ML25079A163
| ML25079A163 | |
| Person / Time | |
|---|---|
| Issue date: | 03/20/2025 |
| From: | Leslie Terry NRC/NRR/DNRL/NCSG |
| To: | |
| Shared Package | |
| ML25079A162 | List: |
| References | |
| Download: ML25079A163 (6) | |
Text
March 20, 2025 U.S. Nuclear Regulatory Commission Public Meeting Summary
Title:
Meeting with the Industry Steam Generator Task Force Meeting Identifier: 20250079 Date of Meeting: March 6, 2025 Location: Webinar Type of Meeting: Observation Meeting Purpose of the Meeting: The purpose of this meeting was for the U.S. Nuclear Regulatory Commission (NRC) staff to discuss steam generator (SG) issues with the industry Steam Generator Task Force (SGTF).
General Details: The industry SGTF met with NRC staff on March 6, 2025, by webinar.
The purpose of the meeting was to discuss a variety of SG issues. The NRC and industry slides are available in the Agencywide Documents Access and Management System (ADAMS) under Accession Nos. ML25062A178 and ML25064A034, respectively. This meeting was noticed as a public meeting and the agenda is available in ML25024A180.
LISTING OF ATTENDEES U.S. NRC MEETING WITH THE INDUSTRY STEAM GENERATOR TASK FORCE March 6, 2025 Participant Affiliation Participant Affiliation Virginia Allen TVA Sean Kil EPRI Isaac Anchondo-Lopez NRC Paul Klein NRC Randall Atkinson EPRI Mike Liu Intertek Sasan Bakhtiari ANL Greg Makar NRC James Benson EPRI Dan Mayes Duke Energy Cotasha Blackburn Southern Company Victor Newman Framatome Steven Bloom NRC Scott Portzline TMI Alert Jasmyn Bone Entergy Paul Rittenhouse DEV Generation - 3 Steve Brown Entergy Miranda Ross NRC Brent Capell EPRI Jay Smith Westinghouse Brad Carpenter Westinghouse Leslie Terry NRC Russ Cipolla Intertek Kester Thompson FPL Helen Cothron EPRI Maxwell Trent TVA Jacob Fakory DEP Nicole Vitale Westinghouse Michael Frotscher Entergy Pat Wagner Wolf Creek Rich Guill EPRI Bill Wiltsey Intertek Andrew Johnson NRC Summary of Presentations: Industry representatives made presentations on recently published Electric Power Research Institute (EPRI) reports, the status of revisions to industry guidelines, recent domestic operating experience, and provided updates on a study related to SG tube wear due to deposits, and on making test samples for eddy current Examination Technique Specification Sheets (ETSSs).
The NRC staff presented feedback on the information being provided by licensees in SG tube inspection reports (SGTIRs) based on the reporting requirements in Technical Specifications Task Force (TSTF) Traveler TSTF-577.
Additional details of the information exchanged during the meeting is provided below.
Recently published EPRI reports on the following topics were discussed:
Noise monitoring of SGs with thermally treated Alloy 690 tubing.
Impacts of diethyl hydroxylamine on resin performance.
Prototype web application and lessons learned related to SG digital twin.
Investigation of the thermal hydraulic condition of helical coil SG tubes and a high-level comparison with conventional recirculating and once-through SGs.
Methods for estimating blockage of broached SG tube support plates.
Characterization of SG deposits as it related to tube wear.
An equivalency study of ETSSs for Alloy 800 SG tubing.
Industry stated that a revision to the Steam Generator Management Program: In Situ Pressure Test Guidelines is in progress and a revision to the Steam Generator Management Program: Pressurized Water Reactor Steam Generator Examination Guidelines (Examination Guidelines) is planned for 2025.
Industry discussed domestic operating experience from fall 2023 related to loss of SG tube integrity at tube support plates and associated Steam Generator Management Program (SGMP) actions. It was stated that the cause was a non-conservative sizing curve used in the 2017 SG tube inspection that resulted in the under sizing of the measured flaws. An industry working group is considering changes to the Steam Generator Management Program: Steam Generator Integrity Assessment Guidelines (Integrity Assessment Guidelines) related to using checklists for operational assessment inputs, checking to ensure flaws are sized accurately, incorporating new SG wear requirements, highlighting sensitive parameters for fully probabilistic operational assessments, and providing more details for probabilistic operational assessments. In addition, the industry discussed tube integrity training that was provided in February 2025 and the development of software to check vendor-supplied operational assessments.
An international utility concluded tube wear has been caused by pieces of oxide scale that spalled from and then lodged against tube surfaces in two units at one plant. The international utility also concluded that dense deposit flakes (i.e., with a porosity level less than 5 percent) and a dense layer thickness of greater than or equal to 0.1mm could potentially result in tube wear. The corrective actions by the international utility include performing chemical cleaning when the dense scale layers exceed 0.1mm and monitoring scale buildup by retrieving scale during each outage to verify scale thickness and to perform wear tests.
In the U.S., the SGMP performed a study that analyzed 65 deposit flake samples removed during SG sludge lancing at 15 different units (6 in the U.S., 6 in Asia, and 3 in Europe) that represent a variety of SG designs, locations, operating chemistry, and tubing materials. The characterization focused on factors related to wear: hardness, composition, porosity, and thickness. Three flakes had porosity less than or equal to 5 percent and one of the three flakes from an international plant contained a dense region of greater than or equal to 0.1mm. No U.S. plants have reported tube wear from tube deposit flakes. In general, chemical cleaning increased flake porosity, decreased flake hardness, and decreased flake thickness. Therefore, the study concluded that chemical cleaning may mitigate the risk of tube wear due to deposit flakes. The study is documented in EPRI Technical Report 3002029280, Steam Generator Deposit Characterization to Address Tube Wear Issues.
An industry representative provided a status update on tube samples with laboratory-produced stress corrosion cracking (SCC). The information from the destructive analyses will be used to support ETSS development and Model Assisted Probability of Detection (MAPOD) calculations. A destructive analysis report on 12 axial primary water SCC (PWSCC) and 17 circumferential outside diameter (ODSCC) samples is complete.
Destructive analysis of circumferential PWSCC is ongoing and is scheduled to be completed in summer 2025. A destructive analysis report on 26 axial ODSCC samples is complete, and additional axial PWSCC and circumferential ODSCC flaw development is ongoing at Dominion Engineering. With regards to ETSS development, the SGMP has completed its quality assurance process, flaw injection process with eddy current data analysis, and tube integrity analysis for the 12 axial PWSCC and 17 circumferential ODSCC samples. A peer review in 2025 is planned for ETSSs 20501.1 (X-Probe) and 20511.1 (+Point') related to axial PWSCC at expansion transitions, and ETSSs 20400.1. (X-Probe) and 21410.1
(+Point') related to circumferential ODSCC at expansion transitions. All four of these ETSSs are being transitioned from Examination Guidelines Appendix H to Appendix I using MAPOD.
Technical specifications require a SGTIR be submitted to the NRC. Appendix G of the EPRI Integrity Assessment Guidelines provides a SGTIR template for plants adopting TSTF-577 or converting to Revision 5 of the Standard Technical Specifications1 (STS). The EPRI Integrity Assessment Guidelines recommends that plants not adopting TSTF-577 consider using the template. Per industry request, the NRC provided feedback on the information being provided by licensees in SGTIRs based on the reporting requirements in TSTF-577.
Specifically, the NRC discussed the nondestructive examination techniques utilized, tube wear at support structures less than 20 percent through-wall (TW), and the number of plugged tubes. In addition, the staff provided SGTIRs examples, including where this information was most clearly reported.
STS Section 5.6.7.c.1, Revision 5, requires the report to include the nondestructive examination techniques used for each degradation mechanism found. The NRC staff noted that SGTIRs identifying the ETSSs used; whether the ETSSs were used for detection or 1 NUREG-1430, Revision 5, Standard Technical Specifications Babcock and Wilcox Plants, dated September 2021 (ML21272A363 (Volume 1) and ML21272A370 (Volume 2).
NUREG-1431, Revision 5, Standard Technical Specifications Westinghouse Plants, dated September 2021 (ML21259A155 (Volume 1) and ML21259A159 (Volume 2).
NUREG-1432, Revision 5, Standard Technical Specifications Combustion Engineering Plants, dated September 2021 (ML21258A421 (Volume 1) and ML21258A424 (Volume 2)).
sizing, or both; and whether technique extension was used to classify degradation outside the bounds of the ETSSs results in a most efficient review. STS Section 5.6.7.c.2, Revision 5, provides the reporting requirements for all service-induced indications detected during the inspection (i.e., location, orientation, if linear, measured size, if available, and voltage response (detailed reporting)). It provides additional information on reporting tube wear at support structure indications less than 20 percent TW. Specifically, licensees that have adopted TSTF-577 or converted to Revision 5 of the STS have the option to provide only the total number of tube wear at support structure indications less than 20 percent TW.
Tube wear at support structures refers to tube wear due to tube contact with support structures (e.g., anti-vibration bars). The NRC staff discussed a recent SGTIR that provided the total number of wear indications at support structures but did not indicate how many of those were less than 20 percent TW. STS Section 5.6.7.c.4, Revision 5, requires the number of tubes plugged during the inspection outage. The NRC staff noted that SGTIRs that report, in addition to the number of tubes plugged due to degradation, the number of preventively plugged tubes and why they were preventively plugged results in a most efficient review.
A member of the public from TMI Alert made comments regarding tube-to-tube wear identified in the replacement SGs at Three Mile Island Nuclear Station, Unit 1, and related to a previously submitted and dispositioned § 2.206 of Title 10 of the Code of Federal Regulations petition (ML19085A335 and ML19189A333).
If you have any questions regarding this meeting summary, please feel free to contact Leslie Terry by phone at 301-415-1167, or by email at Leslie.Terry@nrc.gov.
Attachments:
- 1. Meeting Notice:
- 2. NRC Slides:
- 3. Industry Slides:
- 4. Package:
Package: ML25079A162 Meeting Summary: ML25079A163 Meeting Notice: ML25024A180 NRC Slides: ML25062A178 Industry Slides: ML25064A034 OFFICE NRR/DNRL/NCSG NRR/DNRL/NRLB/LA NRR/DNRL/NCSG/BC NAME LTerry SGreen SBloom DATE 3/20/2025 3/20/2025 3/20/25