CNL-21-037, Request for Alternative, 21-ISI-1 Alternative Repair and Examination for Mid Canopy Seal Weld Replacement
| ML21085A764 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  (DPR-077, DPR-079) |
| Issue date: | 03/26/2021 |
| From: | Polickoski J Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CNL-21-037 | |
| Download: ML21085A764 (8) | |
Text
1101 Market Street, Chattanooga, Tennessee 37402 CNL-21-037 March 26, 2021 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Sequoyah Nuclear Plant, Units 1 and 2 Renewed Facility Operating License Nos. DPR-77 and DPR-79 NRC Docket Nos. 50-327 and 50-328
Subject:
Sequoyah Nuclear Plant, Units 1 and 2, Request for Alternative, 21-ISI-1 Alternative Repair and Examination for Mid Canopy Seal Weld Replacement In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a, Codes and Standards, paragraph (z)(2), Tennessee Valley Authority (TVA) is submitting a request for alternative for the Sequoyah Nuclear Plant (SQN), Units 1 and 2, from the ASME [American Society of Mechanical Engineers] Boiler and Pressure Vessel Code (ASME Code),Section XI, and the original Construction Code requirements to perform defect removal, surface examination of the defect removal area, and a final surface examination on the control rod drive mechanism (CRDM) mid canopy seal welds. Relief is requested on the basis that compliance with the specified requirements would result in hardship (significant radiological exposure and industrial safety risks) without a compensating increase in the level of quality and safety. This request is based on the potential for removal and reinstallation of the seal weld to support replacement of an internal latch assembly, and the potential to modify the seal weld by use of a designed seal weld overlay in the case that a defect identified in the seal weld or seal weld mating base materials cannot be removed.
Because of the physical space limitations, hazardous personnel access aspects, and in consideration of radiation exposure, TVA proposes to perform an enhanced visual examination as an alternative to the liquid penetrant examination required by the original Construction Code for the final weld. In addition, if a modified seal weld overlay is required due to an identified defect that cannot be removed, a weld buildup will be installed over the subject canopy seal weld without removal of the defect.
U. S. Nuclear Regulatory Commission CNL-21-037 Page 2 March 26, 2021 TVA is requesting approval of this relief request for the SQN, Units 1 and 2, for the remainder of the current inservice inspection (ISI) interval. SQN, Units 1 and 2, are currently in the fourth 10-Year ISI interval scheduled to end on April 30, 2025.
The enclosure to this letter provides relief request 21-ISI-1 that requests relief from the requirements of the ASME Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, ASME Code Class 1.
The ASME Code,Section XI, 2007 Edition through 2008 Addenda is the code of record for the SQN, Units 1 and 2.
10 CFR 50.55a(z)(2) authorizes the NRC to grant an alternative to the requirements of paragraphs (b) through (h) of 10 CFR 50.55a when compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
TVA requests approval of this relief request within 12 months from the date of this letter. In the event relief is needed prior to that date (for example, if it is needed to support an outage), TVA will notify the NRC at that time.
There are no new regulatory commitments contained in this letter. If you have any questions regarding this submittal, please contact Kimberly D. Hulvey, Senior Manager, Fleet Licensing at 423-751-3275.
Respectfully, James T. Polickoski Director, Nuclear Regulatory Affairs Enclosure Sequoyah Nuclear Plant, Units 1 and 2, American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Inservice Inspection Program, Request for Alternative, 21-ISI-1 cc:
NRC Regional Administrator - Region II NRC Senior Resident Inspector - Sequoyah Nuclear Plant NRC Project Manager - Sequoyah Nuclear Plant
Enclosure CNL-21-037 E1 of 6 Sequoyah Nuclear Plant, Units 1 and 2, American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Inservice Inspection Program, Request for Alternative, 21-ISI-1 I.
ASME Code Component(s) Affected:
Code Class:
1
Reference:
IWA-4000 Examination Category:
NA Item Number:
NA
==
Description:==
Alternative Requirements for Repair/Replacement of Control Rod Drive Mechanism (CRDM) Canopy Seal Welds in Accordance with IWA-4000 Component Number:
Reactor CRDM Canopy Seal Welds - Class 1 Appurtenance to the Reactor Vessel II.
ASME Code Edition and Addenda:
Sequoyah Nuclear Plant (SQN), Units 1 and 2 are in the fourth inservice inspection (ISI) interval. The ISI Code of Record is the ASME [American Society of Mechanical Engineers]
Boiler and Pressure Vessel Code (ASME Code),Section XI, 2007 Edition through 2008 Addenda, as conditioned by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a.
Throughout this submittal, references to ASME Code Section XI are from the 2007 Edition through 2008 Addenda unless otherwise specified.
The original Construction Code for 53 of the CRDM housings on each unit, including the mid canopy seal welds, is ASME Code Section III, Summer 1968 Edition. An additional four CRDM Housing assemblies on each unit, which were purchased separately, are certified to ASME Code Section III, 1971 Edition through Winter 1972 Addenda. ASME Code Section XI, 1968 Edition, does not include specific fabrication and examination requirements for specially designed seal welds, and ASME Code Section XI, IWA-4221 allows the application of later editions and addenda of the Construction Code. Therefore, for the purpose of this submittal, TVA has applied the technical requirements of the Winter 1972 Addenda of ASME Code Section III to the CRDM housing assemblies for SQN, Units 1 and 2.
III.
Applicable Code Requirement
IWA-4000 of ASME Code Section XI requires that replacements or repairs be performed in accordance with the owner's original Construction Code of the component or system, or later editions and addenda of the Code.
The use of ASME Code Section XI, IWA-4340 is disallowed by 10 CFR 50.55a(b)(2)(xxv)(A) such that identified defects cannot be mitigated by modification, without removal of the defect.
Instead, as noted in ASME Code Section XI, IWA-4412, the requirements of ASME Code Section XI, IWA-4420 must be applied to correct any defect identified in the subject welds.
Enclosure CNL-21-037 E2 of 6 The canopy seal weld is described in ASME Code Section III, and a repair to this weld would require the following activities:
- 1. Excavation of the rejectable indications,
- 2. A surface examination of the excavated areas,
- 3. Rewelding and restoration to the original configuration and materials, and
- 4. Final surface examination.
IV.
Reason for Request
In accordance with 10 CFR 50.55a(z)(2), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Beginning with the SQN, Unit 1 Cycle 24 and Unit 2 Cycle 24 refueling outages, scheduled for April and October 2021 respectively, TVA plans to visually inspect a sample population of CRDM latch assemblies and associated penetration driveline components. The visual inspection will be performed remotely from underneath the reactor vessel head, by inserting a camera into the CRDM guide funnels. The purpose of this non-Code inspection is to assess the condition of the latch assemblies and driveline components currently in operation. In the event that damage exceeding acceptance criteria is identified on a control rod drive (CRD) latch assembly, the affected latch assembly may have to be replaced. The replacement consists of cutting out the mid canopy seal weld (see Figure 1), disconnecting the rod travel housing, and removing and replacing the inner latch assembly. Following replacement, the existing rod travel housing will be reinstalled onto the latch housing and the mid canopy seal weld will be remade to its original designed condition.
Because the latch housing and the rod travel housing are Class 1 pressure boundary items, removal and reinstallation of the mid canopy seal weld is a Repair/Replacement activity, under the jurisdiction of ASME Code Section XI, IWA-4000.
To comply with ASME Code Section XI and the original Construction Code, a surface examination must be performed on the completed seal weld. The CRDM mid canopy seal welds are located above the Reactor Vessel Closure Head, providing extremely limited access for workers to reach the assembled weld for examination. This highly congested area is also subject to high radiation levels. In order to reduce the exposure to personnel involved in the welding process, the repair activities are planned to be performed remotely using robotic equipment to the extent practical. However, the required dye penetrant (PT) examinations would necessitate hands on access to the mid canopy weld. SQN does not have representative radiological surveys at the exact location of this proposed welding activity, and the dose rates vary from each CRDM location based on distance away from the reactor vessel centerline. Past surveys performed at the elevation of the lower canopy seal welds, one to two feet outside the peripheral CRDM housings, are the most representative data available. Assuming the radiation levels at these two locations are similar, and estimating the time required to perform the PT examination for a single seal weld, the anticipated radiological exposure for this nondestructive examination (NDE) activity is estimated to be approximately 1000 to 1500 millirem. In addition, equipment disassembly would be required, only to obtain limited access for an individual working from a suspended bosuns chair harness. These additional support activities would result in further exposure and unnecessary industrial safety concerns. Due to the extremely limited access and high dose rates, compliance with this ASME Code requirement would not
Enclosure CNL-21-037 E3 of 6 meet the intent of the sites as low as reasonably achievable (ALARA) radiological control program, and presents a hardship to the utility and workers.
If the above contingency Repair/Replacement activity is performed, additional non-Code visual examinations will be performed on the seal weld and associated base metal mating surfaces as part of the seal weld disassembly, and non-Code visual and/or surface examinations will be performed on these surfaces of the rod travel housing, once the housing is removed. In the unlikely event that this NDE, or other examinations associated with the Repair/Replacement activity, detects a rejectable flaw in the mid canopy seal weld mating surface materials that cannot be removed due to accessibility or tooling limitations, additional contingency relief is being requested from the defect removal and subsequent surface examination requirements of ASME Code Section XI and ASME Code Section III, such that the affected canopy seal weld can be modified by installation of a weld overlay.
V.
Proposed Alternative and Basis for Use:
If any of the CRDM housing mid canopy seal welds have to be replaced or modified with a seal weld overlay, TVA will perform a remote enhanced visual examination on the completed weld, in lieu of the surface examination required by Code. The new weld will be examined for quality of workmanship and discontinuities will be evaluated and dispositioned to ensure the adequacy of the new leakage barrier. The proposed remote enhanced visual examination would be conducted using a video camera with a minimum of 5X magnification. Lighting and acuity will be verified using ASME Code Section XI, Table IWA-2211-1, requirements for VT-1 note (2).
TVA will perform a demonstration examination for the Authorized Nuclear Inservice Inspector using the remote video equipment system, to verify the alternative technique is adequate to ensure quality of workmanship and detect any relevant discontinuities.
In the event a defect is identified in the seal weld or adjoining base materials that cannot be removed, TVA will replace the original CRDM housing mid canopy seal weld with a modified seal weld overlay. The weld overlay will be analyzed in accordance with the Construction Code as stated in the analysis of record for the existing canopy seal of the CRDM. In accordance with this proposed alternative, this modified seal weld may be installed without removal of the defect, and without the associated surface examination of the defect removal area prior to welding.
TVA will perform a remote enhanced visual examination on the completed weld, in lieu of the surface examination required by Code. The modified seal weld overlay will serve as a full replacement for the secondary leakage barrier, but will not serve any structural function, as the full structural load is still carried by the threaded connection in accordance with the original CRDM housing design.
Basis for Use The threaded joint between the rod travel housing and latch housing provides the primary reactor coolant system Class 1 pressure boundary and structural support for the CRDM.
The canopy seal weld is for secondary leakage prevention only and is not credited in the ASME Code structural analysis. If either of these contingencies are needed, the mid canopy seal weld or modified seal weld overlay will be installed in accordance with the requirements of ASME Section III, using weld procedures qualified in accordance with ASME Code Section IX.
These activities will be performed under the control of the ASME Code Section XI Repair/Replacement program.
Enclosure CNL-21-037 E4 of 6 The proposed alternative enhanced visual examination technique provides higher resolution and consistency than that provided by the requirements of a visually unaided Code VT-1 visual examination, and is comparable to relevant indications detectable using PT surface exam technique. Based on the remote enhanced visual examination systems ability to resolve demonstrated graduation, reasonable assurance of the weld integrity is provided by this proposed alternative.
There is no applicable ASME Code Section XI, Examination Category or Item Number associated with this configuration as canopy seal welds are not subject to Table IWB-2500-1 surface or volumetric examinations.
A VT-2 inservice leakage examination is performed on the Class 1 pressure boundary at the conclusion of each refueling outage, as required by ASME Code Section XI, Table IWB-2500-1, Examination Category B-P, All Pressure Retaining Components, Item Number B15.10. In addition, a non-Code leakage examination is performed through the reactor vessel head shroud duct openings at the beginning of each refueling outage in accordance with the SQN Boric Acid Corrosion Control Program.
In the unlikely event of leakage from the newly installed seal weld, or modified seal weld overlay, these examinations are designed to promptly identify and correct the issue. Because the rod travel housing and latch housing are both made from corrosion resistant material, and the structural threaded connection provides a torturous flow path for leakage, this type of minor leakage will not impact the integrity of any safety related pressure boundary components.
As listed under Section VII, Precedents, this proposed alternative seal weld repair and alternative examination method have been previously implemented on the canopy seal welds at SQN Unit 1, Watts Bar Plant Unit 1, and other utilities. The automated weld repair and alternate visual examination methods result in significantly lower radiation exposure because the equipment is remotely operated.
VI.
Duration of Proposed Alternative:
A sample of CRDM latch assemblies and the associated penetration driveline components will be examined in various outages throughout the ISI interval. Therefore, this alternative is being requested for the remainder of the fourth ISI interval for both SQN, Units 1 and 2. The fourth interval is scheduled to end on April 30, 2025.
VII.
Precedents This request for alternative is similar to the following:
- 1. NRC Letter to Exelon Generation Company, LLC, Braidwood Station, Units 1 and 2 - Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (EPID L-2018-LLR-0033), dated January 17, 2019 (ML18347B419)
- 2. NRC Letter to TVA, Sequoyah Nuclear Plant, Unit 1 - Relief from ASME Code Repair Requirements for Canopy Seal Welds (TAC No. MA9095), dated September 12, 2000 (ML003749067)
- 3. NRC Letter to TVA, Relief from ASME Code Repair Requirements for Canopy Seal Welds at Watts Bar Nuclear Plant (TAC No. MA5051), dated August 25, 1999 (ML073230305)
Enclosure CNL-21-037 E5 of 6
- 4. NRC Letter to TVA, Relief from ASME Code Repair Requirements for Canopy Seal Welds Sequoyah Nuclear Power Plant, Unit 1 (TAC No. M93835), dated April 4, 1996 (LL9604290167)
- 5. NRC letter to Duke Energy Carolinas, LLC, McGuire Nuclear Station, Unit 1, Relief 08-MN-005, for Control Rod Drive Mechanism (CRDM) Canopy Seal Welds (TAC No. MD9875), dated October 14, 2009 (ML092530620)
Enclosure CNL-21-037 E6 of 6 Figure 1. Mid Canopy Seal