The Attenuation of Neutrons in Barite Concrete

Transcription

1 Universities Research Journal 2011, Vol. 4, No. 4 The Attenuation of Neutrons in Barite Concrete Wunna Ko Abstract The aim of this research is to study the neutron attenuation in barite concrete as a result of neutron-moderation that takes place in a medium when neutrons pass through it. Three kinds of concrete samples are prepared. They are (1) sample A (ordinary concrete), sample B (barite concrete without normal aggregate) and sample C (barite concrete with normal aggregate). In this research, two kinds of measurements, the total macroscopic cross-section and the removal cross-section were done. In all measurements, Am-Be neutron source was used and neutron-detection was carried out by using the BF 3 counter. The measured BF 3 spectra were analyzed by using the Gamma Vision 32 software. The graphical representations of the analytical results as well as the required curve fitting were done by using the Excel software. From these, the measured values of the total macroscopic cross-section and the removal cross-section were obtained. Sample B has the greatest values of total cross-section and removal cross-section, and hence, it is superior to other samples for neutron attenuation. Introduction For the selection of neutron shielding materials, it is most important to quickly moderate the neutron to low energies, where it can readily be captured in materials with high absorption cross section. The most effective moderators are elements with low atomic number, and therefore hydrogen containing materials are the major components of most neutron shields. In this application, water, concrete and paraffin are all inexpensive sources of bulk shielding. Because mean free paths of fast neutron typically are tens of centimeters in such material, thickness of 1 m or more are required for effective moderation of almost all incident fast neutron. Because of the low cost and the adaptability to both block and monolithic types of constructions, concrete is more favoured than other materials for shielding purposes. Although ordinary concrete consists mainly of water; the low mass number elements and the moderately high Assistance Lecturer, Dr, Department of Physics, Yadanabon University

2 194 Universities Research Journal 2011, Vol. 4, No. 4 mass number elements, such as calcium and silicon, the other special formulation of concrete, barites concrete and iron concrete which are sometimes called heavy concretes, contain mainly of barium and iron whose atomic masses are 137 and 56 respectively. In fact, heavy elements are not good moderators, but they can reduce energies of very fast neutrons by means of inelastic collision. Sample Preparation All samples prepared for this research were made in Material and Soil-Test Laboratory, Public Works, Mandalay. In this research, three kinds of concrete samples were prepared to study their different attenuation properties. They are (1) Sample A (ordinary concrete) (2) Sample B (barite concrete without normal aggregate) (3) Sample C (barite concrete with normal aggregate) Table 1. Composition of sample A, B and C (kg / 3 m ). Constituents Sample A Sample B Sample C Cement water w / c fine normal aggregate coarse normal aggregate fine barite aggregate coarse barite aggregate

3 Universities Research Journal 2011, Vol. 4, No Experimental Determination of Total Macroscopic Cross-Section by Transmission Method The transmission method for the experimental determination of crosssection is based on measurements of the attenuation of a neutron beam after a passage through a slab of target material of finite thickness. The schematic diagram for the given method is shown in Figure 1. The experiment is based on the well-known equation x I= I0 e Σ where I 0 is proportional to the number of incident neutrons falling on a particular area, I is the number of neutrons which succeed in passing through the thickness x of the material over the same area and Σ is the total macroscopic cross-section of the given sample. The experimental arrangement for the measurement of cross-sections by the transmission method consists of a neutron source and a detector, between which is placed a slab of the material being investigated. By means of a suitable collimating shield, the neutron beam passing through to the detector is restricted to a relatively small solid angle. The cross section determined by in this manner is thus the total cross section. The neutron intensity as observed at the detector without the slab of material is proportional to I 0, where I is the corresponding value with the slab interposed between the same source and detector. For the determination of the cross section by means of above equation, it is not necessary to know the absolute values of I 0 and I, but only their ratio which is obtained in the manner described. [1] In order that this equation apply rigorously, scattered neutrons must not reach the detector, i.e., the experiments must be carried out with the good geometry. Good geometry is best achieved by locating the detector as far from the source as possible and by making the detector and the sample small. [3]

4 196 Universities Research Journal 2011, Vol. 4, No. 4 Experimental Determination of Removal Cross-Section For neutron attenuation calculation, exponential factor has not been found useful, because scattering and slowing down play such important roles. An alternative approach is based on the removal cross section concept; this is strictly applied to a relatively thin slab (or slabs) of a solid material is placed between a fission (fast) neutron source and a moderately thick layer of water. It has been found that under these conditions, the neutron attenuation can be well represented by a simple exponential function of the slab thickness. The physical basic of this situation is that the fast (fission) neutrons are absorbed and scattered in the slabs, the scattered neutrons which are slowed down, are then captured in the water so that they do not reach the detector outside the shield. The net effect is equivalent to exponential absorption in the slab. The thickness of the water about 0.45 m at least, should be sufficient to ensure that all the neutrons scattered in the water layer will be captured. The removal cross-section concept is strictly limited to layer shields of the type referred to above paragraph. Nevertheless, it can be used to provide an approximate measure of the attenuation of fission (fast) neutrons by a homogenous medium, such as concrete. [1] The removal cross section of a given material behaves formally as a macroscopic cross section in determining neutron attenuation, but it is not a cross section in the sense of representing the probability of a particular neutron- nucleus interaction. The removal cross section should be approximately equal to the total cross section of the slab material less part of the scattering cross section for the dominant fission neutrons. The observed value of the removal cross section is, in fact, roughly equal to twothirds of the total (scattering and capture) cross section in the given material for the neutrons having energies in the range of 6 to 8 MeV. Because of the opposing effects, with increasing neutron energy, of increasing penetrability and decreasing proportion in the fission spectrum, the major contribution to the neutrons that penetrate farthest into a water shield is made by those which originally had energies of about 6 to 8 MeV. These are so-called dominant fission neutrons that determined the removal cross section.

5 Universities Research Journal 2011, Vol. 4, No The neutron detector response which is represented by D ( t, z) is given by D (t, z) = D H 2 O(z ) e - Σr t where D H 2 O ( z ) = the detector response with the presence of water layer thickness of z, Σ r = the removal cross section of given material t = the thickness of material The effective removal cross-section (Σ r ) is the probability that a fast or fission energy neutron undergoes a first collision, which removes it from the group of penetrating, uncollided neutrons. If the concrete contains sufficient moderating material, this removal process will determine the attenuation of neutrons. The schematic diagram for finding the removal cross section of concrete sample is shown in Figure 2. [3] Figure (1). The experimental arrangement of finding total cross-section Figure (2). The experimental arrangement of finding removal cross-section

6 198 Universities Research Journal 2011, Vol. 4, No. 4 Experimental Set Up for the Present Work The instruments used for the present work including the Am-Be neutron source are (1) Am-Be neutron source (2) BF 3 counter (3) Detector bias supply (4) Preamplifier (5) Amplifier (6) Multichannel analyzer. Experimental set up for detecting the attenuation of neutron is shown in Figure 3 consisting of a neutron source (Am-Be) and BF 3 tube detector, between which is placed a slab of the material being investigated. By means of a suitable collimating shield, the neutron beam passing through to the detector is restricted to a relatively small solid angle. The Am-Be neutron source used in the present research work, is manufactured by Gammatron Inc., Houston, Texas. The range of its activity is 475 m Ci to 550 m Ci and 6 its yield is given by nps. The specifications of BF 3 counter used for neutron detection are described in Table 4.2. The operating voltage for BF 3 counter is 1700 V and the high voltage is supplied by ORTEC model 659 bias supply. The signal leaving the output of the detector was amplified by using the ORTEC model 671 amplifier. Between the detector and amplifier, ORTEC model 142 PC preamplifier was placed to get the optimized coupling between the detector output and the rest of the counting system and to minimize any sources of noise. At the end of the counting system, MCA (PC with Gamma Vision software) displayed the BF 3 spectrum, from which the required information of the measurements (neutron counts) can be obtained by making suitable analytical method. paraffin Detector Bias Supply ORTEC model 659 Preamplifer ORTEC model 142 PC MCA BF 3 tube Amplifer ORTEC model 671 Figure (3). Electronics of neutron detecting system

7 Universities Research Journal 2011, Vol. 4, No Experimental Procedure All measurements for this research were done in Experimental Nuclear Physics Laboratory at Mandalay University. The experimental procedure for studying neutron attenuation in different concrete samples included two parts. The first was to determine the total macroscopic cross section, and the second, the removal cross section. The experimental arrangements for two measurements were almost the same, but the latter included 45cm thickness of water-layer before the detector. As in Figure 3, the Am-Be neutron source was positioned in a paraffin container. The distance between the source and the moderated BF 3 tube was fixed at 75 cm. After measuring the neutron counts in the absence of the sample, the different thickness of the samples were placed between source and detector in a step of 2 cm at a time. The counting time interval for each thickness was set for 3 min. The measurements of the experiment were carried out until the total sample thickness of 20 cm. The system displayed the results of the measurement on MCA (PC with Gamma Vision- 32 Software) as the spectrum of the BF 3 tube. By analyzing the spectrum the information of the neutron counts can be obtained from the same region of interval. The aim of the measurement concentrated on the number of the neutrons (counts) but not on energy (i.e., the processes which occur due to the neutrons, entering the detector). So, in the spectrum, the gamma portion was discriminated and the information of average counts over the ground state and the excited state of 7 Li associated with the wall effect of the counter were considered. The measured spectrum of the BF 3 counter is shown in Figure (4).

8 200 Universities Research Journal 2011, Vol. 4, No gamma discrimination point Reaction products full energy peaks Counts 40 wall effects Energy (k ev) Figure (4). The measured BF 3 spectrum Results and Discussion The results of the two kinds of measurements, the total cross-section and the removal cross-section, are described in this section. The analytical results are shown as the variation of neutron counts with the sample thickness in the graphs, Figure 5 and 6. The values of the above parameters are obtained by doing curve fitting of these graphs by Excel software. The graphs show the exponential behaviour of the attenuation of neutrons with the thickness of samples and give the attenuation coefficients of the given samples (i.e., required cross-sections). The initial counts of Figure 5. are less than those of Figure 6. because, for finding removal cross-section, a thick layer of water was placed before the detector. The neutrons with the initial energies of 6 to 8 MeV, can pass through this water layer. Hence, the neutrons that did not make any collision in the sample, can reach the detector. However, almost all of the neutrons which were slowed down and scattered in the samples, were absorbed in the thick layer of water. Therefore, only the neutrons with unchanged initial energies can pass through the water layer. Table 2 shows that the macroscopic cross section of sample B is greater than those of A and C (i.e., I \ I 0 value is smaller and it makes the most attenuation after passing through the sample B).

9 Universities Research Journal 2011, Vol. 4, No Hydrogen content (in the form of water) is the same for all samples. Therefore, slowing down of neutrons by elastic scattering reaction of hydrogen nuclei may be same in all samples. In Sample A, slowing down of neutrons is done by hydrogen and other low mass number elements as a result of elastic scattering reaction. In Sample B, hydrogen content is same as Sample A, and slowing down of neutron is done not only by hydrogen but also by the high mass number elements (such as barium) which can slow down as a result of inelastic scattering reaction. The element barium can reduce the energies of very fast neutrons down to about 1MeV. The elastic scattering cross section of hydrogen is also significant (about 4 b at this energy range). So, not only barium can reduce neutron energy but also it can improve the effectiveness of the hydrogen. As a result, the effect of the combination of hydrogen with heavy elements can be found in Sample B. The macroscopic cross section for Sample C, another Sample of barites concrete, is also greater than that of ordinary concrete, Sample A. But it is smaller than that of Sample B because the barite content of Sample C is 50 % less than that of Sample B. sample A sample B sample C 1200 counts I A = e x R 2 = I B = e x R 2 = I C = e x R 2 = sample thickness(cm) Figure (5). Neutron counts Vs concrete thickness (Determination of the total cross-sections of Samples A, B and C)

10 202 Universities Research Journal 2011, Vol. 4, No. 4 Figure (6). Neutron counts Vs concrete thickness (Determination of removal cross-sections of Samples A, B and C) Table 2 The total macroscopic cross-section and the removal cross-section of the concrete samples Samples Σ ( cm ) t 1 Σ ( cm ) A B C r 1 Conclusion The absorption of fast neutrons is a two-step process: (1) the neutrons are slowed down principally by elastic and inelastic collision, with few captures occurring at high energies; and (2) the moderated neutrons are effectively removed by the capture because of the much higher crosssection at lower energies. In this respect, hydrogen and other low mass number elements play an important role in the moderation of fast neutron

11 Universities Research Journal 2011, Vol. 4, No and in the subsequent capture. When they combined with heavy elements, the more neutron attenuation properties can be obtained. Moreover, they can reduce the incoming neutron energy to a few kev by only one or two collisions. Acknowledgements I am grateful to Prof Dr Khin Mg Oo, Rector, Yadanabon University, for his valuable suggestion and advice in preparation of this work. Special thanks should go to Dr Aye Aye Lwin, Professor and Head of Physics Department, for her continuous encouragement and systematic guidance without which this research project would not be possible. References Beckurts, K. H. & K. Wirtz. Neutron Physics. (1964). New York: Springer-Verlag.OHG. El.Sayed Abdo, A. Calculation of the Cross Sections for Fast Neutrons and Gamma-rays in Concrete Shields. (2002). ( / locate / anucence) Glasstone, S. & A. Sesonske, Nuclear Reactor Engineering.(1998). Fourth Edition,Vol 1.