ML080230250

Comments Made by Public Attendees During January 8, 2008 Meeting Between NRC and Entergy/Vermont Yankee

ML080230250
Person / Time
Site:	Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date:	01/11/2008
From:	Hopenfeld J - No Known Affiliation
To:	Rowley J NRC/NRR/ADRO/DLR/RLRB
References
TAC MD2297
Download: ML080230250 (10)

Text

Page 1 of 2 Jonathan Rowley - Re: Questions & Comments - January 8, 2008 meeting with VY/Entergy From:

<Noverflo@aol.com>

To:

<JGR@nrc.gov>

Date:

01/11/2008 12:26 PM

Subject:

Re: Questions & Comments - January 8, 2008 meeting with VY/Entergy CC:

<KTYLER@SDKSLAW.com>, <dlochbaum@ucsusa.org>, <shadis@prexar.com>

Hi Jonathan, thanks for the presentation slides At the end of the subject meeting I submitted to you two hand written questions which I asked the SIA speaker who did not answer my questions. I am resubmitting these questions in case my handwriting was difficult to read.

At the meeting I also wanted to add a comment regarding Union of Concerned Scientists Dave Lochbaum's question relating to oxygen excursions but I was prevented from doing so by Entergy's lawyer who instructed Entergy not to reply to my questions.

My two questions are paraphrased below:

1. What data do you have, to support your statement that the heat transfer coefficients you used to determine the Green functions were conservative (slides 44 and 51) and what do you believe is the degree of conservatism?

2. Does the geometry of the nozzle in slide 9 represent the design geometry of the vessel nozzle or does it represent the "as installed geometry." You have indicated that the validity of the assumption that shear stresses can be neglected is very sensitive to the geometry. It is therefore important to ascertain that the dimensions you used to calculate pipe reactions and peak stress were obtained from the dimensionally correct construction documents and that they also reflect the dimensionally correct geometry at the onset of the extended life period.

Comment Regarding Dave Lochbaum's Question.

A proper NRC reply to Dave.s question is important for the following reasons:

1. The Environmental correction factor, Fen, varies exponentially with the oxygen content.

2.

The experimental fit curve for Fen is based on only one data point at the very low end of the oxygen concentrations.

3. The fatigue usage of each transient is multiplied by its respective Fen. Since the final usage factors are a summation of these products for all transients over the 60 year period one must be assured that the correct oxygen for each transient was employed. This oxygen must reflect the correct concentration at the surface of the given component in question at the time of the transient. It is not clear at all the oxygen concentrations that are measured either daily or bi-weekly at some point in the plant represent the correct value that should be used by the Fen equation in its present form.

file://C:\\temp\\GW}00001.HTM 01/11/2008

Page 2 of 2 I would be grateful if the NRC address the above in the FSAR.

'Sincerely, Joe Hopenfeld Start the year off right. Easy ways to stay in shape in the new year.

file://C:\\temp\\GW}00001.HTM 01/11/2008

NRC FORM 659 F REG,

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I Meeting Date:

Meeting

Title:

01/08/2008 NRC/Entergy Meeting on Metal Fatigue In order to better serve the public, we need to hear from the meeting participants. Please take a few minutes to fill out this feedback form and return it to NRC.

1.

How did you hear about this meeting?

NRC Web Page Radio/TV

] NRC Mailing List r-] Newspaper D Other

2.

Were you able to find supporting information prior to the meeting?

3.

Did the meeting achieve its stated purpose?

4.

Has this meeting helped you with your understanding of the topic?

Yes FD F

A Dý No Somewhat (Please explain below)

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5.

Were the meeting starting time, duration, and location reasonably convenient?

6.

Were you given sufficient opportunity to ask questions or express your views?

D D

7.

Are you satisfied overall with the NRC staff who participated in the meeting?

COMMENTS OR SUGGESTIONS:

A Thank you for answering these questions.

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inue Comments the reverse.

Telephoni'L-No.

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C Check here if you would like a member of NRC staff to contact you.

OMB NO. 3150-0197 Expires: 06/3012009 Public Protection Notification: If a means used to impose an information collection does not display a currently valid OMB control number, the NRC may not conduct or sponsor, and a person is not required to respond to, the information collection.

Please fold on the dotted lines with Business Reply side out, tape the bottom, and mail back to the NRC.

Page 1 of 2 Jonathan Rowley - Re: comments in writing From:

Raymond Shadis <shadis@prexar.com>

To:

"Jonathan Rowley" <JGR@nrc.gov>

Date:

01/10/2008 8:18 AM

Subject:

Re: comments in writing CC:

Dave Lochbaum <dlochbaum@ucsusa.org>, <noverflo@aol.com>, Karen Tyler

<ktyler@sdkslaw.com>, "Ulrich K. Witte" <Ulrich@ulrichwitte.com>

Mr Rowley, Here follows in plain text, my comments. They are also attached in letter form as a MsWord Attachment. Thank you for including them in the record.

Raymond Shadis Raymond Shadis Post Office Box. 76, Edgecomb, Maine 04556 (207) 882-7801 E-mail - shadis@ime.net January 10, 2008 Jonathan Rowley, Project Manager Projects Branch 2 Division of License Renewal Office of Nuclear reactor regulation U.S. Nuclear regulatory Commission Washington, D.C. 20555-0001

Dear Mr. Rowley,

On January 9, 2008, you requested by e-mail that I repeat in writing comments that I made to the Staff following a January 8, 2008 Category I NRC meetingwith Entergy Nuclear Operations to discuss a request for additional information in the license renewal review for Vermont Yankee Nuclear Power Station.

As I was not speaking from notes but simply responding to what I heard regarding stress and aging analysis of VYNPS reactor internal components in the meeting, I cannot reproduce my comments verbatim at this time without referring to sound recordings that were made but which I have yet to access.

I will nonetheless attempt to summarize my comments and hope that helps.

What I observed in the meeting was that the licensee did not appear to offer technically defensible justification of component sample selection, substitution analyses, or assumptions regarding bounding conditions or representative components. It was unclear if stress analyses were based on component design data or on as found condition. It was unclear if reliance on projections from operational history included more than a simple count of transients; for example, were plating programs, shroud cracking, vessel cladding defects, and so on considered.

In any case, it was clear that it would be difficult to record in the review documents the path, choices, and rationale used by the licensee in developing and implementing its analyses. How, for example, would a reviewer validate the licensee's underlying assumptions? How would one assess the licensee's so-called reasonableness standard?

My comment on this can be summed in my concern for the quality of what will be NRC Staff s recorded response to the licensee's analyses and proposed aging management program for reactor internal components. Will NRC Staff provide enough information to allow for technical review of their conclusions and the bases for their conclusions?

file://C:\\temp\\GW}0000l.HTM 01/10/2008

Page 2 of 2 Thank you for your time and attention, Sincerely, Raymond Shadis At 12:03 PM 1/9/2008, you wrote:

Mr. Shadis The NRC appreciates your participation in the meeting on Tuesday (January 8, 2008).

If you would, please put down the comments you made during the comment period of the meeting in writing and send them to me via email. I would liketo make the comments part of the record. The plan is to have all comments made by the public put into an enclosure to the meeting summary.

Thank you!

file://C:\\temp\\GW}00001.HTM 01/10/2008

n51 t Ce BOX 76, EL*ecolnb, aimne 04556 (207) 882-7801 E-matlil - shadisraim.net January 10, 2008 Jonathan Rowley, Project Manager Projects Branch 2 Division of License Renewal Office of Nuclear reactor regulation U.S. Nuclear regulatory Commission Washington, D.C. 20555-0001

Dear Mr. Rowley,

On January 9, 2008, you requested by e-mail that I repeat in writing comments that I made to the Staff following a January 8, 2008 Category I NRC meeting with Entergy Nuclear Operations to discuss a request for additional information in the license renewal review for Vermont Yankee Nuclear Power Station.

As I was not speaking from notes but simply responding to what I heard regarding stress and aging analysis of VYNPS reactor internal components in the meeting, I cannot reproduce my comments verbatim at this time without referring to sound recordings that were made but which I have yet to access.

I will nonetheless attempt to summarize my comments and hope that helps.

What I observed in the meeting was that the licensee did not appear to offer technically defensible justification of component sample selection, substitution analyses, or assumptions regarding' bounding conditions or representative components. It was unclear if stress analyses were based on component design data or on as found condition. It was unclear if reliance on projections from operational history included more than a simple count of transients; for example, were plating programs, shroud cracking, vessel cladding defects, and so on considered.

In any case, it was clear that it would be difficult to record in the review documents the path, choices, and rationale used by the licensee in developing and implementing its analyses. How, for example, would a reviewer validate~the licensee's underlying assumptions? How would one assess the licensee's so-called reasonableness standard?

My comment on this can be summed in my concern for the quality of what will be NRC Staff's recorded response to the licensee's analyses and proposed aging management program for reactor internal components. Will NRC Staff provide enough information to allow for technical review of their conclusions and the bases for their conclusions?

Thank you for your time and attention, Sincerely, Raymond Shadis

NRC FORM 659 NRC ORM65 U.S. NUCLEAR REGULATORY COMMISSION NRC PUBLIC MEETING FEEDBACI Meeting Meeting Date:

01108/2008

Title:

NRClEntergy Meeting on Metal Fatigue In order to better serve the public, we need to hear from the meeting participants. Please take a few minutes to fill out this feedback form and return it to NRC.

1.

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Were the meeting starting time, duration, and location reasonably convenient?

6.

Were you given sufficient opportunity to ask questions or express your views?

7.

Are you satisfied overall with the NRC staff who participated in the meeting?

COMMENTS OR SUGGESTIONS:

N._o Somewhat Yes (Please explain below)

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BUSINESS REPLY MAIL FIRST CLASS MAIL PERMIT NO. 12904 WASHINGTON DC POSTAGE WILL BE PAID BY U.S. NUCLEAR REGULATORY COMMISSION NO POSTAGE NECESSARY IF MAILED IN THE UNITED STATES JONATHAN ROWLEY MAIL STOP O-11F1 OFFICE OF NUCLEAR REACTOR REGULATION U S NUCLEAR REGULATORY COMMISSION WASHINGTON DC 20277-2904 I I III III I III III III I I I I I oil oil I I 111 1111111

~Union of Concerned Scientists Citizens and Scientists for Environmental Solutions January 9, 2008 Jonathan G. Rowley Office of Nuclear Reactor Regulation US Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

QUESTIONS RAISED DURING JANUARY 8,2008, PUBLIC MEETING ON VERMONT YANKEE REACTOR VESSEL NOZZLE FATIGUE

Dear Mr. Rowley:

During public comment period of yesterday's Category I meeting on reactor vessel nozzle fatigue during the proposed license renewal period at Vermont Yankee, I asked three questions. You invited me to submit those questions in writing to ensure they were captured in the NRC's process. It was a fine idea and I am following up on it. Here are my three questions:

1. Early in his presentation, Gary Stevens of Structural Integrity Associates stated that the nozzle fatigue analysis performed for Vermont Yankee included a projection of the water chemistry conditions over the remainder of the plant's operating lifetime. Were the water chemistry conditions assumed in the analysis linked to or more conservative than the technical specification limits?

2.

At slide 17 of the presentation, Entergy's representatives explained that the stress time history for the nozzles had been developed from a thorough accounting of past operational transients.

Were past water chemistry excursions equally captured and accounted for in the analysis?

3. Ken Chang of the NRC staff probed Entergy's representatives at some length regarding the ongoing counting program for operational cycles and the related need to confirm or update the thermal stress calculations. Does a comparable program exist to count water chemistry transients?

Water chemistry is an important factor in nozzle fatigue because it is an input to the F0n term. The Cumulative Usage Factor (CUF) for each nozzle is multiplied by the Fen term.

A very similar issue arose over a decade ago at Nine Mile Point Unit 1 in New York. The issue was reactor vessel core shroud weld cracking rather than reactor vessel nozzle fatigue, but in each case the evaluation of future safety relied heavily on water chemistry assumptions. On April 8, 1997, Niagara Mohawk submitted to the NRC its evaluation (available in the NRC's Public Document Room under Accession No. 9704100242) of the core shroud weld cracking issue. This evaluation relied on a GE analysis of crack growth rates that had assumed water chemistry parameters significantly better than the technical specification limits. On April 17, 1997, UCS submitted a letter (available in the NRC's PDR under Accession No. 9704210098) with the concern that Niagara Mohawk had violated 50.59 by relying Washington Office: 1707 H Street NW Suite 600

• Washington DC 20006-3919

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January 9, 2008 Page 2 of 2 on non-conservative water chemistry parameters that had not been reviewed and approved by the NRC. In short, Nine Mile Point Unit 1 could be operated with water conditions permitted by its technical specifications that would invalidate the basis of its core shroud cracking evaluation. On July 2, 1997, Niagara Mohawk submitted to the NRC a license amendment request (available in the NRC's PDR under Accession No. 9707110350) to incorporate the appropriate water chemistry limits from its core shroud cracking evaluation into the technical specifications.

At this time, I cannot contend that the water chemistry parameters assumed in Entergy's reactor vessel nozzle fatigue assessment are not bound by the water chemistry limits established by Vermont Yankee's technical specifications. Neither can I conclude that the water chemistry assumptions are bound by the technical specification limits.. Unlike Niagara Mohawk, Entergy has not placed the details of its assessment on the docket for the NRC and UCS to independently review.

In asking the questions above, UCS hopes that the NRC staff will ensure the right answers exist before issuing its safety evaluation report on reactor vessel nozzle fatigue.

There was considerable discussiOn between the NRC and Entergy during yesterday's meeting about the future process for monitoring reactor vessel nozzle fatigue. The talk included current practices and future expectations. Absent from this discussion was a vital element - Entergy's legal obligations under 10 CFR 50.71(e) to incorporate information from evaluations performed at the NRC's request into the Vermont Yankee Updated Final Safety Analysis Report (UFSAR). Assuming that Entergy complies with this federal regulation (albeit an unverifiable assumption at this time), a summary of the methodology and results from the reactor vessel nozzle fatigue assessment will be incorporated into applicable sections of the UFSAR. By complying with this federal regulation, the UFSAR will capture and reflect key aspects of the reactor vessel nozzle fatigue assessment, making it more likely that workers five or ten years from now will not inadvertently undermine safety margins.

UCS therefore hopes that the NRC staff will also ensure that Entergy complies with 10 CFRi 50.71(e) by incorporating essential information from the reactor vessel nozzle fatigue assessment into the UFSAR for Vermont Yankee.

Sincerely, David Lochbaum Director, Nuclear Safety Project Washington Office