[en] The present volume contains 56 papers, presented on the sixteenth Symposium of Atomic Energy Research, held in Bratislava, Slovakia, 25-29 September 2006. The papers are presented in their original form, i. e. no corrections or modifications were carried out. The content of this volume is divided into thematic groups: Spectral and Core Calculation, Core Operation Experiments and Code Validation, Fuel Management, Core Surveillance and Monitoring, Reactor Dynamics Thermal Hydraulics and Safety Analysis, Physical Problems of Spent Fuel Decommissioning, and Radwaste, Actinide Transmutation and Spent Fuel Disposal - according to the presentation sequence on the Symposium (Author)

[en] The improvement of nuclear fuel cycle economy results in permanent introducing of new types of fuel. The goal is to maximize the burn up without any increase the leakage of fuel elements. The next generation of fuel with burnable absorber has been operated in Bohunice V-2 NPP since this year. This paper summarizes in brief the whole process of introduction of new fuel generation with burnable absorber Gd and first experience with it (Authors)

[en] The old designed Russian fuel assemblies TVS-M have been under operation for many years at Kozloduy NPP WWER-1000 reactors. It was gained a lot of experience with TVS-M assemblies operation and a distortion tendency was found out. To avoid this problem a decision for a gradually substitution of TVS-M fuel assemblies with advanced ones (TVSA) has been taken. Two different stationary fuel cycles with 42 and 48 fresh TVSA fuel assemblies are presented and discussed in this paper. The KASKAD computer code system is used for development of the fuel loading patterns of Units 5 and 6 at Kozloduy NPP (Authors)

[en] The paper describes methodology for quantification of fuel cladding failure as a result of Loss of Coolant Accident (LOCA) for WWER-440 reactor type. The methodology is based on external coupling of the thermo-hydraulic code RELAP5 and thermo-mechanical code TRANSURANUS. The thermo-hydraulic response of the unit to the accident is simulated by RELAP5 code, providing initial and boundary conditions for the thermo-mechanical simulation by the TRANSURANUS code. Cladding failure criterion of the TRANSURANUS code, derived and implemented into the code in the framework of EXTRA EURATOM fifth Framework Programme is used. Cladding failure probability is evaluated by the Monte Carlo algorithm varying the outer cladding temperature. In the second part of the paper, an example of application of the methodology for typical maximum design accident of the WWER-440 is given, presenting every step of methodology and typical failure rate for this type of accident (Authors)

[en] In this article are analyzed experiments for WWER-440 fuel and compared with theoretical results by new version of the SCALE 5 code: nuclide compositions - measurement in Kurchatov institute for 3.6% - measurement in Dimitrovgrad for 3.6% (project ISTC 2670) The focus is on modules TRITON and ORIGEN-S (Authors)

[en] In the paper some results of validation of the SCALE4.4 depletion control module SAS2H against experimental data are presented. The recently published ISTC report with WWER-440 PIE data for fuel samples from Novovoronez NPP and ORNL publication with WWER-1000 experimental data for fuel samples from Balakovo NPP have been applied for preliminary validation of depletion calculations. The calculated results for 14 actinides and 14 fission products for WWER-440 samples and for 12 actinides and 5 fission products for WWER-1000 samples have been compared with measured isotopic concentrations (Authors)

[en] There is lot of information bringing our attention to the problem of limited raw material resources. Fortunately uranium for nuclear energy is very concentrated source and that is why its transport brings no problems and could be realized from anywhere. Second question is if overall resources are available for current nuclear energy development. Data documenting reasons for nowadays price growth are presenting and it is clearly shown that the most probable explanation is that there is gap in new uranium mines preparation and the lot of smaller mines were closed in the period of low uranium prices. Conclusion is that there is at least for the first half of this century even for thermal reactors enough uranium. Situation could be changed if there will massive production of liquid fuel using hydrogen, produced through nuclear heating. Public information about former military uranium resources are also included. Contemporary about one half of US nuclear power-stations is using high enriched uranium diluted with natural uranium - it is expected to continue this way up to 2012. Uranium is complicated market (Authors)

[en] This article includes a test or in other words data verification of neutron ENDF/B-VIIb2 sub library. The first part consists from the process of preparation ACE files by NJOY 99.90. The starting point of data verification describes needed patches in NJOY 99.90, which are necessary to do for correctly production of ACE files. After the obtaining ACE files follow the test of all ACE files through GODIVA - input file for MCNP. GODIVA is high enrichment sphere of U-235, where every material is added as impurity. The aim of GODIVA test is to obtain a certainty if produced ACE files are able to run through MCNP. The second part of this article begins with choose of benchmarks from 'International Handbook of Evaluated Criticality Safety Benchmark Experiments, 2005'. From this source of criticality experiments were separated some benchmarks for quality verification of ACE files by MCNP (Authors)

[en] In the RRC 'Kurchatov Institute' during 3 last years it was realized work package directed to development of the method of power assemblies correction using thermocouples readings. This entire works include the experimental and calculation ones. The main result of the experiment investigations is fact of incomplete coolant mixing in the head of WWER-440 assembly. Furthermore experimental data were used for validation of CFD code PHOENICS and modernized subchannel code SC-1 (SC-1 is analogue of COBRA-IV code). Modernized SC-1 code is capable to calculate the coolant temperature distribution in not only the fuel part of assembly but the head of assembly too. In the future the modernized SC-1 code was used for calculation analysis of the thermocouples data that were obtained from 3^d and 4^th units of the Kola NPP. At the results of this analysis were developed the method of the moving of incomplete coolant mixing correction in the nozzle of WWER-440 assembly. The developed corrective functions were used for compare of calculated and measured relative assemblies power using thermocouples readings. The main result of such comparison is that the differences between calculated and measured relative assembly powers for representative sample are reduced as a whole. Total amount of the considered variants were more than 1200 (different units, campaigns, assemblies and time moments) (Authors)

[en] An accurate reconstruction of the pin-wise power distribution is an important part of the limit checking of reactor operation. In SCORPIO-WWER core monitoring system, this is provided by the RECON module on the basis of pre-calculated form-functions. A new technique was implemented for improvement of the accuracy of pin-wise power reconstruction via a new methodology of form-functions calculations (Authors)