Anmerkungen:
Quelle: Forschungszentrum Rossendorf; FZR-87 Juni 1995
Beschreibung:
In this report is described the theoretical and experimental determination of all needed parameters for the neutron exposure of the different specimens that were irradiated in the Rheinsberg reactor within the period from 1984 until 1988 to obtain data for neutron embrittlement studies. The methodical approach is represented, possible sources of errors are discussed and all the needed results are collected. The work can be divided into a pure calculation part, into the description of the extraction of the monitors and their gamma spectrometric analysis and into the discussion of the spectrum adjustment procedure which combines experimental and theoretical results. Therefore, this representation deals not only with this special task but gives also a survey about the methodical approach and the state of art in Rossendorf for the general problem of determination of neutron fluences, because the developed methods are generally applicable and not limited to the special problem of neutron embrittlement. Different problems with respect of neutron fluences exist also in Germany. The calculation of fluences is based on an accurate description of the history of the reactor for all irradiation periods, i.e. the time and space dependence of burnup, power and fisson sources. This information was given for a time and space grid of all fuel elements. For the calculation of the fluences were not used time dependent fluxes, but at first integral sources of fission neutrons are calculated for different fissionable isotopes. Using these given integral source distributions the fluences were calculated and the results from the different fissionable source were composed. For the improvement of the comparison between experimental and theoretical results a special method was used calculating special integral sources for the given detector, because each detector according to its decay notes another integral source resp. flux distribution, which has to be theoretically corrected. The basis of all transport calculations was the Monte Carlo method in a special problem adjusted kind. Special procedures were developed and successfully applicated for the reduction of statistical errors. Therefore, also for single specimens results with small statistical errors were obtained. It is remarkable that this accurate method which allows a realistic 3-dimensional description of the system reactor-surroundings could be applied with reasonable calculation times for the great number of needed calculations.