Band 79, Heft 4

Evolution of VVER-440 fuel assembly is characterised by gradual increase of discharged fuel burn-up and prolongation of fuel cycle. Mentioned effects can be reached effectively by fuel enrichment raise and by parasitic absorption reduction. Both methods were used during VVER-440 fuel assembly optimisation, described at this paper. Enrichment interval was enlarged up to 5.95 % 235 U, parasitic absorption was limited by carcass frame of the assembly. Basic characteristics of the best assembly proposals are presented and effects on equilibrium fuel cycle of VVER-440 reactor are characterised.

Because of the difficulties with experimental validation of power distribution predicted by macro-code on the pin by pin level we decided to prepare a calculation benchmark named “FULL-CORE” VVER-440. This benchmark is a two-dimensional (2D) calculation benchmark based on the VVER-440 reactor core cold state geometry with taking into account the geometry of explicit radial reflector. The main task of this benchmark is to test the pin by pin power distribution in fuel assemblies predicted by macro-codes that are used for neutron-physics calculations especially for VVER-440 reactors. The proposal of this benchmark was presented at the 21 st Symposium of AER in 2011. The reference solution has been calculated by MCNP code using Monte Carlo method and the results have been published in the AER community. The results of reference calculation were presented at the 22 nd Symposium of AER in 2012. In this paper we will compare the available macro-codes results of this calculation benchmark.

This paper deals with the development of an approximation method to evaluate the isotopic composition of burnt VVER fuel rods (with U-Gd) by engineering codes (TVS-M, PERMAK-A) and taking into account fuel rods depletion conditions. Burn-up conditions for fuel rods differ from each other during operation due to the presence of water cavities, absorber and other within FA. Therefore, the isotopes evolution during burn-up in different fuel rods goes on in different manner even with identical values of fuel burn-up (MW · day/kg U). The isotopic content evolution can be described in detail by spectral codes. However, a full-size core calculation takes much running time. The approximation method, which takes into account fuel rods spectral differences during operation, was considered. The method is based on spectral functionality (a non asymptotic factor) usage, which allows to take into account the spectral conditions and history of exposure.

The Monte Carlo code SERPENT (http://montecarlo.vtt.fi) is tested and validated for different fuel configurations but not for VVER fuel assemblies with Gd fuel pins and enrichment profiling. The presentation outlines the results obtained during fuel assembly burnup calculations for VVER-440 and VVER-1000 reactor types with Gd fuel pins. The calculations follow the proposal of the benchmark for VVER-440 fuel assemblies specified by Mikolas at the 14 th Symposium of AER in 2004. For fuel assemblies of VVER-1000 reactor types with Gd fuel pins a benchmark by the OECD/NEA was specified in 2002. A set of data for the further verification of the few group data preparation in the framework of the VVER-1000 reactor core burnup benchmark proposed at the 19 th AER Symposium in 2009 is received and analysed. The results presented show sufficient agreement with the reference values.

The CB6 benchmark on VVER-440 final disposal consists of two parts. The decay calculation part is aimed at determination of isotopic composition of irradiated fuel material over the time scale relevant to long-term spent fuel disposal. The second part involves calculation of k eff values in a 3D cask configuration at various time points. In this study, the code Serpent is applied in solution of the benchmark. Serpent is a continuous energy Monte Carlo reactor physics code developed at VTT Technical Research Center of Finland. The decay calculation part of the benchmark is, in addition to Serpent, also carried out using point-depletion calculation code ORIGEN. This enables comparison of the results obtained using the two codes exploiting two different methods in solution of the decay equations. The paper demonstrates the decay and k eff calculation results as determined in the benchmark specification. The differences in results arising between the two calculation codes in decay calculation, and between the ENDF/B-VII and JEFF-3.1.1 based cross section libraries in k eff calculation, are depicted and discussed.

The paper reports about the development and verification of the new nodal methods to be used in the KIKO3DMG code. Two classes of the new methods are presented. The first class makes the treatment of the heterogeneities possible inside the assemblies while the extent of the crucial approximations applied on the node boundaries is more considerable in comparison to those in case of the second type. The nodal methods were validated by two VVER reference problems found in the AER benchmark book (aerbench.kfki.hu/aerbench). The AER-2 and FCM-101 benchmarks correspond to the VVER-440 and VVER-1000 geometry, respectively. It was found that the differences between the converged and the reference solutions are negligible from the practical point of view. The performance characteristics concerning the accuracy and the necessary CPU time – both depending on the mesh refinement – were also compared.

The European version of Supercritical Water Reactors (SCWR), the High Performance Light Water Reactor (HPLWR) operates in the thermodynamically supercritical region of water. Our basic objective was to elaborate a stationary coupled neutronic-thermohydraulic code capable for the calculation of the actual 3-pass core design with fuel assembly clusters. The calculations covered the neutronic transport calculations of HPLWR fuel assemblies, the coupled neutronic-thermohydraulic global calculations and the pin-wise analysis. Applying conservative assumptions, the relation to the linear heat rate and maximum cladding temperature limits was checked for the equilibrium cycle of HPLWR with this new code system.

The paper proposes the assessment techniques for reactor scram effectiveness (including the case of getting stuck CR of the greatest effectiveness) based on the comparison of measured and calculated effectiveness of individual CR groups. The paper assesses the weight coefficients with which the differences between measured and calculated effectiveness of individual CR groups are taken into account in reactor scram efficiency assessment. Methodological uncertainties of this technique are evaluated in the paper.

In this paper a process of determination of methodological component of power distribution engineering factors of the macrocode MOBY-DICK (MD) is presented. This process is based on a direct comparison of measured and calculated data. The directly measured data analyzed in this article are thermocouple (TC) and Rhodium self-powered neutron detector (SPND) readings. Also the normality of deviations between measured and calculated TC temperature rises is discussed and methodological uncertainties of the macrocode MD are then checked by 97.5th percentile. Absolute and relative uncertainties of volumetric (K V ) and radial (K q ) power distributions are determined by the comparisons of power distributions calculated by MD code with either power distributions provided by the Dukovany NPP core monitoring system (SCORPIO) or with power distributions obtained from direct TC and SPND readings. Final determination of engineering factors supposes hyperbolic dependence of MD methodological uncertainties on power peaking.

In this paper steps of calculations for a bare rod bundle and so-called swirl type spacer grid are showed and the effects of choice of model details are examined. The modeled pin bundle and spacer grid geometry are related to rectangular western type PWR fuel assemblies. First a CFD model has been developed for a subchannel of the MATIS-H test facility in order to carry out a mesh independence study. Based on the results a model for the half cross-section of the test bundle has been developed and calculations are carried out with different turbulence models. Translational periodicity is applied in axial direction in order to calculate a fully developed flow, which will be used as inlet boundary condition for spacer grid calculations. Results are compared with LDA (Laser Doppler Anemometry) measurements published in the frame of the OECD NEA MATIS-H benchmark. As a last step a model for rod bundle incorporating swirl type spacer grid is developed and calculations are carried out with different turbulence models using the results of the bare rod bundle simulations as inlet boundary conditions.

The fulfillment of the safety analysis acceptance criteria and the normal operation safety related limitations are usually evaluated by using separate hot-channel or/and hot-assembly thermal hydraulic calculations, especially for the closed VVER-440 assemblies. However, the shroud-less assemblies of the expected new NPP units in Hungary or even those of some recently foreseen VVER-440 are raising several questions; like what is the role of the coolant mixing between the neighboring assemblies and can the closed assembly approximation be regarded as conservative for DBA analyses. Focusing an anticipated transient without scram (ATWS) event, the paper evaluates the thermal hydraulic calculation methodology of the shroud-less assemblies by supplementing the computational domain step by step with the closest pins and subchannels of the neighboring assemblies. We concluded that assembly-wise coolant cross-flows and significant changes in the hot-channel DNBR can appear in case of shroud-less assemblies.

The assessment of the uncertainties of COBRA-IIIC thermal-hydraulic analyses of rod bundles is performed for a 5-by-5 bundle representing a PWR fuel assembly. In the first part of the paper the modeling uncertainties are evaluated in the term of the uncertainty of the turbulent mixing factor using the OECD NEA/NRC PSBT benchmark data. After that the uncertainties of the COBRA calculations are discussed performing Monte-Carlo type statistical analyses taking into account the modeling uncertainties and other uncertainties prescribed in the OECD NEA UAM benchmark specification. Both steady-state and transient cases are investigated. The target quantities are the uncertainties of the void distribution, the moderator density, the moderator temperature and the DNBR. We will see that – beyond the uncertainties of the geometry and the boundary conditions – it is very important to take into account the modeling uncertainties in case of bundle or sub-channel thermo-hydraulic calculations.

The method developed by the Kurchatov Institute's WWER reactor department to detect local coolant boiling in the reactor core on the basis of the neutron noise analysis is currently being implemented at some WWER-1000 units within their regular in-core noise diagnostic systems. Given that WWER-1000 power uprating – as well as new designs being developed, such as AES-2006 and WWER-TOI characterized with higher core power and coolant temperature – would increase the probability of coolant boiling, the efficiency of the available boiling diagnostics method was to be re-evaluated. On the basis of relevant computations, it was suggested to modify the diagnostic model and introduce new parameters to improve the reliability of local boiling detection. Computations simulating in-core boiling were thus performed and showed that the improved method was quite efficient. This paper discusses the possibilities to detect – using the improved method – various in-core coolant boiling cases anticipated for WWERs.

The paper presents an uncertainty and sensitivity (U&S) study of the VVER-1000 reactor hydraulic properties. It is based on the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP) [1]. The novelty of the work consists of taking into consideration all hydraulic uncertainty parameters used in the modeling of the reactor pressure vessel (RPV) internals. A detailed parallel channel ATHLET model of the RPV is developed. It consists of ca. 26 600 control volumes most of them connected with junctions for cross flows. The specific geometry of the gap between upper part of the baffle and upper part of fuel assembly and also a fuel assembly head is taken explicitly into account The influence of the input parameters on the sensitivity and uncertainty of the RPV outlet and inlet temperatures and mass flows as well assembly-wise mass flow and coolant temperature axial distributions is shown.