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Nukleonika

The Journal of Instytut Chemii i Techniki Jadrowej

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0029-5922
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Production of Fission Product 99Mo using High-Enriched Uranium Plates in Polish Nuclear Research Reactor MARIA: Technology and Neutronic Analysis

Janusz Jaroszewicz
  • Corresponding author
  • Nuclear Energy Department, National Centre for Nuclear Research, 7 Andrzeja Sołtana Str., 05-400 Otwock/Świerk, Poland, Tel: +48 22 718 0077
  • Email
  • Other articles by this author:
  • De Gruyter OnlineGoogle Scholar
/ Zuzanna Marcinkowska
  • Nuclear Energy Department, National Centre for Nuclear Research, 7 Andrzeja Sołtana Str., 05-400 Otwock/Świerk, Poland, Tel: +48 22 718 0077
  • Other articles by this author:
  • De Gruyter OnlineGoogle Scholar
/ Krzysztof Pytel
  • Nuclear Energy Department, National Centre for Nuclear Research, 7 Andrzeja Sołtana Str., 05-400 Otwock/Świerk, Poland, Tel: +48 22 718 0077
  • Other articles by this author:
  • De Gruyter OnlineGoogle Scholar
Published Online: 2014-07-08 | DOI: https://doi.org/10.2478/nuka-2014-0009

Abstract

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.

Keywords: fission products; 99Mo production; neutronic calculations; research reactor

References

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About the article

Received: 2013-10-03

Accepted: 2014-04-28

Published Online: 2014-07-08


Citation Information: Nukleonika, Volume 59, Issue 2, Pages 43–52, ISSN (Online) 0029-5922, DOI: https://doi.org/10.2478/nuka-2014-0009.

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© 2014 Janusz Jaroszewicz et. al.. This work is licensed under the Creative Commons Attribution-NonCommercial-NoDerivatives 3.0 License. BY-NC-ND 3.0

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