Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Dose rate estimation of gamma-rays emitted from a LEU miniature plate

Document Type : Research Paper

Authors

S.M. Miremad 1
A.B. Samani 1
S.S. Sayyahi 2

1 Nuclear Fuel Cycle Research School, Nuclear Science and Technology Research Institute, AEOI, P.O.Box: 11365-3486, Tehran - Iran

2 Pars Isotope Company, P.O.Box: 1439955416, Tehran-Iran

https://doi.org/10.24200/nst.2022.1165.1764
Abstract
The LEU miniature plate irradiated in the Tehran Research Reactor, which produces the radioisotope molybdenum-99, is a radioactive source consisting of various radioisotopes. Before any practical application, it is necessary that the dose rate of gamma rays emitted from this source be calculated. This was done for comparison with the recommended dose rates for radioactive materials transport and radiation protection of employees during hot tests. In this paper, the validity of a proposed method for calculating the dose rate of gamma rays emitted from hot targets was investigated, experimentally. The Monte Carlo code MCNPX and multi-step algorithm written in MATLAB were used in the proposed method. The results showed that the calculated dose rate was always lower than the measured dose rate. Therefore, based on a conservative view, it is better to multiply the calculated dose rate values by 2 and then compare them with the allowable limits. This will enable you to determine the appropriate shield thickness.

Highlights

  1. IAEA, Non-HEU Production Technologies for Molybdenum-99 and Technetium-99m, Technical Report No. NF-T-5.4, Vienna, (2013).

 

  1. E. National Academies of Sciences and Medicine, Opportunities and Approaches for Supplying Molybdenum-99 and Associated Medical Isotopes to Global Markets: Proceedings of a Symposium, Washington, DC: The National Academies Press (in English), 86 (2018).

 

  1. S.E. Hosseini, et al, Evaluation of promethium-147 production as a by-product of the fission molybdenum-99 process in Tehran research reactor, Radiochimica Acta, 109(4), 295-300 (2021).

 

  1. S.K. Lee, et al, Development of fission 99Mo production process using HANARO, Nuclear Engineering and Technology, 52(7), 1517-1523 (2020).

 

  1. M.E. de Melo Rego, et al, Study on Shielding Requirements for Radioactive Waste Transportation in a Mo-99 Production Plant–13382, WM2013 Conference, (Phoenix, Arizona USA, 2013), 1-7 (2013).

 

  1. R.G. Abrefah, P.A.A. Essel, H.C. Odoi, Estimation of the dose rate of nuclear fuel of Ghana Research Reactor-1 (GHARR-1) using ORIGEN-S and MCNP 6, Progress in Nuclear Energy, 105, 309-317 (2018).

 

  1. Z. Gholamzadeh, M. Gholshanian, S.M. Mirvakili, ThO2 spent fuel assembly’s gamma dose rate dependency to burnup and cooling time, Radiation Physics and Engineering, 1(3), 43-48 (2020).

 

  1. D.B. Pelowitz, et al., MCNPX 2.7.E Extensions, (2011).

 

  1. Nuclear Energy Agency, JANIS 4.1, https://www.oecd-nea.org/jcms/pl_39910/janis.

 

  1. American National Standards Institute, ANSI/ANS-6.1.1.https://www.webstore.ansi.org/Standards/ANSI/ANSIANS2020.

 

  1. IAEA, Regulations for the Safe Transport of Radioactive Material 2018 Edition Specific Safety Requirements, No. SSR-6(Rev. 1), (2018).

 

  1. N. Kržanović, et al, Development and testing of a low cost radiation protection instrument based on an energy compensated Geiger-Müller tube, Radiation Physics and Chemistry, 164, 108358 (2019).

 

  1. E. Abedi, et al, Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets, Applied Radiation and Isotopes, 118, 160-166 (2016).

 

  1. H. Al Kanti, et al, Conversion coefficients calculation of mono-energetic photons from air-kerma using Monte Carlo and analytical methods, Journal of King Saud University-Science, 32(1), 288-293 (2020).

 

  1. R. Casanovas, E. Prieto, M. Salvadó, Calculation of the ambient dose equivalent H*(10) from gamma-ray spectra obtained with scintillation detectors, Applied Radiation and Isotopes, 118, 154-159 (2016).

Keywords

Dosimetry
LEU target
Transportation container
Shielding
MCNPX
  1. IAEA, Non-HEU Production Technologies for Molybdenum-99 and Technetium-99m, Technical Report No. NF-T-5.4, Vienna, (2013).

 

  1. E. National Academies of Sciences and Medicine, Opportunities and Approaches for Supplying Molybdenum-99 and Associated Medical Isotopes to Global Markets: Proceedings of a Symposium, Washington, DC: The National Academies Press (in English), 86 (2018).

 

  1. S.E. Hosseini, et al, Evaluation of promethium-147 production as a by-product of the fission molybdenum-99 process in Tehran research reactor, Radiochimica Acta, 109(4), 295-300 (2021).

 

  1. S.K. Lee, et al, Development of fission 99Mo production process using HANARO, Nuclear Engineering and Technology, 52(7), 1517-1523 (2020).

 

  1. M.E. de Melo Rego, et al, Study on Shielding Requirements for Radioactive Waste Transportation in a Mo-99 Production Plant–13382, WM2013 Conference, (Phoenix, Arizona USA, 2013), 1-7 (2013).

 

  1. R.G. Abrefah, P.A.A. Essel, H.C. Odoi, Estimation of the dose rate of nuclear fuel of Ghana Research Reactor-1 (GHARR-1) using ORIGEN-S and MCNP 6, Progress in Nuclear Energy, 105, 309-317 (2018).

 

  1. Z. Gholamzadeh, M. Gholshanian, S.M. Mirvakili, ThO2 spent fuel assembly’s gamma dose rate dependency to burnup and cooling time, Radiation Physics and Engineering, 1(3), 43-48 (2020).

 

  1. D.B. Pelowitz, et al., MCNPX 2.7.E Extensions, (2011).

 

  1. Nuclear Energy Agency, JANIS 4.1, https://www.oecd-nea.org/jcms/pl_39910/janis.

 

  1. American National Standards Institute, ANSI/ANS-6.1.1.https://www.webstore.ansi.org/Standards/ANSI/ANSIANS2020.

 

  1. IAEA, Regulations for the Safe Transport of Radioactive Material 2018 Edition Specific Safety Requirements, No. SSR-6(Rev. 1), (2018).

 

  1. N. Kržanović, et al, Development and testing of a low cost radiation protection instrument based on an energy compensated Geiger-Müller tube, Radiation Physics and Chemistry, 164, 108358 (2019).

 

  1. E. Abedi, et al, Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets, Applied Radiation and Isotopes, 118, 160-166 (2016).

 

  1. H. Al Kanti, et al, Conversion coefficients calculation of mono-energetic photons from air-kerma using Monte Carlo and analytical methods, Journal of King Saud University-Science, 32(1), 288-293 (2020).

 

  1. R. Casanovas, E. Prieto, M. Salvadó, Calculation of the ambient dose equivalent H*(10) from gamma-ray spectra obtained with scintillation detectors, Applied Radiation and Isotopes, 118, 154-159 (2016).

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