[en] For the purposes of the waste management of nuclear power plants in the Federal Republic of Germany, the final storage of radioactive wastes is planned in a salt deposit. Since the Government of Lower Saxony has designated the salt deposit Gorleben as a possible site for a final storage facility, this concept is discussed in the public opinion emphatically and controversially. The present contribution to this discussion shall present the special characteristics of rock salt in a generally intellgible manner, which are the reason for the applicability of this rock for a long-term safe final storage. Moreover, open questions and problems will be explained at least by means of examples. (orig.)

[en] Ten different methods for the extraction of tritium from the purge gas of a ceramic blanket are described and evaluated with respect to their applicability for ITER and DEMO. The methods are based on the conditions that the purge gas is composed of helium with an addition of up to 0.1% of H₂ or O₂ and H₂O to facilitate the release of tritium, and that tritium occurs in the purge gas in two main chemical forms, i.e. HT and HTO. Individual process steps of many methods are identical; in particular, the application of cold traps, molecular sieve beds, and diffusors are proposed in several cases. Differences between the methods arise mainly from the ways in which various process steps are combined and from the operating conditions which are chosen with respect to temperature and pressure. Up to now, none of the methods has been demonstrated to be reliably applicable for the purge gas conditions foreseen for the operation of an ITER blanket test module (or larger ceramic blanket designs such as for DEMO). These conditions are characterized by very high gas flow rates and extremely low concentrations of HT and HTO. Therefore, a proposal has been made (FZK concept) which is expected to have the best potential for applicability to ITER and DEMO and to incorporate the smallest development risk. In this concept, the extraction of tritium and excess hydrogen is accomplished by using a cold trap for freezing out HTO/H₂O and a 5A molecular sieve bed for the adsorption of HT/H₂. (orig.)

[en] Out-of-pile experiments were carried out in the SASCHA facility to investigate fission product release and aerosol formation under LWR core melting conditions. Besides fuel temperature, chemical composition of the core melt and test atmosphere were considered as important parameters. The results include the following aspects: Behavior of iodine and cesium during and after release from the fuel; potential of AgI formation in the gas phase; size distribution and chemical composition of aerosol particles; influence of the degree of Zircaloy oxidation on the release; fractional release rate coefficients of 20 elements for 2400degC; release during melt/concrete interaction; estimates of integral values for the release of radioactivity, decay heat, and aerosol mass from a PWR primary system. (orig.)

[en] An experimental program is described which was performed in the frame of a tritium technology task for the NET/ITER fusion fuel cycle. The aim was to investigate commercial gas purifiers containing metallic getters for the purification of gas streams such as the plasma exhaust gas. Five purifiers with up to 3000g of getter material were tested in the PEGASUS facility mainly with respect to the removal of methane, which is known to be much more difficult to remove than other impurities like O₂, N₂, or CO. A proposal for a fuel cleanup method based on a combination of getter beds and Pd/Ag diffusors is presented as the main conclusion of the test program. The discussion of this method includes the aspects of flow rates, tritium inventory, and consumption of getter material. (orig.)

No abstract available

[en] Asea Brown Boveri (ABB) has completed its technology based program. The results developed under Work Breakdown Structure (WBS) 8, concentrated on technology development and demonstration have been partially implemented in newer turbine designs. A significant improvement in heat rate and power output has been demonstrated. ABB will use the knowledge gained to further improve the efficiency of its Advanced Cycle System, which has been developed and introduced into the marked out side ABB's Advanced Turbine System (ATS) activities. The technology will lead to a power plant design that meets the ATS performance goals of over 60% plant efficiency, decreased electricity costs to consumers and lowest emissions

[en] The Tritium Extraction System (TES) is a gas loop with a helium/hydrogen mixture as carrier gas which is used to extract the tritium generated in the blanket by a nuclear breeding process. In addition to a description of the operational requirements of a TES for an ITER test blanket, ten process options are discussed. Then an additional procedure is proposed which includes several process steps of the approaches mentioned above. This method uses a cold trap operated at ≥-100 C to extract HTO and a molecular sieve bed (MSB) operated at -195 C to extract HT/H₂. Both process steps will be tested and optimized in the pilot plant PILATUS at the Tritium Laboratory of Karlsruhe, TLK. The most important objectives of the corresponding tests are: - To prevent the formation of ice aerosol in the cold trap which could lead to a decreasing removal efficiency for H₂O/HTO, - to investigate the adsorption/desorption of H₂, HT, O₂, and N₂ in the MSB under realistic conditions with respect to gas flow rate and gas composition, - to select and test reliable analytical techniques, which are needed to demonstrate the feasibility of the extraction method as well as to supply appropriate methods for process control. (orig.)

[en] A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200⁰C, and the KfK results for release during core meltdown sequences

[en] Investigations on core melt aerosols were performed under largely realistic conditions at the melting facility SASCHA. The following sampling techniques and analytical methods were applied: filtration, cascade impactor, aerosol centrifuge, alpha- and gamma-ray spectrometry, X-ray microanalysis, and Auger electron spectroscopy. The size distribution of the aerosol particles in a steam atmosphere was found to be log-normal with a median-value of the particle diameter of 0,09 μm and a geometric standard deviation of 2,13. Aerosol mass concentrations up to 30 g/m³ were generated along with particle number concentrations in the range of 10⁸ particles/cm³. By microchemical analysis of the mixed aerosol it was found that each single particle contained most of the elements being present in the melt material and that the chemical forms of the particle constituents were mostly oxides but in some cases also hydroxides or the elemental forms. (orig.)

[en] The main task of a Tritium Extraction System (TES) for a helium cooled Li₄SiO₄ DEMO blanket is the tritium recovery from a purge gas stream. On the basis of several TES proposals published for a NET/ITER solid breeder blanket, a new concept has been developed which is especially appropriate to cope with large purge gas streams. As tritium is expected to appear in two chemical forms (HT and HTO) two specific process steps are used for its removal from the primary purge gas loop: a cooler to freeze out Q₂O at 173 K (Q = H,T), and a molecular sieve bed to absorb Q₂ at 78 K. Only these components including some additional devices for the gas pre-conditioning like a compressor and a precooler, are subjected to the high gas flow rates mentioned above. All further processing is done in relatively small secondary loops during and after warmup of the cooler and the molecular sieve bed. Q₂O reduction by using the water gas shift reaction, and separation of Q₂ with Pd/Ag diffusors are the main process steps in the secondary loops. The feasibility of the proposed method is very promising as all process steps are based on well known technical and radiochemical experience. 20 refs., 4 figs., 1 tab