Investigation Of The Effects Of Variation Of Neutron Source-Detector Distance On The Emitted Neutron Dose Equivalent

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1 ISSN: 9- Vol. Issue, June - Investigation Of The Effects Of Variation Of Neutron Source-Detector Distance On The Emitted Equivalent Igwesi, D. I. Physics and Industrial Physics Department, Faculty of Physical Sciences, Nnamdi Azikiwe University, P. M. B. Awka, Anambra State, Nigeria. igwesidavid@yahoo.com Abstract Monte Carlo radiation transport simulations were performed to investigate the effect of variation on the neutron source-detector distances on emitted neutron dose equivalents using pure polythene and borated as shielding materials. The measurements were taken at three different neutron source-detector distances of cm, cm and cm by varying the position of the neutron source inside the water basin. The experimental set up was modelled using MCNP code with a He proportionate counter serving as the detector and different thicknesses of pure polythene and borated polythene as shields. The results from the MCNP simulations were used to calculate the neutron dose equivalents. The calculated results showed that the neutron dose equivalent decreases exponentially as the thickness of both shielding materials increases. The variations observed on the neutron dose equivalents for the three neutron source-detector distances indicated that decreasing the neutron source-detector distance increases the neutron dose equivalent which by implication established that the neutron sourcedetector distance of cm was obtained as the optimum shielding configuration for any occupational workers among the three distances considered. Furthermore,. cm of pure polythene and.7 cm of % borated polythene were obtained as the thickness of the shielding materials that can attenuate % of the neutron dose equivalent. Therefore, the results obtained showed that the effectiveness of any shielding material for high neutron source, such as Am/Be depends on the neutron source-detector distance and the density of the shielding materials. Keywords Am/Be Neutron Source,,, Occupational Workers, MCNP Code, Half Valued Layer.. INTRODUCTION Nuclear science and technology finds applications in many fields such as scientific research, agriculture, industry and medicine and it offers many advantages but not without some difficulties. Since the radiations involved are ionizing that have damaging effects on human health and environment, it is important therefore to evaluate the risks involved and possibly quantify the level of exposure to such ionizing radiations by radiation workers and subsequently develop technological configuration that guarantees the safety of radiation workers. Radiation shielding involves placing a shielding material(s) between the sources of ionizing radiations (such as Am/Be) and the worker or environment. These ionizing radiations which include alpha particles, beta particles, gamma rays, X-rays, neutrons etc, interact differently with shielding materials. Thus, the efficiency and effectiveness of the shielding varies with the types and energy of the radiation to be shielded as well as the shielding materials. The best materials for protection against radiation produced by neutron source are mixture of hydrogenous materials (polythene, water and many plastics), heavy elements, and neutron absorbing elements (such as boron, chromium), because they reduce both the intensity of gamma rays and neutrons. Indeed, hydrogenous materials slow down fast and intermediate neutrons energy via inelastic scattering, and they become thermal neutrons which can easily be absorbed by neutron absorbing materials that have a very high neutron absorption cross-section. One material useful for high energy neutron is polythene. is a good neutron shielding material [], [], [] and more effective when the thickness increases []. However, the shielding effect of polythene can be improved by the addition of boron [], [], [7], [], [9], which is a good neutron absorber. Thus, borated polythene materials shield against neutron source better than the pure polythene materials Unlike other forms of radiations, neutron s shielding introduces some complication because of the wide range of energies needed to be considered. Therefore, fast neutrons are first moderated before being captured. The moderation is achieved by the use of a non-radioactive material of low atomic number, and subsequently, appropriate materials are used to shield the thermalized neutron. Neutrons are associated with significant health issues as they are highly penetrating and can induce secondary deep body ionizing radiation doses. Large amount of neutron exposure biologically affect cells. JMESTN9

2 cm x cm Surface Absorber The affected cells can mutate, and in some cases result in cancer. The purpose of this work therefore is to measure the neutron dose emitted by Am/Be neutron source positioned at three different points within a water bath. The work specifically involves the simulation of different configurations of the source, shielding slabs and detector using MCNP by increasing the thicknesses of polythene and borated polythene slabs. The effective dose equivalent calculated from the MCNP results will be used to estimate the half value layer (HVL) for each of the shielding materials at the three source-detector distance of cm, cm and cm. ISSN: 9- Vol. Issue, June -. METHODOLOGY.MCNP Simulations The MCNP set up (Fig. ) involves an Am/Be neutron source in a water bath of dimension cm by cm, an absorbers (polythene and borated polythene slabs), and a SP9 He- proportional counter of dimensions. cm x. cm x. cm (Fig. ). The atomic percent composition of the water which acts as a moderator is 7% by hydrogen and % by oxygen. Neutron Source-Detector Distance cm cm He- Detector Ci Am-Be Neutron Source Fig. Geometry set up for the Am-Be neutron source shielding. MCNP version was used with cross sectional data obtained from.c series of ENDF/B-IV library while the materials composition of polythene, borated polythene and He- detector were taken from the DLC-/MCNPDATA []. Table shows the elemental composition of the materials used in the work. The Am/Be neutron source emit. x neutron per cm per second. The neutron source, absorbers and detector set up was simulated at three different neutron source-detector distances ( cm, cm, and cm) to ascertain which source-detector distance gives optimum shielding results. In each simulation, neutron histories of x 9 were considered and the number of neutron flux incident on the detector were calculated for each of the set up. Conversions from neutron fluxes to doses in rem hr - /n cm - s - were performed using equation. H = hφ. Table Elemental Composition of the Shielding Materials used Material Density (g/cm ).9 Borated. where H = dose equivalent, Φ = the emitted neutron flux, and h = the flux-to-dose-equivalent conversion which varies with neutron energy and was obtained from ANSI/ANS []. The equivalent dose values in rem hr - /n cm - s - were converted to millisievert (msv) (tables, and ). Constituents MCNP ID Atomic fraction H- C-nat H- C-nat B-.C.C.C.C.C JMESTN9

3 .Theoretical Background The dose equivalent of neutrons, H x, traversing through a thickness, x, of absorber is proportional to the intensity of the neutron source and the thermal neutron total cross section, Σ t, of the absorbing material []: Thus, dh t H.() dx If the equation is integrated, the result shows that the intensity of the uncollided neutrons decreases exponentially with the thickness of the absorber as shown in equation. H( x) H o e t x.() ISSN: 9- Vol. Issue, June - where Σ t is the thermal microscopic cross section of the absorbers. The half value layer was calculated for each neutron source-detector distance. In this work, it was considered that % borated polythene and pure polythene have thermal microscopic cross section of. cm - [] and. cm - [] respectively.. RESULTS AND DISCUSSIONS. Equivalent The results of the dose equivalents presented in tables,, and show that the neutron dose equivalents for the three neutron source-detector distances decreases as the thickness of the absorber increases [], []. From the results calculated, the dose equivalent where H o is the initial neutron dose equivalent and values for pure polythene for a given thickness of H(x) refers to those neutrons dose equivalent that absorber show remarkable high values than the dose penetrate a distance x in an absorber without a equivalent values for borated polythene. For instance, x collision; thus, the attenuation factor, e for a neutron source detector distance of cm, a t neutron dose equivalent of 7. msv/yr was represents the probability that a given neutron travels obtained for a pure polythene of thickness cm while a distance x without an interaction. Therefore, Σ t can for the same thickness of borated polythene, a be regarded as the probability per unit length that a neutron dose equivalent of 7. msv/yr was neutron will undergo an interaction as it moves obtained. This results agree favourably with Karmi through an absorber and be removed from the beam and Greenspan []. However, the difference between by either absorption or scattering []. them are not very wide, supporting the view of Coeck, et al [] who held that radiation reduction power of The half value layer (HVL) which is the thickness of pure polythene and borated polythene is not very the shielding material required to reduce the dose pronounce when dealing with small thickness but equivalent to a half of its initial value was estimated. becomes very noticeable as the thickness of the From equation, the HVL is given as absorbers increase. The dose equivalents for the HVL. 9 three neutron source-detector distances were. () observed to vary significantly. The least values of t dose equivalent were obtained at the neutron sourcedetector distance of cm which indicate that Table equivalents for both shields when Am-Be Source and Detector are cm apart. x -.9 x -. x -. x - 9. x - 7. x x -. x -.9 x -.77 x -. x -.7 x Table equivalents for both shields when Am-Be Source and Detector are cm apart Thickness (cm). x -.7 x -.9 x -.9 x -. x -. x x -.9 x -.7 x -.9 x -.9 x -.7 x JMESTN9

4 Neutron Dos ISSN: 9- Vol. Issue, June - Table equivalents for both shields when Am-Be Source and Detector are cm apart. x -. x -.7 x -. x -.77 x -. x x -.7 x -. x -. x -. x x From figure, it is observed that at neutron source-detector distance of cm, the values of neutron dose equivalents for both pure polythene and borated polythene are less than the International Commission on Radiological Protection (ICRP) recommended dose limit of ms/yr for occupational workers [], [7]. The results therefore showed that the position of the neutron source inside the water bath has significant effect on the number of emitted neutron dose equivalent and that the farther the neutron source-detector distance, the lesser dose equivalent obtained. This result is very useful when considering an optimum position for neutron source for the maximum protection of occupational workers..the half value layer (HVL) The results of the half value layer calculated show that. cm of pure polythene is required to attenuate % of the neutron dose equivalent while.7 cm of borated polythene is required for the same percentage. Thus, making borated polythene a better shielding material than pure polythene []. Pure Fig. Variation of Neutron Source-Detector Distances for shielding materials. CONCLUSION It can be concluded that the thermalization and subsequent absorption of neutrons produced by Am/Be source depends on the neutron sourcedetector distance and the density of the shielding materials. The results obtained show that the increase in the thickness of the shielding material decreases the neutron dose equivalent. It was also observed from the results that decreasing the neutron sourcedetector distance increases the neutron dose equivalent and as such, the neutron source-detector distance of cm produced an optimum shielding configuration for an occupational workers. A shielding thickness of. cm of pure polythene and.7 cm of % borated polythene were obtained to be capable of attenuating % of the neutron dose equivalent. REFERENCE [].D. R. Ochbelagh, H. M. Hakimabadb, and I. Najafabadi, The investigation of Am Be neutron JMESTN9

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