[en] This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and validation of the CHF and post-CHF heat transfer for the RBMK-1500 reactor fuel assemblies employing the VIPRE-02 code. This chapter describes the experiments, which were used for validating the CHF correlations, appropriate for the RBMK-1500 type reactors. These correlations after validation were added to the standard version of the VIPRE-02 code. The VIPRE-02 calculations were benchmarked against the RELAP5/MOD3.3 code. It was found that these user-coded additional CHF correlations developed for the RBMK type reactors (Osmachkin, RRC KI and Khabenski correlations) and implemented into the code by the author, provide a good prediction of the CHF occurrence at the RBMK reactor nominal pressure range (at about 7 MPa). Transition and film boiling are also predicted well with the VIPRE-02 code for this pressure range. It was found, that for the RBMK-1500 reactor applications, EPRI CHF correlation should be used for the CHF predictions for the lower fuel assemblies of the reactor in the subchannel model of the RBMK-1500 fuel assembly. RRC KI and Bowring CHF correlations may be used for the upper fuel assemblies. For a single-channel model of the RBMK-1500 fuel channel, Osmachkin, RRC KI and Bowring correlations provide the closest predictions and may be used for the CHF estimation. For the low coolant mass fluxes in the fuel channel, Khabenski correlation can be applied. The fourth chapter presents the verification of the CORETRAN code for the RBMK-1500 core analysis. The model was verified against a number of RBMK-1500 plant data and transient calculations. The new RBMK-1500 core model was successfully applied in several safety assessment applications. A series of transient calculations, considered within the scope of the RBMK-type reactor Safety Analysis Report (SAR), were performed. Several cases of the transient calculations are presented in this chapter. The HELIOS/CORETRAN/VIPRE-02 core model for the RBMK-1500 is fully functional. The RBMK-1500 CPS logic, added into the CORETRAN provides an adequate response to the changes in the reactor parameters. Chapters 5 and 6 describe the experiments and the analysis performed on the coolability of particulate debris bed and melt pool during a postulated severe accident in the LWR. In the Chapter 5, the coolability potential, offered by the presence of a large number of the Control Rod Guide Tubes (CRGTs) in the BWR lower head is presented. The experimental investigations for the enhancement of coolability possible with CRGTs were performed on two experimental facilities: POMECO (POrous MEdium COolability) and COMECO (COrium MElt COolability). It was found that the presence of the CRGTs in the lower head of a BWR offers a substantial potential for heat removal during a postulated severe accident. Additional 10-20 kW of heat were removed from the POMECO and COMECO test sections through the CRGT. This corresponds to the average heat flux on the CRGT wall equal to 100-300 kW/m². In the Chapter 6 the ex-vessel particulate debris bed coolability is investigated, considering the non-condensable gases released from the concrete ablation process. The influence of the flow of the non-condensable gases on the process of quenching a hot porous debris bed was considered. The POMECO test facility was modified, adding the air supply at the bottom of the test section, to simulate the noncondensable gas release. The process was investigated for both high and low porosity debris beds. It was found that for the low porosity bed composition the countercurrent flooding limit could be exceeded, which would degrade the quenching process for such bed compositions