ML051680050
| ML051680050 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 06/16/2005 |
| From: | Veronica Rodriguez NRC/NRR/DRIP/RLEP |
| To: | Nuclear Management Co |
| Rodriguez V, RLEP/DRIP/NRR, 415-2232 | |
| References | |
| Download: ML051680050 (9) | |
Text
June 16, 2005 LICENSEE: Nuclear Management Company, LLC FACILITY: Point Beach Nuclear Plant, Units 1 and 2
SUBJECT:
SUMMARY
OF TELEPHONE CONFERENCE HELD ON MAY 26, 2005, BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND NUCLEAR MANAGEMENT COMPANY, LLC, CONCERNING OPEN AND CONFIRMATORY ITEMS PERTAINING TO THE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2, SAFETY EVALUATION REPORT The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of Nuclear Management Company, LLC (NMC) held a telephone conference on May 26, 2005, to discuss and clarify the staffs open items (OI) and confirmatory items (CI) concerning the Point Beach Nuclear Plant, Units 1 and 2, safety evaluation report. The conference call was useful in clarifying the intent of the staffs concerns. provides a listing of the meeting participants. Enclosure 2 contains a listing of the OIs and CIs 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/
Veronica M. Rodriguez, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket Nos.: 50-266 and 50-301
Enclosures:
As stated cc w/encls: See next page
DOCUMENT NAME: E:\Filenet\ML051680050.wpd OFFICE PM:RLEP PM:RLEP LA:RLEP SC:RLEP NAME VRodriguez GSuber (V. rodriguez. for) MJenkins SLee DATE 06/11/05 06/11/05 06/14/05 06/16/05 Point Beach Nuclear Plant, Units 1 and 2 cc:
Jonathan Rogoff, Esq. Mr. Jeffrey Kitsembel Vice President, Counsel & Secretary Electric Division Nuclear Management Company, LLC Public Service Commission of Wisconsin 700 First Street P.O. Box 7854 Hudson, WI 54016 Madison, WI 53707-7854 Mr. Frederick D. Kuester David Weaver President and Chief Executive Officer Nuclear Asset Manager We Generation Wisconsin Electric Power Company 231 West Michigan Street 231 West Michigan Street Milwaukee, WI 53201 Milwaukee, WI 53201 James Connolly John Paul Cowan Manager, Regulatory Affairs Executive Vice President & Chief Nuclear Point Beach Nuclear Plant Officer Nuclear Management Company, LLC Nuclear Management Company, LLC 6610 Nuclear Road 700 First Street Two Rivers, WI 54241 Hudson, WI 54016 Mr. Ken Duveneck Douglas E. Cooper Town Chairman Senior Vice President - Group Operations Town of Two Creeks Palisades Nuclear Plant 13017 State Highway 42 Nuclear Management Company, LLC Mishicot, WI 54228 27780 Blue Star Memorial Highway Covert, MI 49043 Chairman Public Service Commission Mr. James Ross of Wisconsin Nuclear Energy Institute P.O. Box 7854 1776 I Street, NW., Suite 400 Madison, WI 53707-7854 Washington, DC 20006-3708 Regional Administrator, Region III Roger A. Newton U.S. Nuclear Regulatory Commission 3623 Nagawicka Shores Drive 801 Warrenville Road Hartland, WI 53029 Lisle, IL 60532-4351 James E. Knorr Resident Inspector's Office License Renewal Project U.S. Nuclear Regulatory Commission Nuclear Management Company, LLC 6612 Nuclear Road 6610 Nuclear Road Two Rivers, WI 54241 Point Beach Nuclear Plant Two Rivers, WI 54241
DISTRIBUTION: Summary of Telecon with NMC concerning Pt. Beach, Dated: June 16, 2005 ADAMS Accession No.: ML051680050 HARD COPY RLEP RF E-MAIL:
RidsNrrDrip RidsNrrDe G. Bagchi K. Manoly W. Bateman J. Calvo R. Jenkins P. Shemanski J. Fair RidsNrrDssa RidsNrrDipm D. Thatcher R. Pettis G. Galletti C. Li K. Winsberg (RidsOgcMailCenter)
R. Weisman M. Mayfield A. Murphy S. Smith (srs3)
S. Duraiswamy Y. L. (Renee) Li RLEP Staff P. Lougheed, RIII J. Strasma, RIII A. Stone, RIII H. Chernoff W. Ruland L. Raghavan T. Mensah OPA
LIST OF PARTICIPANTS FOR TELEPHONE CONFERENCE TO DISCUSS THE POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 LICENSE RENEWAL APPLICATION MAY 26, 2005 Participant Affiliation Bill Herrman Nuclear Management Company (NMC)
Dough Johnson NMC James Knorr NMC John Thorgersen NMC Mark Ortomayer NMC Todd Mielke NMC Gregory Suber U.S. Nuclear Regulatory Commission (NRC)
James Strnisha NRC Kaihwa Hsu NRC Kenneth Chang NRC Kurt Cozens NRC Veronica Rodriguez NRC Enclosure 1
OPEN AND CONFIRMATORY ITEMS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 SAFETY EVALUATION REPORT May 26, 2005 The U.S. Nuclear Regulatory Commission staff (the staff) and representatives of Nuclear Management Company, LLC (NMC) held a telephone conference call on May 26, 2005, to discuss and clarify the staffs open items (OIs) and confirmatory items (CIs) concerning the Point Beach Nuclear Plant, Units 1 and 2, safety evaluation report (SER). The following OIs and CIs were discussed during the telephone conference call.
Section 4.3.13: Metal Fatigue - Crane Load Cycle Limit During the Unit 2 Spring 2005 refueling outage, the staff expressed concerns about the NMCs reactor vessel head replacement plans and the use its polar crane for this activity. The license renewal team requested the applicant to clarify and confirm if the Metal Fatigue - Crane Load Cycle Limit TLAA information provided in the LRA is still valid. During the conversation, the applicant confirmed that the LRA information is still valid and is not expected to change.
Further, the applicant stated that the reactor vessel head lift is not an overrated lift and is within the evaluated limits.
OI 3.3-7 Section 3.3.2.3.3 - Component Cooling Water System - Aging Management Evaluation - Table 3.3.2-2 In LRA Table 3.3.2-2, the applicant proposed to manage cracking due to intergranular attack/intergranular stress corrosion cracking (IGA/IGSCC) of stainless steel material for heat exchanger components exposed to primary treated water with temperature greater than 480 EF using the Water Chemistry Control Program. This line item cites Note 35, which states:
Component/material/environment is not addressed in the corresponding NUREG-1801 Chapter, but the component/material/environment is addressed in another NUREG-1801 Chapter. This line item references AMR line item 3.1.1-36, which provides the following discussion:
Crack initiation growth due to SCC and flaw growth are identified as aging effects requiring management for the reactor vessel nozzle safe ends, CRD housing, and RCS components. Aging management programs credited for managing these effects are the Water Chemistry Program and ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program.
The Note implies that ASME Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program should have also been applied to LRA Table 3.3.2-2. In RAI 3.3-7, dated March 31, 2005, the staff requested the applicant to explain this discrepancy or make a commitment to review the line item in LRA Table 3.3.2-2 to include the Inservice Inspection Program.
Enclosure 2 Discussion: Based on the discussion with the applicant, the staff indicated and the applicant agreed that this open item requires further clarification. During the conversation, the applicant
stated that surrogated inspections will be performed. The applicants clarification will be provided in writing.
CI 3.1.1-1 Section 3.1.2.1.1 - Loss of Fracture Toughness Due to Thermal Aging Embrittlement The staff finds that the use of leak-before-break evaluation method is not equivalent to a flaw tolerance methodology; it assumes through-wall leakage and therefore does not assure the safety function of pressure boundary integrity. In RAI 3.1.1-1, dated March 30, 2005, the staff requested the applicant to clarify how it manages the aging effect of loss of fracture toughness due to thermal aging embrittlement for CASS primary loop elbows. During a telephone conference, the applicant agreed to revise its position and perform flaw tolerance evaluations.
Discussion: Based on the discussion with the applicant, the staff indicated and the applicant agreed that the applicants response requires further clarification. During the conversation, the applicant stated that it will use enhanced volumetric examination to detect and size cracks or a plant- or component-specific flaw. The applicants clarification will be provided in writing.
CI 3.1.1-2 Section 3.1.2.2.10 - Loss of Section Thickness Due to Erosion In RAI 3.1.1-2, dated March 30, 2005, the staff requested the applicant to justify why the steam generator feedrings and associated J-tubes are outside the scope of license renewal. During a telephone conference, the applicant agreed to add the steam generator feedrings and J-tubes to the scope of license renewal and manage the associated aging effects.
Discussion: Based on the discussion with the applicant, the staff indicated and the applicant agreed that the applicants response requires further clarification. During the conversation, the applicant stated that it will provide its plant-specific evaluation with regard to loss of material due to FAC and general corrosion as recommended by the GALL. The applicants clarification will be provided in writing.
CI 3.5-12 Section 3.5.2.2.1 - PWR Containments - Cracking Due to Cyclic Loading and Stress Corrosion Cracking (SCC)
In LRA Section 3.5.2.2.1.7, the applicant stated that SCC is not an applicable aging mechanism for penetration sleeves, bellows, and dissimilar metal welds. Therefore, the applicant did not address cracking due to cyclic loading. In RAI 3.5-12, dated March 30, 2005, the staff requested the applicant to address the difference between its position and the GALL Report recommendation of enhanced inspection methods. The staff noted that the TLAA in LRA Section 4.3.11 does not detect and manage cracking due to cyclic loading. The applicant was requested to provide further clarification for crediting this specific line item to manage cracking due to cyclic loading.
During a telephone conference, the applicant indicated that this is a TLAA and will provide information to confirm that this is adequately addressed in LRA Section 4.3.11.
Discussion: Based on the discussion with the applicant, the staff indicated and the applicant agreed that the applicants response requires further clarification. During the conversation, the applicant stated that, as required by the ASME Code, visual inspections of the penetration, including its vicinity and dissimilar welds, will be performed. The applicants clarification will be provided in writing.
CI 3.5-13 Section 3.5.2.2.1 - PWR Containments - Aggressive chemical attack In LRA Section 3.5.2.2.1.1, the applicant stated that concrete degradation in air due to aggressive rainwater is insignificant and that the below-grade/lake water environment is nonaggressive. In RAI 3.5-13, dated March 30, 2005, the staff requested the applicant to provide sufficient data to support this statement.
Furthermore, during the review, the staff was unable to identify how the LRA addresses the items described in ISG-03. The staff requested the applicant to provide detailed information with regard to how its AMRs address all the items described in ISG-03.
During a telephone conference, the applicant described how it will satisfy the ISG-03 criteria and agreed to provide its most recent data with respect to the below-grade/lake water. The applicant committed to provide a formal response, including a table detailing how it satisfies all the items described in ISG-03.
Discussion: Based on the discussion with the applicant, the staff indicated and the applicant agreed that the applicants response requires further clarification. During the conversation, the applicant provided the percentage of entrapped air in concrete. The applicants statement will be provided in writing.
CI B2.1.11-1 Section 3.0.3.2.11 - Flow-Accelerated Corrosion Program During the audit, the staff noted that for the acceptance criteria program element, it is unclear how the applicant calculates the minimum permitted wall thickness and how it is used in its analysis for flow-accelerated corrosion. In RAI B2.1.11-1, dated March 30, 2005, the staff requested the applicant to clarify its wall thickness calculation and its uses.
The staffs concern was referred to the Region III staff, which performed its AMR/AMP onsite inspection during the weeks of March 7 and 21, 2005. The applicant clarified its methodology.
The applicant stated that the minimum wall calculations are performed using the design pressure, which is greater than the operating pressure and demonstrates that the actual measured wall thickness is greater than the minimum thickness required by the maximum hoop stress. If degradation is detected such that the wall thickness is less than or equal to 87.5 percent of nominal wall thickness for safety-related piping or 60 percent of nominal wall thickness for nonsafety-related piping, additional examinations will be performed in adjacent areas to bound the thinning and assure that the actual minimum wall is measured. In addition, the applicant will provide its justification and confirmation that the minimum wall thickness will be maintain for the period of extended operation.
Discussion: Based on the discussion with the applicant, the staff indicated and the applicant agreed that the applicants response requires further clarification. During the conversation, the applicant stated that for both safety-related and nonsafety-related piping, additional examinations will be performed in adjacent areas to bound the thinning if the remaining service life, based on the code minimum allowable wall thickness, is less than one operating cycle. The sample size will also be expanded for nonsafety-related piping if degradation is detected such that the wall thickness is less than or equal to 60% of nominal wall thickness. This covers situations where the code minimum allowable wall thickness may be less than 60% of nominal wall thickness for nonsafety-related piping. The applicants clarification will be provided in writing.
OI B2.1 Sections 3.0.3.2.1 and 3.0.3.2.2 - ASME Section XI Inspection Programs Relief requests are approved by the NRC as described in 10 CFR 50.55a, Codes and Standards. Relief requests only apply to the current licensing basis (CLB) and are time-limited.
Consequently, citing approved requests cannot be used as a basis for taking exception to the GALL Report since they may not be renewed. Each exception to the GALL Report must be evaluated for NRC approval based on the technical bases that are associated with aging management regardless of whether there is a current, approved, related relief request. Citing a relief request does not provide an acceptable basis to take an exception to the GALL Report.
In RAI B2.1, dated March 30, 2005, the staff requested the applicant to provide its technical bases, as they relate to aging management, and without referencing any relief requests, for the exceptions taken to ASME Code Section XI, Subsections IWB, IWC, and IWD Inservice Inspection Program and ASME Code Section XI, Subsections IWE and IWL Inservice Inspection Program.
Discussion: Based on the discussion with the applicant, the staff indicated and the applicant agreed that this open item requires further clarification. During the conversation, the applicant was requested to provide only those exceptions as they relate to aging management and to specify exceptions taken from the GALL Report. The applicants will provide its formal response in writing.