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TAC:ME6882, Revise Operability Requirements and Actions for RCS Leakage Instrumentation (Approved, Closed) TAC:ME6883, Revise Operability Requirements and Actions for RCS Leakage Instrumentation (Approved, Closed) |
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Category:E-Mail
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[Table view] Category:Request for Additional Information (RAI)
MONTHYEARML23352A3502023-12-19019 December 2023 Dc. Cook Nuclear Power Plant, Units 1 Biennial Licensed Operator Requalification Program Inspection and Request for Information ML23310A1152023-11-0606 November 2023 Notification of the NRC Baseline Inspection and Request for Information, Inspection Report 05000316/2024002 ML22264A1112022-09-21021 September 2022 Notification of Post-Approval Site Inspection for License Renewal - Phase IV; IR 05000315/2023010; 05000316/2023010 and RFI ML22230A4362022-08-19019 August 2022 Licensed Operator Positive Fitness-For-Duty Test ML22049B4532022-02-22022 February 2022 Notification of NRC Baseline Inspection and Request for Information Report 05000315/2022002 ML21307A3352021-11-0303 November 2021 NRR E-mail Capture - D.C. Cook Nuclear Plant Unit Nos. 1 and 2 - Final RAI - License Amendment Request Regarding Containment Water Level Instrumentation ML21215A1242021-08-0303 August 2021 Information Request to Support Upcoming Temporary Instruction 2515/194 Inspection; Inspection Report 05000315/2021013; 05000316/2021013 ML21140A3052021-05-20020 May 2021 NRR E-mail Capture - Final RAI - D.C. Cook 1 & 2 - Relief Request ISIR-4-11, Impractical Examinations for the Fourth 10-Year ISI Interval ML21049A2682021-02-16016 February 2021 NRR E-mail Capture - D.C. Cook Nuclear Plant Unit No. 2 - Final RAI - License Amendment Request for One-Time Extension Containment Type a ILRT ML21026A3472021-01-26026 January 2021 NRR E-mail Capture - Final RAI - D.C. Cook 1 & 2 - Relief Request REL-PP2 for Pump and Valve IST Fifth 10-year Interval IR 05000315/20200112020-09-17017 September 2020 Information Request to Support Upcoming Temporary Instruction 2515/194 Inspection; Inspection Report 05000315/2020011; 05000316/2020011 ML20181A4122020-06-29029 June 2020 NRR E-mail Capture - Final RAI - D.C. Cook 1 - One-Time Extension, Containment Type a ILRT Frequency ML20232B4562020-04-27027 April 2020 Request for Supporting Information for the D.C. Cook SPRA Audit Review - Fragility Questions IR 05000315/20200202020-04-10010 April 2020 Units 1 and 2 - Information Request for NRC Triennial Evaluation of Changes, Tests, and Experiments (50.59) Baseline Inspection 05000315/2020020 and 05000316/2020020 ML20232B4432020-03-25025 March 2020 Request for Supporting Information for the D.C. Cook SPRA Audit Review - Plant Response Model Questions ML19267A1362019-09-24024 September 2019 NRC Information Request (September 23, 2019) - IP 71111.08 - DC Cook U2 - Documents Selected for Onsite Review (DRS-M.Holmberg) ML19204A0962019-07-23023 July 2019 NRR E-mail Capture - D.C. Cook Nuclear Plant Unit Nos. 1 and 2 - Request for Additional Information Related to Unit 2 Leak Before Break Analysis and Deletion of Containment Humidity Monitors for Unit Nos. 1 and 2 ML19211C3032019-07-0303 July 2019 NRR E-mail Capture - D.C. Cook Nuclear Plant Unit Nos. 1 & 2 - Request for Additional Information Related to LAR to Address NSAL-15-1 ML19084A0002019-03-20020 March 2019 NRR E-mail Capture - Resubmitted: Draft Request for Additional Information DC Cook Unit 1 - Leak-Before-Break LAR ML19072A1432019-03-12012 March 2019 NRR E-mail Capture - Resubmitted: Draft Request for Additional Information DC Cook Unit 1 - Leak-Before-Break LAR ML19011A3512019-01-11011 January 2019 NRR E-mail Capture - Request for Additional Information DC Cook Unit 1 Leak Before Break Amendment ML18313A0802018-11-0808 November 2018 NRR E-mail Capture - D.C. Cook Units 1 and 2 - RAI for RPV Threads in Flange Alternative ML18263A1542018-09-14014 September 2018 NRR E-mail Capture - D.C. Cook Unit No. 1 - RAI for Leak-Before-Break LAR ML18204A3722018-07-19019 July 2018 NRR E-mail Capture - D.C. Cook Unit No. 1 - RAI for Leak-Before-Break LAR ML18143B6292018-05-29029 May 2018 Dtf. RAI ML18142B5312018-05-29029 May 2018 Request for Additional Information Concerning 2017 Decommissioning Funding Status Report ML18060A0212018-02-28028 February 2018 Enclosurequest for Additional Information (Request for Additional Information Regarding Indiana Michigan Power Company'S Decommissioning Funding Update for Donald C Cook Nuclear Plant Units 1 and 2 ISFSI) ML18043A0092018-02-0909 February 2018 NRR E-mail Capture - DC Cook, Unit 2 - Request for Additional Information Regarding Reactor Vessel Internals Again Management Program ML17304A0102017-11-0101 November 2017 Unit Nos.1 and 2 - Request for Additional Information Regarding Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools (CAC Nos. MF9444 and MF9445; EPID L-2016-LRC-0001) ML17249A7492017-08-28028 August 2017 E-Mail Sent August 28, 2017, Request for Information for Donald C. 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Cook Nuclear Power Plant, Units 1 and 2 - Request for Information for an NRC Pilot Design Bases Inspection on the Implementation of the Environmental Qualification Program Inspection Report 05000315/2016008; 05000316 ML15267A6832015-10-0505 October 2015 Second Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.8.1 ML15231A5542015-08-18018 August 2015 Ltr. 08/18/15 Donald C. Cook Nuclear Power Plant, Units 1 and 2 - Request for Information for an NRC Triennial Pilot Baseline Component Design Bases Inspection (Inspection Report 05000315/2015008; 05000316/2015008) (Axd) ML15225A5772015-08-13013 August 2015 Information Request to Support Upcoming Problem Identification and Resolution Inspection at D.C. Cook, Units 1 and 2 ML15163A1672015-06-15015 June 2015 Request for Additional Information Regarding License Amendment Request to Revise Technical Specification 3.8.1 ML15149A3832015-05-28028 May 2015 Information Request to Support NRC Annual Baseline Emergency Action Level and Emergency Plan Changes Inspection (Mxg) ML15119A3392015-05-0505 May 2015 Follow-up Request for Additional Information Concerning the Reactor Vessel Internals Aging Management Program Submittal ML15055A5702015-02-24024 February 2015 Cyber Security RFI Temporary Instruction 2201/004 ML15022A1532015-01-21021 January 2015 Dccook ISI Request for Information ML14363A4912015-01-16016 January 2015 DC Cook Nuclear Plant, Units 1 and 2 - Supplemental Information Needed for Acceptance of Requested Licensing Action to Adopt TSTF-490, Rev.0, Deletion of E-Bar Definition and Revision to Reactor Coolant System Specific Activity Technical Sp ML14246A4992014-08-26026 August 2014 NRR E-mail Capture - Donald C. Cook Nuclear Plant, Units 1 and 2 - Draft Requests for Additional Information (Scvb) Change to TS 5.5.14 by Adopting NEI 94-01, Revision 3-A ML14135A3202014-06-0606 June 2014 Request for Additional Information Concerning the Reactor Vessel Internals Afing Management Program Submittal 2023-12-19
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Accession No. ML113000267 From: Tam, Peter Sent: Thursday, October 27, 2011 10:20 AM To: 'hletheridge@aep.com'; mkscarpello@aep.com; 'jrwaters@aep.com' Cc: Tsao, John
Subject:
D. C. Cook Units 1 and 2 - Draft RAI on the Alloy 600 Aging Management Program (TAC ME6882 and ME6883)
Joe:
By letter dated August 17, 2011, Indiana Michigan Power Company (I&M) submitted for NRC review and approval the Alloy 600 aging management program (AMP) for Donald C. Cook (Accession No. ML11238A069). I&M submitted the Alloy 600 AMP to satisfy one of the commitments in its license renewal application. The NRC staff reviewed the proposed AMP in accordance with XI.M11B, Cracking of Nickel-alloy Components and Loss of Material Due to Boric Acid-Induced Corrosion in Reactor Coolant Pressure Boundary Components (PWRs Only) in NUREG-1801, Revision 2, Generic Aging Lessons Learned (GALL) Report. To complete its review, the NRC staff needs the additional information described in this draft RAI. You may choose to accept this e-mail conveying the draft RAI as formal RAI and respond within 45 days of receipt, or you may request to discuss the draft RAI and target response date in a conference call with the NRC staff.
Draft RAI (1) To manage Alloy 600 components, the Program Description section of XI.M11B in NUREG-1801, Revision 2, references American Society of Mechanical Engineers (ASME) Code Case N-729, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1, which is part of the requirements in Title 10, Code of Federal Regulations, Paragraph 50.55a (10 CFR 50.55a). However, Section 1.7 of the proposed Alloy 600 AMP does not reference this code case. Although the licensee has replaced the Units 1 and 2 reactor vessel head with Alloy 690/52/152 material, 10 CFR 50.55a(g)(6)(ii)(D) still applies and thus should be referenced. (a) The licensee needs to cite this code case and 10 CFR 50.55a(g)(6)(ii)(D) or justify why it is not referenced in Section 1.7. (b) Section 1.7 identifies ASME Code Cases N-722-1 and N-770-1. The NRC has imposed conditions on both code cases in 10 CFR 50.55a(g)(6)(ii)(E) and 10 CFR 50.55a(g)(6)(ii)(F), respectively. Section 1.7 also needs to include the reference to these two NRC regulations. As a general rule, whenever these three ASME code cases are mentioned in a regulatory document, they should be followed with a reference to the NRC regulations in 10 CFR 50.55a(g)(6)(ii)(D), (E), or (F) as applicable.
(2) Attribute Item 1 of XI.M11B in NUREG-1801, Revision 2, provides the generic scope of an Alloy 600 AMP. Section 3.5.4 of the proposed AMP states that ...[d]etermine appropriate sample locations, inspection techniques, and acceptance standards in accordance with industry guidelines (a) In addition to industry guidelines, Section 3.5.4 of the proposed AMP needs to state that the sample locations, inspection techniques and acceptance standards are also determined in accordance with 10 CFR 50.55a and relevant NRC generic communications or justify why NRC regulations are not included. (b) Discuss the inspection samples for the components identified in
Attachment 2 (e.g., how many welds and/or components will be inspected) during each inspection. (c) Attachment 1 does not include the following components: reactor coolant pump nozzle welds, and reactor coolant piping branch lines. Attachment 2 does not include the following components: the steam generator shell cladding in Unit 1; the hot and cold leg nozzles, pressurizer nozzle (1-PRZ-23) to safe-end weld, steam generator primary manway diaphragm, steam generator primary nozzle-to-safe-end weld in Unit 2.
Discuss why these components are not covered under the scope of the proposed Alloy 600 AMP.
(3) Attribute Item 3 in XI.M11B in NUREG-1801, Revision 2, of the GALL report specifies parameters that are to be monitored or inspected. (a) Discuss the parameters that will be monitored and inspected under the proposed Alloy 600 AMP. (b) Attachment 2 specifies Visual as the inspection method for various steam generator components such as divider plate, tubesheet cladding, manway diaphragm, nozzle dam rings and shell cladding. Discuss whether the visual inspection is equivalent to ASME Code,Section XI, IWA-2210, visual examination category VT-1, VT-2, or VT-3. If not, explain what the visual inspection is (e.g., what is being examined and what are acceptance criteria). (c) Attachment 2 cited ASME Code Cases N-770-1, N-729-1, and N-722-1 for the required inspection frequency without citing 10 CFR 50.55a(g)(6)(ii)(D), 10 CFR 50.55a(g)(6)(ii)(E), and 10 CFR 50.55a(g)(6)(ii)(F) which impose conditions on these code cases. Include these regulations in Attachment 2 or justify why these NRC regulations are not referenced in Attachment 2.
(4) Attribute Item 4 in NUREG-1801, Revision 2, Detection of Aging Effects, states that Reactor coolant pressure boundary leakage can be monitored through the use of radiation air monitoring and other general area radiation monitoring, and technical specifications for reactor coolant pressure boundary leakage Discuss the capability of the RCS leakage detection systems in Units 1 and 2 with respect to Regulatory Guide 1.45, Revision 1, Guidance on Monitoring and Responding to Reactor Coolant System Leakage.
(5) Attribute Item 5 in NUREG-1801, Revision 2, Monitoring and Trending, states that reactor coolant pressure boundary leakage is calculated and trended on a routine basis in accordance with technical specification to detect changes in the leakage rates.
Explain why this attribute is not included in the proposed AMP.
(6) Attribute Item 6 in NUREG-1801, Revision 2, specifies the acceptance criteria for all indications of cracking and loss of material due to boric acid-induced corrosion. Section 4.7.1 of the proposed AMP states that [i]nspection results that do not meet the acceptance criteria are documented via PMP-7030-CAP-001, Action Initiation The acceptance criteria are not clearly defined in the proposed AMP. The proposed AMP needs to include a section on acceptance criteria for the inspection results. The acceptance criteria section should include the individual criterion for the inspected components as defined in Attachment 2.
(7) Attribute Item 7 of XI.M11B in NUREG-1801, Revision 2, Corrective Actions, specifies that relevant flaw indications of susceptible components found to be unacceptable for further services are corrected through implementation of appropriate repair or replacement as dictated by 10 CFR 50.55a and industry guidelines (e.g., MRP-139). In addition, detection of leakage or evidence of cracking in susceptible components require scope expansion of current inspection and increased inspection frequencies of some
components, as required by 10 CFR 50.55a and industry guidelines (e.g., MRP-139).
Section 4.7.5 of the proposed AMP states that based on the initial inspection results, the need for additional inspections are determined. This information is used to develop future inspection scope and associated inspection intervals. Subsequent inspections may include inspections of the additional locations If the initial inspection discovers degradation in Alloy 600 components, discuss the scope and intervals of the additional inspection with respect to the ASME Code,Section XI requirements.
(8) Attribute Item 8 of XI.M11B in NUREG-1801, Revision 2, Confirmation Process, and Attribute Item 9, Administrative Controls, references 10 CFR Part 50, Appendix B.
Section 4.7.2 of the proposed AMP states that WHEN the test acceptance criteria are not met, THEN an Engineering Evaluation is performed in accordance with 10 CFR Part 50, Appendix B, and documented in accordance with PMP-7030-CAP-002, Condition Evaluation, Action, and Closure, in order to verify that the intended functions of the in-scope components can be maintained consistent with the current licensing basis. [Ref.
5.2.1a] (a) Clarify why 10 CFR Part 50, Appendix B, is used as the basis for the engineering evaluation, in lieu of 10 CFR 50.55a which requires the use of the evaluation in the ASME Code,Section XI. Discuss the procedures that will be used to implement the requirements 10 CFR Part 50, Appendix B. (b) Discuss whether the engineering evaluation would include corrective actions.
Peter S. Tam, Senior Project Manager (for D.C. Cook and Monticello)
Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Tel. 301-415-1451