[en] Presently, new conceptions of nuclear power reactors are being studied for the next generation; among them, high conversion reactors should be mentioned. Their inclusion in an energy production strategy offers the prospects of significant uranium savings, when compared with presently well-known designs, and it can be assumed that they will contribute to a new nuclear era, with better solutions for many of the present problems. It can be foreseen that the new systems, under research and development at the moment, would provide a very big quantity of energy, able to compensate the diminishing of natural resources and to reduce the environmental impact of the energy generation, because of their higher efficiency and better conditions for waste disposal. In the present paper, HCR's main characteristics are described, as well as the most important design problems faced at the moment. Information on the activities connected with the subject under development in the Centro Atomico Bariloche is also given. (author)

[en] The development of high converter reactors (HCR) requires careful burnup calculations because their main goals are reach high discharge burnup levels (Up to 50 GWd/T) and a close to one conversion ratio. Then, it is necessary a revision of design elements used for this type of calculation. In this work, a burnup module (BUM) developed in order to use nuclear data directly from evaluated data files is presented; these was included in the AMPX system. (author)

[en] This paper presents the experimental results of Critical Heat Flux (CHF) measurements which were obtained in 8 mm round tubes using the model fluid Freon-12 under the conditions: pressure from 1.05 to 2.72 MPa, mass flux from 1 to 6 Mg/(sxm²) and inlet temperature from -15deg C to saturation conditions. Reasonable agreement is found between the experimental data and the well known CHF look-up tables. The experimental data are also compared with the data in small diameter tubes. The diameter influence on CHF is discussed. On the average, the experimental data are found to agree well with the predictions of two different CHF-correlations, but the scatter is fairly large. (orig.)

[en] Short communication

[en] After the reflooding tests in an extremely tight bundle (p/d=1.06, FLORESTAN 1) have been completed, new experiments for a wider lattice (p/d=1.242, FLORESTAN 2), which is employed in the recent APWR design of KfK, are planned at KfK to obtain the benchmark data for validation and improvement of calculation methods. This report presents the results of pre-test calculations for the FLORESTAN 2 experiment using FLUT-FDWR, a modified version of the GRS computer code FLUT for analysis of the most important behaviour during the reflooding phase after a LOCA in the APWR design. (orig.)

[en] Nuclear reactor microcell calculations are, normally, carried out using simplified geometrical models which do not include the total number of homogeneous zones actually present. Regarding the particular case of High Conversion Pressurized Water Reactors (HCPWR), the revision of this approximation has been done to determine the sensitivity of its neutronic parameters to the use of these models. The study was performed comparing multiplication factors, reaction rates and neutron spectra, obtained using different geometrical treatments for a HCPWR typical microcell. From the results, it can be asserted that, if only two zones should be used in the calculation, the model which dilutes the clad into the moderator gives best results for the neutron fluxes, but the model that mix it with the fuel is better for k-infinite and reaction rate values. Considering the significance of these parameters on the physical behaviour of the reactor, the last one is recommended for cell calculations. Even if there is a slight difference between the cells considered, some results of this work were also compared with those of the NEACRP HCLWR benchmark with good agreement, so it can be concluded that the methodology here used for data processing and calculations is applicable to HCR's cell studies. (Author)

[en] Recognizing that, in spite of public acceptance problems in most of the industrialized countries, nuclear power will play an important role in supplying energy to a rapidly growing world population, improvement in uranium utilization is an essential element in the further development of light water reactors. A five-year research and development program oriented towards experimental and theoretical investigation of the design limits of a light water high conversion reactor within a cooperative agreement between Siemens, the Karlsruhe Nuclear Research Center, the Paul Scherrer Institute, and the Technical University of Braunschweig was recently completed. Its results clearly indicate the technical feasibility and economic viability of the high conversion reactor concept projected as a logical evolution of the current pressurized water reactor. (orig.)

[en] The integrated program system HEXTIME features a neutronics model for solving the time-dependent few-group diffusion equation in hexagonal-z geometry coupled with an advanced version of the well-known thermal-hydraulics code COBRA-III. Combined with an accurate and efficient pin power reconstruction module the program is thus capable of performing not only global neutronics/thermal-hydraulics calculations but also to evaluate important safety-related parameters like DNB ratios and centerline fuel temperatures. The basic neutronics equations implemented in HEXTIME are essentially the same as those developed for the steady-state code HEXNOD. Therefore, in this paper only the differences to the static formulation are discussed in more detail. Time-integration is based on a fully implicit Euler method in combination with an exponential transformation technique to reduce truncation errors. The reconstruction of pinwise power distributions relies on a hexagonal version of the weak element approximation method which has already been used successfully in Cartesian geometry. The thermal-hydraulic part of the reactor calculation is based on the program COBRA III-C/P. It allows 3-dimensional steady-state and transient full-core as well as subchannel analyses to be performed. The capability to calculate crossflow effects is essential to achieve the high degree of spatial resolution and accuracy aimed at in the coupled system. The calculation procedure between neutronics and thermal hydraulics and of the related thermal margin and DNB evaluation methodology is therefore presented in some detail. A discussion of first results of a steam line break analysis completes this paper. (author). 12 refs, 2 figs

[en] One of the problems which must be solved in severe accidents is the melt concrete interaction which does occur when the core debris penetrates the lower pressure vessel head and contacts the basement. To prevent these accident consequences a core catcher concept is proposed to be integrated into the new PWR design. The core catcher consists of a perforated steel plate covered with a sacrificial layer and is connected with the basement by a supporting structure. In case of an accident the interfacial gap under the bottom plate is filled with sumpwater. In the sacrificial layer melt plugs are inserted which are the extension of the holes in the bottom steel plate. When the core melt erodes the layer, the plugs open by melting, and water/steam flow through the plugs and the melt starts. This continuous cooling stabilizes the melt and finally allows flooding of the upper melt surface. In order to identify the dominant processes during flooding of the melt from the bottom, two prototypic experiments in laboratory scale have been carried out. (orig.)

[en] Twelve different test zones, Cores 7 to 18, have been studied to date in the Phase II program of LWCHR physics experiments at the PROTEUS zero-power facility. This paper reviews the test lattice configurations investigated, the types of integral measurements carried out, experimental techniques and accuracies, the transferability of results to LWHCR design, as also typical comparisons with calculations based on standard LWR methods. It has been shown that the experimental data base provided is broad - in terms of both high converter design characteristics represented and the types of integral data measured. Thus, the experimental program covers changes in moderation ratio (lattice geometry) and effective fissile-Pu enrichment - with investigations of neutron balance components, moderator voidage effects, influence of lattice poisoning, relative control rod worths and core heterogeneity effects. The importance of having such a broad data base is illustrated by the trends currently reported for the C/E (calculation/experiment) variation for reaction rate ratios with degree of moderation. (author). 18 refs, 3 figs, 6 tabs