OPTIMIZATION STUDY OF A TRANSPORTABLE NEUTRON RADIOGRAPHY SYSTEM BASED ON A 252 CF NEUTRON SOURCE. (Received 30 December 2010) Abstract

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1 OPTIMIZATION STUDY OF A TRANSPORTABLE NEUTRON RADIOGRAPHY SYSTEM BASED ON A 252 CF NEUTRON SOURCE J.G. Fantidis 1,*, C. Potolias 1, N. Vordos 1, and D.V. Bandekas 1 1 Dept. of Electrical Engineering, Kavala Institute of Technology, Greece (Received 30 December 2010) Abstract The purpose of this work is the optimization of a transportable thermal neutron radiography system. Neutrons are produced by a 50 mg 252 Cf source. The design was optimized with respect to parameters related with thermal neutron radiography and shielding. Owning to the special collimator design, it was possible to optimize the neutron radiography parameters, while the use of advanced shielding materials reduced the weight and the volume of the unit. The unit was compatible with the European Union Directive on Restriction of Hazardous Substances (RoHS) 2002/95/EC and imulated using the MCNPX code. Keywords: Thermal neutron radiography, MCNP, RoHS Directive 1. Introduction The 252 Cf isotopic neutron source is one of the most important neutron sources. The 252 Cf radioisotope is commonly applied to PGNAA of samples [1] providing a rapid non-destructive elemental evaluate of the principal components of a sample. Other commercial applications include gamma spectroscopy via instrumental neutron activation analysis (INAA) for trace elemental analysis [2], characterization of nuclear materials (fissile material [3], passive-active neutron shufflers [4]) and neutron radiography [5]. Neutron radiography (NR) has been established as a testing technique and as a research tool for over 60 years. The technique is widely used in security applications, engineering studies and industry in order to determine structural defects, geology, medicine and biological research [6, 7]. Of the isotopic sources, 252 Cf is by the far the best for portable NR systems [8-10]. The major objective of this work consists in optimizing the design of the system in terms of NR parameters (such as thermal neutron flux and the number of thermal neutrons within the neutron beam) and shielding. The proposed system is designed according to article 4 of the RoHS Directive 2002/95/EC, regarding the choice of materials. Hence, lead, mercury, cadmium, hexavalent chromium, polybrominated biphenyls, and polybrominated diphenyl ethers have been excluded [11, 12]. 2. Materials and methods 2.1 The neutron source The radioisotope 252 Cf is an intense neutron emitter that is readily encapsulated in compact portable sealed sources. Decay by alpha emission (96.91% probability) and spontaneous fission (3.09% probability) results in an overall half-life of years and neutron emission of with a specific activity of mci/μg. The Maxwellian neutron energy distribution has an average energy of 2.14 MeV and most probable energy of 0.7 MeV [13]. The spectrum of the emitted neutrons extends up to 10 MeV and modeled as a Watt fission spectrum using the

2 Moldavian Journal of the Physical Sciences, Vol. 10, N1, 2011 coefficients provided by the MCNPΧ code [14]. Further to the neutron emission, 252 Cf emits photons s -1 per μg with a mean energy of 0.8 MeV [15] The radiography unit The radiography unit is designed in the form of a sphere with a variable radius (Fig. 1). The 252 Cf neutron source (1) is placed at the centre of the sphere. The first 50 cm of the radius made of high density-polyethylene (HDPE) (2). The fast neutrons from the source are thermalized from HD-PE. The collimator inlet aperture (3) was positioned at 1.5 cm from the source, which is the position of peak thermal flux produced in the HD-PE [16]. Combinations of layers of different materials (4) surround the HD-PE moderator in order to provide the necessary shielding for radiation protection purposes. While the system is in use, the conic shielding part (5) is detached to permit placing the object analyzed in front of the conic collimator. Fig. 1. Side view of the geometric configuration of the irradiating system. The collimator ratio (L/D), which determines the quality of the NR imaging for a given design of the collimator, is given by the following equations: and i (1) 2 L 16 S D u g D, (2) Lf L s where L f is the image surface to object distance, L s is the source to object distance, D is the inlet aperture diameter, φ i is the neutron flux at the image plane, φ a is the neutron flux at the aperture, and u g is the geometric unsharpness. 122

3 J. G. Fantidis, C. Potolias, N. Vordos, and D. V. Bandekas The beam divergence, a significant measure of the usefulness of the beam near its periphery, is described by its half-angle (θ) given by [17] 1 I tan, (3) 2L where I and L are the maximum dimension of the image plane and the length of the collimator. The imaging quality of a system would be additionally characterized by the Thermal Neutron Content (TNC), describing the number of thermal neutrons within the neutron beam thermal neutron flux TNC (4) total neutron flux and the relative intensities of the neutron (n) and the photon (γ) components of the beam (n/γ). Hawkesworth [18] has shown that the least recommended value is about 500 cm -2 msv -1 and typically greater than 10 4 n cm -2 msv -1. Figure 2 shows the proposed collimator system for thermal neutron radiography. The collimator being modeled is made of two parts. The first, which is a HD PE cylinder (2), with a radius of 8 cm and a length of 11 cm, incorporating a conic collimator made either of single sapphire or void. The conic collimator has a length 11 cm and radii of 4 and 2 cm, with the larger radius near to the source. The single sapphire (Al 2 O 3 ) is the best filter for fast neutron filtration [19, 20]. Next to the HD PE cylinder, a divergent collimator is situated, which determines to a great extent the quality of the image for a given source type. A 0.5 mm-thick layer of gadolinium is the lining of the collimator covered by PE B with a depth of 3 cm depth as a shielding against scattered neutrons. Bismuth (Bi) with 1 cm of thickness was chosen as the collimator casing preventing stray photons from arriving at the object. Fig. 2. The collimator geometry for thermal neutron radiography with the sapphire filter in front of the collimator (not in scale). The essential shielding for radiation protection purposes was optimized using the MCNPX code. In order to provide effective shielding against neutrons and stop the gamma rays from the source and from neutron activation, a range of materials were considered (Table 1), which would provide effective shielding while still rendering the unit transportable. The Dose Equivalent Rates (DER) was calculated on the external surface of the sphere. 123

4 Moldavian Journal of the Physical Sciences, Vol. 10, N1, 2011 Table 1. Shielding compositions in mass fraction Element Shielding materials Borated Polyethylene Zirconium borohydride H C O Magnesium borohydride Bismuth Kennertium Bi B Zr Mg Ni 0.09 Cu 0.15 W 0.76 Density (g cm -3 ) Results and discussion The shielding of the unit was designed for a 50-mg 252 Cf source, taking into consideration the RoHS directive, the weight for transportability of the unit and the occupational dose limit of 0.5 Sv (or 25 μsv h -1 ) at the external surface of the sphere. Dose estimates obtained for combinations of different layers of the materials shown in Table 1 are given in Table 2. The total Dose Equivalent Rate (DER) was calculated with the MCNPX Monte Carlo code, using the F2, Fm2 tally and the DE and DF cards. The tally describes the neutron flux in a surface, while the two cards convert the absorbed dose to equivalent dose. The total DER would contain three components, notably from the neutrons (DER1), the gamma-rays emitted by the source (DER2) and induced gamma-rays from the interaction of the neutrons and the moderator material (DER3) [10]. 124

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6 Moldavian Journal of the Physical Sciences, Vol. 10, N1, 2011 Calculations were carried out for up to 10 8 histories per starting particle (NPS) yielding an accuracy in the calculations of < 2%. Although PE B and Bi is attractive for use in shielding design, advanced materials, such as a Kennertium (a machinable Tungsten), Zr(BH 4 ) 4 or Mg(BH 4 ) 2, according to Table 2, show superior shielding capability than the conventional materials. In keeping with the results, the overall weight and volume have improved by factors of 1.75 and 1.42, respectively. According to the Da Silva et al. [10], with the same L/D ratio the minimum values of L (and D) give higher (thermal) neutron flux in the detector position, however, simultaneously, the TNC parameter takes on smaller values. The presence of the first part of the collimator could offer better results both to the f th and the TNC parameters [12]. In order to maximize the f th to the investigation object, a series of calculations were performed. Firstly, we calculated the length of the first part of the collimator in order to maximize the f th at the image plate in the case of L/D = 25 (L = 100 cm and D = 4). In accordance with the results (Fig. 3), the maximum f th occurs for a length of 11 cm. Fig. 3. The thermal neutron flux at the plate in the case of L/D = 25 versus the length of the first part of the collimator. The suggested unit comprises a collimator (L) with a length of 00 cm, a diameter of its aperture next to the image plane (D 0 ) of 16 cm, and a divergence angle of the beam (θ) of 4.57, while the inlet aperture (D) of the collimator has variable values (D = 1 4 cm). The distance between the object and the imaging detector (L f ) was considered at 0.5 cm for better results [17]. The variation of the thermal neutron flux at the field of view at the object position was less than 1%. The calculated thermal neutron flux (f th ), TNC and (n/γ) parameters are shown in Table 3 for different collimator parameters with and without the single sapphire filter for a 50-mg 252 Cf neutron source. 126

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9 J. G. Fantidis, C. Potolias, N. Vordos, and D. V. Bandekas The neutron flux was calculated with the aid of the MCNPX code using the F2 tally. Calculations were performed for up to NPS = histories for neutrons yielding an accuracy < 1 %. The gamma dose in the ratio n/γ calculated with F2, Fm2 tallies and the DE, DF cards. Cutoff (NPS) values up to 10 8 histories were considered with relative error < 2%. In proportion to Table3, the (n/γ) parameter remains in all circumstances above the recommended limit (10 4 n cm - 2 msv -1 ) ranging between and n cm -2 msv -1. The f th, TNC and (n/γ) parameters were defined for different sapphire filter thicknesses in the case of three L/D values (Table 4). The f th varies from up to n cm -2 s -1. The TNC varies from 1.1 up to 31.8%. According to the Gibbs et al., good-quality thermal neutron images require exposures of the order of 10 7 n cm -2 [21]. The exposure time is proportional to the thermal neutron flux. In the case of L/D = 25, the exposure time varies between 1.5 and 2.2 min without and with 11-cm single sapphire filter, respectively. Higher L/D values would require higher exposure time: in the cases of L/D = 50 and 100, exposure times in the range and min would be required, respectively. The exposure time was calculated for the 50-mg 252 Cf source emitting n s Conclusions In this work, the optimization of a transportable system which incorporates a 252 Cf neutron source was studied using the MCNPX Monte Carlo code. The study has provided an improved design which gives more expedient values in the parameters associated with thermal NR. All the materials considered were chosen according to the EU Directive 2002/95/EC. Simple and advanced materials were used with the purpose to ameliorate the shielding design. In accordance with the results obtained from the simulations, the proposed system has dimensions which render it suitable for transportation with a medium size lorry. Acknowledgments The authors thank Dr. G. E. Nicolaou who provided us the computer room from his laboratory with installed MCNPX code. References [1] C. Oliveira, J. Salgado, I.F. Goncalves, et al., A Monte Carlo Study Of The Influence of the Geometry Arrangements and Structural Materials on a PGNAA System Performance for Cement Raw Material Analysis, Appl. Radiat. Isot. 48 (1997) [2] D.P. DiPrete, S.F. Peterson, and R.A. Sigg, Low flux NAA applications at the Savannah River Site, J. Radioanal. Nucl. Chem. 244, (2000) 343. [3] M.M. Chiles, J.T. Mihalczo, and C.E. Fowler, Small, Annular, Double-Contained 252 Cf Fission Chamber for Source-Driven Subcriticality Measurements, IEEE Trans. Nucl. Sci. 40:4, (1993) 816. [4] W. Osborne-Lee and C.W. Alexander, CALIFORNIUM-252: A Remarkable Versatile Radioisotope, Oak Ridge National Laboratory Report ORNL/TM-12706, Oak Ridge, Tennessee,

10 Moldavian Journal of the Physical Sciences, Vol. 10, N1, 2011 [5] J.P. Barton, W.L. Sievers, and A. H. Rogers, Lessons from Pantex Cf-252 Wide Angle Beam Characterization, Nondestr. Test. Eval (2001). [6] B. Schillinger, E. Lehmann, and P. Vontobel, 3D neutron computed tomography: requirements and applications Physica B (2000). [7] J. Bahnart, Advanced Tomographic Methods in Materials Research and Engineering, Oxford University Press, Oxford, [8] R. Reddy and M.V. N. Rao, Neutron Radiography, Def. Sci. J., (1982). [9] J.P. Barton, Arg portable neutron radiography, Los Alamos National Laboratory, [10] Da Silva A.X. and Crispim V.R., Moderator-collimator-shielding design for neutron radiography systems using 252 Cf, Appl. Radiat. Isot (2001). [11] B. D Mellow, D. J. Thomas, M. J. Joyce, et al., The replacement of cadmium as a thermal neutron filter, Nucl. Instr. and Meth. A (2007). [12] J.G. Fantidis, G.E. Nicolaou, and N.F. Tsagas, A transportable neutron radiography system based on a SbBe neutron source, Nucl. Instr. and Meth. A (2009). [13] Martin R. C., Knauer J. B., and Balo P. A., Production, Distribution, and Applications of Californium-252 Neutron Sources, Appl. Radiat. Isot (2000). [14] D.B. Pelowitz (2005) MCNPXTM user s manual Version 2.5.0, University of California at Los Alamos National Laboratory. [15] V.V. Verbinski, H. Weber, and R.E. Sund, Prompt gamma rays from 235 U(n, f), 239 Pu and spontaneous fission of 252 Cf, (1973) Phys Rev C 7(3):1173. [16] J.G. Fantidis, G.E. Nicolaou, and N.F. Tsagas, A transportable neutron radiography system, J Radioanal Nucl Chem (accepted for publication available online from 25 March 2010) [17] J.C. Domanus, Collimators for Thermal Neutron Radiography An Overview, D. Reidel Publishing Company, [18] M.R. Hawkesworth, Neutron radiography: equipment and methods, At. Energy Rev. 15 (1977) 169. [19] D.C. Tennant, Performance of a cooled sapphire and beryllium assembly for filtering of thermal neutrons, Rev. Sci. Instrum (1988). [20] D.F.R Mildner and G.P Lamaze, Neutron transmission of single crystal sapphire, J. Appl. Cryst (1998). [21] K. Gibbs, H. Berger, T. Jones, D. Polansky, J. Haskins, D. Schneberk, and J. Brenizer, in: ASNT Fall Conference, October