ML041480134
| ML041480134 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 05/26/2004 |
| From: | Gregory Suber Division of Regulatory Improvement Programs |
| To: | Forbes J Entergy Operations |
| Suber G, NRR/DRIP/RLEP 301-415-1124 | |
| References | |
| TAC MB8402 | |
| Download: ML041480134 (7) | |
Text
May 26, 2004 Mr. Jeff Forbes Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE ARKANSAS NUCLEAR ONE, UNIT 2, LICENSE RENEWAL APPLICATION (TAC NO. MB8402)
Dear Mr. Forbes:
The U.S. Nuclear Regulatory Commission (NRC) is reviewing a license renewal application (LRA) submitted by Entergy Operators Inc. (Entergy or the applicant) dated October 14, 2003 for the renewal of the operating licenses for Arkansas Nuclear One, Unit 2, pursuant to Title 10 Code of Federal Regulations Part 54 (10 CFR Part 54). The NRC staff has identified, in the enclosure, areas where additional information is needed to complete the review. Specifically, the enclosed request for additional information (RAI) is from Section 3.1 Reactor Vessel, Internals and Reactor Coolant System. These RAIs have been discussed with your staff.
Your response to these RAIs is requested within 30 days of receipt of this letter. If you have any questions, please contact me at (301) 415-1124 or e-mail gxs@nrc.gov.
Sincerely,
/RA/
Gregory F. Suber, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No.: 50-368
Enclosure:
As stated cc w/encl: See next page
May 26, 2004 Mr. Jeff Forbes Vice President, Operations ANO Entergy Operations, Inc.
1448 S. R. 333 Russellville, AR 72801
SUBJECT:
REQUEST FOR ADDITIONAL INFORMATION FOR THE REVIEW OF THE ARKANSAS NUCLEAR ONE, UNIT 2, LICENSE RENEWAL APPLICATION (TAC NO. MB8402)
Dear Mr. Forbes:
The U.S. Nuclear Regulatory Commission (NRC) is reviewing a license renewal application (LRA) submitted by Entergy Operators Inc. (Entergy or the applicant) dated October 14, 2003 for the renewal of the operating licenses for Arkansas Nuclear One, Unit 2, pursuant to Title 10 Code of Federal Regulations Part 54 (10 CFR Part 54). The NRC staff has identified, in the enclosure, areas where additional information is needed to complete the review. Specifically, the enclosed request for additional information (RAI) is from Section 3.1 Reactor Vessel, Internals and Reactor Coolant System. These RAIs have been discussed with your staff.
Your response to these RAIs is requested within 30 days of receipt of this letter. If you have any questions, please contact me at (301) 415-1124 or e-mail gxs@nrc.gov.
Sincerely,
/RA/
Gregory F. Suber, Project Manager License Renewal Section A License Renewal and Environmental Impacts Program Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Docket No.: 50-368
Enclosure:
As stated cc w/encl: See next page DISTRIBUTION: See next page Accession No: ML041480134 Document Name: C:\\ORPCheckout\\FileNET\\ML041480134.wpd OFFICE:
LA:RLEP PM:RLEP SC:RLEP NAME:
M. Jenkins G. Suber S. Lee DATE:
5/21/04 5/24/04 5/26/04 OFFICIAL RECORD COPY
REQUEST FOR ADDITIONAL INFORMATION SECTION 3.1 REACTOR VESSEL, INTERNALS AND REACTOR COOLANT SYSTEM ARKANSAS NUCLEAR ONE - UNIT 2 LICENSE RENEWAL APPLICATION (TAC NO. MB8402)
Section 3.1.1 Reactor Coolant System RAI 3.1.1-1 For Item 3.1.1-2, the applicant identifies the inservice inspection program to manage loss of material due to pitting and crevice corrosion in the steam generator shell assembly. U.S.
Nuclear Regulatory Commission (NRC) Information Notice 90-04 states that the ASME Code Section XI inservice inspection method may not be sufficient to detect general and pitting corrosion in the shell/transition cone welds. The applicant states that the concerns of NRC Information Notice (IN) 90-04 are not applicable to ANO-2 steam generators because they were replaced in 2000 and pitting corrosion of the steam generator shell is not known to currently exist. However, the staff believes that the current operating experience does not provide reasonable assurance that pitting will not occur at the shell assembly in the future. In the absence of corrosion tests to demonstrate that the shell and transition cone would not develop pitting corrosion at the end of extended period of operation, pitting and general corrosion should be assumed and inspection methods should be implemented to detect such corrosion. Clarify whether any inspection procedures in addition to the ASME code will be implemented to inspect the shell assembly, including transition cone, in the ANO-2 steam generators for pitting and general corrosion.
RAI 3.1.1-2 For Items 3.1.1-19 and 3.1.-20, the applicant states that ANO-2 steam generators do not have carbon steel tube support plates and carbon steel tube support lattice bars. Discuss the tube support material and support configuration in the ANO-2 steam generators.
RAI 3.1.1-3 For Item 3.1.1-21, the applicant states that the feedwater ring discussed in generic aging lessons learned (e.g., GALL Section IV.D1.3-a) is applicable to Combustion Engineering System 80 steam generators and is not applicable to the Westinghouse steam generators at ANO-2. However, the staff understands that the ANO-2 steam generators do have a feedwater ring and fittings which have a potential for degradation under adverse operating conditions.
Justify why these components are not included in the scope of the license renewal and not subject to aging management.
RAI 3.1.1-4 For Item 3.1.1-39, the applicant states that loss of material due to erosion affecting steam generator secondary manways and handholds is applicable to once-through steam generators and is not applicable to the Westinghouse steam generators at ANO-2. However, the ANO-2 steam generators do have manways and handholds and will have the potential for erosion.
Justify why loss of material due to erosion is not an applicable aging mechanism to these components in the ANO-2 steam generators.
Section 3.1.2 Reactor Vessel and CEDM Pressure Boundary RAI 3.1.2-5.1 In Table 3.1.2-5, the applicant identifies the steam generator integrity program in license renewal application (LRA) Section B.1.25 to manage cracking in the following components:
anti-vibration bar end caps, peripheral retaining rings, U-bend, and U-shaped retainer bars (Page 3.1-100) and stay rods, stayrod hex nuts, spacer pipes, peripheral backup bars, wrapper, and wrapper jacking screws (Page 3.1-106). (1) Discuss how these components are inspected and the frequency of inspection under the steam generator integrity program and (2) clarify whether the U-bend referred to on (Page 3.1-100) is applicable to the U-bend region of the tube, or, to the U-bend tube supports (e.g., peripheral retaining rings and retainer bars).
RAI 3.1.2-5.2 The applicant identify steam generator instrument nozzles in Table 3.1.1, Item 3.1.1-12; however, this component is not identified in LRA Table 3.1.2-5. Clarify why the instrument nozzles are not included for the aging management in LRA Table 3.1.2-5.
RAI 3.1.2-5.3 The applicant identifies LRA Section B.1.28, System Walkdown Program, to manage loss of material in an air environment for many steam generator components in LRA Table 3.1.2-5.
(1) Clarify how the steam generator components are documented in the system walkdown program because it is not evident that these steam generator components are included in LRA Section B.1.28, and (2) For those steam generator components that are not accessible for the inspection during system walkdown, discuss how those components will be inspected.
RAI 3.1.2-5.4 The applicant identifies several aging mechanisms in the secondary side of the steam generators that contribute to tube degradation. Discuss (1) whether there have been any loose parts in the secondary side of the steam generators; (2) whether there are procedures to retrieve and monitor loose parts in the secondary side; (3) whether sludge lancing is performed periodically, and (4) whether the secondary side inspection procedures are a part of the current licensing basis such that the procedures will be carried over to the period of extended operation.
RAI 3.1.2-5.5 The component type tube plugs is listed in Table 3.1.2-5 (Page 3.1-96). Please discuss (1) the types and materials of tube plugs that have been installed in the ANO-2 steam generator tubes, and (2) whether the following NRC generic communications are applicable to the tube plugs installed in the ANO-2 steam generators: NRC Bulletin 89-01, and associated supplements 1 and 2; NRC Information Notice (IN) 89-33, IN 89-65, and IN 94-87.
RAI 3.1.2-5.6 Page 3.1-96. The applicant identifies the water chemistry control program as an AMP to manage loss of material and cracking in the steam generator tubes. Clarify whether the water chemistry control program follows the guidance in EPRI reports TR-102134 and TR-105714 and identify which revisions of the reports are being used at ANO-2.
RAI 3.1.2-5.7 Pages 3.1-96 and 3.1-97. Industry experience has identified denting of Alloy 600 tubes due to corrosion of carbon steel tube support plate as an aging effect. The applicant does not identify tube denting as an aging effect. Justify why tube denting is not an aging effect for the ANO-2 steam generators.
RAI 3.1.2-5.8 Page 3.1-98. For the 6-inch and 8-inch inspection port covers, the applicant identifies internal treated water as an environment. Clarify why the internal treated water is not identified as an environment for the 3-inch inspection port cover. Additionally, the applicant identifies the diaphragms in the 3-inch inspection port as a component for aging management. Clarify why the diaphragms are not identified in the 6-inch or 8-inch inspection ports.
RAI 3.1.2-5.9 Page 3.1-99. The applicant identifies the inservice inspection program to manage the aging effect of cracking in the anti-vibration bars and tube support plates. (1) Clarify how the inservice inspection program would be used to manage cracking in these components because the staff is not aware of any inservice inspection program that follows ASME Code Section XI and includes the inspection of these steam generators components, and (2) discuss the details of how these two components would be inspected under the steam generator integrity program, including inspection scope, frequency and method.
RAI 3.1.2-5.10 Page 3.1-100. The applicant identifies anti-vibration bar end caps, peripheral retaining rings, U-bend, and U-shaped retainer bars as components in the steam generator that require aging management. (1) Clarify whether the U-bend identified on Page 3.1-100 is the U-bend region of a steam generator tube or U-bend tube support, and (2) discuss where the peripheral retaining rings are located with respect to the tube and discuss their safety function.
RAI 3.1.2-5.11 Page 3.1-102. Feedwater inlet nozzles. (1) Discuss whether a flexitallic gasket is used in the ANO-2 feedwater system; (2) if the flexitallic gasket is used, discuss whether the flexitallic gasket has broken; (3) if the gasket has broken, discuss whether the small pieces of the broken gasket have entered in the steam generator tube bundle and damaged the tube(s); (4) discuss the corrective actions that have been implemented due to the gasket loose parts; and (5) if the flexitallic gasket has not broken, discuss the monitoring procedures to prevent the potential gasket loose parts from entering into the tube bundle.
RAI 3.1.2-5.12 Pages 3.1-102 and 3.1-103. The applicant identifies the inservice inspection program to manage cracking in feedwater inlet nozzles, feedwater thermal sleeves, and flow limiting insert (integral flow restrictors). Discuss the inspection method and frequency for these components.
RAI 3.1.2-5.13 Page 3.1-104. The applicant identifies the steam generator integrity program to manage cracking in key bracket and snubber lugs. Clarify where the key bracket is located and whether it has a safety-related function. The staff is not aware that the steam generator integrity program as specified in GALL is designed to manage the key bracket and snubber lugs. If the ANO-2 plant-specific steam generator integrity program does include these two components, the applicants needs to identify the components in LRA B.1.25.
RAI 3.1.2-5.14 Page 3.1-105. Industry experience has identified wall thinning/flow-accelerated corrosion as an aging effect for the steam outlet nozzle. Please discuss why wall thinning/flow-accelerated corrosion is not identified as an aging effect for the ANO-2 steam outlet nozzle.
DISTRIBUTION: Ltr to Jeff Forbes, Re: Entergy Operations, Dated: May 26, 2004 Accession no.: ML041480134 HARD COPY RLEP RF G. Suber (PM)
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Arkansas Nuclear One, Unit 2 cc:
Executive Vice President
& Chief Operating Officer Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Director, Division of Radiation Control and Emergency Management Arkansas Department of Health 4815 West Markham Street, Slot 30 Little Rock, AR 72205-3867 Winston & Strawn 1400 L Street, N.W.
Washington, DC 20005-3502 Mr. Mike Schoppman Framatome ANP, Richland, Inc.
Suite 705 1911 North Fort Myer Drive Rosslyn, VA 22209 Senior Resident Inspector U.S. Nuclear Regulatory Commission P. O. Box 310 London, AR 72847 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 County Judge of Pope County Pope County Courthouse Russellville, AR 72801 Vice President, Operations Support Entergy Operations, Inc.
P. O. Box 31995 Jackson, MS 39286-1995 Wise, Carter, Child & Caraway P. O. Box 651 Jackson, MS 39205 Garry Young 1448 SR 333 Russellville, AR 72802 Mr. Fred Emerson Nuclear Energy Institute 1776 I St., N.W., Suite 400 Washington, DC 20006-3708