CNL-23-019, Response to Request for Additional Information Regarding Browns Ferry Nuclear Plant, Units 1, 2, and 3, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Request for Alternative BFN-0-ISI-32
| ML23070A004 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/11/2023 |
| From: | Hulvey K Tennessee Valley Authority |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| EPID L-2022-LLR-0062, CNL-23-019 | |
| Download: ML23070A004 (1) | |
Text
1101 Market Street, Chattanooga, Tennessee 37402 CNL-23-019 March 11, 2023 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296
Subject:
Response to Request for Additional Information Regarding Browns Ferry Nuclear Plant, Units 1, 2, and 3, American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Request for Alternative BFN-0-ISI-32 (EPID L-2022-LLR-0062)
References:
- 1. TVA letter to NRC, CNL-22-025, Browns Ferry Nuclear Plant, Units 1, 2, and 3 - American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Request for Alternative BFN-0-ISI-32, dated August 22, 2022 (ML22234A271)
- 2. NRC email to TVA, "Request for Additional Information Related to TVA Alternative Request BFN-0-ISI-32 (CNL-22-025) (EPID L-2022-LLR-0062),"
dated February 9, 2023 (ML23041A002)
In Reference 1, Tennessee Valley Authority (TVA) submitted a request for alternative (RFA) for the Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, from Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Specifically, the RFA proposed an alternative to the inservice inspection (ISI) requirements for volumetric examination of the inner radius of standby liquid control nozzles. The duration of the proposed alternative request is through the remainder of the third, fifth, and fourth ISI intervals for BFN Units 1, 2, and 3, respectively, which are scheduled to end on January 31, 2026.
In Reference 2, the Nuclear Regulatory Commission issued a request for additional information (RAI) and requested that TVA respond by March 11, 2023. The enclosure to this letter provides the response to the RAI.
U. S. Nuclear Regulatory Commission CNL-23-019 Page 2 March 11, 2023 There are no new regulatory commitments contained in this letter. Please address any questions regarding this submittal to slrymer@tva.gov.
Respectfully, Kimberly D. Hulvey Director, Nuclear Regulatory Affairs
Enclosure:
Response to NRC Request for Additional Information cc (Enclosure):
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant NRC Project Manager - Browns Ferry Nuclear Plant Respectfully, Digitally signed by Edmondson, Carla Date: 2023.03.11 17:49:35
-05'00'
Enclosure CNL-23-019 E1 of 4 Response to NRC Request for Additional Information
==
Introduction:==
By letter dated August 22, 2022 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML22234A271), Tennessee Valley Authority (TVA), requested an alternative to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) at Browns Ferry Nuclear Plant (Browns Ferry or BFN),
Units 1, 2, and 3, pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(z)(2). TVA proposed to perform a VT-2 visual examination of the inner radius of the standby liquid control (SLC) nozzle attached to the reactor vessel as part of system leakage testing during plant startup in lieu of the ASME Code required ultrasonic examination for the remainder of the third, fifth, and fourth inservice inspection (ISI) intervals at Browns Ferry, Units 1, 2, and 3, respectively, which are scheduled to end on January 31, 2026.
The NRC staff requests additional information (RAI) to complete its review of the alternative request.
Regulatory Basis:
The regulations at 10 CFR 50.55a(g), Preservice and inservice inspection requirements, require, in part, that the inservice inspection (ISI) of ASME Code Class 1, 2, and 3 components be performed in accordance with Section XI of the ASME Code and applicable edition and addenda.
Pursuant to 10 CFR 50.55a(g)(4), Inservice inspection standards requirement for operating plants, ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements set forth in the ASME Code,Section XI to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(a)(1)(ii),
ASME Boiler and Pressure Vessel Code,Section XI, 18 months prior to the start of the 120-month interval, subject to the conditions listed in 10 CFR 50.55a(b)(2), Conditions on ASME BPV Code,Section XI.
Pursuant to 10 CFR 50.55a(z), Alternatives to codes and standards requirements, alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC if (1) the proposed alternatives would provide an acceptable level of quality and safety or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.
Enclosure CNL-23-019 E2 of 4 Requests for Additional Information:
RAI 1
Issue Section IV of the alternative request states that No leakage from this nozzle inner radius has ever been observed on any unit at BFN However, this statement does not preclude the possibility that repairs have been made to the SLC nozzles.
Section IV states that all three BFN units did not inspect the SLC nozzle inner radii in the first 10-year ISI interval.Section IV also states that with the exception of the BFN Unit 2 second 10-year ISI interval for which relief was requested, ultrasonic examination results have been obtained as required for each successive interval.
Request Discuss whether any repairs have been made to SLC N-10 nozzle inner radii at BFN Units 1, 2, and 3. If so, provide the following:
(a) a brief description of the repair.
(b) discuss how the degradation of the SLC nozzle was detected.
(c) discuss whether extent of condition inspections were performed.
(d) discuss whether the known degradation will affect the adequacy of the proposed alternative of not performing the required volumetric examination of the SLC nozzle inner radius.
TVA Response to RAI 1 There have been no repairs made to the SLC N-10 nozzle inside radii (N10-IR) on any of the three units at BFN. A review of the TVA corrective action program confirmed that no such repairs occurred. Additionally, each of the completed ASME Section XI ultrasonic testing non-destructive examination reports from the N-10 nozzles were reviewed again to confirm that no prior indications have been identified in the N10-IR area.
RAI 2
Issue The alternative request states that performing an ultrasonic examination of the SLC nozzle inner radius location would constitute a hardship or unusual difficulty without a compensating increase in quality or safety.Section IV of the alternative request states that the SLC nozzle is located in the bottom head of the vessel, in an area that is inaccessible without extensive disassembly, hindering the ability to complete the Code required volumetric or approved Code alternative VT-1 examinations from the inside surface of the reactor pressure vessel (RPV).
Request Provide additional information regarding the inaccessibility and disassembly of equipment that constitutes the hardship or difficulty for performing the VT-1 from the inside surface.
Enclosure CNL-23-019 E3 of 4 TVA Response to RAI 2 For BFN Units 1, 2, and 3, the N10-IR surface is located in the RPV lower plenum, below the shroud baffle plate, and directly underneath the non-removable shroud access hole cover. The only way to access this RPV area would require disassembly of a jet pump by removing the hold down beam and removing the inlet mixer portion of the jet pump assembly. Then a camera could be inserted through the jet pump diffuser to gain access to the lower plenum underneath the shroud baffle plate. However, the closest jet pump location where access is possible would be approximately 30 degrees away from the N-10 nozzle. Therefore, even with extensive reactor disassembly, a VT-1 examination would not be possible using readily available camera systems. Based on this hardship, BFN does not implement Nuclear Regulatory Commission endorsed Code Case N-648-1.
RAI 3
Issue Section V of the alternative request states that BFN Units 1, 2, and 3 will perform a VT-2 visual examination of the subject SLC nozzles each refueling outage in conjunction with the Class 1 system leakage test. The alternative request did not describe in detail how the VT-2 examination will be conducted nor the specific ASME Code,Section XI requirements. This information is relevant to the staffs review of the alternative request because it allows the staff to determine the adequacy of the VT-2 examination in lieu of the required ultrasonic examination.
Request (a) State whether the system leakage test and associated VT-2 examination will be performed in accordance with (1) Table IWB-2500-1, Examination Category B-P, Item Number B15.10 and B15.20, and (2) IWA-5000 and IWB-5220 of the ASME Code,Section XI.
(b) Discuss whether the insulation will be removed from the SLC nozzle area prior to the VT-2 examination. If the insulation will not be removed, discuss how the leakage from the SLC nozzle could be detected.
(c) Discuss how the potential leakage from the SLC nozzle can be differentiated from other nozzle leakage if leakage occurs.
TVA Response to RAI 3 (a) VT-2 visual examinations are performed each refueling outage in accordance with ASME Code Section XI, Table IWB-2500-1, Examination Category B-P, Item Number B15.10.
Item Number B15.20 is not applicable to this location because the N-10 nozzle areas are pressurized and tested in accordance with IWB-5222(a), with the plant in the normal operation startup alignment. The VT-2 examination method is used to conduct the system leakage test of this N-10 nozzle location following each refueling outage in accordance with IWA-5000 and IWB-5220.
(b) In accordance with IWA-5240, the system leakage test is conducted without removing the insulation. However, TVA utilizes a site-specific augmented examination requirement for direct visual examination of the N-10 nozzle safe-end to pipe weld, which is located directly adjacent to the subject nozzle-to-vessel inside radius. TVA procedures require the removal of grating, scaffold erection, and removal of the insulation associated with the nozzle
Enclosure CNL-23-019 E4 of 4 safe-end to pipe weld, thus gaining limited access to the nozzle forging. Even if insulation removal was not performed, leakage from this nozzle would still be readily apparent from underneath the vessel because the vessel insulation and vessel nozzle insulation is constructed of overlapping panels that are not sealed. Because of the proximity of this augmented examination location to the subject inside surface, leakage from the nozzle would be even more evident.
(c) System leakage test procedures include steps to require that the source of discovered leakage be identified prior to completion of the test. In accordance with BFN Units 1, 2, and 3 Technical Specification 3.4.4, RCS Operational LEAKAGE, no reactor coolant system pressure boundary leakage is permitted. Indications of leakage during performance of the Class 1 system leakage test VT-2 visual examinations must be reported so the leak source can be positively identified and appropriately corrected. Any leakage identified emanating from inside the reactor vessel insulation or vessel nozzle insulation would require direct access, including insulation removal, if necessary, for identification of the leak source and any necessary corrective measures or repair/replacement activities. Therefore, the leak investigation process would differentiate any SLC nozzle leakage from other leakage sources.