the stress components from the two conditions. Please explain how it could be calculated from stress intensities of the two conditions derived from Greens Functions, especially at locations of geometric discontinuity. Also, please justify the

-vatidity of combining the thermal transient stress intensities with the stress intensities from the external loads and pressure loading.

• ok V( Response SA maThematical proof of the approach is not provided. Rather, to show that the method whereby the Gre unction (GF) approach used in EAF calculations obt irnirsults comparable to results from standard-ASME Qode fatigi."calculations, a

!comparison of calculations performed for the VY feedwater nozzlewas performed.

v The,urrentASMFe Code design fatide calcuiahan.(V -IOQ-3.0D)) 9.hich waSvperformed directly qsing AN$ SSw compared to,.h6 EAF calculation (VY-16Q-301) performeo* g the GF methodology for the turbine roll trh sient, which the most severe design basis transient for fY. To ensure a consistent comparison between the two calculations, the same stress path oc0in were gthe Code fatigue calculation alternating Pstresses (using Sz-Sx) were extracted from the ANSYS model at the base metal rather than at the cladding as was originally performed in VY-10Q-30 In addition, the Code fatigue calculation ANSYS model was re-run with the same heat transfer coefficients and material properties used for the GF calculation.

The comparison showed that the differences in alternating stresses are less than 1% at the both the safe end and blend radius Idcations.

Although this comparison was for the feedwatet nozzle, the results are considered to be equally applicable to all other nozzle locations based on a BWR Vessel and Internals Project (BWRVIP) study (EPRI Report No. 1003557, "BWRVIP-1 08:

BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vesset Shell Welds and Nozzle Blend Radii,"

Final Report. October 2002. St File No. BWRVIP-01-308P).

eJI

. 'C N

z Lii

(~2 C

C' C

C C

Ao'~

/

I.

In BWRVIP-10, 3-0 mo els of four different nozzles were developed and analyze for the BWR fleet. The results of this study showed tha for a range of vessel nozzles. the stress concentration fctors for pressure loading are 2.65 +1-3%. indibating that.

e BWR vessel nozzles have the same geo9metric characteristics for calculating peak stresses in the blend radius.regiions, i The.com'bination of thermal stress intensities with stress j in'ensities from extern'al l-oads and pressure is addressed as f ollows:

SFor the feedwaýter nozZle of another. BWR plant, SI performed l

i-hand calcutatiori for stresses d(ie to mechanical loads (as was done for VY) and benohmarked those calculations against finite i

[element results obtained from applying the mechanical.foads to a finite elem 'ent model. Thehand calculations were performed using Vne same methodology as used for Vy.

The finite element model was an axisymmetric Iwo-dimensional (2-D) finite element model. This model was constructed and meshed in a very similar manner to the VY nozzle FEMs. Non-I symmetric loading elements were used and the shear, moment, axiaL andr torsional loads were applied to the model.

A comparison of the stresses from the hand calculations vs. the FEW is as follows:

Location Stress from r S1ress Di fferen e Hand i From Hand CairQ CalculsUons FEM vs. FEM Safe End 8863 5B62 5t.45%

Uineafized Me-nbrne + 1 B e rnd ing S tressi Safe End,.

8863 7655 12.83%

Total Stres NozzJe Forging -

1042 Lineartzed Membrane + f 769 35,5M%.

i Bendin Sliess-------------

I Nozzle Forging.

1042 554 88.09%

ITotal Stress 0%K As shown. byjthese.results, use of the hand calculations is conservative.compared to the FEM results, especially under the assumption that.the stresses from the hand calculations are treated entirety s membrane in nature.

The-stress range for.pressure. and external loadings range from zero.ica.finite value. The' ress intensities ($1) calculated from extemaI l.oad6ing is conservatively added to the maximum calculated thermal transient Si using the same sign to increase the peaks and valleys and thereby the stress range. The pressure SI value is added to the St from the combined thermal and.external loadings directly as a positive value. since pressure is always positive. As shown in Tables 4 through 6 of calculation VY-16Q-302, for the transient pairs that contribute the Mogt to the CuF at the safe end and blend radius, the stress i inite*sity range is always increased when the mechanical ttresses are added to the thermal stress using the same sign convention and that, for VY, they are small in,magnitude compared to the other stresses.

z CD I

2 Provide justificatioqn for statement on pageK'5of Calculation No. VY-16Q-302.:that 'The Greens Functio&V methodology provides identical.re*u*lts compared to running the input transient through the finite element modeLr."

//

• AJk A verification Calculation (SI calculati0n QA-2000-102) was performed that compared the results of an ANSYS analysis with a Fatigue Pro Green's Function. The calculation was performed on a feedwater nozzle for a turbine roll event. The results showed the stress range difference between the FatiguePro (Green's Function approach) and ANSYS safe end was between

-0.06% and 3.43% and for the blend radius location was between

-1.73% and 1.58%.

Further discussion of Green's Functions and how they are used in a fatigue monitoring system is available in two papers authored by SI (Kuo, Tang and Riccardetla) for the 1M$6 Pressure Vessels and Piping Conference and Exhibition in

z CD 60%

] Chicago Illinois. The papers are titled An On-Line Fatigue

,Monitoring System for Power Plants: Part I - Direct Calculation of Transient Peak Stress Through Transfer Matrices and Green's Functions' and "An On-Line Fatigue Monitoring System for Power.Plants: Part II - Development of a Personal Computer Based System for Fatigue Monitoring".

The intent of the statement in this and other calculations was to.

indicate that equivalent.stress history results are obtained for a giv6n tr~nstent.

3 For the blend radius for the feedwater nozzle in Calculation To maximize stresses in the blend radius: a temperature shock No. VY-16Q-302, Table 4, Page 16: Why are the Total & M+ 1 was performed. using a 1000F temperature at the nozzle and a B stresses for Thermal Transient 3 shown in columns 3 & 4-500'F tempeiratuie at the vessel. Since these temperatures were high at t=0 sec. (zero stress state?) This question also held constant, thestress at 1000 F is high due to the large applies to:

thermal discontinuity.. The stress fTee temperature would be

, Transient 4 at t =1801.9 sec,.

wher"ethe temperature gradient is eliminated (5000F). This Transient 9 at t = 2524 sec..

aplle-to Transient 3,14 and 21-23.

Transient 21-23 at.t= 20144 sec.

This question may also apply to transients 1,.12, and, 14.

Far.Transients 4 and 9 (turbine roll and turbine trip), the stresses are initially much lower at the initial operating temperature that

,. f precedes the temperature shock to the nozzle and increase as

:the nozzle is shocked, consistent with the stress response for the

,Green's Function shown in Figures 18-23 of calculation VY-1O6Q-

/

301. This is also true for Transients 11 (loss of feedwater pumps) and 12 (turbine generator trip),

I To ensure that the overall fatigue usage is conservatively

,N 1 calculated, a zero stress was applied to Transients 1, 2, 24 and

25. to maximize. the stress range.. A check was performed to assess the difference in CUF if the 1 000 F values for transients 3, 14 and 21-23 were changed to zero and the non-required zero stress state conditions that were added to Table 4 were removed, The result of this check was that the difference lh overall CUF was approximately 5%.

4 Explain why there are differences in the calculated CUF values a between Rev. A and Rev. 0 of the Structural Integrity Calculations. Also, why are the CUFs calculated by Structural Integrity different from the CUFs shown in Tables 4.3.1 & 4,3.3 of the Vermont Yankee License Renewal Application?

z CD M

NJ Structural Integrity calculations VY-16Q-301 through VY-16Q-310 issued as Revision A have the Revision Description on each calculation cover sheet labeled as "initial Draft for Review".

These draft calculations were issued for client review and comment. The draft versions of the calculations were never intended to be -the issued version until all external and internal reviews and -comments were incorporated, The revision A.

calculationsawer provided under Entergy's obligation to provide.

all-documents related to Environmentally Assisted Fatigue for NEC Contention 2.

The Revision.0 versions of these calculations were subsequently issued with comments on the draft calculations resolved, Revision 0 (or{later) versions of the calculations are the

,Ct ulalions of Record. The Revision A drafts are no longer applicable.

The COPs shown, in Tables 4.3.1 & 4.3.3 of the Vermont Yankee License Renewal Application Tables are based on the design

..basis fatigue evaluations factored to account for the effects of the 1'20% Extended Power Uprate, or for locations with no plant specific CUEs representative values from NUREG/CR-6260.

The CUFs calculated by Structural Integrity are different from the CUFs shown in the VY LRA due to a number of factors specific to each location, These include: updated finite element modeling and more modern thermal transient analysis than used in the original design, the use of updatedtransient definitions for 60 years of operation shown in Design Input Record (DIR) for EC No. 1773. Rev. 0, "Environmenta! Fatigue Analysis for Vermont Yankee NucleIar Power StationW Revision 1, dated 7/26107' and for the NUREG/CR-6260 locations, new VY plant specific.'ASME I.H fatigue analyses were performed.

I 5

On page 1-1 of Report VY-160-401 it indicates that refined transient definitions 60 years are used in the computation of the CUF including EAF effects. Please explajn the The original Design Transients for the Vermont Yankee Reactor Vessel for a 40 Year design life are given in Section 5.1.8 and MAttachoment 0 ta G&4eraI Electrc PurchaseSpecificationNo.

refinements in the transient definitions, gal/

stZ&o %

-73&`7 1$';

TJU a~~

Cwi1ta z

C) 10

,:21A*tI-'*5"Reactor Pressure Vessel', Revision 4, 10/21(69 and I certified on 10123/69 as contained in the Reactor Pressure Vessel Design Report. Additional clarifications and descriptions

for the design transients were provided by General Electric in GE I Letter.W. J. ZarelIa to DOW. Edwards - Yankee Atomic,

Subject:

I 'V, Y. R:P.V..Temperature Transient I Cycling Events", No. G-HB-5-124. dated:November 5, 1975.

Eailier versions of the specification made reference to a GE Therm..al Cycle Drawing.:N,..D394i,

-.The final version of the Design".Specification relodaýted this cycle information to A"*tt*bh.W.itPtheDesig*Spadificationi~and deleted references to GE dawh':YN&4885D94t:.

Comparisons wer made between the VY Design Specification trans*ients and the design transients shown on Thermal Cycle Drain'hgs from other GE BWR 4 plants of the same and a later vintage:. T'he later plants have a more defined thermal cycle history based on the experiencefrom the earlier GE BWRs.

g-In 9ene7aljVl' is designed for a smaller spectrum of more severe

transle, an the later units. As described in General Electric Letter No. G-HB-5-124. the VY design transients are intended to bound all operating conditions. ý,Forexample, the single severe designrat-Ansientfor 1he- *Y feedwater nozzle of 1500 cycles is

,intendedtoenvelop;All Start-up0, Loss of Feedwater Heater, Scram, and Shutdown events.

To insure a realistic projection of Design Thermal Transient Cycle-d vents for 60 years of operation, the Thermal Cy et i Diagf~nCed at a number of BWR 4 plants were used as a startwn'-j5int. The VY Design Specification transients were mapped onto the typical BWR 4 Transient Diagrams. Then i projections for 60 years were made based on the numbers for 40 I years in the VY Design Specification, the numbers actually analyzed in the VY Design Certified Stress Report for Vermont Yankee Reactor Vessel, Chicago Bridge & Iron, Contract 9-6201, land the number of cycles experienced by VY in approximately 35 he/u A~~;

25tt 01K', -

/

I years of operation..

I The results of the new transient definitions are documented Appendix C of calculation VYC-378 Rev.2 and were provided as input to-the EAF analysis in EN-DC-141 Design Input Record (DIR).for EC No., 1773, Rev. 0, "Environmental Fatigue Analysis for V ar Power Station' Revision,1, dated 1 Thc nAl*bkA.

w"t,

-O.*,Ot K4tt l

dci-Aicc5

-At 4

go 2 For the Feedwater Nozzles there are large differences in tI'*.,._

CUFs without the Fen factors shown in shown in Table 4:3.1 of the Vermont Yankee License Renewal Application and those shown in calculation.VY-160-302. Section 2.0 of the calculation on page 4 of 32 states,

-. several of the conservatisms originally used in the original feedwater evaluation (such as grouping of transients) are removed,.

Please explain what conservatisms were removed.

z CD bb The-orig'ln'al 0 isient for the VY Feedwater Nozzle is givenin ttaohment D to GE Specification No. 21A1 i 15

'Reactor.Pressure Vessel', Revision 4, 10121169. It is a single severe,- design* transient intended to envelop all Start-up, Loss of Feedwater.Heater.' Scram, and Shutdown events. it consists of 1500 cycles of':ý

• a!546F to 1OOF step change with 25% feedwater flow, followed by,

• a:step'change to 260F. followed by,

• ' a ramp from 26OF to 546F at 250F per hour along with increasing feedwater flow 25% to 100% flow.

This transient is equivalent to a Startup and Turbine Roll event combination specified on newer BWR plant Thermal Cycle Diagrams.

As described in GE Letter No. G-HB-5-1 24, dated November 5, 1975, the 1500 such events considered in the design fatigue evaluation of the feedwater nozzle exceed the 518 start up. loss of feedwater heater,: scram, and shut down events listed in the

[original] FSAR.

The. CUF.for the feedwater nozzle shown in Table 4,3.1 ofthe Vermont Yankee License Renewal Application is based on the design basis fatigue evaluations factored to account for the effects of the 120% Extended Power Uprate (EPU). Changes in temperatures for EPU are from GE Nuclear Energy Certified Design Specification No. 26A6019. "Reactor Vessel - Extended

Power Uprate', Rev. 1, 8129/03.

z C,

0

~0

'~0

'C The evaluation of EPU effects on the feedwater nozzle and safe end stress and fatigue analysis is contained in VY Engineering Report,.VY-RPT-05-00100, Rev. 0, "Task T0302 Reactor Vessel Integrity-Stre*s..Evaluation EPU Task Report for ER-04-1409'"

Section 3.3..: 1:.of. GE Report for Task 302, identified the value for the feedwate~r*noh.zzle safe end EPU CUF for 40 years = 0.75.

This is the'value shown in Table 4.3.1.

The 0.75 CUF value is bat6d on the original design report. The original design analysis was performed for "toose fit" feedwater spargers where.the annular cold gap between the stainless steel thermal sleeve-and the nozzle safe end was 0.020 inch. The

feed.water spargers and thermal sleeves were replaced in 1978 wilhh.new "interference fit thermal sleeves. The interference fit therftal.sleeves significantly reduce leakage flow past the therm-.al..sleeve into the bore-region of the nozzles. This reduces the heat transfer from the process fluid to the nozzle base metal,

.thereby reducing thermal stresses during system thermal

• transients,.

Subsequent to the GE report, a re-analysis of the feedwater nozzle was performed. SlA Report No. SIR-04-020 Revision 0.,

March 2004. "Updated Stress and Fatigue Analysis for the Vermont Yankee Feedwater Nozzles" documents a revised ASME Ill Stress and Fatigue Analysis for the feedwater nozzle and safe end. This analysis included effects of the interference fit thermal sleeves. The analysis was performed for both the original licensed power and system flow rates using the enveloping design transient, `Startup; Loss of Feedwater Heaters, Scram & Shutdown', from the original Design Specification and for EPU power and flow conditions as modified per the EPU Design Specification. For the Safe End, the 40 year CUP using 1500 cycles of the enveloping transient and including EPU effects = 0.4513 (as compared to the 0.75 factored GE values used in the L-RA*

/

For the Environmentally Assisted Fatigue (EAF) evaluation, a realistic proiection of Desiqn Thermal Transient Cycles and Thev'

-Y Mttx~ a

(

'f F

CL o{ 4M6.

ivmp(ovtA keM&

,t Co~-l-- coe i; C1

vents for 60 years of operation based on the Feedwatek Nozzle Thermal Cycle Diagram from a typical BWR 4 was used. *The

,aIveloping design transient was mapped to the "Turbine Roll &

Increase to Rated Power" transient. Other transients including 4%..

loss of feedwater heaters and scram events were taken directly i froni theFeedwater Nozzle Thermal Cycles Diagram using VY k

1/2*,

sp.ecific EPU design temperatures. The projections for 60 years

.0 0*.

'were based on the number of events for 40 years in the VY I Desgn Speciication, the numbers analyzed in the VY Design Ceftifie6d Stress Report for.VY. Reactor Vessel, and the number of.ycfes experienced biVY in approximately 35 years of

• The-design transients used in the EAF evaluation for the VY Feedwater Nozile are shown in Attachment I to Design Input Record (DIR) for EC No-1773r Rev. 0, 'Environmental Fatigue Anayis for Vermont Yankee Nuclear Power Station" Revision 1, dated 7126107.

z C7 7

For stainless steel components listed in Table..3-10 in Structural Integrity Report SIR-07--132 (VY-16Q0404), please justify that the calculated-CUFen' values6 are conse.rvative.

Conservatisms used in the evaluation of stainless steel components include; Use of design transients vs. actual operation transients Conservative projections for the numbers of events for60 years of operation based on 35 years of VY operating history Use of bounding Fen values for all transients:.

The Fen factors are calculated using NUREG/CR -5704 (ANL-98131). "Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels', April 1999. The Fen values are maximized by using the highest temperatures, minimum strain rates, and conservative dissolved oxygen values at each location.

l

CORRECTED UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.

(Vermont Yankee Nuclear Power Station)

))

)

)

)

)

Docket No. 50-271 -LR ASLBP No. 06-849-03-LR CERTIFICATE OF SERVICE I, Christina Nielsen, hereby certify that copies of NEW ENGLAND COALITION, INC.'S MOTION TO LATE-FILE REBUTTAL TESTIMONY OF ULRICH WITTE in the above-captioned proceeding were served on the persons listed below, by U.S. Mail, first class, postage prepaid, on the 6th of June, 2008.

Administrative Judge Alex S. Karlin, Esq., Chair Atomic Safety and Licensing Board Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: ask2@nrc.gov Administrative Judge William H. Reed 1819 Edgewood Lane Charlottesville, VA 22902 E-mail: whrcville@embargmail.com Office of Commission Appellate Adjudication Mail Stop: O-16C1 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: OCAAmailknrc.gov Administrative Judge Dr. Richard E. Wardwell Atomic Safety and Licensing Board Panel Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: rew(nrc.gov Office of the Secretary Attn: Rulemaking and Adjudications Staff Mail Stop: O-16C1 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: hearingdocket@nrc.gov Sarah Hofmann, Esq.

Director of Public Advocacy Department of Public Service 112 State Street, Drawer 20 Montpelier, VT 05620-2601 E-mail: sarah.hofmann@state.vt.us Lloyd B. Subin, Esq.

Mary C. Baty, Esq.

Office of the General Counsel Mail Stop 0- 15 D21 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: lbs3@nrc.gov; mcbl@nrc.gov Anthony Z. Roisman, Esq.

National Legal Scholars Law Firm 84 East Thetford Road Lyme, NH 03768 E-mail: aroisman2nationallegalscholars.com

Marcia Carpentier, Esq.

Lauren Bregman Atomic Safety and Licensing Board Panel Mail Stop T-3 F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: mxc7@nrc.gov Lauren.Bregman@nrc.gov Peter C. L. Roth, Esq.

Office of the Attorney General 33 Capitol Street Concord, NH 03301 E-mail: Peter.roth@doj.nh.gov David R. Lewis, Esq.

Matias F. Travieso-Diaz Pillsbury Winthrop Shaw Pittman LLP 2300 N Street NW Washington, DC 20037-1128 E-mail: david.lewis@(pillsburylaw.com matias.travieso-diaza~pillsburylaw.com Diane Curran Harmon, Curran, Spielberg, & Eisenberg, L.L.P.

1726 M Street N.W., Suite 600 Washington, D.C. 20036 E-mail: dcurran@harmoncurran.com Jessica A. Bielecki U.S. Nuclear Regulatory Commission Office of the General Counsel Mail Stop: O-15-D21 Washington, D.C. 20555-0001 E-mail: iessica.bielecki(Qnrc.gov Matthew Brock Assistant Attorney General Environmental Protection Division Office of the Attorney General One Ashburton Place, 1 8th Floor Boston,.MA 02108 E-mail: Matthew.Brockastate.ma.us by:

Christina Nielsen, Administrative Assistant Shems Dunkiel Kassel & Saunders, PLLC

Hearing Docket From:

Sent:

To:

Cc:

Subject:

Attachments:

Good afternoon, Christina Nielsen [cnielsen@sdkslaw.com]

Monday, June 09, 2008 11:25 AM Alex Karlin; whrcville@embarqmail.com; OCAAMAIL Resource; Richard Wardwell; Hearing Docket; sarah.hofmann@state.vt.us; Lloyd Subin; Mary Baty; aroisman@nationallegalscholars.com; Marcia Carpentier; Lauren Bregman; peter.roth@doj.nh.gov; david.lewis@pillsburylaw.com; matias.travieso-diaz@pillsburylaw.com; dcurran@harmoncurran.com; Jessica Bielecki; matthew.brock@state.ma.us Andy Raubvogel; 'Karen Tyler'; shadis@prexar.com New England Coalition Corrected Certificate of Service in Docket 50-271 -LR NEC Certificate of Service (corrected).pdf It has come to our attention that the certificate of service filed Friday with New England Coalition's Motion to Late-File Rebuttal Testimony of Ulrich Witte contained an error. The Motion was served by first-class mail only, and not email. Attached please find a corrected certificate of service.

Thank you, Christina Nielsen Christina Nielsen Shems Dunkiel Kassel & Saunders PLLC 91 College Street Burlington, VT 05401 (802) 860-1003 ext. 108 www.sdkslaw.com 1

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