Pipe breaks in one of the two main circulation loops, the service water system and purification and coolant system as well as steam and feed water line breaks are classified as loss of coolant accidents. The full range of loss of coolant accidents have been assessed. Piping breaks resulting in a loss of coolant from the circuit may occur within the reinforced leak-tight compartments of the ACS or in compartments that are connected to the outside environment. In accordance with regulatory requirements  the following loss of coolant accidents should be analyzed for nuclear power plants with RBMK-type reactors:
The LOCAs addressed in the SAR include the following accidents:
The SAR concluded that the Ignalina NPP is quite well protected against the breaks that occur in the reinforced leak-tight compartments if they do not result in local flow degradation. A prompt activation of the ECCS occurs for breaks with large discharge rates and for breaks with coincident failures that impair global circulation. However, the emergency core cooling system activation is not fast enough to ensure that dangerous, early temperature excursion do not occur following partial breaks in one GDH. However, note that if local deterioration of channel cooling occurs during this LOCA scenario, the contaminated coolant discharges to the ACS. Analysis also shows that four emergency core coolant pumps, i.e. either the ECCS pumps, or the AFWPs, are sufficient for adequate long term cooling.
In the LOCA scenarios analyzed, the peak fuel temperature did not exceed 1200 oC, and the fuel cladding oxidation did not reach the maximum allowable levels. The fuel cladding failure criterion of 700 oC is exceeded in the following LOCA scenarios: full break of the pressure header accompanied with multiple failures, full break of the GDH, and partial break of the GDH. Analysis shows that, except for the last case, the fuel cladding failure criteria are violated for only a very short period of time during the initial phase of accident. Thus, fuel cladding failure is not expected in the first two cases. In the LOCA scenario with flow stagnation conditions in one GDH, fuel elements could fail in several channels. Design modification to improve the activation of the short-term ECCS was recommended and accepted by the Ignalina NPP. This improvement would be implemented during implementation of the SIP-2.
The SAR analysis shows that for all LOCAs which occur inside the reinforced leak-tight compartments, pressure tube temperatures do not exceeded the failure criterion of 650 oC. Results of analysis also states that for all breaks inside the reinforced leak-tight compartments, the existing prescribed public dose limits would not be exceeded.
However, for breaks outside the ACS, especially for main steam line breaks, peak cladding and pressure tube temperatures as well as doses could exceed acceptance criteria. The main reason of this is that breaks outside of the reinforced leak-tight compartments do not trip the reactor nor do they activate the ECCS. Violation of acceptance criteria could also result due to a large number of pre-existing cladding failures permitted during normal operation, and due to a limited drainage capacity in the vented compartments. The SAR analysts propose a number of hardware modifications and changes in regulations and procedures to overcome the design weaknesses and to better protect the surrounding population against radiological exposure after steam rupture events. First of all an additional early reactor trip and emergency coolant injection for all break locations, based on the dP/dt measurements in steam separators should be installed. This modification will be implemented in the immediate future at the Ignalina NPP. The SAR also recommended as a safety enhancement measure to keep the number of pre-existing fuel rod failures as low as achievable. Means to rapidly remove the contaminated water from compartments that are in direct communication with the environment will be developed and implemented.
Downcomer breaks outside the ACS do not result in violation of safety criteria. However, reactor hall over-pressure protection may not be sufficient to prevent the release of contaminated coolant to the environment and provisions to improve the reactor hall over-protection will be installed during implementation of the SIP-2.