Karlsruhe Nuclear Research Center Research Articles

Fabrication of microstructures with high aspect ratios and great structural heights by synchrotron radiation lithography, galvanoforming, and plastic moulding (LIGA process)

Under the LIGA process for fabricating microstructures having high aspect ratios and great structural heights, synchrotron radiation lithography produces a primary template which is filled with a metal by electrodeposition. The metallic structure so produced is used as a mould insert for fabricating secondary plastic templates which, in mass production, replace the primary template. This is a report about the status of work performed by the Karlsruhe Nuclear Research Center with the cooperation of Siemens AG and the Fraunhofer Institute for Solid State Technology: By irradiation and development of polymethyl methacrylate (PMMA) plates, primary templates were produced which, for structural heights of several hundred μm, exhibit deviations in critical dimensions of less then some 0.1 μm. The best results were obtained with an X-ray mask consisting of a 25 μm thick beryllium foil and 18 μm thick absorber structures consisting of copper and gold. Practically perfect metallic replicas were obtained by electrodeposition of nickel in the PMMA microstructures and even details in structure of less than 0.1 μm size were reproduced. Moulding was done with a methacrylate-based casting resin with an internal mould release agent. By electrodeposition of nickel in the secondary templates, secondary metal structures were produced which practically do not differ from the primary structures. The LIGA process can be expected to be superior to other methods for fabricating microstructures with high aspect ratios if, in series production of microstructures with complex shapes, stringent requirements are imposed on the resolution, the aspect ratio, the structural height, and the parallelism of the structural walls.

Fuel Reprocessing Versus Direct Disposal of Spent Fuel—A Comparison from the Standpoint of Radiological Safety

Within the framework of the fuel cycle evaluation that has been conducted between 1981 and 1985 by Karlsruhe Nuclear Research Center in the Federal Republic of Germany, radiological safety has been considered an essential assessment criterion. The collective doses from normal operation turned out to exceed markedly the accidental doses associated with all stations at the back end of the fuel cycles. Both occupational and nonoccupational doses in the fuel cycle based on fuel reprocessing are higher than those from the once-through cycle, but radiological exposure of the population consists mainly of small individual doses that are only a small fraction of the doses due to natural background radiation.

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Alternative Fuel Cycle Evaluation in the Federal Republic of Germany

In compliance with a 1979 governmental decision, Karlsruhe Nuclear Research Center managed the research and development (R&D) program “Alternative Entsorgun,” which had two major goals: (a) development of technical concepts for direct disposal of spent fuel and (b) comparison of the traditional back end of the nuclear fuel cycle based on fuel reprocessing with direct disposal. With respect to radiological safety, this comparison revealed that direct disposal scores better than the reprocessing path, but the difference is judged minor when compared to doses from natural radiation background. From the standpoint of cost, direct disposal proves more economical by a margin of 30% even though the reference system of direct disposal was not cost optimized. Finally, some unresolved but not unresolvable issues exist with regard to safeguarding a repository for spent fuel. In its decision in January 1985, the government of the Federal Republic of Germany stated that the results of the R&D program indicated that the legal requirement to ensure Entsorgung were better fulfilled by fuel reprocessing than by direct disposal of spent fuel.

Entwicklung des Trenndüsenverfahrens zur Uran‐Anreicherung

AbstractDevelopment of the nozzle separation process for enrichment of uranium. In the nozzle separation process an enrichment of 235U is achieved by extremely high centrifugal forces in a curved flow of UF6 diluted with a light gas. In Brazil a cascade section consisting of 24 nozzle separation stages has been constructed and has been in operation since late 1985. To produce nozzle separation systems with extremely small characteristic dimensions permitting plant operation at atmospheric pressure, a manufacturing process has been developed which relies on combined X‐ray lithography and microgalvanic deposition (LIGA‐technology). Using this technology the prototype separation stages constructed at the Karlsruhe Nuclear Research Center enable commercial nozzle separation plants with capacities of up to 3 Mio. SWU/year to be realized.

Assessment of the Risk Resulting from Postulated Accidents in Nuclear Power Plants: An Interim Report on Advanced Studies

As a continuation of the “German Risk Study: Phase A,” further plant-oriented analysis has been performed and the off-site accident consequence model has been partially improved. The plant-oriented analysis carried out at the Karlsruhe Nuclear Research Center has been focused on two release categories, namely FK2 (core meltdown followed by immediate release of radionuclides from the leaking containment) and FK6 (core meltdown followed by late release of radionuclides as a result of failure due to overpressure in the containment). The thermohydraulic processes in the molten mass and the behavior of the fission and activation products released from the molten mass are considered in a more realistic way. The improvements of the off-site accident consequence model relate mainly to a more realistic modeling of the deposition and resuspension processes, to the ingestion submodel, and to dose factors. The results show that the improvements of the off-site accident consequence model do not change the final results dramatically; the model rather draws a more realistic picture of the interrelated processes, and consequently allows the application of the model to other problems in the nuclear field as well. From the investigations belonging to the plant-oriented analysis it appears, however, that the releases to be expected from postulated accidents are remarkably lower. Consequently, the risk is lower than assessed in Phase A of this study.

Conceptual design of the tritium laboratory of the Karlsruhe nuclear research centre

Conceptual design of the tritium laboratory of the Karlsruhe nuclear research centre

Experimental Determination of the Atmospheric Dispersion Parameters at the Karlsruhe Nuclear Research Center for Emission Heights of 60 and 100 m

Experiments were carried out at the Karlsruhe Nuclear Research Center to determine the atmospheric diffusion of pollutants. The influence on atmospheric diffusion by conditions specific to the site...

Fusion technology in the Karlsruhe Nuclear Research Center

The Karlsruhe Nuclear Research Center was founded in 1956 in order to resume activities in nuclear research and technology in post-war Germany. The Center is a Government founded organization with holdings of 90% by the Federal Government and of 10% by the state of Baden-Württemberg. The total funding is about 500 millions of DM per year. The total number of staff is about 3400, with about 60% working at the about 25 scientific institutes and departments, 30% in supply services and 10% in the administration.

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The main acitivities of the Karlsruhe nuclear research center

The Karlsruhe Nuclear Research Center in West Germany was founded in 1956 in order to resume activities in nuclear research and technology in post-war Germany. The Center is a limited liability company with holdings of 90% by the Federal Government and of 10% by the state of Baden-Wuerttemberg. The total funding is about 500 millions of DM per year. The total number of staff is about 3,400, with about 60% working at the about 25 scientific institutes and departments, 30% in supply services and 10% in the administration. Past and present activities as well as equipment are reported.

The Role of Aerosol Behavior in Light Water Reactor Core Melt Accidents

A review of the processes important to the behavior of aerosols during a severe reactor accident involving core melting shows processes leading to particle size change (agglomeration, condensation, and evaporation) and processes leading to removal of particles from the atmosphere (diffusion, sedimentation, thermophoretic, and inertial deposition). The NAUA model and computer code developed at the Karlsruhe Nuclear Research Center treats these processes in a hypothetical core melt accident.The NAUA code is based on first principles, without further restrictions. Its application to such an accident in a pressurized water reactor (Biblis B) shows that the mass of aerosol leaked from a containment building during an accident is strongly dependent on the aerosol source from the core and the existing steam conditions. Condensing steam is effective in reducing leaked aerosol mass. Most of the leakage would occur during the first 12 h of an accident; such leakage is not directly proportional to the aerosol source strength but tails off significantly as the initial aerosol concentration increases.

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Experimental investigation of the aerosol-catalyzed oxidation of SO 2 under atmospheric conditions

One of the most important processes affecting the chemical and physical form of atmospheric sulfur dioxide is the interaction of gaseous SO 2 with solid or liquid aerosol particles. The investigation of the catalytic activity of some synthetic aerosols and dust particles of different industrial origin under real atmospheric conditions (i.e. a wide range of relative humidity and temperature) have been investigated in the course of an experimental research program at Karlsruhe Nuclear Research Center, Germany. The experiments were carried out in a 4.5 m 3 climate-controlled reaction chamber especially designed for SO 2-experiments in the ppm-concentration range. Starting with SO 2- concentrations of 3–5 mgm −3 the removal of SO 2 from gas phase in the presence of moist aerosol-free air and moist aerosol-polluted air was investigated as was the change of chemical composition of aerosols due to SO 2-adsorption and catalytic oxidation. The results of these experiments with synthetic MnSO 4 aerosols and fly-ash particles of a hard coal fired power plant are presented. They indicate a strong dependence of the catalytic oxidation rate on the relative humidity but no disappearance of the catalytic activity at low relative humidities.

25. Cryopumped hydrogen cluster beam source

Summary Cluster beams are of growing interest in plasma physics for heating and fueling fusion devices. At the Institute of Nuclear Engineering of the Karlsruhe Nuclear Research Centre an injector is under construction, which aims at the production of high power beams of electrically accelerated cluster ions of hydrogen. A hydrogen cluster is a condensed aggregate of molecules held together by the van der Waals forces. A cluster beam is obtained by rapid expansion of gaseous hydrogen from a cooled nozzle into vacuum. The outstanding properties of cluster beams are their high particle flux densities, an almost uniform particle velocity, a very sharp spatial limitation and a broad distribution of the particle size. An essential part for cluster beam generation is, besides the expansion nozzle and proper stagnation conditions, an adequate pumping system for the extreme gas load which occurs in cluster beam formation. The portion of cluster beam amounts to about 10% of the entire nozzle mass flux; therefore, 10 times of the cluster beam mass flow have to be disposed of. A cluster beam of 100 A equivalent current (= 10 torr l/s−1) calls for a pumping capacity of 100 torr 1/s−1 at a pressure of about 10−4. A cryosystem using liquid helium is described which provides the necessary pumping speed by cryopumps. The same cryostat yields also a cryogenic capacity for nozzle cooling and, in the case of the cluster injector mentioned above, the pumping speed required for the ionization regime and the accelerating tube. The production of intense cluster beams, cryopumping under extremely high gas load and typical properties of the cluster beams source with integrated cryopumps are discussed.

Plutonium in Gemüse, Obst und Böden

Plutonium in vegetables, fruits, and soilsThe plutonium content of fruits and vegetables from the Karlsruhe area in the years of 1971 ‐ 1975 was determined. In vegetables the Pu‐239 and Pu‐240 activities were highly different, fluctuating between 64 fCi/kg fresh weight and values below the detection limit of about 6 fCi/kg. The activity of most of the vegetables and of all fruits was not higher than 10 fCi/kg. The nuclide Pu‐238 could be identified in a qualitative way only, i. e. all measuring values remained below the detection limit of about 17 fCi/kg. The total‐alpha‐activities which were simultaneously determined in the vegetable and fruit ashes were in the range of 0.5 ‐19 pCi/kg fresh weight. Soil samples taken from the same region had a mean Pu‐239 and Pu‐240 activity of 5.9 ± 1.4 fCi/g and a total‐alpha‐activity of (10 ± 1.5) · 103 fCi/g. Samples taken from the immediate vicinity of the Karlsruhe Nuclear Research Centre did not show increased activity levels.

Experience in the analysis of atmospheric aerosols at the Karlsruhe Nuclear Research Center

In the Karlsruhe Nuclear Research Center over a 2-year period 24 monthly average concentration values of 45 elements in air dust aerosols have been determined. Significant correlations could be found between individual elements. A comparison with the measured values from other measurement stations located in Germany and abroad is given.

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High-Pressure Capsule for the Irradiation of Fuel Swelling Specimens in the FR2 Reactor

Designing advanced fuel element concepts for fast breeder reactors and assessing by models the fuel rod behavior over the intended in-pile time require information about creeping and swelling in the nuclear fuels under in-pile conditions. For this reason, a number of in-pile facilities that allow such material data to be determined have been used at the Karlsruhe Nuclear Research Center. These data depend not only on the properties of the fuel, above all the type of fuel and its density, but also, and in a very decisive way, on such parameters as specimen temperature, specimen loading, and burnup. A new series of experiments serves the concrete purpose of a quantitative assessment of the dependence on these parameters of carbides as potential fast breeder fuels, carbides being more susceptible to swelling than oxides. The loading of the specimen is given by the cladding restraint with the swelling fuel, and does not cause any undue expansion of the cladding of realistic fuel rods under operating conditions. This permissible contact pressure is generally below 300 bar (30 MPa).In an in-pile capsule recently developed, fuel specimens having the form of thin annular pellets may be exposed to omnilateral loads of up to 50 N/mm2 by an aerostatic gas pressure of 500 bar (50 MPa) maximum at temperatures of up to ~1800°C. The swelling of the fuel after various burnups under different test conditions will be measured in a postirradiation examination. Because of the extraordinary stress exerted upon these high-pressure capsules under irradiation conditions, extensive preliminary tests and trials were necessary. A first prototype rig with two swelling specimen capsules containing five annular U02 pellets each has now been successfully tested in the FR2 reactor over an in-pile period of 170 days. In this case, the fuel temperatures were ~1350°C, and the specimen stresses were 45 N/mm2. Further experiments are planned with (U, Pu)C-fuel up to 1800°C and 50 N/mm2.

西ドイツ・カールスルーエ原子力研究所における放射性物質による身体汚染者の処置法

The method practiced in Karlsruhe Nuclear Research Center in West Germany for the treatment of persons contaminated with radioactive substances is presented, classifying the contamination into three categories: 1) contamination of the surface of the body, 2) inhalation exposure and 3) wound contamination. The estimation of body burden of plutonium and the assessment of effectiveness of DTPA in removing incorporated plutonium after DTPA therapy performed in an accident which occurred in the above-mentioned Research Center are also described.