[en] As part of the European Technology Fusion Programme, the irradiation embrittlement characteristics of the more ductile and isotopic grades of beryllium manufactured by Brush Wellman has been investigated using modern powder production and consolidation techniques . This study was initiated in support of the development and evaluation of beryllium as a neutron multiplier for the solid breeder blanket design concepts proposed for a DEMO fusion power reactor. Four different species of beryllium: S-200 F (vacuum hot pressed, 1.2 wt% BeO), S-200FH (hot isostatic pressed, 0.9 wt% BeO), S-65 (vacuum hot pressed, 0.6 wt% BeO), S-65H (hot isostatic pressed, 0.5 wt% BeO) have been compared. Three batches of the beryllium have been investigated, a neutron batch, a thermal control batch and a reference batch. Neutron irradiation has been performed at temperatures between 175 and 605 degrees Celsius up to a neutron fluence of 2.1 10²⁵ n.m^-2 (E> 1 MeV) or 750 appm He. The results of the tensile tests are summarized

[en] The Ag(NO₃)⁺ species is particularly useful to mineralize organic nuclear wastes at room temperature. The process produces no ashes and allows to confine the contaminants in a nitric acid solution. A successful analytical method for the determination of the speciation of silver in nitric acid solutions was studied. Since no suitable analytical standard for Ag²⁺ is available, an indirect method was used. First, Ag(NO₃)⁺ was reduced by an excess of Ce³⁺. Next, the resulting Ce⁴⁺ was measured spectrophotometricaly at 343 nm. Finally, Ag(NO₃)⁺ was determined directly by spectrophotometry by measuring the absorbance at 390 nm and by associating the latter to the concentrations obtained by the cerium method. The stability of the Ce⁴⁺ solutions in nitric acid has also been studied as a function of time. It was observed that the absorbance measured for these solutions decrease with time when the solutions are exposed to light. Conserving the standards in a dark environment allows to maintain a long term reproducibility

[en] The dismantling of a nuclear reactor results in large amounts of contaminated waste. The amount of waste is minimized by so-called decontamination processes. These decontamination processes result in decontaminated metal as well as contaminated effluents, containing acid and metal salts. Standard treatment of these effluents involves neutralisation of the acid using sodium hydroxide, resulting in the precipitation of hydroxides; and evaporation of the solvent. However, if the acid is extracted out of the decontamination effluents prior to neutralisation, this acid can be re-used in the decontamination process, substantially reducing the final waste. This report investigates the possibility of using Electro Electrodialysis (EED) for the extraction of 900 moles of sulphuric acid out of a 1 m³ contaminated solution, containing 22 kg of ferrous, chromous, cerous and nickel ions and 1000 moles of sulphuric acid. EED removes acids out of acid/salts solutions, using an electrolyser equipped with an anionic membrane: the acid-anion is displaced by migration through the membrane while the proton is displaced by electrolysis. This report gives a full description of the technique, the lab-scale apparatus, the analyses involved in studying the technique on lab-scale or monitoring it on full scale, the choice of electrode and membrane and of process parameters such as temperature and current density. On lab scale, EED can displace 90% of the acid out of the acid/salt solution just described, at a current density as high as 0.14 A.cm^-2, at 50 degrees Celsius, if the cathode and the anode are of smoothed platina or of platinised titanium, and the anion selective membrane is the Neosepta AMH -membrane of Tokuyama. The study of EED still needs to be completed with a further study of metal deposition on the cathode, and a comparison of the technique to other (electro)membrane techniques

[en] Activities at the Belgian Nuclear Research Centre SCK-CEN related to the optimization of its multi-entry model for computing decommission costs are described

[en] The results of seven years of total energy policy in the Netherlands are presented and the main objectives of the dutch energy policy are reviewed and evaluated. The reduction of energy consumption by making use of the total energy principle will contribute to the reduction of the emission of greenhouse gases such as CO₂. The objective of the dutch energy saving policy for the time -period 1990-2000 is to improve the annual efficiency with 2 percent. An interim evaluation of the energy saving policy indicates that in practice, efficiency will be improved by 1.8 percent per year. In the short-term, the total energy principle remains the main tool for energy saving. The expected total amount of energy, produced by the total energy principle (including district heating) varies from 3285 Mw e in 1990 to 4085 MW e in 1995 and 55000 Mw e in 2000. (A.S.)

[en] The economic aspects of energy production from waste wood are evaluated. Heating systems based on the incineration of wood have been considerably improved recently. Several aspects of the incineration of waste wood are reviewed: the implications with regard to the greenhouse effect, the calorific value of wood, the incineration process, and the cost price calculation of energy production by waste wood incineration. In conclusion is stated that energy production by waste wood incineration is a valuable economic alternative for heat production by oil products, especially in view of the current anti-pollution taxes in Belgium. (A.S.)

[en] In order to extend the UOP₂-PuO₂ irraditaion induced densification data necessary for the evaluation of the fuel performance, BN and CEN/SCK have decided in the framework of plutonium recycling in LWR's to carry out a research program to characterize the densification behavior of plutonium fuel fabricated by the manufacturing plant of Belgonucleaire. This programme includes the irradiation in BR2 of mixed oxide fuel rods typical of BWR's (P17 program) and PWR's (P20 program), the fuel rod behavior during irradiation being followed by fuel rod neutronographies, after successive BR2 irradiation cycles. After the fuel description, of the fuel rod and irradiation devices as well as irradiation conditions and neutrographies the results of the non-destructive examinations are given. The analysis of the neutrography techniques leads to an estimate of the errors introduced in the fuel column length calculations. A shrinkage of the fuel column length has been observed during irradiation, a maximum shrinkage of 1% being reached after 8000MWd/tM for the P17. A temperature dependence of the densification is noticed in the P20 irradiations: a temperature increase of the fuel accelerates the densification for a given burnup. The maximum densification is also temperature dependent. (A.F.)

[en] PU is considered as being a very toxic substance but it doesn't constitute an exception if the other industry-produced toxic substances are considered well. The movements of Pu within the biosphere as well as its intestinal absorption in man are strongly restrained by its chemical properties. The more frequent contamination pathway is represented by inhalation of small particulates of Pu. That may take place when a failure in the protective barriers is shown up during the retreatment of nuclear fuel, but such a kind of accident is scarcely likely. In normal circumstances, Pu is therefore thought to be scarcely human health harmful. It is true that Pu, entering, the organism is retained for longtime in bones and liver and after inhalation in lungs. In the case of high die exposure, Pu might theoretically be expected to induce an increase of cancer incidence in these organs. The risk associated with a Pu exposure is and will very probably remain an essentially occupational risk. The population runs fewer risks with Pu than with the products of other energy sources. (G.C.)

[en] Since 1963 the material testing reactor BR2 at Mol is operated for the realisation of numerous research programs and experiments on the behavior of materials under nuclear radiation and in particular under intensive neutron exposure. During this period special irradiation techniques and experimental devices were developed according to the desiderata of the different experiments and to the irradiation possibilities offered at BR2. The design and the operating characteristics of quite a number of those irradiation rigs of proven reliability may be used or can be made available for new irradiation experiments. A brief description is given of some typical irradiation devices designed and constructed by CEN/SCK, Technology and Energy Dpt. They are compiled according to their main use for the different research and development programs realized at BR2. Their eventual application however for different objectives could be possible. A final chapter summarizes the principal irradiation conditions offered by BR2 reactor. (author)

[en] A calorimeter designed for standard measurement of absorbed doses in graphite by photons with energies above 0.6 MeV and electrons with energies above 2 MeV is described. In the dose range above 50 Rad the total uncertainty is less than 0.5% if due attention is paid to the necessary corrections. Temperature drift rates can be reduced below 10^-7 6K/min. Three different methods using the calorimetric, ionometric and chemical principles to implement the unit of absorbed doses in water are compared to find out systematic errors for the direct determination of absorbed doses in water by calorimetry. Some preliminary experiments are caried out. (author)