[en] The LOBI experimental facility and the BT01 experiment is described. The experiment represents a small break transient in the secondary side (steam line) followed by special conditions for the establishment of pressurized thermal shock and accident management procedures. The computational analysis has been performed by the CATHARE thermal hydraulic system code. The results of calculations are in satisfactory agreement with the experimental values. A comparison has been made with a secondary side break test performed on the PMK-2 facility. (author). 14 refs., 26 figs., 6 tabs

[en] The CODEX-2 experiment was performed with a VVER-440 type electrically heated 7-rod bundle. The bundle with UO₂ pellets was fixed by three spacer grids and covered by a hexagonal shroud. The facility was heated up in argon atmosphere, then steam was added to the coolant and the electrical power was increased. Temperature escalation occurred in the upper part of the bundle which was destroyed during the temperature transient. Post-test examination of the bundle cross sections was carried out after the experiment. (author)

[en] The LOBI experimental facility and the BT17 experiment. This experiment represents a loss-of-feedwater transient with feed and bleed procedure. The computational analysis has been performed by the CATHARE thermal hydraulic system code. The results of calculations are in satisfactory agreement with the experimental values. A comparison has been made with a Loss-of-Feedwater test performed on the PMK-2 facility. (author). 16 refs., 22 figs., 5 tabs

[en] The RIA and LOCA fuel safety criteria are under revision in the international community of fuel suppliers, authorities and research organizations. The main criteria will be reviewed in the paper for WWER fuel. Experimental data on the fuel failure behaviour under reactivity-initiated accident (RIA) conditions produced in the last decade in French and Japanese test reactors indicated low failure enthalpy for high burnup fuel compared to fresh fuel. However the high burnup was not the only phenomenon influencing the fuel failure. The oxide scale on the external surface of the fuel rod, hydrogen content of the Zr cladding and the local hydriding seemed also be responsible for the failure at low enthalpy. Furthermore differences have been found between Western design fuel and Russian type WWER fuel. The burnup dependence of fuel failure for WWER fuel was found much less, probably due to the low oxidation during normal operational conditions compared to other PWRs. The recently published Vitanza and KAERI correlations for RIA failure enthalpy have been applied to 23 WWER tests. Experimental data from Russian IGR and BIGR reactors have been used. The calculations have shown that both burnup and cladding oxidation effects must be considered, however the pulse width dependence of failure enthalpy has not been confirmed. During loss of coolant accidents (LOCA) the peak cladding temperature and local oxidation criteria have to be met. The oxidation criterion is under discussion today in many laboratories. The AEKI carried out several experimental series with Zr1%Nb cladding used in WWER reactors. The paper will describe the main results of the tests and present the limit for ductile-brittle transition derived from ring compression test. The behaviour of Zr1%Nb (E110) and Zircaloy-4 claddings under LOCA conditions will be compared as well. (author)

[en] A series of ballooning experiments was performed at the KFKI-AEKI in order to compare the mechanical behaviour and strength of Zircaloy-4 and Zr1%Nb claddings. The effects of temperature, oxidation and iodine absorption on deformation and burst pressure was investigated in almost 100 biaxial tests. Numerical post-test analyses have also been performed with the stand-alone fuel module of the French CATHARE code and the US fuel behaviour code FRAP-T6. Comparing the experimental and the analytical results, relevant differences of high temperature strength due to different α-β phase transition temperature were revealed between the investigated cladding materials. (author)

[en] Experimental investigation of the interaction of high temperature VVER bundle with cooling water under severe accident conditions was carried on the CODEX facility. The test section of the facility included a seven-rod hexagonal bundle with electrical heating. During the tests the bundle was heated up and quenched by cold water. Two experiments were carried out with the same bundle, but with different conditions: preoxidation with quenching at 1150 deg C (CODEX-3/1) and quenching of the pre-oxidized bundle at 1500 deg C (CODEX-3/2). The tests provided an unexpected result, as very limited temperature increase and hydrogen production was observed. (author)

[en] The behaviour of failed fuel rods includes several complex phenomena. The cladding failure initiates the release of fission product from the fuel and in case of large defect even urania grains can be released into the coolant. In steady state conditions an equilibrium - diffusion type - release is expected. During transients the release is driven by a convective type leaching mechanism. There are very few experimental data on leaking WWER fuel rods. For this reason the activity measurements at the nuclear power plants provide very important information. The evaluation of measured data can help in the estimation of failed fuel rod characteristics and the prediction of transient release dynamics in power plant transients. The paper deals with the simulation of leaking fuel rods under steady state and transient conditions and describes the following new results: 1) A new algorithm has been developed for the simulation of leaking fuel rods under steady state conditions and the specific parameters of the model for the Paks NPP has been determined; 2) The steady state model has been applied to calculation of leaking fuel characteristics using iodine and noble gas activity measurement data; 3) A new computational method has been developed for the simulation of leaking fuel rods under transient conditions and the specific parameters for the Paks NPP has been determined; 4) The transient model has been applied to the simulation of shutdown process at the Paks NPP and for the prediction of the time and magnitude of ¹²³I activity peak; 5) Using Paks NPP data a conservative value has been determined for the upper limit of the ¹²³I release from failed fuel rods during transients

[en] The activity release from damage fuel rods after the Paks-2 cleaning tank incident was estimated on the basis of coolant activity concentration measurements and chimney activity data. The release rates indicated that there were significant differences between the release of different elements. The typical release rate of noble gases, iodine and cesium was 1-2 %, but e.g. the release rate of ruthenium was only ∼ 0.01-0.04%. The dissolution rate of the uranium-dioxide was 0.3% during the first year, after the incident. (author)

[en] In the framework of the contract between IAEA and the KFKI Atomic Energy research Institute, numerical models have been developed for the simulation and thermal-hydraulic behaviour of CASTOR type spent fuel storage constructed at Dukovany NPP and MVDS type spent fuel storage operated by Paks NPP with WWER-440 fuel. The model is based on the code COBRA-SFS which is well validated for spent fuel storage system with western PWR and BWR type fuels

[en] A short summary is given on the so called thermal hydraulic system codes. Introducing this topic some elementary information is presented on thermal hydraulic phenomena occurring during accidents. Special attention will be paid to two-phase (gas-liquid) flow behavior and to the calculation of flows like that. The basic models of two-phase flow simulation is shown, and making use of the examples of SMABRE and CATHARE codes the main features of thermal hydraulic system codes are listed, including the basic equations, steps of code validation and verification and the cycle of code development strategy. In the conclusions an attempt will be made on the state-of-art description of the limitations and capabilities of thermal hydraulic codes. (author). 5 refs., 7 figs