05000440/LER-2024-003, Technical Specification Required Shutdown Due to Increase in RCS Unidentified Leakage
| ML24198A025 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 07/15/2024 |
| From: | Penfield R Vistra Operations Company |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-24-168 LER 2024-003-00 | |
| Download: ML24198A025 (1) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(i)(A), Completion of TS Shutdown |
| 4402024003R00 - NRC Website | |
text
L-24-168 July 15, 2024 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Perry Nuclear Power Plant, Unit 1 Docket Number 50-440, License Number NPF-58 Licensee Event Report 2024-003-00 Perry Nuclear Power Plant Rod Penfield Site Vice President 10 Center Road Perry, OH 44081 440-280-5382 724-462-0816 ( cell) 10 CFR 50.73(a)(2)(i)(A)
Enclosed is Licensee Event Report (LER) 2024-003-00, "Technical Specification Required Shutdown Due to Increase in RCS Unidentified Leakage." This event is being reported pursuant to 10 CFR
- 50. 73( a)(2) (i)(A ).
There are no regulatory commitments contained in this letter or its enclosure. If there are any questions or if additional information is required, please contact Mr. Robert W. Oesterle, Manager, Regulatory Compliance, at (419) 321-7462.
Sincerely, L
Rod L. Penfield Enclosure: LER 2024-003-00 cc:
NRC Region III Administrator NRC Resident Inspector NRR Project Manager
Abstract
While operating in MODE 1 at 100% power on May 23, 2024 at 0000, Perry Nuclear Power Plant (PNPP) operators entered Technical Specification (TS) 3.4.5 due to the unidentified reactor coolant system (RCS) leak rate exceeding 5 gpm. PNPP operators commenced a manual reactor shutdown in accordance with site procedures at 0053 on May 23, 2024. Unable to reduce the unidentified RCS leakage, the shutdown transitioned to a TS required shutdown at 0400 and reported as Event Number 57136. The required safety systems and shutdown equipment performed as expected, resulting in an uncomplicated shutdown. The source of the unidentified RCS leakage was the reactor recirculation loop 'A' suction valve packing leak-off line via a failed sight glass flange. The leak-off line was isolated and sight glass repaired.
The root causes were failing to maintain and operate the leak detection system leak-off lines as designed because of unrecognized potential consequences from a failure in inaccessible areas and not prioritizing permanent repairs for long standing equipment degradation during outage scoping.
This event is being reported pursuant to 10 CFR 50.73(a)(2)(i)(A) as completion of a plant shutdown required by the Technical Specifications.
APPROVED BY 0MB: NO. 3150--0104 EXPIRES: 04/30/2027
- 2. DOCKET NUMBER YEAR 440 2024
- 3. LER NUMBER SEQUENTIAL NUMBER 003 REV NO.
00 Energy Industry Identification System (EIIS) codes are identified in the text as [XX].
Background:
The Perry Nuclear Power Plant (PNPP) reactor coolant recirculation system [AD] is designed to provide a forced coolant flow through the core to remove heat from the fuel. The pressure containing components of the reactor coolant system (RCS) and the portions of connecting systems out to and including the isolation valves define the reactor coolant pressure boundary (RCPB). During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage through either normal operational wear or mechanical deterioration. Limits on RCS operational leakage are required to ensure appropriate action is taken before the integrity of the RCPB is impaired. The safety significance of leaks from the RCPB varies widely depending on the source, rate, and duration; therefore, detection of leakage in the drywell is necessary. Methods for quickly separating the identified LEAKAGE from the unidentified LEAKAGE are necessary to provide the operators quantitative information to permit them to take corrective action should a leak occur detrimental to the safety of the facility or the public. A limited amount of leakage inside the drywell is expected from auxiliary systems that cannot be made 100% leak tight. Leakage from these systems should be detected and isolated from the drywell atmosphere, if possible, so as not to mask RCS operational LEAKAGE detection. Twenty-one drywell valves, including the reactor coolant recirculation pump suction valves [AD-V], have packing leak-off lines routed to the drywell equipment drain sump [WK]. Leakage directed to the drywell equipment drain sump is considered identified leakage while leakage collected through the floor drains to the floor drain sump is unidentified. Both leakage rates display and alarm in the Control Room. Each valve packing leak-off line has a temperature transmitter [AD-TT] that provides an alarm in the Control Room on high temperature in the line. A manually operated solenoid valve [AD-TSV] can then be closed from the Control Room to isolate the line. Located in the drywell and inaccessible while at power, the leak-off lines have a manual isolation valve [AD-ISV] upstream and a sight glass [AD-FG] downstream of the solenoid valves.
Technical Specifications:
PNPP Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.5, "RCS Operational LEAKAGE" applies in MODES 1, 2, and 3 and requires: no pressure boundary LEAKAGE; less than or equal to 5 gpm unidentified LEAKAGE; less than or equal to 30 gpm total LEAKAGE over the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period; and less than or equal to 2 gpm increase in unidentified LEAKAGE within the previous 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period in MODE 1. When unidentified or total leakage is not within the limits, TS 3.4.5 Action B.1 requires the leakage to be within limits in 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and Action C.1 requires verification that the unidentified LEAKAGE increase is not service sensitive austenitic material within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. If the required actions and associated completion times are not met, TS 3.4.5 Action D.1 requires the plant to be placed in MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and Action D.2 requires the plant to be placed in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
DESCRIPTION OF EVENT
With PNPP operating in MODE 1 and 100 percent power, the drywell unidentified rate of change high alarm was received at 0000 on May 23, 2024. The drywell floor drain sump rate indicated 7.7 gpm. Condition B of TS 3.4.5 was entered to reduce the unidentified leakage to less than or equal to 5 gpm within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and lowering reactor power to achieve MODE 3 was started, in accordance with site procedures. Subsequently, the operations crew received the valve stem leak-off temperature alarm, indicating at least one of the 21 drywell valve leak-off line temperatures were greater than or equal to180 degrees F. Four lines were identified as potential sources.
Simple troubleshooting found none of the four manually operated solenoid valves functioned. At 0400 on May 23, 2024, the drywell floor drain sump rate of 5.6 gpm was still above the TS operational limit; therefore, TS 3.4.5 Condition D was entered with required actions to place the plant in MODE 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. On May 23, 2024, PNPP entered MODE 3 at 1338 hours0.0155 days <br />0.372 hours <br />0.00221 weeks <br />5.09109e-4 months <br /> and MODE 4 at 2134 hours0.0247 days <br />0.593 hours <br />0.00353 weeks <br />8.11987e-4 months <br />. During the initial drywell entry at 1124 hours0.013 days <br />0.312 hours <br />0.00186 weeks <br />4.27682e-4 months <br /> on May 23, 2024, no RCPB leakage was identified; however, the reactor recirculation loop 'A' suction valve packing leak-off line sight glass flange was found failed. The failed flange allowed the reactor recirculation loop 'A' suction valve packing leakage to enter the drywell floor drain sump instead of the drywell equipment drain sump and increase the unidentified leakage.
CAUSE OF EVENT
Reactor recirculation loop 'A' suction valve packing degradation was initially identified because of high stem leak-off line temperature during the 2013 refueling outage start-up. The direct cause of the unidentified leakage increase was failure to close the reactor recirculation loop 'A' suction valve packing leak-off line solenoid valve in 2013 through 2024 resulting in challenging downstream piping performance, causing sight glass flange leakage, and immediately exceeding unidentified RCS leakage TS limits.
The root causes of this event were:
- 1) Failing to maintain and operate the leak detection system leak-off lines as designed because of unrecognized potential consequences from a failure in inaccessible areas.
- 2) Inadequate risk recognition and consequence-bias during the Outage Scope Review Board and Engineering Review process led to not prioritizing permanent repairs for long standing equipment degradation.
ANALYSIS OF EVENT
The operating crew responded correctly to the event. The applicable abnormal response procedure was properly entered, and performance met expectations. There were no safety systems inoperable, and no safety system failures related to this event. Since no RCPB leakage was identified, there were no actual safety consequences from this event. The leakage from the reactor recirculation loop 'A' suction valve packing leak-off line sight glass flange was contained within the drywell, and the plant reached a safe shutdown condition. There were no system failures that prevented the safe shutdown of the plant. It is therefore concluded that even if a design basis accident had occurred concurrent with this event, all safety systems would have operated to safely mitigate the event. Based on the above considerations, the safety significance of this event is very low, and the event did not pose a threat to the health and safety of the public or plant personnel.
Reportability Discussion:
The initiation of any nuclear plant shutdown required by the plant's Technical Specifications is reportable within four hours of the event in accordance with 10 CFR 50.72(b)(2)(i). Since a manual shutdown transitioned to a TS 3.4.5 required shutdown on May 23, 2024, at 0400 hours0.00463 days <br />0.111 hours <br />6.613757e-4 weeks <br />1.522e-4 months <br />, the event was reported to the NRC Operations Center at 0425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br /> as Event Number 57136.
This event is being reported in accordance with 10 CFR 50.73(a)(2)(i)(A) as a plant shutdown required by the Technical Specifications.
00
CORRECTIVE ACTIONS
Completed Actions:
APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 04/30/2027
- 2. DOCKET NUMBER YEAR 440 2024
- 3. LER NUMBER SEQUENTIAL NUMBER 003 REV NO.
00 The increased unidentified RCS leakage was stopped by closing the recirculation pump 'A' suction valve packing leak-off line manual isolation valve on May 23, 2024 at approximately 1600 hours0.0185 days <br />0.444 hours <br />0.00265 weeks <br />6.088e-4 months <br />. Additionally, the manual solenoid valve was closed on May 24, 2024 at approximately 0200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />, the sight glass was repaired on May 24, 2024 at approximately 0500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, and the recirculation pump 'A' suction valve was back seated.
Scheduled Actions:
Future corrective actions include: developing and implementing a site valve packing preventive maintenance program for inaccessible locations while online, modifying the leak detection system design and updating the drywell valve packing designs, and revising outage scoping documents with specific guidance for leaks in inaccessible areas while online.
PREVIOUS SIMILAR EVENTS
There have been no PNPP Licensee Event Reports related to a plant shutdown required by Technical Specifications in the past three years. Page!of!