ML081550244
| ML081550244 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/02/2008 |
| From: | Subin L NRC/OGC |
| To: | Atomic Safety and Licensing Board Panel |
| SECY/RAS | |
| References | |
| 50-271-LR, ASLBP 06-849-03-LR, RAS M-72 | |
| Download: ML081550244 (31) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
ENTERGY NUCLEAR VERMONT YANKEE,
)
Docket No. 50-271-LR LLC, 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 the NRC STAFF REBUTTAL TESTIMONY CONCERNING NEC CONTENTION 4 and NOTICE OF APPERANCE of Susan L. Uttal in the above-captioned proceeding have been served on the following by electronic mail with copies by deposit in the NRCs internal mail system or, as indicated by an asterisk, by electronic mail, with copies by U.S. mail, first class, this 2nd day of June, 2008.
Alex S. Karlin, Chair Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: ask2@nrc.gov Office of the Secretary Attn: Rulemakings and Adjudications Staff Mail Stop: O-16G4 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: hearingdocket@nrc.gov William H. Reed*
Administrative Judge Atomic Safety and Licensing Board 1819 Edgewood Lane Charlottesville, VA 22902 E-mail: whrcville@embarqmail.com Marcia Carpentier, Law Clerk Atomic Safety and Licensing Board Panel Mail Stop: T-3F23 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: mxc7@nrc.gov Richard E. Wardwell Administrative Judge Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: rew@nrc.gov Lauren Bregman, Law Clerk Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Mail Stop: T-3 F23 Washington, D.C. 20555-0001 E-mail: lauren.bregman@nrc.gov Office of Commission Appellate Adjudication Mail Stop: O-16G4 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 E-mail: OCAAmail@nrc.gov Peter C.L. Roth, Esq*
Office of the Attorney General 33 Capitol Street Concord, NH 3301 E-mail: peter.roth@doj.nh.gov Ronald A. Shems, Esq.*
Karen Tyler, Esq.
Shems Dunkiel Kassel & Saunders, PLLC 91 College Street Burlington, VT 05401 E-mail: rshems@sdkslaw.com Ktyler@sdkslaw.com Anthony Z. Roisman, Esq.*
National Legal Scholars Law Firm 84 East Thetford Rd.
Lyme, NH 03768 E-mail: aroisman@nationallegalscholars.com David R. Lewis, Esq.*
Matias F. Travieso-Diaz, Esq Elina Teplinsky, Esq Blake J. Nelson, Esq Pillsbury Winthrop Shaw Pittman LLP 2300 N Street, NW Washington, DC 20037-1128 E-mail: david.lewis@pillsburylaw.com matias.travieso-diaz@pillsburylaw.com elina.teplinsky@pillsburylaw.com blake.nelson@pillsburylaw.com 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 Diane Curran*
Harmon, Curran, Spielberg, & Eisenberg, LLP 1726 M Street N.W., Suite 600 Washington, D.C. 20036 E-mail: dcurran@harmoncurran.com James R. Milkey*
Assistant Attorney General, Chief Environmental Protection Division Office of the Attorney General One Ashburton Place, 18th Floor Boston, MA 02108 E-mail: jim.milkey@state.ma.us
/RA/
Lloyd B. Subin Counsel for NRC Staff
June 2, 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)
)
NRC STAFF REBUTTAL TESTIMONY CONCERNING NEC CONTENTION 4 INTRODUCTION Pursuant to 10 C.F.R. § 2.1207(a)(2) and the Initial Scheduling Order (Nov. 17, 2006)
(unpublished), the staff of the U.S. Nuclear Regulatory Commission (Staff) hereby files rebuttal testimony of Kaihwa R. Hsu, a supporting affidavit and exhibits in response to New England Coalition, Inc.s (NEC) initial statement of position and testimony.1 For the reasons set forth in the rebuttal testimony, the Staff again submits that NECs challenge to the Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, Inc.s (Entergy) application for renewal of the Vermont Yankee operating license cannot be sustained.
DISCUSSION The issue in this rebuttal testimony addresses NEC contention 4, which is a challenge to Entergys plans for aging management of plant components subject to [flow-accelerated corrosion] FAC. LBP-06-20, 64 NRC 131, 194 (2006). The Staff maintains its position that 1 New England Coalition, Inc. Initial Statement of Position, Direct Testimony and Exhibits (Apr. 28, 2008).
Entergys program for monitoring FAC is adequate. See NRC Staff Initial Statement of Position on NEC Contentions 2A, 2B, 3, and 4 (May 13, 2008), at 21.
Respectfully submitted,
/RA/
Lloyd B. Subin Counsel for NRC Staff Dated at Rockville, Maryland this 2nd day of June, 2008
June 2,2008 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of 1
)
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)
)
NRC STAFF REBUTTAL TESTIMONY OF KAIHWA R. HSU CONCERNING NEC CONTENTION 4 Q1.
Please state your name, occupation, and by whom you are employed.
Al.
My name is Kaihwa R. Hsu ("Hsu"). I am employed by the US Nuclear Regulatory Commission ("NRC") as a senior mechanical engineer in the Engineering Division in the Office of New Reactors. Previously I was employed as a materials engineer in the office of Nuclear Reactor Regulation Division of License Renewal. A statement of my professional qualifications was attached to the staff's "Affidavit of Kaihwa R. Hsu, Jonathan G. Rowley, and Thomas G. Scarbrough Concerning NEC Contention 3 (Steam Dryer)," filed May 13, 2008.
Q2.
What is the purpose of this testimony?
A2.
The purpose of this rebuttal testimony is to address the pre-filed written testimony and exhibits of Dr. Hausler, Mr. Witte, and Dr. Hopenfeld regarding New England Coalition, Inc.'s ("NEC") Contention 4, which were submitted on behalf of NEC on April 28, Q3.
In his "Discussion of the Empirical Modeling of Flow-Induced Localized Corrosion of Steel under High Shear Stress" ("NEC-RH-03"), Dr. Hausler expressed his concern for uncertainties in the methodology of ultrasonic thickness ("UT") measurements due to: 1) "[tlhe inherent variability of the instrument with which the measurements are being made"; and 2)
"[tlhe inherent difficulty of placing the handheld UT probe at exactly the same location for repeat measurements one-and-a-half to two years apart." Exhibit NEC-RH-03 at Appx. A. Do you agree that these are areas of concern?
A3.
No, I do not agree that these are areas of concern for UT measurements. First, inherent variability of the instrument is not a concern because in the nuclear industry, standards require that UT instruments be properly calibrated and UT technicians be trained to perform UT measurements. In addition, information regarding the accuracy of handheld UT probes is provided to the user by the manufacturer. Recent UT wall thickness technology has demonstrated that UT measurements are capable of attaining measurement accuracy for a high frequency UT transducer of +I- 0.01 mm, which is significantly lower than +I-1 Oh to 2% of wall thickness claimed by Dr. Hausler, see id.
Second, any inherent difficulty in placing the probe in the same location for temporally separate repeat measurements has been eliminated because the plant has painted a permanent grid on the outside surface of the pipes. This permanent grid provides assurance that the probe will be placed in the same location for repeat measurements.
Q4.
In NEC-RH-03, Dr. Hausler concluded "that the absolute minimum number of thickness measurements required for reasonably accurate prediction of failure is three, if an assessment of the confidence limits of the resulting trend is to be made." Id. at Appx. A. Do you agree with this statement?
A4.
The Staff agrees that three measurements are required. In order for the aging management program to be consistent with GALL Report's recommendation, "limited baseline inspections to determine the extent of thinning at these locations" can be accomplished by performing two measurements, and "follow-up inspections to confirm the predictions" can be accomplished by performing a third measurement. IVRC Staff Initial Statement of Position on
hIEC contentions 2A1 2B1 3, and 4 (May 13, 2008), Exhibit 7 at XI M-61.
Although Dr. Hausler stated "that at least two measurements are needed to determine the rate of deterioration," he concludes that a minimum of three measurements are required for a reasonably accurate prediction of failure, if an assessment of the confidence limits is required.
Exhibit NEC-RH-03 at Appx. A. This notion is based on his concern of uncertainty in the methodology of measurement. As described above in A3, the uncertainty has been essentially eliminated due to the accuracy of the nieasuring equipment. Therefore, the uncertainty concerns are not valid.
Q5.
Mr. Witte, in his report regarding proposed aging management programs for flow-accelerated corrosion ("FAC"), stated that a concern "regarding deficiencies in implementation of the program brings into question the results of FAC inspection during RFO 25 and RFO 26...." Exhibit NEC-UW-03 at 2. Do you agree with this statement?
A5.
No, I do not agree with this statement. There is no basis to question the results of the FAC inspection during RFO 25 and RFO 26. The results of these FAC inspections are actual UT wall thickness measurements that are independent of the CHECWORKS software or model update. CHECWORKS is used to manage and evaluate the actual wall thickness data to help trendjpredict pipe failure due to FAC.
Q6.
In his report regarding proposed aging management programs for FAC, Mr.
W itte stated that
[wlith the exception of VY1s [Vermont Yankee's] strength in reactively replacing piping or components with FAC-resistant material during repairs or maintenance, the program itself was not effective as a predictive modeling tool. Simply stated, once something ruptured or was found to be outside of its design margin, it was replaced in a reactive management approach. Proactive management of the program to predict failures has been inadequate in the FAC Program....
Exhibit NEC-UW-03 at 7 (emphasis in original). Do you agree with this statement?
A6.
No, I do not agree with this statement. The Staff's position is that a FAC
program that performs inspections successfully, identifies critical FAC-susceptible components, and allows for replacement of piping and components with FAC-resistant material to prevent failure due to FAC, is effective.
Q7.
In "Review of License Renewal Application for Vermont Yankee Nuclear Power Station: Program for Management of Flow-Accelerated Corrosion," Dr. Hopenfeld stated that
[alccording to NUREG/CR-6936, Probabilities of Failure and Uncertainty Estimate Information for Passive Components - a Literature Review (May 2007) at Table 5.15, there were 250 through-wall pipe failures from FAC in BWRs and PWRs between 1988 and 2005, compared to 183 failures that occurred between 1976 and 1987. On a yearly basis, this represents a reduction of 2 failures per year during 1988-2005 period compared to the previous period, disregarding the number of reactors and their age. Since the CCC codes were introduced in 1987, one could attribute the 10% reduction to the CCC codes.
Exhibit IVEC-JH-36 at 9. Do you agree with this statement?
A7.
No, I do not agree with this statement. Dr. Hopenfeld's conclusions and use of the data is incorrect. NLIREGJCR-6936 does not support his testimony.
Dr. Hopenfeld relies on data reported in NUREGJCR-6936, which was extracted from Appendix D of NLIREG-1829 (Exhibit A, Estimating Loss-of-Coolant Accident (LOCA)
Frequencies Through the Elicitation Process, (Apr. 2008)). However, the service data extracted from NUREG-1829, as recorded in PIPExp. (a proprietary database for pipe failure experience),
included applicable worldwide service experience, not just US nuclear industry data.
NUREG-1829 states that 0.3.2 2 FW Piping Service Experience - Figures D.9 and D.10 summarize the service experience with FW piping. With respect to plant designed by General Electric, the Code Class I portion of BWR carbon steel feedwater piping has performed well in the field.
There are no reported leaks in medium-or large diameter RCPB piping. Foreign plants have experienced (and in some cases, continue to experience) thermal fatigue damage due to thermal mixing and stratification. In fact, 80% of the degradation of the RCPB portions of FW piping has occurred in foreign plants with a piping system design that differs from that of U.S. BWR plants.
The U.S. service experience includes a few instances of non-through wall cracking of FW nozzle-to-safe-end (bimetallic) welds. The root cause of the cracking is attributed to weld defects from original construction. As documented in Information Notice 92-35 [D. 191, Susquehanna Unit I has experienced flow-accelerated corrosion damage about 250 mm (10 inches) from a weld connecting NPS12 piping to a 20-inch by 12-inch reducing tee.
There have been no reported flaws in any U.S. plant beyond T = 15 years of operation.
Id. at D-22 (emphasis added). Contrary to Dr. Hopenfeld's statement, this information indicates that the frequency of FAC related events at US nuclear plants has declined significantly and there have been no FAC-related injuries at US nuclear plants since improved FAC programs (e.g., EPRl guidelines, CHECWORKS) have been used in the US nuclear industry.
Q8.
Mr. Witte stated in his report regarding proposed aging management programs for FAC that "VY is the first plant modified to achieve Constant Pressure Power Up-rate to 120% power and only one other plant out of the fleet of 104 was licensed to 120% increase in power in one step. Given the uniqueness of the design of VY's power up-rate, CHECWORKS has little industry benchmarking data, and is of marginal use." Exhibit NEC-UW-03 at 8. Is this a correct statement?
A8.
No, this is not a correct statement. There is enough industry data regarding BWR extended uprates to demonstrate CHECWORKS benchmarking. For example, Dresden Units 2 & 3 extended 17% of their power from 2527 MWt to 2957 MWt; Quad Cities Units 1 & 2 extended 17.8% of their power from 251 1 MWt to 2957 MWt; and Clinton extended 20% of its power.from 2894 MWt to 3473 MWt. Exhibit 6, Approved Application for Power Uprates, http://www.nrc.gov/reactors/operating/licensing/power-uprates/approved applications.html (last visited June 2, 2008). In comparison, VY extended 20% of its power from 1593 MWt to 1912 MWt. Id. The original power levels of Dresden, Quad Cities, and Clinton are much greater than VY's extended power level. The Staff's position is that the above plants are comparable to VY, and therefore, there is enough industry data to demonstrate benchmarking for extended power uprates.
CHECWORKS was designed for FAC prediction at the power levels at which the plant is being operated. The program does not recognize whether the power levels have been uprated or remain at lower levels, the use of the program remains unchanged. Data from plants that have a power level much higher than VY's extended power level have already been considered in the CHECWORKS development. Therefore, it is not accurate to say CHECWORKS only has a marginal use.
Q9.
In "Review of License Renewal Application for Vermont Yankee Nuclear Power Station," Dr. Hopenfeld asserts that "[tlo account for local turbulence..., the grid should be kept to below 1" x 1" inch." Exhibit NEC-JH-36 at 15. Do you agree with this statement?
A9.
No, I do not agree with this statement. If the grid is kept to below 1 inch by 1 inch, then inspection of more than 6,000 points for a 30 inch diameter, long radius elbow would be required. This is not necessary or feasible. The FAC failure cases have demonstrated that a large bore pipe failure occurs over much more than a 1 inch x 1 inch area.
Q10. Dr. Hopenfeld has previously stated that "[ilt is important to realize that wall thinning rate from FAC is not necessarily constant with time, and therefore a considerable number of cycles are needed to establish the FAC rate on a given component at a particular plant." Exhibit C, Petition for Leave to Intervene, Request for Hearing and Contentions (May 26, 2006) at Exhibit 7 fl 24 (ADAMS ML061640032). In contrast, NEC's pre-filed exhibit, NEC-,IH-37, indicates that there is a linear relationship between FAC degradation and time. Do you agree with Dr. Hopenfeld's previous statement or with the information submitted in NEC Exhibit NEC-JH-37?
A10.
I agree with the information in NEC-JH-37. The laboratory data and plant data shown in Figures 4 and 5 of NEC-JH-37 clearly demonstrate that the FAC degradation is linear with time. This is the basis for trendinglpredicting FAC failure date. This algorithm has been adopted for all plants, including those plants not using CHECWORKS.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
1 ENTERGY NUCLEAR VERMONT YANKEE, LLC )
Docket No. 50-271 -LR AND ENTERGY NUCLEAR OPERATIONS, INC. )
1 ASLBP NO. 06-849-03-LR (Vermont Yankee Nuclear Power Station) 1 AFFIDAVIT OF KAIHWA R. HSU I, Kaihwa R. Hsu, do hereby declare under penalty of perjury that my statements in the foregoing testimony are true and correct to the best of my knowledge and belief.
KAIHWA R. HSU Executed at Rockville, MD this 2nd day of June, 2008
EXHIBIT A
United States Nuclear Regulatory Commission Protecting People and the Environment Estimating Loss-of-Coolant Accident (LOCA)
Frequencies Through the Elicitation Process Main Report Office of Nuclear Regulatory Research
United States Nuclear Regulatory Commission Protecting People and the Environment Estimating Loss-of-Coolant Accident (LOCA)
Frequencies Through the Elicitation Process Main Report Manuscript Completed: March 2008 Date Published: April 2008 Prepared by R. Tregoning (NRC), L. Abramson (NRC)
P. Scott (Battelle-Columbus)
A. Csontos, NRC Project Manager Office of Nuclear Regulatory Research
United States Nuclear Regulatory Commission Protecting Peopk and the Environment Estimating Loss-of-Coolant Accident (LOCA)
Frequencies Through the Elicitation Process Appendices A through M Office of Nuclear Regulatory Research
United States Nuclear Regulatory Commission Protecting People and the Environment Estimating Loss-of-Coolant Accident (LOCA)
Frequencies Through the Elicitation Process Appendices A through M Manuscript Completed: March 2008 Date Published: April 2008 Prepared by R. Tregoning (NRC), L. Abramson (NRC)
P. Scott (Battelle-Columbus)
A. Csontos, NRC Project Manager Office of Nuclear Reg~~latory Research
APPENDIX D PIPING BASE CASE RESULTS OF BENGT LYDELL An Application of the Parametric Attribute-Influence Methodology to Determine Loss of Coolant Accident (LOCA) Frequency Distributions Report No. 2 to the NRC Expert Panel on LOCA Frequency Distributions Prepared for U.S. Nuclear Regulatory Commission Washington (DC)
June 2004
0.3.2.2 FW Piping Sewice Experience - Figures D.9 and D. 10 summarize the service experience with FW piping. With respect to plant designed by General Electric, the Code Class 1 portion of BWR carbon steel feedwater piping has performed well in the field. There are no reported leaks in medium-or large-diameter RCPB piping. Foreign plants have experienced (and in some cases, continue to experience) thermal fatigue damage due to thermal mixing and stratification. In fact, 80% of the degradation of the RCPB portions of FW piping has occurred in foreign plants with a piping system design that differs from that of U.S. BWR plants.
The U.S. service experience includes a few instances of non-through wall cracking of FW nozzle-to-safe-end (bimetallic) welds. The root cause of the cracking is attributed to weld defects from original construction. As documented in Information Notice 92-35 [D. 191, Susquehanna Unit I has experienced flow-accelerated corrosion damage about 250 mm (10 inches) from a weld connecting NPS 12 piping to a 20-inch by 12-inch reducing tee. There have been no reported flaws in any US, plant beyond T = 15 years of operation.
b 6 0 O
qb,46 QO O%
1" 1%,db,d6,dO,aP,$@,a@,9B
,a% $9,a
,a
,a
,o,o
\\a
,929 Figure D.9 Service Experience with FW Piping (i)
0 2
3 6
1 4, 2 + * ~, 4, +,, !, 3, p, $ $
Years of Operation Figure D.10 Service Experience with FW Piping (ii)
D.3.3 Review of PWR-Specific Piping Service Experience Limited to the PWR Base Case systems, this section summarizes the service experience with Code Class 1 piping. The results of this review are input to the pipe failure rate estimation.
0.3.3.1 RC & HPI/NMU Piping Service Experience - There have only been a limited number of events involving through-wall cracks in the large-diameter RC piping and the Class 1 portion of SVCV piping.
Evidence of axial primary water stress corrosion cracking (PWSCC) in the bimetallic safe-end to RPV nozzle welds of the RC-HL piping has been reported at Ringhals [D.20] and V.C. Summer [D.21].
During an eight-year period, the now decommissioned Trojan nuclear power plant experienced pressurizer surge line movement, which was attributed to thermal stratification [D.22]. In response, the NRC issued Bulletin 88-1 1 in December of 1988 [D.23] requesting that licensees perform visual inspections of the pressurizer surge line at the first available cold shutdown. Purpose of the inspections was to determine presence of any "gross discernible distress or structural damage in the entire pressurizer surge line, including piping, pipe supports, pipe whip restraints, and anchor bolts."
The current version (June 2004) of the PIPExp database includes four records associated with degradation of pressurizer surge lines:
Record # 19849; during the Three Mile Island-1 2003 Refueling Outage (18-Oct-2003 to 3-Dec-2003), a UT examination found an axial flaw about 13 rnm (0.5-inch) deep in the surge line nozzle-to-safe end interface in dissimilar metal weld No. SR0010BM. This weld connects a 10-inch Schedule 140, carbon steel nozzle to stainless steel safe end.
Record # 19736; in November 2002 during UT examination of RC piping in the Belgian plant Tihange-2 (a 900 MWe series plant designed by Framatome), code rejectable indications were
EXHIBIT B
NRC: Approved Applications for Power Uprates Page 1 of 4 Index I Site Map I FAQ I Facility Info ( Reading Rrn I New I Help I Glossary I Contact Us IGaoSleCur(om mj search options w
v About NRC Nuclear Nuclear Radioactive Nuclear Public Meetings Reactors Materials Waste Securitv
& Involvement Power Uprates Home > Nuclear Reactors > Operating Reactors > Licensing > Power Uprates > Approved Applications Approved Applications Pending Applications Approved Applications for Power Uprates Expected Applications The following power uprates have been reviewed and accepted by the NRC. The licenses for the following plants have been amended to reflect the increase in power level shown in the table.
(TYPE -- S = Stretch; E = Extended; MU = Measurement Uncertainty Recapture)
The following links on this page are to documents in our Agencywide Documents Access and Management System (ADAMS). ADAMS documents are provided in either Adobe Portable Document Format (PDF) or Tagged Image File Format (TIFF). To obtain free viewers for displaying these formats, see our Plugins, Viewers, and Other Tools. I f you have problems with viewing or printing documents from ADAMS, please contact the Public Document Room staff.
NO.
1 2
3 4
5 6
7 8
9 10 11 12 13 14 15 16 17 18 19 20 21 22 PLANT Calvert Cliffs 1 Calvert Cliffs 2 Millstone 2 H. B. Robinson Fort Calhoun St. Lucie 1 St. Lucie 2 Duane Arnold Salem 1 IVorth Anna 1 NorthAnna2 Callaway TMI-1 Fermi 2 Vogtle 1 Vogtle 2 Wolf Creek Susquehanna 2 Peach Bottom 2 Limerick 2 Susquehanna 1 Nine Mile Point 2 OIo UPRATE 5.5 5.5 5
4.5 5.6 5.5 5.5 4.1 2
4.2 4.2 4.5 1.3 4
4.5 4.5 4.5 4.5 5
5 4.5 4.3 MWt 140 140 140 100 80 140 140 6 5 73 118 118 154 3 3 137 154 154 154 148 165 165 148 144 DATE APPROVED 09/26/77 11/08/77 06/25/79 06/29/79 081 15/80 11/23/81 03/01/85 03/27/85 02/06/86 08/25/86 08/25/86 03130188 07/26/88 09/09/92 03/22/93 03/22/93 11/10/93 0411 1/94 10/18/94 02/16/95 02/22/95 04/28/95 TYPE S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S ACCESSION #
7907240100*
7907180064*
8008280223*
ML013530273 ML013600080 ML021890435 IYL011660249 ML013460131 ML013460131 ML021650524 ML003779786 ML020720520 ML012330056 ML012330056 ML022030519 ML010170334 IY LO1 1490143 ML011560773 9503070354*
9505090259*
NRC: Approved Applications for Power Uprates Page 2 of 4 23 24 25 26 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 53 54 55 56 57 58 59 60 61 62 WNP-2 Peach Bottom 3 Surry 1 Surry 2 Hatch 1 Hatch 2 Limerick 1 V. C. Summer Palo Verde 1 Palo Verde 2 Palo Verde 3 Turkey Point3 Turkey Point 4 Brunswick 1 Brunswick 2 Fitzpatrick Farley 1 Farley 2 Browns Ferry 2 Browns Ferry 3 Monticello Hatch 1 Hatch 2 Comanche Peak 2 LaSalle 1 LaSalle 2 Perry River Bend Diablo Canyon 1 Watts Bar Byron 1 Byron 2 Braidwood 1 Braidwood 2 Salem 1 Salem 2 San Onofre 2 San Onofre 3 Susquehanna 1 Susquehanna 2 4.9 5
4.3 4.3 5
5 5
4.5 2
2 2
4.5 4.5 5
5 4
5 5
5 5
6.3 8
8 1
5 5
5 5
2 1.4 5
5 5
5 1.4 1.4 1.4 1.4 1.4 1.4 163 165 105 105 122 122 165 125 76 76 76 100 100 122 122 100 138 138 164 164 105 205 205 34 166 166 178 145 73 48 170 170 170 170 48 48 48 48 48 48 05/02/95 07/18/95 08/03/95 08/03/95 0813 1/95 0813 1/95 01/24/96 041 12/96 05/23/96 05/23/96 05/23/96 09/26/96 09/26/96 1110 1/96 1110 1/96 12/06/96 04/29/98 04/29/98 09/08/98 09/08/98 091 16/98 10/22/98 10/22/98 09/30/99 05/09/00 05/09/00 06/01/00 10/06/00 10/26/00 01/19/01 05/04/0 1 05/04/01 05/04/01 05/04/01 05/25/01 05/25/01 0701610 1 07/06/01 07/06/01 070/6/0 1 S
S S
S S
S S
S S
S S
S S
S S
S S
S S
S E
E E
MU S
S S
S S
MU S
S S
S MU MU MU MU MU MU ML022120154 ML021580312 ML012710328 ML012710328 ML013020073 ML013020073 ML011560244 ML012320013 ML021710572 ML021710572 ML021710572 ML013390234 ML013390234 9611070136*
9611070136*
9612180303*
ML012140259 ML012140259 ML042670045 ML042670045 ML020920138 ML013030084 ML013030084 ML021820306 ML003716743 ML003716743 ML003724441 ML003762072 ML003764792 ML010260074 ML011420274 ML011420274 ML011420274 ML011420274 ML011350051 ML011350051 ML012180231 ML012180231 ML011760551 ML011760551
NRC: Approved Applications for Power Uprates Page 3 of 4 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 79 80 8 1 82 83 84 85 86 87 88 89 90 9 1 92 93 94 95 96 97 98 99 100 101 102 Hope Creek Beaver Valley 1 Beaver Valley 2 Shearon Harris Comanche Peak 1 Comanche Peak 2 Duane Arnold Dresden 2 Dresden 3 Quad Cities 1 Quad Cities 2 Waterford 3 Clinton South Texas 1 South Texas 2 ANO-2 Sequoyah 1 Sequoyah 2 Brunswick 1 Brunswick 2 Grand Gulf H. 8. Robinson Peach Bottom 2 Peach Bottom 3 Indian Point 3 Point Beach 1 Point Beach 2 Crystal River 3 D.C. Cook 1 River Bend D.C. Cook 2 Pilgrim Indian Point 2 Kewaunee Hatch 1 Hatch 2 Palo Verde 2 Kewaunee Palisades Indian Point 2 1.4 1.4 1.4 4.5 1.4 0.4 15.3 17 17 17.8 17.8 1.5 20 1.4 1.4 7.5 1.3 1.3 15 15 1.7 1.7 1.62 1.62 1.4 1.4 1.4 0.9 1.66 1.7 1.66 1.5 1.4 1.4 1.5 1.5 2.9 6
1.4 3.26 46 3 7 3 7 138 4 7 13 248 430 430 446 446 5 1 579 5 3 5 3 21 1 44 44 365 365 6 5 3 9 56 56 42.4 21.5 21.5 24 54 5 2 5 7 3 0 43 2 3 41 4 1 114 99 35.4 101.6 ML011910345 ML012490569 ML012490569 ML012830516 ML012550246 ML012550246 ML013050342 ML013540187 ML013540187 ML013540222 ML013540222 ML020910734 ML021650543 ML020800263 ML020800263 ML021130826 ML021220060 ML021220060 ML021440346 lY LO21440346 ML022630304 ML023100365 ML031000317 ML031000317 ML023290636 ML023370133 ML023370133 IY LO23380800 ML023470126 ML030340294 ML030990129 ML031220007 ML031420375 ML031530734 ML032590944 ML032590944 ML032720538 ML040430633 ML040970622 ML042960007 07/30/01 09/24/01 09/24/0 1 10/12/01 101 1210 1 10/12/01 11/06/01 12/21/01 12/21/01 12/21/01 12/21/01 03/29/02 04/05/02 04/12/02 04/12/02 04/24/02 04/30/02
' 04/30/02 05/31/02 05/31/02 10/10/02 11/05/02 11/22/02 11/22/02 11/26/02 11/29/02 11/29/02 12/04/02 12/20/02 01/31/03 05/02/03 05/09/03 05/22/03 07/08/03 09/23/03 09/23/03 09/29/03 02/27/04 06/23/04 10/28/04 MU MU MU S
MU MU E
E E
E E
MU E
MU MU E
MU MU E
E MU MU MU MU MU MU MU S
MU MU MU MU MU MU MU MU S
S MU S
NRC: Approved Applications for Power Uprates Page 4 of 4 VermontYankee 1
20 1
319 1
03/02/06 1
E I ML060050024 Waterford Palo Verde 1 Palo Verde 3 Sea brook I
1.7 1
61 1
05/22/06 I MU I ML061430044 S
S 02/28/05 03/24/05 Ginna 1
16.8 0711 1/06 I E
( ML061380133 ML050140453 ML050600380 176 148.6 Sea brook Indian Point 3 8.0 2.9 2.9 I Beaver Valley 1 I 8
07/19/06 1
E 1 ML061720274 5.2 4.85 112 Beaver Valley 2 1
8 211 07/19/06 275 114 114 I
116 I Susquehanna 2 13 463 01/30/08 E
041 1 510 5 11/16/05 11/16/05 114 115 Capacity Recapture Power Uprates for Provisional Operating License Plants are not included in this table. These are Haddam Neck uprate of 24% in 1969, Oyster Creek uprate of 14% in 1971, Palisades uprate of 15% in 1977, Ginna uprate of 17% in 1984, Maine Yankee uprate of 10% in 1989, and Indian Point 2 Uprate of 11% in 1990.
E S
S Crystal River 3 Susquehanna 1 118 N0TE:'rhe NRC staff approved an MUR power uprate for Fort Calhoun on January 16, 2004, which authorized an increase in the licensed thermal power limit to 1,524 megawatts-thermal.
The Omaha Public Power District was subsequently informed by Westinghouse that the potential instrument inaccuracies in the Advanced Measurement and Analysis Group (AMAG) ultrasonic flow meter would not allow implementation of the MUR power uprate at Fort Calhoun. As a result, on May 7, 2004, prior to implementation of the MUR power uprate, the Omaha Public Power District submitted an exigent license amendment request to return Fort Calhoun's licensed thermal power limit to 1,500 megawatts-thermal, the pre-MUR level. On May 14, 2004, the NRC staff approved this license amendment.
ML051030068 ML053130286 ML053130286 Privacy Policy I Site Disclaimer Wednesday, May 21, 2008 1.6 13
- Documents can be requested from the Public Document Room Vogtle 2 Total MWt Total MWe 4 1 463 1.7 12/26/07 01/30/08 60.6 15788.2 5263 MU E
02/27/08 ML073610197 MU ML080350345
EXHIBIT C
12A5 SHEMS DUNKIEL KASSEL & SAUNDERS P L L C RONALD A. SHEMS BRIAN S. DUNKIEL*
JOHN B. KASSEL MARK A. SAUNDERS GEOFFREY H. HAND KAREN L. TYLER ASSOCIATE ATTORNEYS ANDREW N. RAUBVOGEL EILEEN 1. ELLIOTT OF COUNSEL DOCKETED USNRC May 30, 2006 (3:30pm)
May 26, 2006 OFFICE OF SECRETARY RULEMAKINGS AND ADJUDICATIONS STAFF Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555-000 1 Attn: Rulemaking and Adjudications Staff Re:
In the matter of ENTERGY NUCLEAR VERMONT YANKEE, LLC and ENThRGY NUCLEAR OPERATIONS, INC., Vermont Yankee Nuclear Power Station License Renewal Application, Docket No. 50-271
Dear Sir or Madam:
Please find enclosed for filing in the above stated matter New England Coalition'ss Petition for Leave to Intervene, Request for Hearing, and Contentions; and the Notice of Appearance on behalf of New England Coalition of Shems Dunkiel Kassel & Saunders PLLC by attorneys Ronald A. Shems, and Karen Tyler.
Thank you for your attention to this matter.
Sincerely, Karen Tyler Enclosures cc: see attached Certificate of Service 9 1 COLLEGE STREET BURLINGTON.
VERMONT 0540 1
802 8 60 1 003 E102 / 860 I1208 - www.sdkslaw.com T--r,77 1 0, +ý
%C / -0'31 SLac1 'I o
'Also admitted in the District of Columbi
EXHIBIT 7 UNITED STATES NUCLEAR REGULATORY COMMISSION In the maqtter Of ENTERGY NUCLEAR VERMONT YANKEE, LLC
)
and ENTERGY NUCLEAR OPERATIONS, INC.
)
NO. 50-27 1 Vermont Yankee Nuclear Power Station)
DECLARATION OF DR. JORAM HOPENFELD
- 1.
My name is Dr. Joram Hopenfeld. The New England Coalition (NEC) has retained me as an expert witness in proceedings concerning the application of Entergy Nuclear Operations, Inc. ("Entergy") to renew its operating license for Vermont Yankee Nuclear Power Station ("Vermont Yankee") for twenty years beyond the current expiration date of March 21, 2012.
- 2.
I am a mechanical engineer and hold a doctorate in engineering. I have 45 years of professional experience in the fields of instrumentation, design, project management, and nuclear safety, including 18 years in the employ of the U.S. Nuclear Regulatory Commission. My curriculum vitae is attached to this declaration as Attachment A.
- 3.
I have reviewed Entergy's License Renewal Application, and such publicly available documents as are relevant to the subjects of my declaration.
CONTENTION TWO
- 4.
Paragraphs 4 - 14 of this declaration concern NEC's "Contention Two." I refer to the following documents:
- 22.
The Entergy program to manage the effects of Flow-Accelerated Corrosion (FAG) is based on NUREG 1801 § XI.M17 and EPRI Report NSAC-202L-R2. License Renewal Application Table 3.4.11¶3.4.1-29, and Appendix B § B.1. 13. These guidance documents recommend use of a computer code, CHEC WORKS, to recommend the scope and frequency of in-service inspections. It can be reasonably deduced that Entergy proposes to use the CHECWORKS code to manage FAG during the new license term.
- 23.
Because Entergy has recently increased the operating power level of its plant by 20%, CHEC WORKS would require additional inputs before it can be used at the VY plant as an adequate FAG management tool. Consequently, the proposed program, as presented in the Entergy Application, will not be valid throughout the entire period of the extended plant operation.
- 24.
The theoretical basis of FAC is not completely understood; however, it is well established that turbulence intensity, steam quality, material compositions, oxygen content and coolant pH are the main variables that affect FAG. The CHECWORKS computer code is not a mechanistic code; it is an empirical code that must be updated continuously with plant-specific data. Inspection results are routinely used as inputs to the code. The code can be used to predict pipe wall thinning as long as plant parameters (velocity, coolant chemistry, etc.) do not change drastically and the data has been collected for a long period of time. It is important to realize that wall thinning rate from FAG is not necessarily constant with time, and therefore a considerable number of cycles are needed to establish the FAC rate on a given component at a particular plant. Since Vermont Yankee has recently increased the coolant flow rate by 20%, which also
significantly accelerates local wall thinning, it would take at least 10- 15 years before CHEC WORKS can be benchmarked with the Vermont Yankee inspection data.
- 25.
The inability to reliably predict wall thinning from FAC has been very costly. In 1986, a feed water pipe elbow ruptured at the Surry Nuclear plant. There were several fatalities and the reactor was down for several months. The accident resulted from severe pipe walls thinning due to FAC (References a & b). In 1991 and in 1993, the feed ring and the J tubes at San Onofre's steam generators (References c and d) failed from FAC.
In 1997, extraction steam piping ruptured at the Fort Calhoun Station (Reference e). In July 2004, several workers were killed at the Mihama nuclear power plant due to FAG in the secondary loop (Reference f).
- 26.
The above is only a partial list of the failures that occurred from FAG in nuclear plants. This list alone, however, is sufficient to demonstrate that CHEC WORKS (developed in 1987) has not been successful in averting major catastrophes and costly outages. The prediction of FAG is an art not a science and must be obtained empirically and with expert engineering judgment. The above plant experience indicates that a lack of proper FAG controls can lead to very senious consequences.
- 27.
The key to a valid FAG program is the ability to adequately specify the frequency of inspections. CHEGWORKS cannot be used to provide adequate guidance regarding inspe ction frequencies. Therefore, Entergy cannot assure the. public that the minimum wall thickness of carbon steel piping and valve components will not be reduced by FAG to below the ASME code limits during the period of extended operation.
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
ENTERGY NUCLEAR VERMONT
)
Docket Nos. 50-271-LR YANKEE, LLC, and ENTERGY
)
NUCLEAR OPERATIONS, INC
)
ASLBP No. 06-849-03-LR
)
(Vermont Yankee Nuclear Power Station) )
NOTICE OF APPEARANCE Notice is hereby given that the undersigned attorney herewith enters an appearance in the captioned matter in accordance with 10 C.F.R. ' 2.314(b).
Name:
Susan L. Uttal Address:
U.S. Nuclear Regulatory Commission Office of the General Counsel Mail Stop: O-15-D21 Washington, D.C. 20555-0001 Telephone Number:
(301) 415-1582 Fax Number:
(301) 415-3725 E-Mail Address:
susan.uttal@nrc.gov Admissions:
New Jersey; Pennsylvania; U.S. Court of Appeals 3rd Circuit Name of Party:
NRC Staff Respectfully submitted,
/RA/
Susan L. Uttal Counsel for NRC Staff Dated at Rockville, Maryland this 2nd day of June, 2008