Helmuth Böck

Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz.

PURPOSE In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. METHODS The depletion calculation code origen2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition of the current core was used in a mcnp5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established attila model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. RESULTS The mcnp5 simulated neutron spectrum and source strength are found to be in good agreement with the previous attila model whereas the photon production is much lower. Both mcnp5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. CONCLUSIONS A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the mcnp5 simulations and experiments demonstrates that the attila model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural modifications in the thermal column irradiation channel or the use of different irradiation sites than the thermal column, e.g., the beam tubes.

Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10(9)cm(-2)s(-1) at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation.

Low-cost activation analysis at small research reactors

A software implementation of a loss-free counting multichannel analyzer, storing immediately into the multimegabyte memory of a low-cost 486 or Pentium type PC, enables the real-time control of a rabbit system as well as the collection of up to 1000 pairs of simultaneously recorded loss-corrected and non-corrected spectra of 16 k channels each, in a true sequence without time gaps in between, at throughput rates of up to 200 kc/s.1 Intended for activation analysis of short-lived isomeric transitions, the system renders possible peak to background optimizations and separations of lines with different half-lives without an a priori knowledge of sample composition by summing up appropriate numbers of spectra over appropriate intervals of time. By automatically adapting the noise filtering time to individual pulse intervals, the Preloaded Digital Filter (PLDF) combines low- to medium-rate resolutions comparable to those of high-quality Gaussian amplifiers with throughput rates of up to 100 kc/s, and high-rate resolutions superior to those of state-of-the-art gated integrator systems. In contrast to commercially available digital filters, the PLDF in its new implementation performs pulse shortening as well as pole zero cancellation in the analog domain. This not only results in a simpler digital core but also, for the first time, makes possible the use of a low-cost ADC in a spectrometric application.2 A simple but highly effective rabbit system is made from an aluminum incore part, inexpensive plastic tubing and an industrial pressurized air generator. Large sample containers have a volume of 25 cm3. Smaller containers of 5 cm3 are automatically separated from a transport capsule. A transport time of below 0.5 s enables activation analysis of short-lived isomeric transitions. The combination of rabbit system, PLDF and software based multichannel analyzer provides a low-cost but powerful solution for NAA at Triga reactors in developing countries, a forthcoming research project of the IAEA, Vienna.

The status and patterns of nuclear education in an anti-nuclear environment, Austria

This paper describes the current status and basic patterns of nuclear education in Austria and highlights the Nuclear Knowledge Management (NKM) activities of Atominstitute (ATI) through national and international networks. This institute hosts a TRIGA Mark II research reactor for research, training and education purposes. Austria has no nuclear power plant in operation because of the legislation prohibiting nuclear power production within its boundaries. Therefore, its nuclear education focuses on safety issues, i.e. environmental, health and engineering. This paper presents the increasing trend of students in nuclear subjects since 1995. The atomic institute is playing an active role in NKM through the European Nuclear Education Network (ENEN) association. Being the closest facility to the International Atomic Energy Agency (IAEA) headquarter, it also facilitates the international training and educational activities regarding nuclear science and technology.

Investigations on irradiated fission chamber electrodes

Abstract Analysis of some failed in-core fission chambers in a power reactor made a series of investigations on fission chamber electrodes necessary. Two types of electrodes with different uranium coating techniques were used. One type of electrodes was first exposed to maximum neutron doses of 2 × 10 20 nvt at steady-state operation. Both were exposed to several reactor power pulses at the TRIGA reactor Vienna. Changes in the uranium layer on the electrode surface have been observed during steady-state operation depending on the neutron dose. Complete destruction of the uranium layer was observed after pulse irradiation with one electrode type, while the other electrode type remained in good condition.

Nuclear education trends after Fukushima nuclear accident

Nuclear education has seen many ups and downs since the birth of nuclear technology. Owing to the highest prior support from all stakeholders during 1960s, nuclear education along with its technology witnessed its peak time. But during late 1990s, it was felt that the amount of nuclear knowledge would become sub-critical if no countermeasures were taken to arrest the declining situation of nuclear education. Many local, regional and international efforts were put together to restore nuclear education. These painstaking efforts started nuclear education growing again but the Fukushima nuclear disaster in 2011 seems to create serious obstacles in the restoration of nuclear education in many countries. This paper studies the influence of Fukushima nuclear accident on nuclear education. For this purpose, a worldwide survey has been conducted through a well-designed questionnaire asking maximum possible information about the trends of nuclear education. This survey covers Europe, Asia and America continents. Based on the results of this international survey, this research work highlights the impacts of Fukushima nuclear accident on nuclear education in European, Asian and North American countries separately.

Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation

Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr95. This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes.