RS-22-033, Response to Request for Additional Information Regarding Relief Request I4R-13 Relief from Code Examinations for 1B33-F060A and 1B33-F060B Repairs

Response to Request for Additional Information Regarding Relief Request I4R-13 Relief from Code Examinations for 1B33-F060A and 1B33-F060B Repairs

ML22062B658
Person / Time
Site:	LaSalle
Issue date:	03/03/2022
From:	Lueshen K Constellation Energy Generation
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-22-033
Download: ML22062B658 (7)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-22-033 10 CFR 50.55a March 3, 2022 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit 1 Renewed Facility Operating License No. NPF-11 NRC Docket No. 50-373

Subject:

Response to Request for Additional Information regarding Relief Request I4R-13 Relief from Code Examinations for 1B33-F060A and 1B33-F060B Repairs

References:

1. Letter from K. Lueshen (Constellation Energy Generation, LLC) to Nuclear Regulatory Commission (NRC), Relief Request I4R-13 Relief from Code Examinations for 1B33-F060A and 1B33-F060B Repairs, " dated March 2, 2022.

2. Letter from B. Vaidya (NRC) to J. Taken (Constellation Energy Generation, LLC), "LASALLE UNITS 1 AND 2 -REQUEST FOR ADDITIONAL INFORMATION (RAI) RE: Relief Request I4R-13 Relief from Code Examinations for 1B33-F060A and 1B33-F060B Repairs (EPID-L 2022-LLR-0028)," dated March 3, 2022.

In Reference 1, Constellation Energy Generation, LLC (CEG), in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(2), Constellation Energy Generation, LLC (CEG) requested NRC approval of a relief request associated with the fourth Inservice Inspection (ISI) interval for LaSalle County Station (LSCS), Unit 1.

In Reference 2, the NRC requested additional information. A clarification call between the NRC and CEG was held on March 3, 2022 to ensure a common understanding of the information requested in Reference 2.

The attachment to this letter contains the CEG response to the requested information.

CEG requests authorization of the proposed alternative and relief from radiography requirements by March 7, 2022.

March 3, 2022 U.S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Jason Taken at (630) 657-3660.

Respectfully, Digitally signed by Lueshen, Kevin Lueshen, Kevin Date: 2022.03.03 13:11:28 -06'00' Kevin Lueshen Sr. Manager - Licensing Constellation Energy Generation, LLC

Attachment:

Response to Request for Additional Information regarding Relief Request I4R-13 Associated with Alternative Examination Requirements for Repairs of Reactor Recirculation Flow Control Valves cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - LaSalle County Station NRC Project Manager, NRR - LaSalle County Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT Response to Request for Additional Information regarding Relief Request I4R-13 Associated with Alternative Examination Requirements for Repair of Reactor Recirculation Flow Control Valves 1B33-F060A and 1B33-F060B

Response to Request for Additional Information regarding Relief Request I4R-13 Associated with Alternative Examination Requirements for Repairs of Reactor Recirculation Flow Control Valves ATTACHMENT By letter to the U.S. Nuclear Regulatory Commission (NRC) dated March 2, 2022, Constellation Energy Generation, LLC (the licensee) proposed an alternative to certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),

Section III and Section XI, related to the repair of degraded Reactor Recirculation Flow Control Valves at LaSalle County Station, Unit 1 (LaSalle Unit 1). The licensee submitted Relief Request No. I4R-13, for the proposed alternative Enhanced Visual testing (EVT) examination requirements in lieu of the required Penetrant Testing (PT) for the repairs.

The NRC has determined that the following additional information is necessary to complete its review and make a regulatory decision.

RAI #1 The 2021 Relief Request I4R-12 for LaSalle Unit 2 (ADAMS Accession No. ML21085A874) discussed the unusual wear of similar valve bodies in Unit 2 and stated that it was attributed to an atypical low-power operating condition that occurred in 2015. Relief Request 14R-13, for Lasalle Unit 1, described wear in the valve bodies but does not give any specific reason for the wear. Discuss whether the root cause of the wear for LaSalle Unit 1 is understood to be the same as for Unit 2, or whether a different cause of the wear is present for Unit 1.

CEG RESPONSE In 2021, CEG performed an investigation of the root cause for the damage that occurred on Unit 2 that resulted in repairs of the 2B33-F060A and 2B33-F060B and ultimately the 2021 relief request. The investigation concluded that extended operation with a known failure mechanism of high flow vibration without a mitigating strategy was the cause.

Further, a contributing cause for the as-found damage of the 2B33-F060A/B was extended low flow operations without adequate assessment of the impact on flow control valve health.

As part of that root cause investigation for Unit 2, CEG reviewed plant operations under low flow conditions for the past three cycles on Unit 1. The low flow conditions reviewed were conditions in which the Reactor Recirculation (RR) pumps were operating at high speed with the flow control valve 70% or less open. The investigation found that although Unit 2 operated at this condition for a significantly longer time than Unit 1, Unit 1 still operated in this condition for over 600 hours0.00694 days <br />0.167 hours <br />9.920635e-4 weeks <br />2.283e-4 months <br /> during the past three cycles.

Given the similarities in operating history to the Unit 2 flow control valves, CEG expected that some wear could also exist on the Unit 1 flow control valves. Therefore, CEG performed inspections of the Unit 1 flow control valves during the current outage (L1R19). The as-found condition of the 1B33-F060A and 1B33-F060B valves reasonably supports the conclusion that the periods of operation at high RR pump speed conditions with the flow control valves at less than 70% open contributed to the degradation discovered during the current Unit 1 refuel outage. This condition is 1 of 4

Response to Request for Additional Information regarding Relief Request I4R-13 Associated with Alternative Examination Requirements for Repairs of Reactor Recirculation Flow Control Valves ATTACHMENT consistent with the findings of the root cause investigation conducted for Unit 2 and the findings from the last Unit 2 outage. Corrective actions from the Unit 2 root cause investigation have been implemented on both Unit 1 and Unit 2 to limit plant operation in this low flow operation condition described previously. This ensures plant operation in this low flow region going forward on both units will be significantly less than what it has been in the past several years.

RAI #2 In Relief Request 14R-13, the wear rate in the valves after the repairs is expected to be 0.00015 inches per operating cycle. Please provide details on this wear rate calculation and the uncertainties in the calculation.

CEG RESPONSE Kalsi Engineering Inc. (KEI) prepared calculations for wear rates of the RR flow control valve components. These valves require frequent flow modulation causing the ball to rotate repeatedly within its bearings. The relative motion between the valve ball shaft and the bearings would cause wear of the mating surfaces. Wear calculations were performed for lower plug-to-body contact (self-mated 316 stainless steel).

The fluid media for the subject valves is hot water that is considered to be a poor lubricant. KEI references provide a wear coefficient for an identical metal on metal case for poor lubricant. For the lower plug-to-body contact the wear coefficient for an identical metal on metal case for poor lubricant case was used. This wear coefficient was adjusted to a higher value (conservative) based on the mass percentage of the major elements in 316 stainless steel.

The average radial force on the lower shaft bearing was calculated using the differential pressure (DP) values and the resultant force coefficients (CR) obtained from KEI proprietary analyses. Assuming that the total bearing load is equally shared by the upper and lower bearings, half of the average load was used for wear calculation for the valve operation between 20° and 85° ball position. The weight of components like the lower plug guide was conservatively assumed in excess of actual weight by about 20%.

For the lower plug-to-body contact in the radial direction, half of the maximum bearing radial force was conservatively used for the wear calculation. The calculation uses a bearing coefficient of friction (COF) higher than used in the design report, therefore the maximum bearing load calculation is conservative.

Other analyses include review of the sliding distances traveled by the rotating ball shaft of the Unit 2 valves for the past two consecutive operational cycles, including transient conditions such as startup and shutdown conditions, as well as the period during the 2 of 4

Response to Request for Additional Information regarding Relief Request I4R-13 Associated with Alternative Examination Requirements for Repairs of Reactor Recirculation Flow Control Valves ATTACHMENT 2015 low power event. Weld stress analyses, fatigue life, and flow inducted vibration assessments were also performed.

The component geometries and material properties are known design inputs. The uncertainties of the calculation are documented as analysis assumptions with the various operational loadings. KEI utilized proprietary, empirical data to support the basis for the assumed loads selected in the wear rate calculation. Other uncertainties related to component weight, coefficient of friction, and contact height are addressed through conservative assumptions as discussed above.

RAI #3 In Relief Request 14R-13, the licensee stated that, notwithstanding the observed wear, Valve Bodies 1B33-F060A and 1B33-F060B remain above the minimum wall thickness. It is stated that at this time 1B33-F060B does not require welding, however, in the event that the repair plans require welding, welding would be conducted in accordance with the ASME Code. Valve Body 1B33-F060A requires repair welding and machining.

In the Relief Request I4R-12 for Unit 2, it was stated that surface conditioning of the valve surfacing prior to welding reduced the total thickness to below the minimum required wall thickness. Please specify if surface conditioning for Unit 1 will be performed prior to welding on these valves, and what plans are in place if the wall thickness is reduced to below the minimum wall thickness.

CEG RESPONSE CEG confirms that surface conditioning will be performed for Unit 1 prior to welding operations. Surface conditioning is required to smooth out horizontal surfaces that exceed desired flatness tolerances as well as to enlarge the radial bore of both valves.

The as-found bore geometry for both Unit 1 valves was found to be oval shaped due operational wear. The oval shaped bores were machined, only the minimal extent needed to become circular again. The internal geometry features were able to be measured and there was no gross damage identified in areas outside of where the internal components normal interface. This is contrary to the as-found conditions of the Unit 2 valves in February of 2021. Three-dimensional (3D) laser scanning technology allowed LSCS to directly measure the thicknesses of the Unit 1 valve bodies and understand the cross-sectional wear patterns.

CEG developed the respective repair plans for the Unit 1 valves based on the 3D scans described in Reference 1. Based on the 3D scan results, the as-found conditions of both 1B33-F060A and 1B33-F060B valves were both above minimum wall thickness. If during machining operations the wall thickness is reduced below the minimum wall 3 of 4

Response to Request for Additional Information regarding Relief Request I4R-13 Associated with Alternative Examination Requirements for Repairs of Reactor Recirculation Flow Control Valves ATTACHMENT thickness, then weld repairs will be conducted consistent with ASME code and I4R-13 requirements to restore minimum wall thickness to the planned thicknesses identified in Reference 1.

REFERENCES

1. Letter from K. Lueshen (Constellation Energy Generation, LLC) to Nuclear Regulatory Commission (NRC), Relief Request I4R-13 Relief from Code Examinations for 1B33-F060A and 1B33-F060B Repairs, " dated March 2, 2022.

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