L-25-219, Report of Facility Changes, Tests, and Experiments
| ML25349B332 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 12/15/2025 |
| From: | Hicks J Vistra Operations Company |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-25-219 | |
| Download: ML25349B332 (0) | |
Text
Jack Hicks Senior Manager, Fleet Licensing Vistra Operations Company LLC P.O. Box 1002 6322 FM 56 Glen Rose, TX 76043 10 CFR 50.59(d)(2)
L-25-219 December 15, 2025 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
SUBJECT:
Perry Nuclear Power Plant, Unit No. 1 Docket No. 50-440, License No. NPF-58 Report of Facility Changes, Tests, and Experiments Pursuant to 10 CFR 50.59(d)(2), Vistra Operations Company LLC hereby submits the Perry Nuclear Power Plant, Unit No. 1 Report of Facility Changes, Tests, and Experiments. The attached report covers the period from October 9, 2023 to October 9, 2025.
There are no regulatory commitments contained in this submittal. If there are any questions or if additional information is required, please contact, please contact Mr. Jack Hicks, Senior Manager, Fleet Licensing, at (254) 897-6725 or Jack.Hicks@vistracorp.com.
Sincerely, Jack Hicks
Attachment:
Perry Nuclear Power Plant Report of Facility Changes, Tests, and Experiments for the Period October 9, 2023 to October 9, 2025 cc:
NRC Region III Administrator NRC Resident Inspector NRR Project Manager
Attachment L-25-219 Perry Nuclear Power Plant, Unit No. 1 Report of Facility Changes, Tests, and Experiments for the Period October 9, 2023 to October 9, 2025 Page 1 of 2
Title:
Install Circuit to Facilitate Controllability of flow Control Valve A due to Degraded RVDT Activity
Description:
The Perry Nuclear Power Plant has two reactor recirculation loops, each with its own flow control valve (FCV).
The FCVs control core flow through the reactor and thus can be used to control reactor power. During plant operation, Reactor Recirculation Flow Control Valve A, 1 B33F0060A, position indication showed the valve moving without operator input, first opening and then returning to its original position. Following that event, core plate differential pressure and reactor thermal power were lower than values prior to the start of the event.
During problem solving activities, it was determined that the rotary variable differential transformer (RVDT), 1 B33N0202A, which provides a signal corresponding to the position of FCV A was failing and giving invalid position indication. The FCV and RVDT are located in the drywell. Plant computer data shows the indicated position, and therefore the position signal supplied to the Flow Control Valve control system, was no longer accurate, causing this event.
The FCV A was hydraulically locked in accordance with Technical Specification 3.4.2 pending a resolution of the issue. Without being able to move the FCV A, it limits the plant's ability to maintain full power and to execute pre-scheduled control rod adjustments. Full repair to the RVDT cannot be completed until entry can be made into the drywell, so the basis of this temporary modification is to provide a method for controlling the valve until those repairs are made.
Temporary modification 25-1003 was designed to mitigate the effects of the failed valve position feedback into the control circuit. The RVDT A will be temporarily disconnected from the FCV A control circuity until it can be replaced. Due to the RVDT being failed, the computer point, B33EA062, is no longer accurate. A byproduct of disconnecting the RVDT at the valve position control card is that the computer point will also be disconnected.
The rate of change lockup that is driven by the erroneous valve position feedback is also being disabled. In the interim, a resistor-capacitor (RC) network will be installed between the operator's desired position input and the feedback to the position controller. The RC network will allow the valve to move when given an open or closed command by the operator and then null the demand signal after a time constant, determined by the capacitance and resistance values of the RC network. The reactor recirculation system is mechanically safety-related since it is part of the primary coolant pressure boundary. However, the power and electronics are non-safety related since it is not required or credited under accident conditions or to protect the health and safety of the public. These temporary changes will allow controlled valve movement that will be similar to, but slower than, the typical valve control from the operator's perspective.
Summary of Evaluation:
The temporary modification was determined to degrade the automatic recirculation runback feature, specifically to the FCV A, which acts to reduce the power to within the capacity of the remaining feedwater pumps when one of the reactor feed pump turbines (RFPTs) trip. With the temporary modification installed the runback signal will continue to occur for the B reactor recirculation loop FCV. It will not sufficiently run back the A reactor recirculation loop FCV due to actual valve position being removed from the feedback loop. The 10 CFR 50.59 Screen determined the temporary modification is adverse with respect to the runback feature described in Updated Safety Analysis Report (USAR) since it will not operate as designed and if the action is not successfully accomplished, would initiate a transient that the plant is required to withstand, which in this case is a SCRAM on low reactor water level.
Attachment L-25-219 Page 2 of 2 The as-modified reactor recirculation flow control valve A control system will continue to perform the same essential function, to allow reactor power manipulations by varying core flow, as the existing control system but with the credited runback feature degraded. The reactor recirculation system as a whole and the flow control valves specifically do not directly control any system, structure, or component (SSC) that mitigates the consequences of an accident. In the event the runback feature is not successful or if the temporary modification fails in a manner that drives the valve open, the SSCs that are credited with mitigating those conditions are unaffected by this change. The failure modes of the temporary modification have been analyzed in the Failure Modes and Effects Analysis included in this 10 CFR 50.59 Evaluation and there are no new failure effects that would initiate an accident or create an accident of a different type. Any failures are bounded by the RECIRCULATION FLOW CONTROL FAILURE WITH INCREASING FLOW transient described in USAR section 15.4.5 and 15A.6.3.3 - Event 11 which, at worst, results in a reactor SCRAM.
The reactor recirculation system and the associated A flow control valve control system are considered important to safety, as varying core flow changes reactor power; however, any failure of the system is already bounded by the transients analyzed in the USAR for the FCV failing open and failing closed. Therefore, the temporary modification to the FCV A control system does not (1) increase the likelihood of occurrence of previously evaluated malfunctions of an SSC important to safety, (2) increase the consequences of a malfunction of an SSC important to safety, or (3) introduce a malfunction of an SSC important to safety with a different result. The temporary modification to the FCV A logic does not require a departure from a method of evaluation described in the USAR and the FCV control system is not credited with directly or indirectly protecting a fission product barrier.
In conclusion, the proposed activity does not meet any of the criteria in paragraph (c)(2) of 10 CFR 50.59 and therefore the evaluation of the proposed activity has determined that a license amendment is not required.