Small-break loss of coolant accident

Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications, January... 

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 original test programme, including three AP-600 major design basis transients, small break loss of coolant accident, steam generator tube rupture and main steam line break, was concluded in October 1994. Two new tests were also performed to verify the facility repeatability and the effect of a more severe cooldown. 

LOCA = loss of coolant accident SBO = station blackout SLOCA = small break loss of coolant accident T = transient initiator. 

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. 

NUREG-0635, Generic assessment of small break loss-of-coolant accidents in combustion engineering designed operating plants, January 1980. 

After the incident at the Three Mile Island nuclear plant in 1979, the nuclear industry as a whole around the world intensified experimental and analytical investigations into all aspects of namral-circulation thermal-hydraulics. The small-break loss of coolant accident (SBLOCA) natore of the incident revealed the critical importance of natural circulation to understanding the response of such systems under these conditions. The entire industry worked to ensure that understanding of all aspects SBLOCAs, for existing and future systems, was correct and complete. 

SBLOCA Small-Break Loss of Coolant Accident... 

Aksan, N., 2008. International standard problems and small break loss-of-coolant accident (SBLOCA). Science and Technology of Nuclear Installations, http //dx.doi.org/10.1155/ 2008/814572. Article ID 814572, 22 p. 

The success-path definition should make sure that the most likely scenarios of plant disturbance after a design-exceeding seismic event are covered. At least two success paths - leading to a durable safe state of the plant - need to be defined, and one of the two success paths should include the mitigation of a small break loss of coolant accident (LOCA), unless even a small break LOCA can be excluded, e.g., due to a particularly high quality of the supports of small-bore primary piping such as measurement lines. 

The risk posed by severe LWR accidents is considered to be dominated by transient and small-break loss-of-coolant accident sequences in which the core is uncovered only after a prolonged boiloff of reactor coolant. The discussions presented in this module presume, for the most part, that the reactor vessel is pressurized. However, the potential for temperature-induced failures of the reactor coolant system pressure boundary is addressed. In addition, the discussion presumes that reactor shutdown (scram) successfully terminates the fission process, so that decay heat drives the core-damage process. Most of the processes discussed in the context of pressurized, decay-heat driven accidents would exist in unpressurized and/or ATWS sequences as well although such sequences would differ in timing, rates and extent of core heating and oxidation, thermal-hydraulic conditions including the presence of water in the lower plenum, and other factors. 

MEASURES TO MITIGATE SMALL BREAK IN LOSS-OF-COOLANT ACCIDENTS AND LOSS-OF-FEEDWATER ACCIDENTS—IE BULLETINS... 

The use of short and small-diameter nozzles to connect the main equipment units and elimination of lengthy large-diameter pipelines in the primary circuit, to exclude loss of coolant accidents initiated by large and medium-breaks in the primary circuit ... 

The low-pressure scenario which is initiated by a large-sized break in the primary circuit, that is an event similar to that of the loss-of-coolant design basis accident described in Section 6.2.1.. In the severe accident scenario it is additionally postulated that, after the action of the accumulators and the borated water storage tanks, the sump water recirculation pumps will fail to operate. Thus, the decay heat cannot be removed from the reactor core vnth the consequence that the water volume present inside the reactor pressure vessel (RPV) begins to boil off at about atmospheric pressure. The AB sequence of WASH-1400 describes such a large-break scenario. In this low-pressure scenario, the treatment of fission product behavior inside the primary circuit is comparatively simple the probability of occurrence of such an accident, however, is extremely small. 

LOCA (Primary) Loss of Primary Coolant Accident (small pipe break) LOCA (Secondary) Secondary Pipe Rupture (water or steam)... 

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