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[en] Curium isotopes are generated hitherto as a waste product in nuclear energy production. Exposure in humans has occured mainly via inhalation. After reprocessing of spent reactor fuel these nuclides represent the highest alpha activity during the first 60 years. Therefore it is necessary to study the resulting radiation exposure in man after a accidental contamination. Lung tissue sections were analysed for histological characteristics by means of adaptive pattern recognition methods, using an electronic image analyzer. Alpha particle tracks were superimposed and interaction with cellular structures was simulated. Cell frequency distribution, along with specific hit-probability is used to assess quantitatively the resulting energy deposition in the single cells
[en] The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature gradients within the fuel assemblies would be too high, and fuel economy is poor. Two improved fuel concepts are proposed: (1) a redesign of the classic TRISO coated particle fuel, and (2) an innovative hollow sphere design. Both fuel elements are used in a core design based on direct cooling of the coated particle fuel. To increase the neutronic margins and obtain adequate self-breeding capabilities, the proposed reactor has 2400 MWth power output and a power density of 50 MW/m3. With both types of fuel, it is possible to obtain a closed fuel cycle. Long irradiation intervals (several years) are possible with a low burnup reactivity swing, which reduces the required over-reactivity of the fresh core and reduces control rod requirements during operation. In the closed fuel cycle it is important to be able to predict whether a certain initial fuel composition will in fact yield a new fuel, after irradiation, cool down and reprocessing, with which the reactor can be restarted. A theoretical framework is presented in this thesis which allows calculation of the ‘Breeding Gain’ (BG) of the reactor. The BG quantifies the performance of the fuel for batch i + 1 as a function of the composition of the initial fuel of batch i. If this BG can be made equal to zero, both fuel compositions give the same nuclear performance. To be able to calculate the fuel performance, the reactivity weight, i.e. the contribution of each isotope to the overall reactivity of the reactor, needs to be estimated. It is proposed in this thesis to calculate these reactivity weights using a first-order eigenvalue perturbation calculation. It is shown that this approach yields an expression which reduces to a well-established formula for reactivity weights. All steps in the fuel cycle, i.e. irradiation, cool down and reprocessing, have to be taken into account to calculate the Breeding Gain for the closed fuel cycle. First order nuclide perturbation theory provides an efficient method to calculate the effects of small variations of the initial fuel composition on the performance of the closed fuel cycle. The theory is applied to the closed fuel cycle of a 600MWth Gas Cooled Fast Reactor. The result is that the closed fuel cycle can be obtained if the reprocessing is efficient enough in retrieving the transuranics from the irradiated fuel (> 99%). Calculations were done adding extra MA to the GCFR fuel, to estimate the transmutation potential of the GCFR concept. Extra MA in the fuel improve the Breeding Gain, and reduce the burnup reactivity swing. The GCFR core power density is high in comparison to other gas cooled reactor concepts. Like all nuclear reactors, the GCFR produces decay heat after shut down, which has to be transported out of the reactor under all circumstances. The layout of the primary system therefore focuses on using natural convection Decay Heat Removal (DHR) where possible, with a large coolant fraction in the core to reduce friction losses. However, due to the combination of high power density and low thermal inertia in the core, transients in the GCFR core may lead to high temperatures. To protect the reactor under all circumstances during transients, passive reactivity control devices are researched. These devices control the reactor power under off-nominal conditions when all other control devices fail. The proposed devices use liquid 6Li as an absorber, which is passively introduced into the core. Activation of the device is by freeze seals, which melt when the core outlet temperature is too high. These devices can be integrated into the normal control assemblies of the reactor while still keeping enough room available for the regular control elements. The passive devices are shown to adequately limit the power production of the GCFR core. It is also shown that natural circulation DHR is possible under pressurized core conditions.
[en] These volumes present the results of a study undertaken for the Commission of the European Communities. The aim was to review available data concerning the movement of radionuclides through the environment and to recommend values of parameters for use in environmental transport models. The elements reviewed all have radioactive isotopes which could contribute significantly to the radiological impact of chronic releases of radioactivity from nuclear installations within the countries of the European community, i.e. the major activation and fission products. In dividing these elements between volumes an effort has been made to take account of the method of production of their major radioisotopes, together with their chemical similarities and environmental interactions. This volume covers the radionuclide distribution of americium and curium. The main areas which are covered include the deposition of radionuclides on plants and soils, transport in soils, uptake and translocation in plants via the roots and foliage, metabolism in domestic animals and radionuclide transfers through the main physical and biotic components of the aquatic environment. In reviewing these subject areas, account has been taken not only of the literature relating to specific radionuclides, but also of the literature relating to the stable element of which they are radioisotopes. (Auth.)
[en] Prediction error was evaluated for decay heat and nuclide generation in spent mixed oxide (MOX) fuels on the basis of error files in JENDL-3.3. This computational analysis was performed using SWAT code system, ORIGEN2 code, and ERRORJ code. The results of nuclide generation error evaluation were compared with some discrepancies in the calculated values to experimental values (C/E ratio) which were already published and were obtained by analysis of post irradiated experiments (PIE) data. Though the discrepancies of some C/E values, especially those of americium and curium isotopes, ranged from a half to twice, the present error evaluation based on the error file of nuclide generation became 10% or less. We conclude that the discrepancy between calculation and the PIE data is almost factor 5 larger than that evaluated from the covariance data in JENDL-3.3. Therefore the practical error value of total decay heat should be 20% or more on 1 σ basis. (authors)
[en] Various research collaborations with foreign organizations such as IAEA, OECD, ORNL, BNL, etc. have been performed and strengthened. By actively participating in meetings that are sponsored by IAEA and OECD, we could make a decision on the research directions for nuclear data and offer our nuclear data to the international community without difficulty. The evaluation and validation of Np-237, Pu-240, and Cm isotopes have been carried out in collaboration with ORNL and the evaluation of Fe-56 were performed in collaboration with BNL. The resonance region uncertainty analysis code has been improved and the uncertainty analyses for the structural materials and Np-237 have been carried out. The revision of the evaluated library for minor actinides has been carried out and the evaluated files for 11 Curium isotopes have been adopted into new JEFF-3.2 Library released in March 2014. New evaluation of the resolved and unresolved resonance region for U-238 capture cross sections has been carried out with EC-JRC-IRMM at Belgium. The evaluation methodology for the structural materials is being developed and the new evaluated files for future nuclear fusion reactors will be produced. The electron-impact ionization (EII) cross sections for Wq+ and the electron-impact recombination cross section for W17+ have also been calculated. The validation system of nuclear reaction/covariance data for Np-237, Pu-240 and Cm-isotopes was established, comparative analysis of recent covariance data and validation of the data through nuclear data sensitivity/uncertainty analysis for some criticality benchmark problems was performed, and contribution to validation of COMMARA-2.0 covariance data through some fast reactor benchmark analyses under International Nuclear Data Evaluation Co-operation managed by OECD/NEA was made. Research for building time-of-flight facility has been performed. The shielding analysis and architectural design for the building have been performed, and the heat and structural analysis and manufacture of main part for the liquid lead target have been done. The nuclear reaction cross sections were measured at various domestic and overseas facilities including KIRAMS, KOMAC, Pohang electron accelerator, HZDR, IRMM and IMAC.
[en] A method based on the statistical Hauser-Feshbach theory for the calculation of fission cross sections was described. The main values, the level densities and the transmission coefficients in the different reaction channels are discussed in detail. As an example, the results of Cm(n,f) reaction calculations are presented. (author). 25 refs, 8 figs, 2 tabs
[en] Using the developed analytical method for Pu, a technician is able to perform 4 analyses per day. Chemical yield: 70-80%, Lower detection limit: < 0,1 fCi/g. The monthly determined releases of plutonium in the gaseous effluents of Karlsruhe reprocessing plant range over several orders of magnitude. The released Pu-238 and Pu-239+240 remains below the releases of gross alpha activity. The plutonium concentrations in liquid effluents range in the average between 0,1-1 pCi Pu/1. Concentrations of Pu were measured in: soil, plants, sediments, air, animals and water. (orig./RW)
[de]Mit der Analysenmethode fuer Pu koennen von einem Laboranten 4 Analysen pro Tag durchgefuehrt werden. Chemische Ausbeute: 70-80%. Nachweisgrenze: < 0,1 fCi/g. Die monatlich gemessenen Pu-Emissionen mit der Abluft der WAK schwanken ueber mehrere Groessenordnungen. Die Gesamtemission von Pu-238 und Pu-239+240 liegt wesentlich unter der Gesamtalphaemission. Im Abwasser werden im Durchschnitt 0,1-1 pCi Pu gefunden, Pu-Konzentrationen wurden gemessen in Boden-, Pflanzen-, Sediment-, Luft-, Tier- und Wasserproben. (orig./RW)
[en] A new measurement program was set up at SCK.CEN to determine the thermal neutron-induced fission cross section of a number of Cm isotopes. These transuranium isotopes are produced in nuclear reactors and are candidates for transmutation. This paper presents preliminary results of our 245Cm(n, f) cross-section measurement. (authors)