Operating nuclear power plants require a safety analysis report which confirms the original design basis and describes the behavior of the plant for all potential accidental conditions. In accordance with regulatory requirements, the safety analysis should be based on the current status of the systems, structures and components of the NPP, and should consider all the modifications carried out during upgrading outages including those changes which are committed for implementation. For the Ignalina NPP this information is presented in several reports. This includes the TOB , the Safety Analysis Report  and additional anticipated transients without scram analyzed in the SAR .
The initial safety studies were performed by the Russian design institute, RDIPE . For the evaluation of break flow of a steam-water mixture from the rupture an equilibrium two-phase flow model taking into account the hydraulic losses along the pipeline length was used. Critical flow of subcooled water through the break was calculated using a non-equilibrium flow model, which approaches the equilibrium model as the degree of subcooling is reduced. Calculations of the transient pressure response were performed using quasi-static correlations for energy and mass transfer processes. The RDIPE calculations ware performed before 1989 and therefore used the design thermal power level of 4800 MW. However, after the Chernobyl accident the maximum permissible thermal power level of Ignalina reactors was reduced up to 4200 MW.
The SAR computations reflect the present operational power level of about 4200 MW. The accident analysis performed in the SAR were undertaken using Western state-of-the-art computer codes. System codes such as RELAP5 and ATHLET were used for thermal-hydraulic analyses and modern Russian codes such as the 3-dimensional codes SADCO and MOUNT which incorporate coupled neutronic-thermal-hydraulic calculations were used for evaluating reactivity initiated accidents. A review of the verification and validation studies which had been performed for each of these codes was undertaken as part of the quality assurance program. The Western codes had been validated extensively for PWR and BWR reactors but had only limited validation for conditions relevant to the RBMK. The Russian codes had undergone varying degrees of verification. In order to compensate for this lack of extensive verification, the codes were used cautiously when any of the critical and unverified regimes were encountered.
A number of accidents sequences which have to be analyzed in accordance with current Lithuanian regulations were not explicitly addressed either in the Ignalina TOB  or in the SAR . As noted in Section 10, the SAR was initially conceived as a Western-style safety analysis report, but the completion of such a SAR would have consumed several times the resources budgeted for the in-depth safety assessment of Ignalina NPP. The scope, especially the scope of the accident analysis, was therefore defined as including assessment specific essential items . A list of 23 accidents was developed which was intended to cover the “worst case” for each accident category in the sense that these sequences bounded those accidental events which were not included. In order to ensure that no important sequence was omitted an assessment was made by Task Group which undertook the development of a Fault Schedule. The goal of this task was to prepare a summary of all the accidental conditions which can be identified as having the potential to lead to fuel damage or a release of radioactivity from the plant. However, a thorough comparison of the accidents considered in the Ignalina SAR with initiating events of an extended Fault Schedule showed that they are bounding most of the credible events and no sequences were found which would have required a modification of the essential items list of accidents specified in the Guidelines for production and review of Ignalina SAR .
This Section incorporates material from the SAR Report  and Barselina Phase 4 Report .