ML062720218
| 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, 2006 LICENSEE:
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 REGULATORY COMMISSION AND ENTERGY NUCLEAR OPERATIONS, INC., CONCERNING INFORMATION PERTAINING TO THE VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of Entergy Nuclear Operations, Inc., held a telephone conference call on August 10, 2006, to discuss and clarify the staffs requests for additional information concerning the Vermont Yankee Nuclear Station license renewal application. The conference call was useful in clarifying the staffs questions. provides a listing of the conference call participants. Enclosure 2 contains a listing of 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 Manager License Renewal Branch B Division of License Renewal Office of Nuclear Reactor Regulation Docket No. 50-271
Enclosures:
As stated cc w/encls: See next page
ML062720218 OFFICE PM:RLRB LA:RLRB BC:RLRB NAME Jrowley /RA/
I. King /RA/
Jzimmerman
/RA DJM for/
DATE 09/28/06 09/28/06 09/29/06
Vermont Yankee Nuclear Power Station cc:
Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr. David R. Lewis Pillsbury, Winthrop, Shaw, Pittman, LLP 2300 N Street, N.W.
Washington, DC 20037-1128 Mr. David OBrien, Commissioner Vermont Department of Public Service 112 State Street Montpelier, VT 05620-2601 Mr. James Volz, Chairman Public Service Board State of Vermont 112 State Street Montpelier, VT 05620-2701 Chairman, Board of Selectmen Town of Vernon P.O. Box 116 Vernon, VT 05354-0116 Operating Experience Coordinator Vermont Yankee Nuclear Power Station 320 Governor Hunt Road Vernon, VT 05354 G. Dana Bisbee, Esq.
Deputy Attorney General 33 Capitol Street Concord, NH 03301-6937 Chief, Safety Unit Office of the Attorney General One Ashburton Place, 19th Floor Boston, MA 02108 Ms. Carla A. White, RRPT, CHP Radiological Health Vermont Department of Health P.O. Box 70, Drawer #43 108 Cherry Street Burlington, VT 05402-0070 Mr. James M. DeVincentis Manager, Licensing Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500 Resident Inspector Vermont Yankee Nuclear Power Station U. S. Nuclear Regulatory Commission P.O. Box 176 Vernon, VT 05354 Director, Massachusetts Emergency Management Agency ATTN: James Muckerheide 400 Worcester Rd.
Framingham, MA 01702-5399 Jonathan M. Block, Esq.
Main Street P.O. Box 566 Putney, VT 05346-0566 Mr. John F. McCann Director, Licensing Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Gary J. Taylor Chief Executive Officer Entergy Operations 1340 Echelon Parkway Jackson, MS 39213
Vermont Yankee Nuclear Power Station cc:
Mr. John T. Herron Sr. VP and Chief Operating Officer Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Oscar Limpias Vice President, Engineering Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Christopher Schwartz Vice President, Operations Support Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Michael J. Colomb Director of Oversight Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Travis C. McCullough Assistant General Counsel Entergy Nuclear Operations, Inc.
440 Hamilton Avenue White Plains, NY 10601 Mr. Theodore Sullivan Site Vice President Entergy Nuclear Operations, Inc.
Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500 Mr. James H. Sniezek 5486 Nithsdale Drive Salisbury, MD 21801 Mr. Garrett D. Edwards 814 Waverly Road Kennett Square, PA 19348 Ms. Stacey M. Lousteau Treasury Department Entergy Services, Inc.
639 Loyola Avenue New Orleans, LA 70113 Mr. Norman L. Rademacher Director, NSA Vermont Yankee Nuclear Power Station P.O. Box 0500 185 Old Ferry Road Brattleboro, VT 05302-0500 Mr. Raymond Shadis New England Coalition Post Office Box 98 Edgecomb, ME 04556 Mr. James P. Matteau Executive Director Windham Regional Commission 139 Main Street, Suite 505 Brattleboro, VT 05301 Mr. William K. Sherman Vermont Department of Public Service 112 State Street Drawer 20 Montpelier, VT 05620-2601 Mr. Michael D. Lyster 5931 Barclay Lane Naples, FL 34110-7306 Ms. Charlene D. Faison Manager, Licensing 440 Hamilton Avenue White Plains, NY 10601 Mr. James Ross Nuclear 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 REGULATORY COMMISSION AND ENTERGY NUCLEAR OPERATIONS, INC., CONCERNING INFORMATION PERTAINING TO THE VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION HARD COPY DLR R/F JRowley E-MAIL:
JFair RWeisman 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 CALL TO DISCUSS THE VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION August 10, 2006 PARTICIPANTS AFFILIATIONS Jonathan Rowley U.S. Nuclear Regulatory Commission (NRC)
Ronald Young NRC Ganesh Cheruvenki NRC Jim Nicholas Pacific Northwest National Laboratory Mike Hamer Entergy Nuclear Operations, Inc. (ENO)
Andy Taylor ENO Lori Potts ENO Allan Cox ENO VERMONT YANKEE NUCLEAR POWER STATION LICENSE RENEWAL APPLICATION August 10, 2006 The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of Entergy Nuclear Operations, Inc., held a telephone conference call on August 10, 2006, to discuss and clarify the staffs 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.1.2-1 The applicant states that the Control Rod Drive (CRD) return line nozzle has been capped at the 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 with the 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 at capped 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 repaired extensively. 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 the Pilgrim LRA. The staff pointed out the differences. The applicant indicated that the question is clear.
RAI-B.1.2-2 The 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 inspection results for these welds indicated no cracking, the applicant proposed to delete the inspection requirements for these welds during the extended period of operation. In order to effectively evaluate 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 welded to the 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 per the 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 not performing 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 in these 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, the techniques 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 this RAI.
RAI 2.3.3.2a-1 License 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 locations with respect to the applicable requirements of Title 10 of the Code of Federal Regulations Section 54.4(a) (10 CFR 54.4(a)).
Discussion: The applicant indicated that the question is clear.
RAI 2.3.3.2a-2 LR 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 a different location of the LRA where the information could be found. The staff was requested to determine if the information in Table 2.3.3.13-B addresses the concern. The staff will evaluate the information in the table and withdraw the RAI if found adequate.
RAI 2.3.3.3-1 LR Drawing, LRA-G-191159-SH-05-0, Location P-10 at Valve 29 is within the scope of LR. This section 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 LR boundary 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 incorrectly referenced the drawing of interest. The staff will issue this RAI with the correct drawing reference of LRA-G-191159-SH-03-0.
RAI 2.3.3.6-1 LR Drawing, LRA-G-191162 Sheet 2, provides details about the fail open (FO) supported emergency 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 also supported by the FO system, e.g., details about the transfer system between the 75,000 gallon fuel 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 portable tank 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 diesel system to establish the relationship to the FO system and to clarify where and what the AMR tables should include in both Sections 2.3.3.6 and 2.3.3.12. Also, provide additional information for 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 the applicant to describe the system rather than having to provide a system drawing.
RAI 2.2-2 The Nuclear System Leakage Rate Limits and Leakage Detection Systems are described in Section 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 other supporting systems. There were no LRA sections identified to address the leakage detection methods and systems described in UFSAR Section 4.10. The applicant is asked to clarify if the leakage detection systems and components are included in the LR scope or the basis for their exclusion.
Discussion: The applicant indicated that the question is too general and requested the staff be more specific. The staff will reword the RAI accordingly prior to submittal.