05000387/LER-2022-002, Automatic Reactor Scram Due to High Reactor Vessel Pressure Following Inadvertent Closure of Inboard Main Steam Isolation Valve
| ML22202A279 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 07/21/2022 |
| From: | Cimorelli K Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| PLA-8018 LER 2022-002-00 | |
| Download: ML22202A279 (4) | |
| Event date: | |
|---|---|
| Report date: | |
| Reporting criterion: | 10 CFR 50.73(a)(2)(iv)(A), System Actuation 10 CFR 50.73(a)(2)(iv)(B), System Actuation |
| 3872022002R00 - NRC Website | |
text
Kevin Cimorelli Site Vice President Susquehanna Nuclear, LLC 769 Salem Boulevard Berwick, PA 18603 Tel. 570.542.3795 Fax 570.542.1504 Kevin. Cimorell i@TalenEnergy.com July 21, 2022 Attn: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 SUSQUEHANNA STEAM ELECTRIC STATION LICENSEE EVENT REPORT 50-387 /2022-002-00 UNIT 1 LICENSE NO. NPF-14 PLA-8018 TALEN.
ENERGY 10 CFR 50.73 Docket No. 50-387 Attached is Licensee Event Report (LER) 50-387/2022-002-00. The LER reports an event involving an automatic scram due to a Reactor Protection System actuation on high reactor pressure as a result of a Main Steam Isolation Valve closure. The condition is being rep01ied in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of a system listed in 10 CFR 50.73(a)(2)(iv)(B).
There were no actual consequences to the health and safety of the public as a result of this event.
- y commitments.
K. Cimorelli Attachment: LER 5 0-3 8 7 /2022-002-00 Copy: NRC Region I Mr. C. Highley, NRC Senior Resident Inspector Ms. A. Klett, NRC Project Manager Mr. M. Shields, PA DEP/BRP
NRG FORM366 (08-2020)
U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY 0MB: NO. 3150-0104 EXPIRES: 08/31/2023 Susquehanna Steam Electric Station, Unit 1 05000387
- 13. Page 1 of 3
- 4. Title Automatic Reactor Scram due to High Reactor Vessel Pressure following Inadvertent Closure of Inboard Main Steam Isolation Valve
- 5. Event Date Month Day Year 05 23 2022
- 9. Operating Mode
- 6. LER Number Year Sequential Rev Number No.
2022 -
002 00 1
- 7. Report Date Month Day Year Facility Name 07 21 2022 Facility Name
- 10. Power Level
- 8. Other Facilities Involved 100 Docket Number 05000 Docket Number 05000
)
28 I
2022
Abstract
At approximately 17: 16 on May 23, 2022, Susquehanna Steam Electric Station, Unit 1, experienced an automatic reactor scram. Reactor Protection System (RPS) was actuated on a high reactor vessel pressure signal due to inadvertent closure of an inboard Main Steam Isolation Valve (MSIV) (HV141F022D). All control rods inserted, and operators placed the mode switch to shut down. All safety systems responded as designed during this event.
Event Notification 55909 reported this event in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). This event is also reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of a system listed in 10 CFR 50.73(a)(2)(iv)(B).
The cause of the scram was due to a loss of pneumatic pressure resulting in closure of an inboard MSIV (HV141 F022D) which resulted in a high reactor vessel pressure signal and valid RPS actuation. The loss of pneumatic pressure was the result of a Containment Instrument Gas tubing failure. Further failure analysis is in progress and will be provided in a supplement to this LER. Affected tubing was replaced and the weld on 3/8" tubing to HV141 F022D was repaired.
There were no actual consequences to the health and safety of the public as a result of this event.
I I
I I
I I
I I
I I
05000-387
CONDITIONS PRIOR TO EVENT
Unit 1 - Mode 1, approximately 100 percent Rated Thermal Power (RTP)
Unit 2 - Mode 1, approximately 100 percent RTP
EVENT DESCRIPTION
YEAR SEQUENTIAL REV NUMBER NO.
2022 002 00 At approximately 17: 16 on May 23, 2022, Susquehanna Steam Electric Station, Unit 1, experienced an automatic reactor scram. Reactor Protection System (RPS) [EIIS System Code: JC] was actuated on a high reactor vessel pressure signal due to inadvertent closure of an inboard Main Steam Isolation Valve (MSIV)
(HV141 F022D) [EIIS System/Component Codes: SB/ISV] following loss of pneumatic pressure. All control rods inserted, and operators placed the mode switch to shut down.
Containment isolations [JM] and both Reactor Recirculation Pump [AD/P] trips occurred as reactor water level dropped below the Anticipated Transient Without Scram - Reactor Pump Trip logic setpoint. High Pressure Coolant Injection [BJ] and Reactor Core Isolation Cooling [BN] initiated as designed when the reactor water level lowered. Subsequently, operators maintained reactor water level within normal band using Reactor Feedwater [SJ]. All other safety systems responded as designed.
Event Notification 55909 reported this event in accordance with 1 O CFR 50.72(b)(2)(iv)(A),
10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). This event is also reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of a system listed in 10 CFR 50. 73(a)(2)(iv)(B).
CAUSE OF EVENT
The cause of the scram was due to a loss of pneumatic pressure resulting in inadvertent closure of an inboard MSIV (HV141 F022D) which resulted in a high reactor vessel pressure signal and valid RPS actuation. The loss of pneumatic pressure was caused by a Containment Instrument Gas [LK] tubing failure. Further failure analysis is in progress and will be provided in a supplement to this LER.
ANALYSIS/SAFETY SIGNIFICANCE
The actual consequence of this event was an automatic reactor scram. All safety systems responded as designed during this event. Residual Heat Removal [BO] and Residual Heat Removal Service Water [Bl]
remained available to remove residual heat. No fuel or clad damage occurred during the scram, as evidenced by Main Steam Line and Off-Gas radiation levels decreasing post-scram. All applicable systems were available to control the release of any radioactive material. All safety systems were available to mitigate the consequences of an accident.
The condition described herein did not result in a safety system functional failure. Accordingly, this event will not be counted as a safety system functional failure in the Reactor Oversight Process Performance Indicator.
There were no actual consequences to the health and safety of the public as a result of this event.
CORRECTIVE ACTIONS
05000-387 YEAR SEQUENTIAL REV NUMBER NO.
2022 002 00 The sheared tubing was replaced, and pneumatic control restored to the inboard MSIV (HV141F022D).
Additionally, an extent of condition visual exam and liquid penetrant examination were completed on the other three (3) Unit 1 inboard MSIVs. The outboard MSIVs are a different configuration, as such this failure mode is not applicable. Additional actions will be taken to perform an extent of condition review on the Unit 2 applicable MSIVs as well.
Any additional corrective actions will be provided in a supplement to this LER.
COMPONENT FAILURE INFORMATION
Component Name-Containment Instrument Gas supply line from header JCD-115 to line HCC-137 Component Identification - JCD-118 Part Number - SS-600-6 Manufacturer - Swagelok
PREVIOUS OCCURRENCES
None. Page.l._ of.l._