A calibration method for realistic neutron dosimetry in radiobiological experiments assisted by MCNP simulation

Journal of Radiation Research, Vol. 57, No. 5, 2016, pp. 492–498 doi: 10.1093/jrr/rrw063 Advance Access Publication: 5 July 2016 A calibration method for realistic neutron dosimetry in radiobiological experiments assisted by MCNP simulation Mehrdad Shahmohammadi Beni1, Dragana Krstic2, Dragoslav Nikezic1,2 and Kwan Ngok Yu1,3* 1 Department of Physics and Materials Science, City University of Hong Kong, Tat Chee Avenue, Kowloon Tong, Hong Kong 2 Faculty of Science, University of Kragujevac, Serbia 3 State Key Laboratory in Marine Pollution, City University of Hong Kong, Tat Chee Avenue, Kowloon Tong, Hong Kong *Corresponding author. Department of Physics and Materials Science, City University of Hong Kong, Tat Chee Avenue, Kowloon Tong, Hong Kong. Tel: +852-3442-7812; Fax: +852-3442-0538; Email: Received January 28, 2016; Revised March 23, 2016; Accepted May 9, 2016 ABSTRACT Many studies on biological effects of neutrons involve dose responses of neutrons, which rely on accurately determined absorbed doses in the irradiated cells or living organisms. Absorbed doses are difficult to measure, and are commonly surrogated with doses measured using separate detectors. The present work describes the determination of doses absorbed in the cell layer underneath a medium column (DA) and the doses absorbed in an ionization chamber (DE) from neutrons through computer simulations using the MCNP-5 code, and the subsequent determination of the conversion coefficients R (= DA/DE). It was found that R in general decreased with increase in the medium thickness, which was due to elastic and inelastic scattering. For 2-MeV neutrons, conspicuous bulges in R values were observed at medium thicknesses of about 500, 1500, 2500 and 4000 μm, and these were attributed to carbon, oxygen and nitrogen nuclei, and were reflections of spikes in neutron interaction cross sections with these nuclei. For 0.1-MeV neutrons, no conspicuous bulges in R were observed (except one at ~2000 μm that was due to photon interactions), which was explained by the absence of prominent spikes in the interaction cross-sections with these nuclei for neutron energies <0.1 MeV. The ratio R could be increased by ~50% for small medium thickness if the incident neutron energy was reduced from 2 MeV to 0.1 MeV. As such, the absorbed doses in cells (DA) would vary with the incident neutron energies, even when the absorbed doses shown on the detector were the same. KEYWORDS: neutrons, Monte Carlo, MCNP, radiation dosimetry I N T RO D U C T I O N Biological effects of neutrons are relatively less studied and less well understood compared with other types of ionizing radiations such as high-energy photons and heavy ions. Neutron-induced bystander effects (NIBEs) were in general not demonstrated in early in vitro or in vivo studies [1–3]. Only recently, NIBEs were demonstrated in zebrafish embryos [4]. Similarly, results on the neutron-induced radioadaptive response (RAR) have been equivocal. Wiencke et al. [5] and Ng et al. [6] demonstrated that neutrons failed to induce a RAR in human lymphocytes and zebrafish embryos, respectively. In contrast, Marples and Shov [7] revealed a neutron-induced RAR in Chinese hamster V79 cells. Interestingly, Gajendiran et al. [8] examined whole blood samples collected from 10 people, but detected a neutron-induced RAR in the samples from only one donor. Although the discrepancies between some of these results were explained in terms of mitigation of neutron-induced damages by the γ rays that were emitted together with the neutrons from the neutron sources [4, 6, 9–11], these might also have arisen because of the obscurity in the definition of the absorbed neutron dose. Many studies on the biological effects of neutrons involve neutron dose responses, which can only be established with accurately determined absorbed doses in the irradiated cells or living organisms. Unfortunately, it is practically difficult to directly measure the absorbed doses in cells or in living organisms, and as such these are © The Author 2016. Published by Oxford University Press on behalf of The Japan Radiation Research Society and Japanese Society for Radiation Oncology. This is an Open Access article distributed under the terms of the Creative Commons Attribution Non-Commercial License (http://creativecommons.org/licenses/ by-nc/4.0/), which permits non-commercial re-use, distribution, and reproduction in any medium, provided the original work is properly cited. For commercial re-use, please contact • 492 Calibration of neutron dosimetry by MCNP simulation commonly surrogated with the doses measured using some separate radiation detectors such as an ionization chamber. However, it might not be straightforward to ascertain the ‘conversion coefficients’ between the neutron doses (DE) recorded by such radiation detectors placed in the ambient environment and the neutron doses (DA) actually absorbed in the exposed cells or living organisms, since both DE and DA critically depend on the dimensions, geometries and densities of these various exposed targets, the materials surrounding the exposed targets, and the energy of the incident neutrons. The present study used in vitro experiments with cells as an example to demonstrate how the conversion coefficients R (= DA/DE) could be determined through computer simulation using the MCNP (Monte Carlo N-Particle) code [12]. In fact, the task was similar to the development of concepts in the field of radiation protection. Traditionally, for radiation protection purposes, three categories of ‘quantities’ have been defined, namely (i) ‘physical quantities’ such as air kerma for photons and absorbed dose for β particles; (ii) ‘protection quantities’ (or ‘primary limiting dose quantities’) such as organ absorbed dose, organ equivalent dose, and effective dose; and (iii) ‘operational quantities’ such as the ambient dose equivalent H*(d), the directional dose equivalent H’(d,Ω) and the personal dose equivalent Hp(d) [13] defined using the ICRU sphere phantom with a diameter of 30 cm built with a tissue-equivalent material (density = 1 g cm−3; mass composition: 76.2% oxygen, 11.1% carbon, 10.1% hydrogen and 2.6% nitrogen). While the protection quantities were defined to characterize the risk of exposures to ionizing radiations, these were in general not measurable. As such, operational quantities were required that characterized the external exposures, either to an area or to an individual. Operational quantities could be calculated from physical quantities using the ICRU sphere phantom, while protection quantities could also be calculated from physical quantities using anthropomorphic phantoms together with the radiation weighting factors WR and the tissue weighting factors WT. As such, measurements on operational quantities could provide information • 493 on the protection quantities. The conversion coefficients for these quantities have been published by ICRP [14, 15]. The similarities b (...truncated)