[en] The fifth three-dimensional hexagonal benchmark problem continues a series of the international benchmark problems defined during 1992-1996 in the international VVER cooperation forum AER. The initial event of the fifth AER benchmark is a symmetrical break in the middle part of the main steam header at the end of the first fuel cycle and under the hot shutdown condition with one stuck control rod group. The main difference from previous benchmark is that the system works of the primary and secondary sides are considered in this benchmark. The main aim of the benchmark is a calculation of the transient after the recriticality had achieved. The solution of the fifth three-dimensional hexagonal dynamic AER benchmark problem obtained by code package ATHLET/BIPR8KN is presented. The used reactor scheme is described including the description of the core, primary and secondary side. The amount of necessary tuning and tools of tuning to achieve a requested in the definition of the problem reference values are considered. Comparative analysis of the results obtained by using a different detalization schemes are carried out. (author)

[en] In this paper main ways for WWER-440 fuel assemblies (FA) and fuel cycle improving are discussed and various examples are shown. Based on the presented results, the following conclusions have been made: 1) Operational reliability of this new fuel has the same level, as traditional fuel. 2) Technical solutions laid in the design of second-generation fuel assemblies, were proven and confirmed by the results of trial and commercial operation. 3) Development of a technical proposal on a third-generation fuel assembly is currently underway. 4) Profiled fuel with burnable absorber based on gadolinium makes it possible to realize full-scale 5-year fuel cycles with average fuel enrichment reduced from 4.4-4.38% to 4.25% (in second-generation assemblies). 5) Present-day fuel cycles for WWER-440 developed on the base of new FA constructions (second-generation fuel) ensure considerable increase of nuclear fuel utilization efficiency. 6) Present-day fuel cycles for WWER-440 make it possible to realize various operational fuel load lifetimes. This allows optimal adapting of the unit's electricity production to the specific energy system requirements. 7) Fuel cycles for WWER-440 using modernized construction of fuel were developed. These cycles make it possible to work at excess reactor power of up to 105-110%. 8) In case the modernized fuel is used, present-day fuel cycles for WWER-440 also make it possible to realize the maneuvering reactor operation mode, when the reactor power varies in wide frames during a short time period

[en] The solution of the sixth three-dimensional hexagonal dynamic AER benchmark problem obtained by the code package ATHLET/BIPR8KN is presented. A report contains the descriptions of the plant model, have been chosen the solution of the benchmark problem. Models and approximations in use at the problem solution are given (Authors)

[en] A detailed model of the primary and secondary loop of the new design of WWER-1200 NPP is being created for the coupled system code ATHLET/BIPR-WWER. On the basis of the previously gained experience, a very detailed 3D modeling of the reactor pressure vessel (RPV) and of the steam generators (SG) is being successfully applied. The nodalization schemes of these objects are chosen to be optimal ones concerning the fidelity to the real geometry and the needed CPU time. The thermal fluid objects are modeled in ATHLET as objects of type 'pipe' most of them connected with cross flows, that allow to describe the mixing phenomena in RPV and in steam generators near to reality. A pre-processor system supports automatically to prepare the complex and great number of nodalized volumes for the ATHLET input. A detailed modeling of the control and safety systems covers a wide spectrum of initiating events. Generic design data are used to model the 3D neutron-kinetics in BIPR, applying the modernized fuel assembly design for WWER-1200. As a demonstration of the simulation capabilities of the coupled system code ATHLET/BIPR-WWER for the new reactor design WWER-1200 a RIA transient is analysed. A rod ejection within 0.1 s is simulated at nominal reactor power. The results are visualized with a special 3D graphical system developed in RCI KI. This advanced tool allows through rotating the whole model or only selected parts of it to visualize the thermal-hydraulic values at any location. By applying the 'cut' function the internal volumes of the 3D objects can be selected and visualized too. The coupled system code ATHLET/BIPR-WWER is being applied successfully by performing calculations analysing transients for the new design of a NPP with WWER-1200 reactor (Authors)

[en] The article describes methods for determination of the engineering margin factors currently used to estimate the uncertainties of the VVER reactor design parameters calculated via the KASKAD software package developed at the National Research Center Kurchatov Institute. These margin factors ensure the meeting of the operating (design) limits and a number of other restrictions under normal operating conditions.

[en] The new solution of the sixth three - dimensional hexagonal dynamic AER benchmark problem obtained by the code package ATHLET/BIPR8KN is presented. The main differences from the previous one consist in applying the new model of the steam generator. A report contains the descriptions of the plant model, have been chosen for the solution of the benchmark problem. Models and approximations in use at the problem solution are given (Authors)

[en] This paper describes the objectives of code development for coupling 3D neutronics codes with thermal-hydraulic system codes. The present status of coupling ATHLET with three 3D neutronics codes for VVER- and LWR-reactors is presented. After describing the basic features of the 3D neutronic codes BIPR-8 from Kurchatov-Institute, DYN3D from Research Center Rossendorf and QUABOX/CUBBOX from GRS, first applications of coupled codes for different transient and accident scenarios are presented. The need of further investigations is discussed

[en] The seventh dynamic benchmark is a continuation of the efforts to validate systematically codes for the estimation of the transient behavior of VVER type nuclear power plants. This benchmark is a continuation of the work in the sixth dynamic benchmark. It is proposed to be simulated the transient - re-connection of an isolated circulating loop with low temperature or low boron concentration in a VVER-440 plant. It is supposed to expand the benchmark to other cases when a different number of loops are in operation leading to different symmetric and asymmetric core boundary conditions. The purposes of the proposed benchmark are: 1) Best-estimate simulations of an transient with a coolant flow mixing in the Reactor Pressure Vessel of WWER-440 plant by re-connection of one coolant loop to the several ones on operation, 2) Performing of code-to-code comparisons. The core is at the end of its first cycle with a power of 1196.25 MWt. The basic additional difference of the 7-seventh benchmark is in the detailed description of the downcomer and bottom part of the reactor vessel that allow describing the effects of coolant mixing in the Reactor Pressure Vessel without any additional conservative assumptions. The burn-up and the power distributions at this reactor state have to be calculated by the participants. The thermohydraulic conditions of the core in the beginning of the transient are specified. Participants self-generated best estimate nuclear data is to be used. The main geometrical parameters of the plant and the characteristics of the control and safety systems are also specified. Use generated input data decks developed for a WWER-440 plant and for the applied codes should be used. The behaviour of the plant should be studied applying coupled system codes, which combine a three-dimensional neutron kinetics description of the core with a pseudo or real 3D thermohydraulics system code. (Authors)

[en] The aim of the performed study is to find a reliable correlation between the measured thermocouples values at core outlet of reactor VVER-1000 with the coolant temperatures at the assemblies outlets at these positions. Studies have been performed with the system coupled code ATHLET/BIPR-VVER, which allows to model the thermal-hydraulics and kinetics processes in VVER reactors in a 3D manner. The analysis is based on local in-core measurements data collected during the commissioning phase of VVER-1000 NPP Kalinin. The studied transient is a switch off of one main circulation pump at nominal power while the other there pumps remain in operation

[en] The question of intra-reactor vessel modeling with the aim of the calculation of the three-dimensional (coarse mesh) coolant parameters distribution in the reactor pressure vessel within the frame of the ATHLET/BIPR8KN application are considered. The turning on of the inactive loop of WWER-440 reactor of Kola NPP is presented as an example (Authors)