Residual Heat Removal (RHR) aka residual heat removal system (RHS) is a mode of Low Pressure Injection System which is meant to cool the reactor core after its shutdown. Since fuel will still be reactive to some degree it will continue to give off heat. The RHR prevents the water covering the pool from getting hot and boiling off.
Modes
Shutdown Cooling
The shutdown cooling (SDC) mode of RHR is meant for long term reactor vessel to RHRSW heat exchangers.
Low Pressure Cooling Injection
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| | Site | Start date | Title | Description |
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| ENS 56977 | Brunswick | 19 February 2024 04:25:00 | Automatic Start of Emergency Diesel Generator | The following information was provided by the licensee via phone and email:
At approximately 2325 EST on February 18, 2024, with Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 100 percent power, emergency diesel generator 2 automatically started due to the unexpected loss of AC power to emergency bus E2 during a planned transfer of E2 DC control power from normal to alternate for the 1B-1 battery. In addition, the unexpected loss of AC power to E2 resulted in Unit 1 primary containment isolation system (PCIS) partial Group 2 (i.e., drywell equipment and floor drain, residual heat removal (RHR), discharge to radioactive waste, and RHR process sample), Group 6 (i.e., containment atmosphere control/dilution, containment atmosphere monitoring, and post accident sampling systems), and partial Group 10 (i.e., air isolation to the drywell) isolations.
Emergency diesel generator 2 automatically started and re-energized the E2 bus as designed when the loss of E2 signal was received. The PCIS actuations were as expected for the outage plant line up on Unit 1 at the time. The cause of the loss of electrical power to emergency bus E2 is under investigation at this time.
This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of emergency diesel generator 2 and PCIS.
There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
This event will be entered into the plant's corrective action program. | | IR 05000333/2023004 | FitzPatrick | 7 February 2024 | Integrated Inspection Report 05000333/2023004 and Independent Spent Fuel Storage Installation Inspection Report 07200012/2023001 | | | IR 05000220/2023004 | Nine Mile Point | 1 February 2024 | Integrated Inspection Report 05000220/2023004 and 05000410/2023004 | | | IR 05000317/2023004 | Calvert Cliffs | 1 February 2024 | Integrated Inspection Report 05000317/2023004 and 05000318/2023004 | | | IR 05000321/2023004 | Hatch | 31 January 2024 | Integrated Inspection Report 05000321/2023004 and 05000366/2023004 | | | NUREG-2194 Volume 2, Rev. 1, Standard Technical Specifications, Westinghouse Advanced Passive 1000 (AP1000) Plants, Volume 2: Bases | | 31 January 2024 | NUREG-2194, Vol. 2, Rev. 1, Standard Technical Specifications, Westinghouse Advanced Passive 1000 (AP1000) Plants, Volume 2: Bases | | | IR 05000395/2023004 | Summer | 31 January 2024 | Integrated Inspection Report 05000395/2023004 | | | ENS 56936 | Peach Bottom | 29 January 2024 17:02:00 | Automatic Reactor Scram | The following information was provided by the licensee via email:
At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram caused by a main turbine trip. Investigation is still ongoing.
The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance:
All control rods were fully inserted. The licensee indicated that the turbine trip may have been caused by a power load imbalance, however the cause of the incident is under investigation. The scram was not complex.
Decay heat is currently being removed thru bypass valves dumping to the main condenser. Initially unit 2 lost the use of the bypass valves due to lack of condenser vacuum. Unit 2 used the high pressure coolant injection (HPCI) system in the condenser storage tank (CST) to CST mode to remove decay heat. Residual heat removal was used to keep the torus cool. Condenser vacuum was regained and unit 2 is back to removing decay heat with the turbine bypass valves.
There was no impact to unit 3.
The licensee confirmed there was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
- * *UPDATE ON 01/29/24 AT 1935 EST FROM PAUL BOKUS TO NATALIE STARFISH* * *
The following information was provided by the licensee via email:
Licensee adds 8-hour non-emergency 10 CFR 50.72(b)(3)(iv)(A) specified system actuation report to original 4-hour non-emergency 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation report.
At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram by a main turbine trip. All control rods inserted. Reactor core isolation cooling system (RCIC) was manually initiated for level control. HPCI was manually initiated for pressure control. Primary containment isolation system (PCIS) Group II and III isolations occurred (specified system actuation). Investigation is ongoing.
The NRC Resident Inspector has been notified. | | IR 05000397/2023004 | Columbia | 29 January 2024 | Integrated Inspection and Independent Spent Fuel Storage Installation Report 05000397/2023004 and 07200035/2023001 | | | ML24026A011 | Ginna | 26 January 2024 | R. E. Ginna Nuclear Power Plant, Relief Request Associated with Inservice Testing of B Auxiliary Feedwater Pump | |
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