5. Gamma and neutron shielding characteristics of concretes containing different colemanite proportions

Transcription

1 Transworld Research Network 37/661 (2), Fort P.O. Trivandrum Kerala, India Nuclear Science and Technology, 2012: ISBN: Editor: Turgay Korkut 5. Gamma and neutron shielding characteristics of concretes containing different colemanite proportions Department of Civil Engineering, Faculty of Engineering, Bartin University, Bartin, Turkey Abstract. Radiation dose above the maximum permissible limit is harmful to environment and bodies. Study of radiation absorption in material has become an important subject to protect living creature and environment from harmful effect of radiation. Concrete is one of the most important materials used for radiation shielding in facilities containing radioactive sources and radiation generating equipment where radioactive impermeability is required. Nonetheless, measurement and method of radiation attenuation characteristic of a shielding barrier has become important. Sometimes, to do physical tests can be difficult and give insufficient or discrepancy results. Monte Carlo simulation method is a numerical technique that, beside other applications, offers numerical solutions to radiation transport problems that are either too complex or impractical to be solved analytically. In this respect, this study presents the irradiation measurement and Monte Carlo simulation results for attenuation of photons and neutrons by colemanite-based concrete samples. The results for neutrons agree reasonably while the limited number of measurements on photons reveals discrepancies that can be attributed to the divergence of the experimental setup from the thin target conditions. Correspondence/Reprint request: Dr., Department of Civil Engineering, Faculty of Engineering Bartin University, Bartin, Turkey. osmangencel@gmail.com

2 42 Introduction Proper absorption of ionizing radiation by shielding materials is a practical concern in radiation applications. The material of choice for radiation shielding is usually concrete, due to being inexpensive and effective for shielding both photons and neutrons. This attenuating medium is usually prepared using various materials of different densities as aggregates, which make up the largest proportion (about 70-80% of total weight) and play an essential role in modifying the mechanical properties as well as the shielding characteristics of concrete [1, 2]. Consequently, different concrete mixes have very different attenuation characteristics. Typical concrete mixtures consist of about 80 percent by weight of oxygen and silicon, with the rest of the composition comprising of calcium, aluminum and lesser quantities of sodium, potassium and iron. For attenuating neutrons, the hydrogen content, which makes up less than one percent by weight of most concrete, is very crucial [3] as well as other low Z elements such as boron. Turkey has a significant abundance of boron minerals possessing about 60% of the world reserves. Commercial boron ores of the country are in the form of colemanite, tincal, and ulexite. This material is used in various industrial applications requiring better absorption capabilities for neutrons such as being used as control rods in nuclear reactors and as a constituent material for neutron shields because of its high absorption cross section. Although considerable experiences have been gained in the past regarding the production of different types of concretes for shielding purposes, using the right ingredient usually relates to locally available materials that will provide the sought characteristics [4]. Some researchers have reported that boron and its various compounds have been used in cement production to enhance the shielding performance [5]. Okuno has produced polymer-based shielding slabs using colemanite [6]. Gencel et al. have investigated the engineering properties of concrete containing colemanite at different proportions [7]. Gencel et al. investigated protective effect of concrete produced with colemanite as biologic shield on rat. In that work, rats were housed in the cage concretes containing colemanite and then irradiated with 7 Gy gamma rays from Elekta SLi-25 Linear Accelerator (Siemens, Germany) twice over a week [8]. This study investigates the effect of colemanite proportion on neutron and gamma radiation transmission properties of concrete using irradiation measurements and Monte Carlo calculations.

3 Radiation shielding properties of colemanite loaded concrete 43 Materials and methods Sample preparation Concrete is one of the most important construction materials used for radiation shielding in facilities which employs a radiation generating equipment and therefore radiation impermeability is required. The shielding properties of concrete may be enhanced by changing its composition, especially the amount of aggregates included in the mixture. Colemanite ore (obtained from ETI Mine Works Inc., Turkey; density: 2.42 g/cm 3 ; chemical composition given in Table 1), was incorporated into the mixtures as aggregate as explained by Gencel et al. [7]. Five different concrete samples were produced for this study varying in colemanite proportion: 10% (CC10), 20% (CC20), 30% (CC30), 40% (CC40) and 50% (CC50). Table 1. Chemical composition of the colemanite ore by weight percentage. The Portland cement used in all the mixtures was manufactured according to the European Standards EN (1994) and EN (2000) and labeled as CEM II/A-M (P-LL) 42.5N. A unique water to cement ratio was selected as 0.42 and the cement content in each mixture was fixed to be 400 kg/m 3. A detailed explanation of the physical and mechanical properties of the concrete mixtures produced following this methodology can be found Table 2. Elemental weight percentages and densities of the samples.

4 44 in Gencel et al. [7]. Slabs of cm dimension were fabricated to be later used for measuring the radiation transmission or absorption properties. In addition, a plain concrete sample (PC00) that contains only the limestonebased aggregates (no colemanite aggregate addition) with three different grain sizes were prepared for comparison purposes. Table 2 presents the chemical composition of the concrete samples. Experimental setup The sample slabs were irradiated under the irradiation conditions depicted in Fig. 1. To follow the good geometry setup, a collimator was placed between the source and the detector 11 [9]. First, three counts were read without the sample in place and the average of these readings was taken to be I 0 (the incoming intensity). Then, three more counts were measured with the sample between the source and the detector (as shown in Fig. 1). The average of these values was taken to represent the transmitted intensity I. For neutron measurements, an Am-Be source irradiated the samples and the readings were carried out using a BF3 counter. On the other hand, a Co-60 source was utilized as a photon source and an ionization chamber was used for photon measurements. After the readings were taken, the Beer-Lambert law, I = I 0 e -ax (1) where x is the thickness of the slab in cm. Figure 1. Geometry setup.

5 Radiation shielding properties of colemanite loaded concrete 45 The parameter a (in units of cm -1 ) represents the absorption properties of the attenuating medium. For photons, it is referred to as the linear attenuation coefficient and is denoted as µ. It is a function of the incoming energy of the photons and the elemental composition of the attenuating material. For neutrons, the parameter of interest is called the fast removal cross section and is denoted as Σ R which is dependent on the incoming neutron energy and the chemical composition of the absorber. Monte Carlo simulations The Monte Carlo method is a numerical technique that, beside other applications, offers numerical solutions to radiation transport problems that are either too complex or impractical to be solved analytically. Particle interactions in material media are treated statistically and quantities such as the transferred energy, position of interaction, flight directions, etc. are estimated from appropriate probability distributions. The final answer for the quantity of interest is always derived by averaging the outcomes of many trials [10]. There are many computer software packages that handle radiation transport problems by the Monte Carlo technique. MCNP is one such code that is accepted as the industrial standard and is widely used by engineers and researchers in the field [11]. It can tackle particle interactions in threedimensional geometries and complex radiation sources such as line, surface or volume sources can be modeled to yield results for particle fluence, energy absorption, or dose. In this study, MCNP version 5 was employed. The irradiation geometry was modeled as close as possible to the thin target geometry (Fig. 2), where a disc (with 10 cm radius and 1 cm thickness) located at the center of the coordinate system represented the samples. Figure 2. MCNP geometry.

6 46 The particle source is a point source in air that is situated 20 cm away from the slab (target) along y and emits mono-energetic photons along +y direction. The detector volume, which records particle current across a surface (F1 tally in MCNP), is located on the other side of the slab 20 cm away from the slab s surface. The densities and material composition information for the samples were taken from Table 2. Only photons were tracked by MCNP; no secondary particles from photon interactions were followed in simulations. Results and discussion In order to investigate the radioactivity content of the samples, activity concentrations for gross alpha, gross beta and gamma sources were measured using the techniques outlined in [12]. The results are listed in Table 3 and are comparable to the radioactivity levels observed in environmental samples reported in literature [12]. Table 3. Radioactivity concentrations for the samples. The effect of the different colemanite addition on the specific weight of the hardened concrete specimens is seen in Fig. 3. The density of colemanite aggregates (2.4 g/cm 3 ) is lower than that of the lime based aggregate (2.7 g/cm 3 ). The reason behind the unit weight losses is the differences of specific gravity of aggregates. It is obvious from Fig. 3 that specific weight decreases with the increase in the added portion of colemanite into the concrete. But as can be noticed, the loss in specific weight is small for addition up to 40% colemanite. After that ratio, the loss is more observable.

7 Radiation shielding properties of colemanite loaded concrete 47 Figure 3. Unit weights of concrete samples. Table 4. The measured and computed values of fast removal cross section (Σ R ) for neutrons. Figure 4. Linear attenuation coefficients of samples for 30keV-30MeV energy range.

8 48 Table 4 lists the results of irradiation measurements and Monte Carlo simulations for fast removal cross sections of neutrons (Σ R ; in units of cm -1 ). The results agree within 10% except for the sample concrete CC50, where the disagreement may be attributed to some experimental error since Σ R values tend to decrease with density as mentioned above. Table 5. The measured* and computed values of linear attenuation coefficient (µ) for photons.

9 Radiation shielding properties of colemanite loaded concrete 49 Table 5 provides the results of Monte Carlo simulations for linear attenuation coefficient for photons (µ; in units of cm -1 ) in the energy range 30 kev 30 MeV. The data is plotted in Fig.4 as a function of photon energy and depicts a similar decreasing behavior observed in other materials. There is only one irradiation measurement for each sample listed at the bottom of Table 5, which corresponds to 1.25 MeV (average photon energy of a Co-60 source). The observed discrepancies partly result from the experimental setup, which to some extent diverges from the good geometry conditions (2 cm sample thickness in measurements as opposed to 1 cm thick targets in simulations). Conclusion The results from this study reveals that the colemanite-based concretes have the desired neutron absorption capabilities and it is vital to follow the good geometry conditions in measuring the attenuation characteristics of a shielding material. References 1. Kharita, M.H., Yousef, S., Al Nassar, M. 2009, Progr. Nucl. Energ., 51, Malhotra, V.M. and Kumar, M.P. 1996, Pozzolanic and Cementitious Materials, Gordon and Breach Science Publisher, SA. 3. Kase, K.R., Nelson, W.R., Fasso, A., Liu, J.C., Mao, X., Jenkins, T.M., Kleck, J.H. 2003, Health Phys., 84, Kharita, M.H., Takeyeddin, M., Al Nassar, M., Yousef, S. 2008, Progr. Nucl. Energ., 50, Demir, D. and Keles, G. 2006, Nucl. Instrum. Meth. B 245, Okuno, K. 2005, Radiat. Prot. Dosim., 115, Gencel, O., Brostow, W., Ozel, C., Filiz, M. 2010, Int. J. Phys. Sci. 5 (3), Gencel, O., Naziroglu, M., Celik, O., Yalman, K., Bayram, D. 2010, Biol. Trace. Elem. Res.,135, Cember, H. and Jhonson, T.E. 2008, Introduction to health physics, 4 th Ed., McGraw-Hill, NY. 10. Bielajew, A.F. 2001, Fundamentals of the Monte Carlo method for neutral and charged particle transport, The University of Michigan Press. 11. Briesmeister, J.F. 2000, MCNP-A general Monte Carlo N-particle transport code, Version 4C. Technical Report No. LA M, Los Alamos National Laboratory, New Mexico. 12. Bozkurt, A., Yorulmaz, N., Kam, E., Karahan, G., Osmanlioglu, A.E. 2007, Radiat. Meas. 42, EN 196-1, (1994) Test Methods for Cement, CEN TC 51. EN 197-2, (2000) Conformity evaluation, CEN TC 30.