L-23-229, Request for Additional Information Regarding the Spring 2023 Generic Letter 95-05 Voltage-Based Alternate Repair Criteria and Steam Generator F-Star Reports
| ML23333A208 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/29/2023 |
| From: | Blair B Energy Harbor Nuclear Corp |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-23-229, EPID L-2023-LRO-0058 | |
| Download: ML23333A208 (1) | |
Text
energy harbor Barry N. Blair Site Vice President, Beaver Valley Nuclear November 29, 2023 L-23-229 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Beaver Valley Power Station, Unit No. 2 Docket No. 50-412, License No. NPF-73 Energy Harbor Nuclear C01p.
Beaver Valley Power Station P.O. Box 4 200 State Route 3016 Shippingport, PA 15077 724-682-5234 REQUEST FOR ADDITIONAL INFORMATION REGARDING THE SPRING 2023 GENERIC LETTER 95-05 VOLTAGE-BASED ALTERNATE REPAIR CRITERIA AND STEAM GENERATOR F-STAR REPORTS (EPID: L-2023-LRO-0058)
By letter dated August 7, 2023 (Accession No. ML23219A059), Energy Harbor Nuclear Corp. submitted the Generic Letter (GL) 95-05 Voltage-Based Alternate Repair Criteria (ARC) and Steam Generator (SG) F-Star (F*) Reports for Beaver Valley Power Station, Unit No. 2, associated with the spring 2023 refueling outage (2R23).
By email dated October 18, 2023, the Nuclear Regulatory Commission (NRC) staff requested additional information to complete its review of the reports. The Energy Harbor Nuclear Corp. response to the NRC request is enclosed.
There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact Mr. Phil H. Lashley, Manager - Fleet Licensing, at (330) 696-7208.
Barry N. Blair
Beaver Valley Power Station, Unit No. 2 L-23-229 Page 2
Enclosure:
Response to Request for Additional Information cc: NRC Region I Administrator NRC Resident Inspector NRR Project Manager Director BRP/DEP Site BRP/DEP Representative
Enclosure L-23-229 Response to Request for Additional Information (9 pages follow)
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 1 of 8 Revision 0 Westinghouse Electric Company LLC DMW-NRCD-RF-LR-000003-NP Revision 0 Beaver Valley Power Station, Unit 2 - Responses to Request For Additional Information -
Refueling Outage 23 Generic Letter 95-05 and Steam Generator F-Star (F*) Reports November 2023 Author: Electronically Approved
- Jay R. Smith*
Component Design and Management Programs Verifier: Electronically Approved
- David A. Suddaby*
Component Design and Management Programs Reviewer: Electronically Approved
- John S. Rees*
Component Engineering & Chemistry Operations Reviewer: Electronically Approved
- Gary W. Whiteman*
Licensing Engineering Approved: Electronically Approved
- Robert S. Chappo, Jr.*, Manager Component Design and Management Programs
©2023 Westinghouse Electric Company LLC All Rights Reserved
- Electronically approved records are authenticated in the Electronic Document Management System.
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 2 of 8 Revision 0 Beaver Valley Power Station, Unit 2 - Responses to Request For Additional Information -
Refueling Outage 23 Generic Letter 95-05 and Steam Generator F-Star (F*) Reports
Background
By letter dated August 7, 2023 (ML23219A059), Energy Harbor Nuclear Corp. submitted the Spring 2023 Generic Letter (GL) 95-05 Voltage-Based Alternate Repair Criteria (ARC) and Steam Generator (SG) F Star (F*) Reports for Beaver Valley Power Station, Unit 2 (Reference 1). The SG tube inspections were performed during refueling outage 23 (2R23). When the voltage-based ARC and the F* methodology have been applied, Technical Specification (TS) Sections 5.6.6.2.2 and 5.6.6.2.4, respectively, require that a report be submitted within 90 days after the initial entry into hot shutdown (MODE 4) following completion of an inspection of the SGs performed in accordance with TS Section 5.5.5.
Responses to Request for Additional Information To complete its review of the inspection, the U.S. Nuclear Regulatory Commission (NRC) staff requests the following additional information:
RAI 1
During the baseline PlusPoint' inspections of the cold leg top of tubesheet region of tubes that were de-plugged during refueling outage (RFO) 23, two circumferential Outer Diameter Stress Corrosion Cracking (ODSCC) indications were detected in two tubes in SG B at the cold leg expansion transition region. These two tubes were originally plugged in 2008 and 2014. Following expansion of the cold leg tubesheet inspection scope, include all previously plugged tubes and about 2100 other tubes in SG A and SG B, one additional tube in SG A was found with circumferential ODSCC at the cold leg expansion transition. This tube had been previously plugged but was returned to service during refueling outage 22. Historical data review of the RFO 22 PlusPointTM data for this tube showed a small precursor signal. The staff recognizes identification of precursor signals is significantly improved knowing the location of an indication detected during a subsequent outage.
- a. Given the knowledge gained during the RFO 23 inspections, would an indication similar to the RFO 22 precursor signal be detected during RFO 23, or are the precursor signal characteristics only able to be detected in hindsight?
Response
The subject tube, R27C39 in SG A, was found to have a small circumferential ODSCC indication reported at the cold leg top of tubesheet expansion transition during the RFO 23 inspection (spring 2023). This tube was originally plugged during the RFO 13 outage in 2008 due to a circumferential ODSCC indication at the hot leg tubesheet and was returned to service in the RFO 22 (Fall 2021) outage as part of the tube recovery program that replaced tube plugs with sleeves. Baseline eddy current inspections were performed on the de-plugged tube with qualified probes at all locations that could be susceptible to degradation to demonstrate acceptability of the tube to be returned to service. The baseline inspection with a PlusPointTM probe reported no degradation at the cold leg expansion transition or anywhere else in the tube and the tube was placed into service. Following discovery of the circumferential indication at the cold leg expansion transition in this tube during the current RFO 23
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 3 of 8 Revision 0 inspection, a historical data review was performed for the RFO 22 inspection data. The review showed a small precursor signal that was outside the flaw phase plane window. The signal (phase angle) changed during the Cycle 23 operation which led to the indication being reported in RFO 23. Industry experience has shown that retrospective review of historical data of known indications significantly increases the identification of small signals that could be described as precursor signals that are at or below the threshold of detection.
With discovery of this new degradation mechanism, measures were completed to improve the reporting of circumferential indications at the cold leg expansion transition region over that from prior inspections, these being analyst awareness, supplemental training, and additional analyst data reviews.
Analysts performing data analysis of de-plugged tubes received supplemental training specific to the signals found at the cold leg expansion transition, including the RFO 22 precursor signal. The training raised awareness for the potential of finding small indications in de-plugged tubes and to recognize the signal formations of actual signals found. The primary and secondary data analysis of all de-plugged tubes was performed by a team of Resolution Data Analysts rather than the Primary and Secondary Production Analysts. Resolution Data Analysts typically have more experience and knowledge in dispositioning complex signals. A separate team of Resolution Analysts acted as the resolution team.
The Independent Qualified Data Analyst (IQDA) and the contracted Energy Harbor Level III Data Analyst reviewed the raw eddy current data of all de-plugged tubes. It is believed that these measures provided improved detection capabilities and if an indication similar to the RFO 22 precursor signal were present, it would have been detected during RFO 23.
- b. Discuss future cold leg tubesheet inspection plans for previously plugged tubes that have been returned to service.
Response
All tubes that have been previously plugged and returned to service in future and past outages are planned to be inspected at the cold leg tubesheet expansion transition region with the PlusPointTM in each SG every outage.
RAI 2
The Unit 2 F* (F Star) Report for RFO 23 discusses an evaluation of the stresses in the cold leg top of tubesheet region in a plugged tube and surrounding tubes that are assumed to be locked in the tube support plate (TSP) prior to the tube being plugged. This evaluation concluded that the lower temperature of the plugged tube could cause up to two times the axial tensile stress in the plugged tube, relative to the active tubes.
- a. Please provide additional details concerning the evaluation of stress at the cold leg top of tubesheet in the plugged tubes.
Response
An evaluation was performed that evaluated the stress condition at the top of the cold leg tubesheet joint in a plugged tube. In this evaluation, it was assumed that a plugged tube and surrounding tubes were locked at the first TSP prior to the tube being plugged. The Beaver Valley Unit 2 SGs contain
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 4 of 8 Revision 0 carbon steel drilled hole TSPs which are susceptible to corrosion from the secondary side conditions within a TSP to tube intersection. The corrosion of the carbon steel support plate causes corrosion product build-up within the crevice between the tube and the support plate. The corrosion products can cause the tube to become locked or even dented from the external forces acting on the tube circumference. Beaver Valley Unit 2 corrosion induced denting has occurred at TSP elevations on both the hot leg cold tube sections, and therefore, locked tube conditions are also likely in inservice and plugged tubes.
For inservice tubes that are not locked, the tube will expand axially without restriction due to thermal expansion and pressure growth. The axial stress induced at the tubesheet joint is caused by end cap loading and hoop stress caused by internal pressure. The elevation of the supports relative to the tube is determined by the thermal expansion of tube, stayrods, wrapper, and lower shell, each having material of different thermal expansion coefficients than the tubes. The RFO 23 affected tubes were originally plugged after operating 13 fuel cycles or more and therefore, it is likely that the tubes became locked prior to the tubes being plugged. For this stress evaluation, it was assumed that the plugged tube and surrounding inservice tubes became locked during operation (i.e., hot conditions) before the affected tube was plugged. Tube plugging is performed at cold conditions when the plant has been shut down. When the plant is restarted and returns to normal operating conditions, the inservice tubes, as well as the stayrods, wrapper and lower shell will return the TSPs (and locked inservice tubes) to the same elevation and stress conditions as in previous cycles.
Since the plugged tube has no primary side coolant, and no internal pressure, the tube will not expand due to internal pressure and will not have end cap loading. The inservice tubes have internal primary coolant inside and will have a temperature between the primary coolant and secondary coolant temperature, while the plugged tube will be at the lower secondary coolant temperature. The plugged tube will have less thermal expansion displacement than the inservice tubes. Therefore, the plugged tube will undergo axial tension during operation when returning the tube to the elevation before it was locked, thereby inducing a tensile stress at the tubesheet joint expansion transition.
The tubes in this stress evaluation are modeled as thin-walled cylinders with the standard axial and hoop stress equations:
=
2
=
Where P is the tube pressure, r is the tube radius, and t is the tube thickness.
Hookes Law defines the relationship between stress and strain (i.e., displacement) in the equation form of:
=
=
Where is displacement, F is force, A is the area that the force is acting upon, L is the material length, and E is the material modulus of elasticity.
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 5 of 8 Revision 0 The thermal expansion or contraction displacement is modeled using the standard equation:
= ()
Where is thermal displacement, is the coefficient of thermal expansion, L is the tube length from the top of the tubesheet to the TSP that is locked, and T is temperature.
The above equations are used to determine the stresses in locked inservice tubes and locked plugged tubes. It was conservatively assumed that the tubes were locked at the first TSP above the flow distribution baffle (FDB), as tube locking at the FDB is not likely due to the large gap between the tube and the support.
The stress evaluation results determined that the axial stress in a locked plugged tube is 2 times higher than in an adjacent locked tube and the hoop stress is compressive since there is no internal pressure and only external pressure in a plugged tube. The additional axial stress increases the susceptibility of circumferential ODSCC at the tubesheet expansion transition in a locked tube.
- b. Are there any additional tube degradation concerns (e.g., fatigue) if the tubes are assumed to be locked into the tube supports?
Response
When an inservice tube becomes locked into a tube support, changes occur to the tubes operational and shutdown stress conditions, as well as changes to the tubes natural frequency and stiffness. A change in the tube operational and shutdown stress condition produces an axial alternating stress that may result in fatigue related circumferential cracking. A change in the tubes natural frequency and stiffness can impact the fluid-elastic tube vibration response due to the hydrodynamic excitation from the secondary fluid on the outside of the tube. These effects from a locked tube condition are evaluated below.
Locking of SG tubes within a carbon steel drilled hole TSP may occur during elevated temperature conditions during normal operation. The thermal expansion coefficient of the Alloy 600 tube is higher than the carbon steel TSP support structures (i.e., stay rod, wrapper and vessel shell). Additionally, an inservice tube operates with internal pressure and at a higher temperature than the TSP support structures with no pressure effects. Consequently, a locked tube can have additional axial stress at shutdown due to the differential expansion between the tube and stayrod and the axial expansion of the tube due to internal pressure. This causes a fluctuating stress between shutdown and operating conditions. The alternating stress between these conditions is well below the ASME Code fatigue limit stress at 1,000,000 cycles for the tube material. Given that the SG is designed for 200 cycles between shutdown and operating temperatures, the added fatigue due to the tube being locked is negligible and has no effect on fatigue-related circumferential cracking. A similar fatigue evaluation showed similar results for the locked tube condition in a plugged tube. The results also concluded that there is a negligible fatigue impact due to a locked tube condition in a plugged tube.
Non-uniform anti-vibration bar (AVB) insertion depths or improperly supported U-bend tubes have caused high cycle fatigue failure in lower row U-bend tubes, as described in NRC Bulletin 88-02 (Reference 2) based on the North Anna Unit 1 tube rupture event. U-bend fatigue analyses were
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 6 of 8 Revision 0 performed for Beaver Valley Unit 2 in accordance with Reference 2 at the time of publication (Reference 3). These analyses had addressed and included the locked tube condition at the top TSP in the fatigue evaluations.
A tube locked within a TSP can alter the tubes natural frequency and stiffness which may impact the fluid-elastic tube vibration response due to hydrodynamic excitation by the secondary fluid on the outside of the tubes. This has the potential to impact the bending stress due to flow-induced vibration but will not impact the axial membrane stress because the axial stresses due to vibration are negligible for either the locked or unlocked tube conditions. When it is assumed that a tube becomes locked in the support system due to corrosion product build-up between the tubes and supports, the frequency and stiffness of the tube increases and the damping decreases. The overall impact of these factors results in a net decrease in the tube stability ratio and an increase in bending stress, however, this stress is noted in one study can lead to a fatigue usage about 30 times lower than the limit under conservative assumptions of tube clamping. A review of detailed analyses for Westinghouse model feedring SGs that considered postulated clamping (i.e., locking) conditions with reduced damping resulted in a maximum tube bending stress of 2.0 ksi. Further, in another Westinghouse study which addressed the effect of an extended power uprate on SG tube wear, it was noted that the thermal-hydraulic effects that result in tube vibration and wear are small in the tube U-bend region and bounds the straight-leg region.
Therefore, a locked tube condition within tube supports presents no concern for flow induced vibration and wear mechanisms. The tube stress levels are still maintained below the endurance limits and fatigue remains within acceptable levels.
A tube may degrade after it has been removed from service via tube plugging, either by continued degradation of the mechanism that caused the tube to be plugged or by initiation of other mechanisms after the tube had been plugged. A plugged tube may experience secondary fluid flow effects such as flow induced vibration, and therefore, may be susceptible to wear or fretting type degradation from support structures or from foreign material to the same degree as for an inservice tube. It is likely that a locked tube condition may have existed in the Beaver Valley Unit 2 SGs for many years, given the experience of similarly designed SG models. Through routine inspections, existing wear mechanisms have also been monitored for many years. The initiation rate and growth rate of wear mechanisms in inservice tubes have remained low with few tubes requiring tube plugging, even with the expectation that tube may be locked within the TSPs. This experience for inservice tubes can be applied to plugged tubes since a plugged tube is subject to the same secondary side hydrodynamic conditions as in-service tubes.
When a tube is plugged, primary coolant fluid does not enter the tube resulting in a tube temperature that is much lower than in an inservice tube. The temperature of the plugged tube is essentially at the temperature of the secondary coolant fluid. Typically, at these temperatures, stress corrosion cracking is not of concern (i.e., less than 590 degrees F), except when additional stress conditions are present.
As shown in the RAI #1b response and discussed within this response above, a tube locked at the first TSP above the FDB may experience additional stresses at the top of the tubesheet. The three cold leg top of tubesheet circumferential indications reported in RFO 23, were very small upon initial detection despite having been plugged in 2008 or 2013 suggests that the growth rate in plugged tubes are small, likely due to the lower temperature of a plugged tube. The indication in tube SG-A R27C39 exhibited
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 7 of 8 Revision 0 little to no change to the signal voltage amplitude or length compared to its precursor from historical data review in the prior outage, which also suggests a possible effect from the lower temperature.
Beaver Valley Unit 2 is licensed to repair tubes using Alloy 800 mechanical sleeves (Reference 4).
WCAP-15919-NP (Reference 5) provides the technical justification and qualification of the sleeve and repair process. The locked tube condition was considered in the qualification and analyses of the sleeve.
The sleeve-to-tube joint deflection capability is sufficient for thermal expansion effects with non-severed or severed parent tube conditions even if the tube is locked with tube supports. Reference 5 concludes that there is no degradation of leak limiting or structural load capabilities for the worst-case thermal expansion cycles with locked or non-locked tubes.
With consideration from the above discussions, and with primary-to-secondary leakage monitoring in compliance with the plant technical specifications, it can be concluded that there are no additional degradation concerns with operation with either active or plugged tubes locked into the tube supports.
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
Westinghouse Non-Proprietary Class 3 DMW-NRCD-RF-LR-000003-NP Page 8 of 8 Revision 0
References:
- 1. Energy Harbor Nuclear Corporation Letter L-23-168, Beaver Valley Power Station, Unit No. 2, Docket No. 50-412, License No.NPF-73, Steam Generator Reports, August 7, 2023.
- 2. NRC Bulletin 88-02, Rapidly Propagating Fatigue Cracks in Steam Generator Tubes, February 5, 1988.
- 3. Westinghouse WCAP-12142, Revision 0, Beaver Valley Unit 2 Evaluation for Tube Vibration Induced Fatigue, January 1989.
- 4. NRC Safety Evaluation Report, Beaver Valley Power Station, Unit No. 2 - Issuance of Amendment 201 RE: Revision of Technical Specifications Related to Steam Generator Tube and Repair Methods (EPID L-2020-LLA-0140), June 2021. (ADAMS Ascension Number ML21153A176)
- 5. Westinghouse WCAP-15919-NP, Revision 2, Steam Generator Tube Repair for Westinghouse Designed Plants with 7/8 Inch Inconel 600 Tubes Using Leak Limiting Alloy 800 Sleeves, January 2006.
- This record was final approved on 11/10/2023 08:51:11. (This statement was added by the PRIME system upon its validation)
DMW-NRCD-RF-LR-000003-NP Revision 0 Non-Proprietary Class 3
- This page was added to the quality record by the PRIME system upon its validation and shall not be considered in the page numbering of this document.**
Approval Information Author Approval Smith Jay R Nov-08-2023 17:05:33 Verifier Approval Suddaby David A Nov-08-2023 20:46:51 Reviewer Approval Whiteman Gary Nov-09-2023 06:09:06 Reviewer Approval Rees John Nov-09-2023 11:31:34 Manager Approval Chappo jr. Robert Nov-10-2023 08:51:11 Files approved on Nov-10-2023