ML062720218

From kanterella
Jump to navigation Jump to search
2006/09/29-Summary of a Telephone Conference Call Held on August 10, 2006, Between the U.S. NRC and Entergy Nuclear Operations, Inc., Concerning Information Pertaining to the Vermont Yankee Nuclear Power Station License Renewal Application
ML062720218
Person / Time
Site: Vermont Yankee File:NorthStar Vermont Yankee icon.png
Issue date: 09/29/2006
From: Rowley J
NRC/NRR/ADRO/DLR/RLRB
To:
Entergy Nuclear Operations
Rowley J, NRR/DLR/RLRB, 415-4053
References
%dam200611
Download: ML062720218 (10)


Text

September 29, 2006LICENSEE:Entergy Nuclear Operations, Inc.FACILITY:Vermont Yankee Nuclear Power Station

SUBJECT:

SUMMARY

OF A TELEPHONE CONFERENCE CALL HELD ON AUGUST 10, 2006, BETWEEN THE U.S. NUCLEAR REGULATORYCOMMISSION AND ENTERGY NUCLEAR OPERATIONS, INC., CONCERNING INFORMATION PERTAINING TO THE VERMONT YANKEE NUCLEARPOWER STATION LICENSE RENEWAL APPLICATIONThe U.S. Nuclear Regulatory Commission staff (the staff) and representatives of EntergyNuclear Operations, Inc., held a telephone conference call on August 10, 2006, to discuss and clarify the staff's requests for additional information concerning the Vermont Yankee Nuclear Station license renewal application. The conference call was useful in clarifying the staff's questions.Enclosure 1 provides a listing of the conference call participants. Enclosure 2 contains a listingof the issues discussed with the applicant, including a brief description on the status of the items.The applicant had an opportunity to comment on this summary.

/RA/Jonathan Rowley, Project ManagerLicense Renewal Branch B Division of License Renewal Office of Nuclear Reactor RegulationDocket No. 50-271

Enclosures:

As statedcc w/encls: See next page

ML062720218OFFICEPM:RLRBLA:RLRBBC:RLRBNAMEJrowley /RA/ I. King /RA/ Jzimmerman/RA DJM for/DATE09/28/0609/28/0609/29/06 Vermont Yankee Nuclear Power Station cc:

Regional Administrator, Region IU. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415Mr. David R. LewisPillsbury, Winthrop, Shaw, Pittman, LLP 2300 N Street, N.W.

Washington, DC 20037-1128Mr. David O'Brien, CommissionerVermont Department of Public Service 112 State Street Montpelier, VT 05620-2601Mr. James Volz, ChairmanPublic Service Board State of Vermont 112 State Street Montpelier, VT 05620-2701Chairman, Board of Selectmen Town of Vernon P.O. Box 116 Vernon, VT 05354-0116Operating Experience CoordinatorVermont Yankee Nuclear Power Station 320 Governor Hunt Road Vernon, VT 05354G. Dana Bisbee, Esq.Deputy Attorney General 33 Capitol Street Concord, NH 03301-6937Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108Ms. Carla A. White, RRPT, CHPRadiological Health Vermont Department of Health P.O. Box 70, Drawer #43 108 Cherry Street Burlington, VT 05402-0070Mr. James M. DeVincentisManager, Licensing Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500Resident InspectorVermont Yankee Nuclear Power Station U. S. Nuclear Regulatory Commission

P.O. Box 176 Vernon, VT 05354Director, Massachusetts Emergency Management Agency ATTN: James Muckerheide 400 Worcester Rd.

Framingham, MA 01702-5399Jonathan M. Block, Esq.Main Street

P.O. Box 566 Putney, VT 05346-0566Mr. John F. McCannDirector, Licensing Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Gary J. TaylorChief Executive Officer Entergy Operations 1340 Echelon Parkway Jackson, MS 39213 Vermont Yankee Nuclear Power Station cc:

Mr. John T. HerronSr. VP and Chief Operating Officer Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Oscar LimpiasVice President, Engineering Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Christopher SchwartzVice President, Operations Support Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Michael J. ColombDirector of Oversight Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Travis C. McCulloughAssistant General Counsel Entergy Nuclear Operations, Inc.

440 Hamilton Avenue White Plains, NY 10601Mr. Theodore SullivanSite Vice President Entergy Nuclear Operations, Inc.

Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500Mr. James H. Sniezek5486 Nithsdale Drive Salisbury, MD 21801Mr. Garrett D. Edwards814 Waverly Road Kennett Square, PA 19348Ms. Stacey M. LousteauTreasury Department Entergy Services, Inc.

639 Loyola Avenue New Orleans, LA 70113Mr. Norman L. RademacherDirector, NSA Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500Mr. Raymond ShadisNew England Coalition Post Office Box 98 Edgecomb, ME 04556Mr. James P. MatteauExecutive Director Windham Regional Commission 139 Main Street, Suite 505 Brattleboro, VT 05301Mr. William K. ShermanVermont Department of Public Service 112 State Street Drawer 20 Montpelier, VT 05620-2601Mr. Michael D. Lyster5931 Barclay Lane Naples, FL 34110-7306Ms. Charlene D. FaisonManager, Licensing 440 Hamilton Avenue White Plains, NY 10601Mr. James RossNuclear Energy Institute 1776 I Street, NW, Suite 400 Washington, DC 20006-3708 Vermont Yankee Nuclear Power Station cc:

Diane Curran, Esq.Harmon, Curran, Spielberg &

Eisenberg, L.L.P 1726 M Street, NW, Suite 600 Washington, DC 20036 Note to Entergy Nuclear Operations, Inc., from Jonathan Rowley dated September 29, 2006

SUBJECT:

SUMMARY

OF A TELEPHONE CONFERENCE CALL HELD ON AUGUST 10, 2006, BETWEEN THE U.S. NUCLEAR REGULATORYCOMMISSION AND ENTERGY NUCLEAR OPERATIONS, INC., CONCERNING INFORMATION PERTAINING TO THE VERMONT YANKEE NUCLEARPOWER STATION LICENSE RENEWAL APPLICATIONHARD COPYDLR R/FJRowleyE-MAIL:JFairRWeisman AMurphy RPettis GGalletti DShum GBagchi SSmith (srs3)

SDuraiswamy YL (Renee) Li RidsNrrDlr RidsNrrDlrRlra RidsNrrDlrRlrb RidsNrrDlrRlrc RidsNrrDlrReba RidsNrrDlrRebb RidsNrrDe RidsNrrDci RidsNrreEemb RidsNrrDeEeeb RidsNrrDeEqva RidsNrrDss RidsNrrDnrl RidsOgcMailCenter RidsNrrAdes


JRowley RLaufer JShea CAnderson, RI RPowell, RI DScrenci, RI MModes, RI DPelton, Sr. Resident MYoung RidsOpaMail LIST OF PARTICIPANTS FOR TELEPHONE CONFERENCE CALLTO DISCUSS THE VERMONT YANKEE NUCLEAR POWER STATIONLICENSE RENEWAL APPLICATIONAugust 10, 2006PARTICIPANTS AFFILIATIONSJonathan RowleyU.S. Nuclear Regulatory Commission (NRC)Ronald YoungNRC Ganesh CheruvenkiNRC Jim NicholasPacific Northwest National Laboratory Mike HamerEntergy Nuclear Operations, Inc. (ENO)

Andy TaylorENO Lori PottsENO Allan CoxENO VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATIONAugust 10, 2006The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of EntergyNuclear Operations, Inc., held a telephone conference call on August 10, 2006, to discuss and clarify the staff's requests for additional information (RAIs) concerning the Vermont Yankee Nuclear Power Station (VYNPS) license renewal application (LRA). The following issues were discussed during the telephone conference call:RAI-B.1.2-1The applicant states that the Control Rod Drive (CRD) return line nozzle has been capped atthe VYNPS unit. The staff requests that the applicant provide the following information regarding the cap and the weld:(1)Describe the configuration, location and material of construction of the capped nozzle. This should include the existing base material for the nozzle, piping (if piping remnants exist) and cap material, and any welds.(2) Describe how the aging effects for this weld and the cap are managed in accordance withthe guidelines of Boiling Water Reactor Vessel and Internals Project-75 (BWRVIP-75),

"BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedule." (3) Discuss whether the event at Pilgrim Nuclear Power Station (Pilgrim)(leaking weld atcapped nozzle, September 30, 2003) is applicable to VYNPS. The staff issued Information Notice 2004-08, "Reactor Coolant Pressure Boundary Leakage Attributable to Propagation of Cracking in Reactor Vessel Nozzle Welds," dated April 22, 2004, which states that the cracking occurred in an Alloy 182 weld that was previously repairedextensively. Discuss experience with previous leakage at the VYNPS capped nozzle, if any. Include in your discussion the past inspection techniques applied, the results obtained, and mitigative strategies imposed. Provide information as to how the plant-specific experience related to this aging effect impacts the attributes specified in Aging Management Program B.1.2, "BWR CRD Return Line Nozzles."Discussion: The applicant believed that this question was similar to an RAI issued for thePilgrim LRA. The staff pointed out the differences. The applicant indicated that the question isclear.RAI-B.1.2-2The applicant also states that CRD return lines are connected to the reactor water clean-up(RWCU) piping system which is classified as a non-safety system. Since previous inspectionresults for these welds indicated no cracking, the applicant proposed to delete the inspectionrequirements for these welds during the extended period of operation. In order to effectivelyevaluate this proposal, provide the following information regarding the welds between the CRD return line and the RWCU piping: (1) Provide information regarding the total number of the CRD return lines that are wel ded tothe RWCU piping.(2) Provide a drawing or a sketch indicating the safety/non-safety boundary of these welds. Additionally, the applicant should confirm that all safety-related welds will be inspected perthe American Society of Mechanical Engineers Code,Section XI, Inservice Inspection program. Since the applicant intends to not perform inspections of these welds during the extended period of operation, the applicant should provide a technical justification for notperforming the inspections of these welds taking into account the effects of aging degradation and effective techniques that will be used to mitigate the aging process.(3) Provide information regarding the type of base metal and weld metal that are used inthese welds. If austenitic stainless steel weld metal is used for these welds, the applicant should provide information regarding the amount of delta ferrite that is present in these welds. This information is necessary in assessing the susceptibility of these welds to intergranular stress-corrosion cracking (IGSCC). If the predominant aging mechanism in these welds is other than IGSCC, the applicant should identify the aging mechanism, thetechniques it intends to use in mitigating the aging degradation, and aging monitoring program for these welds.Discussion: The applicant indicated that the question is not clear. The staff will reword thisRAI.RAI 2.3.3.2a-1License Renewal (LR) Drawing LRA-G-191159-SH-01-0, Location H-12, depicts pipe section 2"-SW-566C to be within the scope of LR. Upstream from where 2"-SW-566C enters the RX Building from outside there is no drawing continuation to depict the LR boundary. Provide details for the continuation of 2"-SW-566C to the LR boundary and justify the boundary locationswith respect to the applicable requirements of Title 10 of the Code of Federal RegulationsSection 54.4(a) (10 CFR 54.4(a)).Discussion: The applicant indicated that the question is clear.RAI 2.3.3.2a-2LR Drawing LRA-G-191159-SH-01-0, at Location H-11, Drawing Note 16 indicates piping section(4"-SW-567) and supports on the Reactor Building Air Conditioning supply piping where the vacuum breakers tie in are structure and components 2 for structural integrity. This pipe section from Valve 23D through Valves RBAC-1A, 1B, 1C and 1D is not within the scope of LR. Failure of this section of pipe could have an effect on the LR intended pressure boundary function for the service water piping. Provide additional information on why this section of pipe and components are not within the scope for LR and justify the boundary locations with respect to the applicable requirements of 10 CFR 54.4(a).Discussion: The applicant indicated that the question is clear and directed the staff to adifferent location of the LRA where the information could be found. The staff was requested todetermine if the information in Table 2.3.3.13-B addresses the concern. The staff will evaluatethe information in the table and withdraw the RAI if found adequate. RAI 2.3.3.3-1LR Drawing, LRA-G-191159-SH-05-0, Location P-10 at Valve 29 is within the scope of LR. Thissection of pipe is the RBCCW return to the Alternate Cooling System. There is not a drawing continuation provided. Provide details for the continuation of this piping section to the LRboundary and justify the boundary locations with respect to the applicable requirements of 10 CFR 54.4(a).Discussion: The applicant indicated that the question is not clear. The staff incorrectlyreferenced the drawing of interest. The staff will issue this RAI with the correct drawingreference of LRA-G-191159-SH-03-0.RAI 2.3.3.6-1LR Drawing, LRA-G-191162 Sheet 2, provides details about the fail open (FO) supportedemergency diesel generators, diesel-driven fire pump, and house heating boiler systems.However, the drawing does not provide details about the John Deere Diesel system that is alsosupported by the FO system, e.g., details about the transfer system between the 75,000 gallonfuel oil storage tank and the day tanks for the two John Deere diesels and single fire pump diesel which are required to provide an intended function for 10 CFR 54.4 (a)(3) in support of the fire protection regulation (10 CFR 50.48). The LRA text mentions only that a 500-gallon portabletank is used to transport fuel oil to those diesels. Typical aging management review (AMR) components for diesels like the day tank, strainer, etc., for the John Deere Diesel are not covered. Provide the FO system drawings to include a schematic of the John Deere dieselsystem to establish the relationship to the FO system and to clarify where and what the AMRtables should include in both Sections 2.3.3.6 and 2.3.3.12. Also, provide additional informationfor the LR boundary and justify its location with respect to the applicable requirements of 10 CFR 54.4(a).Discussion: The applicant indicated that the question is clear. The staff agreed to allow theapplicant to describe the system rather than having to provide a system drawing.RAI 2.2-2The Nuclear System Leakage Rate Limits and Leakage Detection Systems are described inSection 4.10 of the VYNPS Updated Final Safety Analysis Report (UFSAR). Identified and unidentified leakage rates are an important aspect of plant operation and Technical Specifications. Section 4.10 of the UFSAR does not specifically address a single "leakage detection system," but describes a "leakage detection method" with references to othersupporting systems. There were no LRA sections identified to address the leakage detectionmethods and systems described in UFSAR Section 4.10. The applicant is asked to clarify if theleakage detection systems and components are included in the LR scope or the basis for theirexclusion.Discussion: The applicant indicated that the question is too general and requested the staff bemore specific. The staff will reword the RAI accordingly prior to submittal.