Loss of coolant accident LOCA

Safety. A large inventory of radioactive fission products is present in any reactor fuel where the reactor has been operated for times on the order of months. In steady state, radioactive decay heat amounts to about 5% of fission heat, and continues after a reactor is shut down. If cooling is not provided, decay heat can melt fuel rods, causing release of the contents. Protection against a loss-of-coolant accident (LOCA), eg, a primary coolant pipe break, is required. Power reactors have an emergency core cooling system (ECCS) that comes into play upon initiation of a LOCA. 

Reactor coolant pump (RCP) seal failures that lead to a loss of coolant accident (LOCA)... 

Loss, in lasers, 14 664-666 Loss factor, monitoring, 10 15 Loss-in-weight method, 26 248 Loss-in-weight systems, 26 249 Loss modulus, 20 346 21 722-723 Loss-of-coolant accident (LOCA), 17 577, 582, 595, 596... 

AP600 Passive Safety System Details, These features reduce operator responsibilities and add an extra margin of safety over contemporary PWR designs. See Fig. 37 on p. 1121. Large volumes of water stored in the containment eliminate the need for operator action to assure make-up water, either for small leaks that may occur during normal operation or for a major loss of coolant accident (LOCA). A passive plant is a system llial assures public safety even if the operators fail to act... 

LWR tests-to-failure had been performed to evaluate accident scenarios involving loss of coolant accident (LOCA) events such as occurred in the Three Mile Island incident. The power burst tests in a 20 MWt PWR have created fuel failures and defined the initiating conditions. The LOCA tests with a 50 MWt... 

In the event of a malfunction of the reactor or safety related equipment, a condition known as a Loss of Coolant Accident (LOCA) may occur within the Primary Containment Structure. Should this happen, the environment would become dramatically altered in a matter of seconds and result in the escalation of temperature and pressure to dangerous levels. To counteract this condition, large quantities of water with chemical additives are automatically directed onto all surfaces by means of high pressure spray systems. 

Electric wires and cables used in nuclear power plant are exposed by low dose rate irradiation during the life time of the plant. In addition, loss of coolant accident (LOCA) that is a design basis accident brings about the degradation by heat and radiation on the electric wire and cables. The dose varies from plant to plant. IEEE std.323-1974 (/) estimated the dose 0.5 MGy for the period of the life time and 1.5 MGy for LOCA as one of the example. The standard estimated 40years and one year, for the lifetime of plant and duration of LOCA respectively. 

The earthquake caused extensive damage to the structures of the Fukushima—Daiichi power plant and knocked out the pump systems that supply cooling water to the reactors and the spent fuel pools. This is known as a Loss of Coolant Accident (LOCA) takes place. 

Loss of coolant accidents (LOCAs) as initiators of pressurized thermal shock (PTS). 

A system to provide makeup inventory to the RCS following small break loss of coolant accidents (LOCAs) and maintain adequate core cooling such that reactor conditions do not exceed acceptable safety limits (outlined below). 

The HPI system at TMI-2 correctly came into operation because the system was undergoing a loss of coolant accident (LOCA) because of the stuck open relief valve. But at the time, the operators did not know that yet. They had neither diagnosed a LOCA nor its cause, because the control room pressurizer water level instrumentation indicated a level that was higher than normal. 

When a loss of coolant accident (LOCA) occurs in the primary circuit the water level inside the RPV nill decrease. Due to the integrated arrangement of the primary circuit and all penetration of small pipes located at the upper part of the vessel the reactor core will never be uncovered. But as a result of the water lev el decrease the natural circulation of the primary circuit might be interrupted. In this case the residual heat of the reactor will be transported by vapor condensed at the uncovered tubes of the primary heat e.xchangers. 

This is one of the engineered safety features provided to mitigate the consequences of Loss of Coolant Accident (LOCA) in the event of a break in Primary Heat Transport (PHT) system boundary. The ECCS is designed to provide enough coolant to the PHT system and to transport heat from the core to the ultimate heat sink in such a way as to ensure adequate core cooling during all phases of LOCA. 

The core of a light water reactor must always be kept covered with water, to prevent a melt down and the consequent release of radioactive products. The safety injection system task is to ensure that the core will remain covered with water in the event of a Loss of Coolant Accident (LOCA). In this event, the pressure vessel water level will begin to decrease. Before the water level is low enough to endanger the core, the injection system is triggered and floods the pressure vessel vwth water, preventing core uncovering. 

Generic Safety Issue (GSI) 023 in NUREG-0933 (Reference 1), addresses the high rate of Reactor Coolant Pump (RCP) seal failures that challenge the makeup capacity of the ECCS in PWRs which could result in a small-break loss-of-coolant-accident (LOCAs) and possibly result in core damage. 

Unresolved Safety Issue (USI) A-02 in NUREG-0933 (Reference 1), addresses asymmetric blowdown loads imposed on the reactor vessel (RV) as a result of a design basis loss of coolant accident (LOCA). The resultant forces from these loads could affect reactor vessel support integrity, thus jeopardizing plant safety. 

Typical environmental conditions (temperature, pressure, humidity, integrated radiation dose, and exposure to chemicals) are given in CESSAR-DC Appendix 3.11A and cover the 60-year design lifetime. Conditions are tabulated for normal operation in and outside of containment, and for loss-of-coolant accident (LOCA) and main steam line break inside containment (MSLB). 

Unresolved Safety Issue (USI) A-31 in NUREG-0933 (Reference 1), addresses the safe shutdown of the reactor, following an accident or abnormal condition other than a Loss of Coolant Accident (LOCA), from a hot standby condition (i.e., the primary system is at or near normal operating temperature and pressure) to a cold shutdown condition. Considerable emphasis has been placed on long-term cooling which is typically achieved by the residual heat removal system which starts to operate when the reactor coolant pressure and temperature are substantially lower than the hot-standby values. 

Since 1974 the NRC requirements for performing a loss-of-coolant-accident (LOCA) licensing analyses (ECCS analyses) have been specified in 10 CFR 50.46 Appendix K (Reference 2). During the years since 1974, extensive research has been conducted on the various aspects of a LOCA. Because of this research, 10 CFR 50 now states that "...It is now confirmed that the methods specified in Appendix K, combined with other analysis methods currently in use, are conservative and that the actual cladding temperature would be much lower than that calculated using Appendix K methods". 

Generic Safety Issue (GSI) C-10 in NUREG-0933 (Reference 1) is concerned with the effectiveness of various containment spray solutions in removing airborne radioactive materials present in the containment after a loss-of-coolant accident (LOCA). Also of concern is the possible damage to equipment in the containment caused by the solutions in an inadvertent actuation of the spray system. 

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