Volume 71, Issue 4

A new multipurpose thermalhydraulic test facility TOPFLOW (TwO Phase FLOW) was built and put into operation at Forschungszentrum Rossendorf in 2002 and 2003. Since then, it has been mainly used for the investigation of generic and applied steady state and transient two phase flow phenomena and the development and validation of models of Computational Fluid Dynamic (CFD) codes in the frame of the German CFD initiative. The advantage of TOPFLOW consists in the combination of a large scale of the test channels with a wide operational range both of the flow velocities as well as of the system pressures and temperatures plus finally the availability of a special instrumentation that is capable in high spatial and temporal resolving two phase flow phenomena, for example the wire-mesh sensors.

The uncertainty in the calculation of neutron induced reaction cross-sections using modern nuclear models and codes has been investigated. The cross-sections have been calculated with the help of the TALYS code and the modified ALICE code using different models for the calculation of nuclear level density. The experimental data from EXFOR for neutron induced reactions for nuclei from 27 Al to 209 Bi and incident neutron energies above 0.1 MeV have been used for the comparison with calculations. The results obtained give the possibility to find the best approaches for the cross-section calculation for nuclei from different mass ranges.

In this paper an exposure build-up factor formulation is reported by solving the photon transport equation in a heterogeneous slab by the LTS N method, assuming the Klein-Nishina scattering kernel as the scattering differential cross section as well as the multigroup model in the wavelength variable. Numerical simulations and comparisons with available results in the literature are presented for different compositions containing water, iron and lead.

A simple model has been developed to predict the transient behaviour of research reactors with simple means. The developed model uses a lumped parameters approach for the kinetics and heat transfer modeling with continuous feedback reactivities. The model is used for the analysis of the International Atomic Energy Agency (IAEA) 10MW MTR research reactor benchmark problem. Transient responses to reactivity insertion and loss of coolant flow are presented and analyzed. The model predictions are accurate enough compared with the calculations conducted in various institutions using different codes. It is shown that the model can provide accurate predictions as long as the clad temperature does not exceed the ONB (Onset of Nucleate Boiling) temperature. However, the results are very encouraging and the model is useful for practical purposes.

The Critical Heat Flux (CHF) is an important parameter, which limits the thermal hydraulic performance of the nuclear fuel bundle. The tools available for the prediction of the CHF are empirical in nature and are valid for their experimental range only. However, the recently developed Look-Up Table (LUT) approach has emerged to be a promising tool for predicting CHF in a tubular geometry over a wide range of parameters, which can be extended to the rod bundle geometry considering correction factors for the rod bundle effects. The error statistics of the present assessment confirms the values provided with the LUT for the HBM approach for the tubular application. However, the error statistics by DSM (not provided with the LUT development) is found to be quite different from that of the HBM. It is found that CHF in the rod bundle can also be predicted with a good accuracy using the Heat Balance Method (HBM). The proposed correction factor is found to improve the prediction accuracy of the LUT for the rod bundle application. This paper deals with the assessment of the CHF prediction by LUT for the tubular and bundle geometry and evaluation of the correction factor for rod bundle at the normal operating pressure of BWR (70 bar) using the experimental data base.

A static strength analysis finite-element model has been developed using the ANSYS computer code in order to simulate the axial compression in CANDU type fuel bundle subject to hydraulic drag loads, deflection of fuel elements and stresses and displacements in the end plates. The validation of the finite-element model has been done by comparison with the out-reactor strength test results. Comparison of model predictions with the experimental results showed very good agreement. The comparative assessment reveals that SEU43 and SEU43L fuel bundles are able to withstand high flow rate without showing a significant geometric instability.

The effect of fuel element packing density on the effective multiplication factor, k eff , in a typical pebble-bed reactor (PBR) is studied using MCNP and WIMS computer codes. It is observed that at normal operating conditions k eff increases with packing density. Increase in packing density of 1% results in approximately 0.0015 increase in k eff . A negative temperature effect is observed at all feasible packing densities. The effect of secondary coolant (water) ingress in the core at accidental conditions is studied at various packing densities. The effect of effective water density on k eff depends on packing fraction, fuel enrichment and fuel-to-graphite ratio. It is strongly positive in highly enriched fuel at low water density regardless of packing fraction. For fuel enrichment 8% and normal packing density, a maximum positive effect of approximately +0.10 on k eff is observed at a water density of ∼0.1 g/cm 3 . The positive effect is reduced at low enrichments. Above a water density of 0.5 g/cm 3 , the effect on k eff becomes negative regardless of enrichment and packing density.

Spent nuclear fuel resulting from reactor operation must be safely stored and managed prior to reprocessing and/or final disposal of high-level waste. Any spent fuel storage system must provide for safe receipt, handling, retrieval, and storage of spent fuel. In order to achieve the safe storage, the design should primarily provide for radiation protection, subcriticality of spent fuel, and removal of spent fuel residual heat. This article is focused on the design of a metal-shielded dry-cask storage system, which will host spent LWR fuels burned to 33000, 45000, and 55000 MWd/t U and cooled for 5 or 10 years after discharge from reactor. The storage system is analyzed by taking into account radiation protection, subcriticality, and heat-removal aspects; and appropriate designs, in accordance with the international standards.

The main characteristics, such as temperatures of the fuel rod cladding and cask surface, dose rates at the surface and at the some distance for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks loaded with spent nuclear fuel are presented. These casks are used for an interim dry storage of spent nuclear fuel at Ignalina Nuclear Power Plant. Numerical modeling (calculation of the equivalent dose rates, activities of nuclides, etc.) and experimental measurements of the equivalent dose and gamma spectrum on the cask surface at the dry storage facility were performed for assessment of radiation characteristics. Temperatures were evaluated using only numerical modeling. Rather good agreement between experimentally determined and calculated dose rates for CASTOR RBMK-1500 and CONSTOR RBMK-1500 casks was obtained. Also it was revealed that maximum fuel rod cladding temperature is higher for CONSTOR RBMK-1500 cask, but never exceeds the maximum allowable value. The cask surface temperatures are similar for both cask types.

Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena.