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Concept of a BNCT line with in-pool fission converter at MARIA reactor in Swierk

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Pytel K., Andrzejewski K., Golnik N., Osko J.
Polish Journal of Medical Physics And Engineering
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2009
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vol. 15
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issue 4
209-214
EN
BNCT facility in the Institute of Atomic Energy in Otwock-Swierk is under construction at the horizontal channel H2 of the research reactor MARIA. Measurements of the neutron energy spectrum performed at the front of the H2 experimental channel, have shown that flux of epithermal neutrons (above 10 keV) at the BNCT irradiation port was below 109 n cm-2 s-1 i.e. it was too low to be directly used for the BNCT treatment. Therefore, a fission converter will be placed between the reactor core and the periphery of the graphite reflector of MARIA reactor. The uranium converter will be powered by the densely packed EK-10 fuel elements with 10% enrichment. Preliminary calculations have shown that the total neutron flux in the converter will be about 1013 n cm-2 s-1 and flux of epithermal neutrons at the entrance to the filter/moderator of the beam will be about 2·1013 n cm-2 s-1.
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Production of Fission Product99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

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Jaroszewicz J., Marcinkowska Z., Pytel K.
Nukleonika
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2014
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vol. 59
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issue 2
43-52
EN
The main objective of 235U irradiation is to obtain the 99mTc isotope, which is widely used in the domain of medical diagnostics. The decisive factor determining its availability, despite its short lifetime, is a reaction of radioactive decay of 99Mo into 99mTc. One of the possible sources of molybdenum can be achieved in course of the 235U fission reaction. The paper presents activities and the calculation results obtained upon the feasibility study on irradiation of 235U targets for production of 99Mo in the MARIA research reactor. Neutronic calculations and analyses were performed to estimate the fission products activity for uranium plates irradiated in the reactor. Results of dummy targets irradiation as well as irradiation uranium plates have been presented. The new technology obtaining 99Mo is based on irradiation of high-enriched uranium plates in standard reactor fuel channel and calculation of the current fission power generation. Measurements of temperatures and the coolant flow in the molybdenum installation carried out in reactor SAREMA system give online information about the current fission power generated in uranium targets. The corrective factors were taken into account as the heat generation from gamma radiation from neighbouring fuel elements as well as heat exchange between channels and the reactor pool. The factors were determined by calibration measurements conducted with aluminium mock-up of uranium plates. Calculations of fuel channel by means of REBUS code with fine mesh structure and libraries calculated by means of WIMS-ANL code were performed.
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