Nuclear Science and Technology by Research Signpost

Nuclear Science and Technology Editor

Turgay Korkut Faculty of Science and Arts, Department of Physics, Ağrı İbrahim Çeçen University 04100, Ağrı, Turkey

Transworld Research Network, T.C. 37/661 (2), Fort P.O., Trivandrum-695 023 Kerala, India

Published by Transworld Research Network 2012; Rights Reserved Transworld Research Network T.C. 37/661(2), Fort P.O., Trivandrum-695 023, Kerala, India Editor Turgay Korkut Managing Editor S.G. Pandalai Publication Manager A. Gayathri Transworld Research Network and the Editor assume no responsibility for the opinions and statements advanced by contributors ISBN 978-81-7895-546-9

Preface Since the discovery of radioactivity by Becquerel in 1896, nuclear sciences have began to take place in the world of science. Altough existing of many of studies, this branch of physics is still full of unknowns. Nowadays, nuclear technologies have a large application area. Some of these fields are nuclear power centrals, nuclear medicine and defense industries. There are many advantages of the use of nuclear technologies: Nuclear technology is used for diagnosis and treatment of many diseases, especially cancer. Nuclear power plants as a clean energy source is another area. In addition to these advantages, increasing needs to nuclear technology creates a host of problems. Security is the most important one of them against possible nuclear damages. Studies have been made in this book is about nuclear fusion, fusion reactors, radiation damage, electronic damage by particles and radiation shielding topics. There are two methodologies used in the studies as experiments and Monte Carlo simulations (FLUKA and MCNP). I hope that readers will find quite relevant papers, precious results and useful comments in here. Turgay Korkut

Contents

Chapter 1 The use of MCNP code in APEX fusion reactor technology Aybaba Hançerlioğulları Chapter 2 Neutronic calculations at uranium powered cylindrical reactor by using Bessel differential equation Aybaba Hançerlioğulları Chapter 3 Experiments on neutron transmission and Monte Carlo simulations on production of radioisotopes through 4, 5 MeV neutrons on several boron compounds Turgay Korkut and Abdulhalik Karabulut Chapter 4 A new neutron absorber material: Oil loaded paraffin wax Bünyamin Aygün and Gökhan Budak Chapter 5 Gamma and neutron shielding characteristics of concretes containing different colemanite proportions Osman Gencel

Chapter 6 Estimation of neutron irradiation damages in Ni/n-GaAs Schottky contact layers via FLUKA Monte Carlo simulations Hatun Korkut, Turgay Korkut, Hülya Doğan and Abdülmecit Türüt Chapter 7 Measurement of mass attenuation coefficients by Si(Li), NaI(Tl) and Cd(Tl) detectors Mustafa Recep Kaçal, İbrahim Han and Ferdi Akman

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Nuclear Science and Technology, 2012: 1-13 ISBN: 978-81-7895-546-9 Editor: Turgay Korkut

1. The use of MCNP code in APEX fusion reactor technology Aybaba Hançerlioğulları

Kastamonu University, Kastamonu Arts & Sciences Faculty, Physics Dept. 37100 Kastamonu, Turkey

Abstract. In this study, new APEX hybrid model was developed by the way of using the APEX fusion technology. The superiority of the APEX fusion technology from the other fusion technologies is that a fluid wall was used in reactor, which flows instead the first solid wall. The advantage of this fluid wall is to extend the life the structural material of the reactor by reducing the rate of damage on the structrual material.It also allows high neutron wall loads.The meausures for the APEX hybrid model has been taken from the ARIE-RS reactor design which was made in the framework of studies.In the APEX studies, the conventional first solid wall facing with the plasma is replaced with fast flowing thin liquid wall layer. Free-surface first liquid wall concept is a revolutionary concept. The first liquid wall flows very fast and detains charged particles, and followed by the thick liquid wall (blanket) which flows slowly and absorbs generated energy and converts it to heat. In the study, the flowing molten salt (first wall and blanket) composed of Flibe (Li2BeF4) was considered as the main constituent mixed with different mole fractions (0-12%) of heavy metal salt ThF4 to increase the energy multiplication. Self sufficient Tritium Breeding Ratio (TBR>1,05) has been taken into account to determine the upper limit of the fraction of heavy metal salt in the mixture. Design and calculations of APEX were carried out as 3-D torus by using MCNP-4B computer code. Correspondence/Reprint request: Dr. Aybaba Hançerlioğulları, Kastamonu University, Kastamonu Arts & Sciences Faculty, Physics Dept., 37100 Kastamonu, Turkey. E-mail: aybaba@kastamonu.edu.tr

Aybaba Hançerlioğulları

Introduction In a commercially available fission reactor, only a few percent of uranium is utilized for energy generation. More than 97% of uranium fuel is removed from the reactor as spent fuel. Hence uranium is not utilized at its full potential by fission reactors. The situation for thorium is worse than uranium; despite there has been interest in utilizing thorium as a nuclear fuel over the last 30 year. The 2009 IAEA-NEA “Red Book” gives a figure of 4.5 million tones of thorium reserves and additional resources, but points that this excludes data from much of the world [1]. Thorium, like Uranium-238 is fertile. Thorium (Th-232) absorbs a neutron to produce Uranium-233, which is fissile. These fertile materials can also make fission with high energy neutrons. The Fusion reactor has a good potential to utilize uranium and thorium in the future. The term hybrid reactor refers to nuclear reactors which are driven by a fusion neutron source and include fertile or fissile material. The general idea of a hybrid reactor is to have fusion component to provide a source of high energy fusion neutrons which are to interact with a sub-critical fission component located adjacent to plasma. The main products of hybrid reactors are fissile fuel and/or energy. Plasma was designed as neutron source that the inner surface of first liquid wall exposed to neutrons homogeneously and the calculations were conducted with the fusion neutron spectrum of D-T reaction.

First wall in APEX fusion technology The primary objective of APEX is to identify and explore novel, possibly revolutionary, concepts for the Chamber Technology that can substantially improve the attractiveness of fusion energy systems. The primary objective of APEX is to identify and explore novel, possibly revolutionary, concepts for the Chamber Technology that can substantially improve the attractiveness of fusion energy systems. A number of promising ideas for new innovative concepts have already emerged from the first phase of the APEX study [2-5]. While these ideas need extensive research before they can be formulated into mature design concepts, some of them offer great promise for fundamental improvements in the vision for an attractive fusion energy system. These ideas fall into two categories. The first category seeks to totally eliminate the solid “bare” first wall. The most promising idea in this category is a flowing liquid wall concept. The liquid wall idea is “concept rich”. These concepts vary from “liquid first wall”, where a thin layer (< 2 cm) of liquid is flown on the plasma-side of the first wall, to “thick liquid wall”, where an all-flowing thick (> 40 cm) liquid serves as liquid wall/liquid blanket. Liquid walls offer many potential advantages that

MCNP code in APEX

represent an excellent opportunity to substantially enhance the attractiveness of fusion energy systems. The replacement of the first wall with a flowing thick liquid offers the potential advantages of high power density, high reliability and availability (due to simplicity and low failure rates), reduced volumes of radioactive waste, and increased structure lifetime. All these advantages make the thick liquid wall approach a strong candidate in the APEX study. The second category of ideas focuses on extending the capabilities, particularly the power density and temperature limits, of solid first walls. A promising example is the use of high temperature refractory alloys (e.g. tungsten) in the first wall together with an innovative heat transfer and heat transport scheme based on vaporization of lithium.

APEX fusion reactor model The liquid wall idea evolved during the APEX study into a number of concepts that have some common features but also have widely different issues and merits. These concepts can be classified, as shown in Table -1, according to: thickness of the liquid, type of liquid used, and the type of restraining force used to control the liquid flow [6,17,18]. The primary objective of Advanced Power Extraction (APEX) study is to explore innovative concepts for fusion power technology that can tremendously enhance the potential of fusion as an attractive and competitive energy source. Table 1. APEX liquid wall alternatives [2].

Aybaba Hançerlioğulları

One of the promising idea for new innovative concepts emerged from the APEX study seeks to totally eliminate the solid first wall. This most promising idea is a flowing liquid wall concept. The concept varies from “liquid first wall”, where a thin layer (< 2 cm) of liquid is flown on the plasma-side of the first wall, to “thick liquid wall”, where an all-flowing thick (> 40 cm) liquid serves as liquid wall/liquid blanket. Liquid walls offer many potential advantages that represent an excellent opportunity to substantially enhance the attractiveness of fusion energy systems. The replacement of the first wall with a flowing liquid offers the potential advantages of high power density, high reliability and availability reduced volumes of radioactive waste, and increased structure lifetime APEX/ARIES-RS modifications In the APEX study and adapt the GMD (gravity momentum driven) geometry concept, several changes were required to the baseline ARIES-RS design. The ARIES-RS consists of high temperature shield following first wall in the inner blanket where breeding blanket does not exist. The outher blanket has an advanced “dual cooled ” breeding blanket with flowing lityum and Hecooled ferritic steel structures [8, 9,14]. First the density was approximately doubled to obtain the correct surface heat flux and neutron wall load specified by APEX design goals. A list of the ARIES-RS Parameters and APEX modifications are listed in Table 2. The majority of the work reported here was carried out for the tokamak. Specifically, the ARIES-RS geometry was utilized whenever possible, with modifications for the unique structures and high flow rates required for CLIFF (Convective Liquid Flow First Wall). This means, however, that the ARIES-RS fusion power needs to be scaled-up 4500 MW to give the 10 MW/m2 peak neutron wall load and 2MW/m2 peak surface heat flux goals of the APEX study. Tokamak present a difficult challenge for liquid walls due to the fact that the plasma chamber is relatively closed with short scrape-off lengths, and so, vaporized liquid wall material must be screened by the edge plasma to keep it from penetrating to the core. Tritium breeding and energy multiplication Tritium breeding ratio, TBR, is defined as the ratio of the rate of tritium production in the system to the rate of tritium burned in plasma. In order to provide adequate tritium breeding, the flowing liquid must be a lithium containing medium. The tritium production reactions are as follows; 6

Li + 1n → 3H + 4He + Q (4,784 MeV)

Li + 1n → 3H + 4He + 1n + Q (-2,467 MeV)

MCNP code in APEX

Then the only practical liquids for first wall and blanket are lithium, leadlithium, Flibe, and Sn-Li. Flowing liquid metals may require the use of electrical insulators to overcome the MHD drag, while for Flibe free surface flows, MHD (Magneto hydrodynamics) effects caused by the interaction with the mean flow are less significant. In case Flibe, TBR is maximum with natural lithium-6 enrichment and it is reduced with Li-6 enrichment. Hence, Flibe has advantage of utilizing lithium without enrichment. The Energy Multiplication Factor (M) is defined as the ratio of the total energy deposited in the system to the incident neutron energy. About 80% of fusion energy, 14.1 MeV, is carried with neutron that penetrates the first wall and blanket and dissipates its energy through exothermic nuclear reactions. The presence of Uranium or Thorium in the in the liquid first wall and blanket on the other hand, provides additional energy generation through fission reactions with fusion neutrons. Table 2. ARIES-RS parameters and APEX modifications [2]. Structural Modification Major radius(m)

ARIES-RS Reactor 5,52

Minor radius(m)

APEX reactor 5,52

1,38

Plasma aspect ratio

Number of sectors

Fusion Power(MW)

2171

~4000

Neutron Power(MW)

1736

~3400

Alpha Power(MW)

433

~600

Fusion Power density(MW/m3)

6,38

~12

Average neutron load (MW/m3)

4,03

Peak neutron load(MW/m2) Average FW surface heat flux(MW/m2) Peak FW surface heat flux(MW/m2)

5,67

0,4

1,5

0,47

Design for APEX by MCNP-4 code MCNP is a general-purpose Monte Carlo n-particle code that can be used for neutron, photon, electron or coupled neutron/photon/electron transport, which is developed by Los Alamos National Laboratory nowadays, the complexity in the nature of the industrial problems

Aybaba HanĂ§erlioÄ&#x;ullarÄą

unfortunately, makes analytical solution impossible. On contrary to the analytical approaches, simulation models are more successful in modeling and solution of complicated problems [6, 11, 12]. Monte Carlo technique is randomly number selection technique from one or more probabilistic distribution in a special trial or simulation study. The complexity in the nature of the industrial problems unfortunately makes analytical solution impossible. The nature of problems becomes complicated and the number of integrated systems increases very fast with the technological developments on contrary to the analytical approaches, simulation models is more successful in modeling and solution of complicated problems. It is easier to follow the interactions between the variables in simulation designs. But, it requires too much computer usage. It is aimed to get numerical results by applying the data collected from the reel system to the model developed on the computer. By evaluating and interpreting the results, some estimates are done for system performance criterions. By using simulation models the worst condition scenarios can also be investigated. Calling the simulation technique as Monte-Carlo technique was done by Von Neumann and Ulam, and first applications was carried out in neutron diffusion problems. Monte-Carlo technique is randomly number selection technique from one or more probabilistic distribution in a special trial or simulation study. The method was then adopted easily for solution of much more complicated and non-statistical problems such as Integra -differential evaluation problems [6, 12]. Some authors suggested classification of the method for using only for sampling works of variance reduction techniques. However, the usage of the method nowadays is generally in selection of values randomly from the probabilistic distributions. APEX fusion reactor used in the study was designed by using MCNP-4 computer code, using Monte-Carlo technique, as 3-D torus. The dimensions for the APEX reactor has been taken from the ARIES-RS reactor design which was made in the framework of APEX studies. In this model, the radius of torus is 552 cm and minor radius starting from inner surface of first wall is 143 cm. The height of torus starting from center of first wall is 250 cm. The radius and thicknesses in one dimension are shown in detail in Figure 1. APEX fusion reactor used in the study was designed by using mcnp-4 computer code, using Monte Carlo technique, as in three dimensional torus the dimensional for the apex reactor has been taken from the ARIES-RS reactor design which was made in the framework of apex studies. in this model, the major radius of torus is 552 cm and minor radius starting from inner surface of first wall is 143 cm. the height of torus starting from center of first wall is 250 cm. A radius and thicknesses in one dimensionally are shown detail in figure 1 in APEX model; temperature

MCNP code in APEX

values at various points for Flibe liquid flow are given as oC (degrees centigrade). According to this when the entrance degree of the liquid to the system is 500 oC, the surface temperature at the exit is around 600 oC [2]. in the apex studies liquid walls concept, although containing many common aspects, the variability related to the liquid used, liquid thickness and the methods utilized to control the liquid flow have been shown at table. A hybrid reactor is based on either magnetic fusion energy (MFE) or inertial fusion energy (IFE) the neutron source is volumetric in magnetic fusion energy systems, whereas the target represents a point neutron source in plants. The (D, T) fusion neutron driver of MFE for hybrid reactor has been evaluated for 10MW/m2. Hence, this corresponds to the fusion neutron flux of 1014 (14.1 MeV) n/cm2.s at FW for conventional (D, T) driven hybrid reactor.

Numerical calculations Cross-sectional view of APEX designed by using MCNP-4B computer code is shown in Figure 2. The inner region is consisting of plasma and vacuum. Following this, first liquid wall, blanket, ferritic steel, shield, stainless steel and ferritic steel zone take place. One of the main neutronic parameters for a fusion or hybrid reactor is the energy multiplication factor (M). Fusion neutron energy can be amplified in the blanket by the fissions of 233 U and 232Th mainly. Exothermic and endothermic neutron capture reactions by 6Li and 7Li, respectively, also affect the M values. These reactions are given as follows:

M can simply be defined as below:

where dE dV is total integral fission rate, T6=∫∫Φ•Σ(n,α)T dE dV on 6Li and T7=∫∫Φ•Σ(n,n′α)T dE dV on 7Li represent the integral fission rate per D-T fusion neutron. One can observe that the addition of 233U to the salt improves the fission rate significantly, as expected, since 233U, having much higher fission cross sections, is very effective to enhance fission reactions compared to 232 Th.

Aybaba Hançerlioğulları

Figure 1. One-dimensional APEX model (a) inboard, (b) outboard [2].

Figure 2. Cross-sectional view of APEX fusion reactor model designed in MCNP-4B [6].

MCNP code in APEX

The effect of Uranium and Thorium on energy multiplication was investigated by using APEX model. The molten salt ThF4 and UF4 were separately added to the first liquid wall and blanket up to 12%. The total tritium production amount per source neutron (TBR) in first liquid wall, blanket and shield zones was calculated with respect to percentage of heavy metal content in the mixture. Considering Thorium, up to 9% ThF4 content in the mixture, TBR meets the requirement of TBR > 1, 05 which are necessary for self sufficient fusion reactor. On the other hand for Uranium, TBR requirement are met even at 12% UF4 content [14, 15, 17]. The results showed that by using 12% of natural Uranium in the molten salt mixture, the generated energy in the hybrid reactor is increased about 35% in comparison with the pure fusion reactor. Energy multiplication increases with heavy metal salt content. The rate of increase for UF4 is much higher comparing that of ThF4.

Tritium breeding There are only two candidate liquids that might meet all the criteria, especially that of being able to breed enough tritium: Li, Flibe (Li2BeF4)[2,3,5]. A commercial fusion reactor must have a tritium breeding ratio of (TBR>1.05) self sustaining. Fig 3 shows the variation of TBR with the heavy metal content in mole % in the flowing liquid. As expected, the TBR decreases with increased heavy metal content. Tritium self sufficiency has been maintained in the range of molten salt mixtures. The tritium production reactions are as follows 6

Li +

n → 3H + 4He + (4,784 MeV)

n → 3H + 4He+1n (−2,467 MeV)

Figure 3. The variation of TBR values versus the heavy metal fraction in the liquid medium.

Aybaba Hançerlioğulları

The effect of Uranium and Thorium was investigated by using the APEX model. The molten salt ThF4 and UF4 were separately added to the first liquid wall and blanket up to 12%.The total tritium production amount per source neutron (TBR) in first liquid wall, blanket and shield zones was calculated with respect to percentage of heavy metal content in the mixture. Considering Thorium, up to 9% ThF4 content in the mixture, TBR meets the requirement of TBR > 1.05 which is necessary for a self sufficient fusion reactor. On the other hand for Uranium, TBR requirement are met even at 12% UF4 content. The results showed that by using 12% of natural Uranium in the molten salt mixture, the generated energy in the hybrid reactor is increased about 35% in comparison with the pure fusion reactor.

Energy multiplication The Energy Multiplication Factor (M) is defined as the ratio of the total energy deposited in the system to the incident neutron energy. A Pure fusion reactor of APEX design using Flibe as a liquid wall has a fusion power of 4000 MW a blanket energy multiplication of M=1.72 and produces ~4880 MW total power. The change in energy multiplication with respect to heavy metal content is illustrated in Fig 4. At the left ordinate. M has a very similar shape to the fission rate [21]. The ThF4 content has a very minor effect on energy multiplication, whereas UF4 can lead to remark able energy amplification. The nuclear heat production in the ThF4 and lithium are fairly comparable so that a gradual replacement of lithium by ThF4 does not vary the gross plant power remarkably. One of the main neutronic parameters for fusion reactor is the energy multiplication factor. Fusion neutron energy can be multiplied in the blanket by the fissions of 233U and 232Th mainly about 80% of fusion energy, 14.1 MeV, is carried with neutron that penetrates the first wall and blanket and dissipates its energy through exothermic nuclear reactions. The presence of Uranium or Thorium in first liquid wall and blanket on the other hand, provides additional energy generation through fission reactions with fusion neutrons. One of the main neutronic parameters for a fusion or hybrid reactor is the energy multiplication factor M which is defined below: M, energy multiplication factor is defined as the ratio of the ratio of the total energy release in the blanket to the incident fusion neutron energy. Total energy release in blanket can be calculated as Total energy release in blanket=200*<ΣF*Φ>+4,784 T6−2,467 T7 M=

MCNP code in APEX

Figure 4. The variation of M with the heavy metal fraction in the liquid medium.

Where, Σf is total fission rate, T6 and T7 are tritium produced by 6Li(n,t) and 7 Li(n,n’,t)reactions, Respectively <ΣF*Φ>=∫ ∫ΣF*Φ dE dV=total integral fission rate, T6= ∫∫Φ* Σ(n,α)T dEdV T7= ∫∫Φ *Σ(n,n’α)T dE dV represents the integral fission rate per D-T fusion neutron [10,12]. One can observe that the addition of 233U to the salt improves the fission rate significantly, as expected, since 233U, having much higher fission cross sections, is very effective to enhance fission reactions compared to 232Th. Energy multiplication increases with heavy metal salt content. The rate of increase for UF4 is much higher compared with that of ThF4. Fissile fuel production The Fissile fuel production calculations are done for a neutron wall load of 10 MW/m2 fissile fuel production rates of 238U(n, γ)239Pu increase almost linearly with increased heavy metal content as can be seen in Fig 5. A substantial amount of fissile fuel would be produced by using heavy metal molten salt. The fissile fuel production result from the fertile-fissile

Aybaba Hançerlioğulları

conversion with (n ,γ) reaction in the fertile blanket, and tritium breeding takes place in the tritium breeding zone, which is positioned behind the fuel layer and contains Li2BeF4. The production of fissile fuel from the fertile fuel in the fuel zone of the blanket result from the following fission reaction. The neutron, which is on the left-hand side of equation, n starts the reaction, is a fast neutron, is the D –T fusion reaction. Utilization of the molten salt mixture Flibe+ThF4 produces a precious nuclear fuel 233U .The 239 Pu fissile fuel can be produced by 238U(n, γ)239Pu reaction Fertile fuel +n →Fissile Fuel +νn’ +200MeV

Figure 5. Fissile fuel production in the liquid wall versus the heavy metal percentage.

Conclusions and discussion In a commercially available fission reactor, only a few percentage of Uranium is utilized for energy generation. More than 97% of Uranium fuel is removed from the reactor as spent fuel. Hence, Uranium is not utilized at its full potential by fission reactors. The situation for Thorium is worse than Uranium; despite there has been interest in utilizing Thorium as a nuclear fuel over the last 30 years The 2009 IAEA-NEA “Red Book” gives a figure of 4.5 million tons of Thorium reserves and additional resources, but this excludes data from much of the world [1]. Thorium, like Uranium-238 is fertile. Thorium (Th-232) absorbs a neutron to produce Uranium-233, which is fissile. These fertile materials can also make fission with high energy neutrons. In this study, the main neutronic parameters, energy multiplication and fissile fuel breeding were examined for the APEX fusion reactor with various thorium and uranium molten salts. In addition to this, the new APEX

MCNP code in APEX

hybrid model has been developed by the way of using the APEX fusion technology and this model on the first liquid wall, blanket and shield zones fertile which changes between %0-12 is used together with (%100 flibe). Sufficient trityum amount is needed for the reactor to work itself. In the TBR>1.05 APEX fusion model TBR ( trityum breeding rate) nearly 1.22 TBR values decreases when UF4 and ThF4 proportion increases between %0-12 range. This decrease for ThF4 is faster when compared with the UF4 and M the energy production factor is nearly 1.74 these values are for % 100 natural Flibe. In APEX model Fisil material production speed per fusion neutron increases linear to heavy metal salt percentage. For further studies radiation damage to structural materials, in particular inner and outer first walls can be investigated futher.

References 1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13. 14. 15. 16. 17. 18.

“Uranium 2009: Resources, Production and Demand (“Red Book”)”, OECD NEA & IAEA, 2009. Abdou, M.A. 1999, Fusion Engineering and Design, 145, 16745. Sarer.B., Hançerlioğullari,A., Savruk,N. 2005, G.U. Journal of Science, 18(1), 17. Şahin, H.M. 2007, Ann. of Nucl. Energy, 34, 861. Şahin, S., Übeyli, M. 2004, Energy Conversion and Management, 45, 1497. Bremister J. 1993, Mcnp-4a General Monte Carlo Code N-Particle Transport Code, Version 4a, La-12625, New-Mexico. Ping, X., Luo, T.Y., Tong, L. 2005, Fusion Engineering and Design, 1275, 75. Moir, R.W. 1997, Nuclear Fusion, 37, 557. Abdou, M.A., Ying, A., Morley, N. 2001, Fusion Engineering and Design, 54, 181. Mccarthy, K.A., Petti, D.A., Moore, R.L., Merrill, B.J. 2000, Fusion Engineering and Design, 549, 51. Şarer, B., Günay, M., Korkmaz, M.E., Hançerlioğullari, A. 2007, Fusion Science and Technology, 52. Johston, R. A. 1963, A General Monte Carlo Neutronics Code, LAMS-2856, Los Alamos. Tillack, M.S., Wang, X.R., Pulsifer, J.,et al. (ARIES Team). 2009, Fusion engineering and Design, 49, 689. Übeyli, M. 2003, G.U. Journal of Science, 16, 387. Yalçın S., Übeyli, M., Acır, A. 2005, Sadhana, 585-600. Ying A. 1999, Thick Liquid Blanket Concept, APEX Interim Report, University of California. Übeyli, M ., Yalçın, Ş. 2008, Energy Conversion and Management, 49, 947. Şahin, S., Şahinaslan, S., Kaya, A. 1998, Fusion Technology, 34, 95.

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Nuclear Science and Technology, 2012: 15-24 ISBN: 978-81-7895-546-9 Editor: Turgay Korkut

2. Neutronic calculations at uranium powered cylindrical reactor by using Bessel differential equation Aybaba Hançerlioğulları Kastamonu University, Kastamonu Arts & Sciences Faculty, Physics Dept. 37100 Kastamonu, Turkey

Abstract. Nuclear reactors are the complex machine-equipment systems constructed through the use of advanced engineering technologies. Fission-type reactors are devices developed to generate energy at a stable power by taking the chain reaction under control. Therefore, K, neutron multiplication coefficient is an important nuclear parameter for a nuclear reactor to sustain generating energy by itself at a stable power. K, is the ratio of the number of neutrons generated in a generation to the number of neutrons absorbed in the previous generation. In this study, It has been used different cylindrical fuel configurations. These 233 U, %1,24 234U, configurations are Uranium isotopes (%98,13 235 238 %0,03 U, %0,60 U), respectively. For these configurations it has been calculated criticality (KEFF) using Bessel differential equation, Neutron flux (Ø) averaged over cylindrical surface and total fission energy deposition over cylindrical can be applied. The Bessel equation is formulated as follows and this equation is a special case of Bessel’s equation [1, 2]. Correspondence/Reprint request: Dr. Aybaba Hançerlioğulları, Kastamonu University, Kastamonu Arts & Sciences Faculty, Physics Dept., 37100 Kastamonu, Turkey. E-mail: aybaba@kastamonu.edu.tr

Aybaba Hançerlioğulları

Introduction In a commercially available fission reactor, only a small percentage of Uranium is utilized for energy generation. More than 97% of Uranium fuel is removed from the reactor as spent fuel. In the study Bessel differential equations are used for the calculations of neutron flux (Ø) and criticality coefficient (K) and cylindrical geometric structure are taken into account as the reactor geometry. Neutron flux (Ø) in the reactor changes according to the geometry of the reactor cylindrical the type of the fuel used and physical properties of the reactor. One -group method can be applied fairly effectively to the determination of the critical size of a fast reactor, provided that properly averaged cross-section values for the neutron spectrum are used [5]. The principle task of a reactor control system is to maintain control over the chain reaction, that is, to control the number of neutrons in one generation relative to the number of neutrons in the previous generation. With all the above conditions and simplifications, the Neutron Diffusion Equation for one energy group bares an infinite critical reactor model. In our day, the production of nuclear energy is mostly met by light and heavy water reactors [3]. Fossil fuel energy source, which can be divided with normal neutrons, is used in these reactors. Reaction effect sections generated with neurons in fusion and fusion energy reactors has an important place in reactor design. During the generation of these important reactions, not only the structural endurance of the materials but also their geometrical design is of importance. As part of nuclear energy raw materials, U233, U235, Th232 and Pu-239 cores take their place as fuel. In this study, by taking certain percentages from the cores mentioned, we have calculated the total flux in the core of the reactor and its reflector, the gain of the reflector, its critical volume and its total power at the heart of the reactor. While making neutronic calculations, we have made use of R radius reflective spherical reactor geometrical structure. We have taken advantage of spherical Bessel Differential Equation, which was modified under one group method approach becomes [1, 2, 8]. This cylindrical reactor has radius R. In this reactor flux depends only on the distance r from the axis.

Bessel differential approach to solution The Bessel equation is formulated as follows and this equation is a special case of Bessel’s equation, d2ø/dr2 + 1/r.dø/dr + (B2-n2/r2). Ø =0

(1)

Neutronic calculations for reactors

In which n is integer(n=0,1,2,3…) , if we let r ⇒ x, φ ⇒ y , and α mn = B 2 , after multiplication by r. Using these approaches, we can reach the balance equation. From our recent discussions, we recognize this as Bessel’s differential equation.

x 2 y '' + xy ' + ( x 2 − n 2 ) y = 0

(2)

The solutions of this equation are called Bessel Functions of order n. Since Bessel's differential equation is a second order ordinary differential equation, two sets of functions, the Bessel function of the first kind Y1=A Jn(x) and Y2=CYn(x) are the solutions to the above formulated equation. Y1 and Y2 are respectively called as the functions of the Bessel function of the first kind and the Bessel function of the second kind. The solution to (*) y ( x) = AJ n ( x) + CYn ( x)

(3)

Equation Bessel function of the first kind of order can be expressed as a series of gamma functions. The Bessel function of the second kind of order can be expressed in terms of the Bessel function of the first kind. As illustrated in Fig.1 and the Bessel function of the first kind and second kind. J n ( x) =

⎤ ∞ (−1) k ( x / 2) n + 2 k ⎡ xn x2 x4 1 ... − + − ⎥=∑ ⎢ 2 n Γ(n + 1) ⎣ 2(2n + 2) 2 x 4(2n + 2)(2n + 4) ⎦ k =0 k!Γ(n + k + 1)

⎡⎛ 1 1⎞ ⎛ 1 1 ⎞⎤ ( −1) m −1 ⎢⎜ 1 + + ... + ⎟ + ⎜ 1 + + ... + 2m−n ⎟ m⎠ ⎝ 2 m + n ⎠ ⎥⎦ ⎡ x ⎤ 2 1 ∞ ⎛ x ⎞ 1 n −1 (n − m − 1)! ⎛ x ⎞ ⎝ 2 ⎣ + ∑ Yn ( x) = J n ( x) ⎜ ln + y ⎟ − ∑ ⎜ ⎟ π π m=0 m! m !( m + n)! ⎝ 2 ⎠ π m=0 ⎝2⎠ ⎣⎢ 2 ⎦⎥ J p ( x) cos pπ − J − p ( x) = lim n. p →n sin pπ

(4) 2m+ n

The Bessel function of the second kind of order can be expressed in terms of the Bessel function of the second kind also known as the Weber Function. Bessel Equation can be expanding into series. Br 1+ 2 k ) 2 J n ( Br ) = ∑ k = 0 k! ( n + k )! ∞

(

Thus, J n (z ) which satisfies Bessel’s equation is a cylinder function.

Aybaba Hançerlioğulları

That for the real physical system, the neutron flux must be real and nonnegative, and the only Eigen function that is positive over the full domain, 0<r<R is related to the fundamental mode since the first zero of the J0(x) Bessel function occurs at X1 = 2.4048, the real neutron flux is the distribution in the physical system [6, 7, 12]. ⎛ 2 .4 0 5 ⎞ r ⎟ R ⎝ ⎠

φ (r ) = φ M AX J 0 ⎜

(5)

Determination of the maximum neutron flux The simplest form of the neutron balance equation is called Bessel Differential Equation. One of the steady state critical ideal reactor geometries that can be treated via analytical means is a long cylindrical core model, as illustrated in Fig.1. All the adjectives used to describe the system are needed to reduce the general, very complicated, particle balance equation into a form that can be treated analytically. The infinite homogeneous description implies that the axial height is large relative to the radius and that the neutron density in the axial and azimuthally directions is negligible, leaving a functional dependence material properties are constant throughout the system.

Figure 1. Basic geometry for the cylindrical critical reactor model A and B.

These conditions suggest that the variation of the only one variable or φ ( r , θ , z ) ⇒ φ ( r ) , where φ is the symbol used to represent the neutron flux. Also, symmetry in the system suggests that the neutron population will be the largest in the center of the reactor, which implies that the flux gradient is zero at r =0 critical reactor boundary conditions as shown in Fig.1.

Neutronic calculations for reactors r = 0,

dφ dr

=0 r =0

and at r = R , φ ( R ) = 0 , in mathematical terms, this system is a 2nd

order homogeneous boundary. In other words, K multiplication coefficient is the ratio of the previous neutron generation to the next neutron generation. Neutron flux ( φ ) in the reactor changes according to the geometry of the reactor. The reason why Bessel differential equations are used in the study is that the Bessel differential equations are relatively easier to solve and dependent on boundary value data than other mathematical equations and that other equation modeling. Offer advanced mathematical solutions (integral + differential). For critical reactor, “P” is the possibility of not leaking and expressed as following formula [4, 5]. P=

K eff K

∞

Rate of neutron absorption Rate of neutron absorption + Rate of neutron loss

(6)

As the number of neutrons will be steady-state in finite medium when the reactor is totally critical which is studied, following solution can be shown according to one-group diffusion method [3] D∇ 2ϕ − ∑ a ϕ + S = 0

(7)

In the equation, the first term refers to rate of the production of neutrons in Volume, and the second term refers to rate of absorption of neutrons in volume, and the third term refers to the rate of leakage of neutrons from Volume. As the rate at which neutrons are lost by absorption per cm3 / sec is equal to S = K ∞ ∑a φ

(8)

Thus, by using diffusion coefficient, we can find L, diffusion length. In this situation, the formula the formula B 2 =

B2 =

Kα − 1 =1 1 + L2 .B 2

Kα − 1 L2

is called as the buckling of the reactor or

can be used. This equation is generally called as

“Critical Equation”. For the infinite reactor, the formula must be the following B2 = (

2.405 2 ) R

Aybaba Hançerlioğulları

If the reactor is the composition of the core and the reflector, the 9th equation can be defined in two different ways. These equations are known as one-group modeling Critical specifications of some moderators which are used in the reactor are shown on the Table 1. Table 1. Critical specifications of some moderators [5].

A separate calculation using the total power of the reactor P, which is a design criterion, should be carried out. In practice, the total power for a noboiling reactor can be easily determined by measuring the flow rate of the coolant as well as its inlet and outlet temperatures to the reactor core. Let us determine now the maximum thermal neutron flux for a homogeneous reactor equipped with axial and radial reflectors. Using Eq-7, we can write the total power of the reactor as P = E R Σ f ∫ φ ( r )dV

where dV is the differential volume element. In the view of the geometry of the problem, dV is the given by dV=2πrdr for the infinite cylinder. This integral can be evaluated using the formulas below P = 2π E R Σ f R 2 A1 J 1 ( 2 . 405 ) / 2 . 405 = 1,35 E R Σ f R 2 A

Final expression for the flux (Ø(r)) is then. φ (r ) =

0, 738 P J 0 (2.405 r / R ) ER ∑ f R 2

For the finite cylinder, the buckling of the reactor and finite flux is shown using the formulas below, B2 = (

2.405 2 π 2 ) +( ) R H

Neutronic calculations for reactors

φ ( nz ) = φmax J 0 ( Br ) cos(

π .z H

)

(9)

Numerical calculations In this study, by benefitting from Numerical Bessel Equation, critical calculations are calculated using the four different variations of uranium material. In Table 2, microscopic cross section values are shown as “barn” type. Table 2. Microscopic cross section values of fuel material [5].

For K critical calculation, KEFF values are given in Table 4. The number of the total neutron number crossing over the surface of cylinder. During the study, we used ‘Matemice-7 Wolfram and Numeric Bessel’s programmers. The values are in accordance with those from Monte Carlo Mcnp\code system of ENDF-V-VI (RSIC computer code collection Mcnp-4b). In Fig.2, KEFF values are reflected in accordance with crash*length of the way/ absorption of active track number. The results are compared using the KEFF results which are gained using analytic Bessel function with those from MCNP4b/ ENDF-V-VI [13, 14]. Table 3. The Flux Distribution (H=4.81, R=2.405).

Aybaba Hançerlioğulları

Table 4. KEFF, active track number.

1,05

1,00

Keffect

0,95

0,90

0,85

0,80

Bessel Calc. Monte Carlo Calc.

0,75 0

Track number

Figure 2. KEFF values compared track number.

Figure 3. Flux rate comparison for infinite cylindrical reactor.

100

Neutronic calculations for reactors

In Table 3, H=1 and R=1are considered to be a reference, when we compare the flux value in finite and infinite cylinder, as seen on Fig.3, n=0 Bessel function is equal to at the Ccritic=2.405. This shows that the first critical length doesn’t change. However, as seen in Fig.3, another harmonics of Bessel, flux rates can change in these lengths (n=1, 2, 3..) That is, flux value depends on geometrical structure and used fusion fuel variation. In Table 4, active track number of the reactor compared with those of calculated (KEFF) values. As seen in Fig.2, between the track numbers 0-20 and 60-80, Bessel calculations and Monte Carlo Calculations are almost the same and KEFF of the reactor is Fig.1.

Conclusions and discussion In the study, the reason why Bessel differential equations are used in the study is that the Bessel differential equations are relatively easier to solve and dependent on boundary value data than other mathematical equations and that other equation modeling offer advanced mathematical solutions (integral + differential). Bessel differential equations are second order ordinary differential equations and they offer solutions in the cylindrical, spherical and polar coordinates easily and also required physical parameters in the reactor can easily be obtained through the use of Bessel differential equations. Bessel differential equations are used for the calculations of neutron flux (φ) and criticality coefficient (K) and cylindrical geometric structure is taken into account as the reactor geometry. Bessel differential equation of higher order can be expressed by Bessel function of lower orders. Keeping the first terms in the series expansions the behavior of a Bessel function at small or large can be captured and expressed as elementary functions which are much easier to be understood and calculated than the more abstract symbols [8]. K, multiplication coefficient is the ratio of the previous neutron generation to the next neutron generation. Neutron flux (φ) in the reactor changes according to the geometry of the reactor (the type of the fuel used and physical properties of the reactor. Several mathematical equations are required to obtain criticality coefficient (K) regarding the changes mentioned. The important equations and theories used are as follows; Bessel differential equations, the Monte Carlo method, general diffusion equations, Fourier and Taylor Series, the perturbation theory.

References 1. 2.

Gray, A., et al. 1992, Bessel functions London. Watson, G.N. 1922, Theory of Bessel functions, Cambridge.

3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13. 14.

Aybaba Hançerlioğulları

Özemre, A.Y. 1959, Nukleonik, 1, 347. Lamarsh, J.R. 1983, Introduction to nuclear engineering 2nd Edition, AddisonWesley. Aybers, N. 1988, Nükleer Güç Reaktörlerinin Termodinamik Analizi, Istanbul University, Institue of Nuclear Science Publish. White, J.R. 1998, Mathematical Methods the strum lioville problem neutron diffusion in nuclear’Lecture notes, Lowell. Byerly, W.E. 1895, Fourier’s series and spherical, Boston. Kuzmin, R.O. 1930, Bessel functions. Lebedev, N.N. 1972, Special Functions and their applications, Paperback. Horie, J. 1974, J. Nucl. Sci. Technol., 11(9), 359. RSIC computer code collection Mcnp 4B. 1995, Press, London. Horie, J., Nishihara, H. 1975, J. Nucl. Sci. Technol., 12(9), 531. Thomas, J.W., Downar, T.J.2007, Reactor physics simulation with coupled Monte Carlo calculation and computational fluid dynamics, International Conference on Emerging Nuclear Energy Systems, Istanbul, Turkey. Shayesteh, M., Hahriari, M .S. 2009, Ann. of Nucl. Energy, 36(8), 901.

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3. Experiments on neutron transmission and Monte Carlo simulations on production of radioisotopes through 4, 5 MeV neutrons on several boron compounds Turgay Korkut1 and Abdulhalik Karabulut2

Faculty of Science and Arts, Department of Physics, Ağrı İbrahim Çeçen University, 04100, Ağrı Turkey; 2Faculty of Science, Department of Physics, Atatürk University, 25040, Erzurum, Turkey Erzurum Technical University, Erzurum, Turkey

Abstract. In this work, four boron compounds (MgB2, NaBH4, H3BO3 and KBH4) were irradiated 4,5 MeV neutrons. Neutron transmissions were measured using by an equivalent dose rate detector. Experimental results were evaluated depending on number of boron atoms per unit volume of compounds. Also produced total radioisotopes per primary neutrons emanated interactions between neutrons with compound atoms were simulated using by FLUKA Monte Carlo code. We have found that neutron absorption capability is relate to density of boron atoms per unit volume. Total numbers of radioisotopes per primary neutrons are direct proportion of hydrogen content of compounds.

Introduction Neutrons and protons make up the nucleus of an atom. Because neutrons don’t have an electric charge, they don’t influence by Coulomb force. So neutron radiation is hazardous a lot. To preserved neutron radiation several Correspondence/Reprint request: Dr. Turgay Korkut, Faculty of Science and Arts, Department of Physics, Ağrı İbrahim Çeçen University, 04100, Ağrı, Turkey. E-mail: turgaykorkut@hotmail.com

Turgay Korkut & Abdulhalik Karabulut

shield materials have been used up to now. Studies about neutrons are become widespread recently. For piperazinium hexachlorodicuprate under hydrostatic pressure, a neutron scattering study was made [1]. Mat'as et al. implemented a neutron diffraction study for KEr(MoO4)2 material [2].Fast neutron shielding characteristics of three boron compounds were investigated [3]. There are some studies on neutron transmission measurements in literature. Neutron transmission through pyrolytic graphite crystals was measured [4]. High-resolution neutron transmission and capture measurement results of Pb-206 nucleus were obtained [5]. Sato et al. designed a new material evaluation method by using a pulsed neutron transmission with pixel type detectors [6]. Korkut et al. determined neutron dose transmissions of several new concrete samples [7]. Epithermal neutron capture and transmission measurements of natural molybdenum were performed for resonance parameters and uncertainties analyses [8]. Monte Carlo simulation is a method for iteratively evaluating a deterministic model using sets of random numbers as inputs. These methods are used for designing detectors, comprehensing their conduct and comparing experimental results to theory, or on more large scale of the galaxy modeling in experimental particle physics [9]. FLUKA is a fully integrated Monte Carlo simulation code. FLUKA code has several applications in high energy physics and physical engineering, shielding processes, detectors and telescope design, cosmic ray physics, dosimetric studies, medical applications and biological physics [10]. There are several FLUKA simulation studies in literature. Gamma radiation absorption properties of amethyst mine were simulated by using FLUKA code [11]. A FLUKA simulation on parameterization of muon and electromagnetic signals was performed [12]. In this paper, interactions between 4.5 MeV neutrons with four boron bearing compounds are investigated. Neutron dose equivalent transmissions tested using by a neutron detector. Radioisotope nuclei production simulations were achieved by using FLUKA MC code.

Experimental details In this study, we used a NP-100B neutron detector. It provides to detect slow and fast neutrons. Tissue equivalent dose rates of the neutron field can be measured by it. Our neutron detector contains a proportional counter which produces pulses resulting from neutron interactions within it. The probe includes components to alleviate and attenuate neutrons. Because of a neutron particle has no charge, it can only be detected indirectly through nuclear reactions that create charged particles. As the conversion target, the

Neutron experiments and simulations for three boron compounds

NP100B detector uses 10B isotopes. Detector characteristics are shown in Table I. Fig.1 reveals energy response curve of detector. We used 241Am-Be neutron source which emits 4.5 MeV effective energetic neutron particles. 241 Am-Be neutron source is compacted mixture of americium oxide with beryllium metal in a cylindrical shape. Alpha particles sending from 241Am isotope have near to 5.5 keV maximum energy.

Figure 1. Energy response curve of detector. Table 1. Characteristics of Canberra NP100B Neutron Detector.

Turgay Korkut & Abdulhalik Karabulut

We have elaborated four disc samples; MgB2, NaBH4, H3BO3 and KBH4. It is necessary to pre-concentrate and homogenize samples and prepare a uniform specimen before detection process. Our boron bearing samples were then ground and sieved to a mesh size of 300 and then mixed the mixing time was 10 min. The samples have prepared as pellets with radius and thickness of about 1 and 1.5 cm, respectively. 241Am-Be neutron source was coated using collimator box included paraffin and lead conformably to data sheet of source. A shield stick used to avoid interactions between neutrons with air molecules in collimator box. During measurements samples were fixed in collimator box as adjacent neutron source. The Canberra NP100B series neutron detector was fixed combined with the exit of collimator box. It is satisfied that neutrons passed from samples arrived to detector probe. Detector was positioned front of circular hole where is surface of collimator box. The background measurements were done 100 times. And then samples were put between collimator and detector, respectively. By this means 100 counts were done for boron compounds samples. Measurement results were read on 606M model transportable rate meter and system PC via RADACS software. Experimental design is illustrated in Fig.2.

Figure 2. Experimental design.

Monte Carlo simulations FLUKA is a general aim code for simulations of particle transport and interactions with matter. It is useful for simulate all particle physics, nuclear

Neutron experiments and simulations for three boron compounds

physics, medical physics, accelerator physics applications. FLUKA Monte Carlo tool simulates with high precision the interaction and spread in matter of about 60 different particles as photons, electrons, neutrinos, muons etc [10]. In our simulations, four boron compounds were irradiated 4,5MeV neutron particles and radioisotope production outputs were considered. Firstly, atomic structures and densities of samples were written in MATERIAL and COMPOUND cards. Then primary neutron energy (BEAM card), transmission geometry parameters (sample dimensions, distances source-sample and sample-detector), detector properties were typed in FLUKA input file. Secondly, simulation has been started for 1.000.000 primary neutron particles. After running, number of total radioisotopes, mass numbers of produced radioisotopes and their efficiencies read from RESNUCLEI output.

Results and discussion In this section experimental results and outputs of FLUKA simulations are presented. In Fig.3, neutron equivalent dose rate transmissions are evaluated to number of boron atoms per unit volume for each compound. As can be seen in Fig.3, transmission is a negative function of number of boron 0,50 2

Linear Fit (R =0.99031)

0,45

MgB2 0,40

B atoms/cm

0,35 0,30

NaBH4

0,25 0,20

H3BO3 0,15

KBH4 0,10 0,800

0,805

0,810

0,815

0,820

0,825

0,830

Transmission

Figure 3. Neutron equivalent dose rate transmissions.

0,835

Turgay Korkut & Abdulhalik Karabulut

atoms. That is to say, neutron absorption capacity increases with number of boron atoms in a compound. If we want to enhance neutron absorption capability of compounds, we should applied boronization process. Table 2. Mass numbers of produced isotopes and isotope efficiencies of MgB2.

Table 3. Mass numbers of produced isotopes and isotope efficiencies of NaBH4.

Table 4. Mass numbers of produced isotopes and isotope efficiencies of H3BO3.

Neutron experiments and simulations for three boron compounds

Table 5. Mass numbers of produced isotopes and isotope efficiencies of KBH4.

Table.2, Table.3, Table.4 and Table.5 demonstrate mass numbers and isotope efficiencies of MgB2, NaBH4, H3BO3 and KBH4 respectively. Judging by these tables, KBH4 has maximum number of radioisotope. Total number of radioisotopes per primary neutrons as a function of hydrogen percentages is shown in Fig.4. As can be seen in Fig.4, total number of radioisotopes increases with increasing hydrogen content. If compounds which have high hydrogen amount are used to shield 4.5MeV neutrons, number of radioisotopes increases the same amount. 0,10

Number of Radioisotopes per primary n

0,09 0,08

NaBH4

0,07 0,06

KBH4 2

Linear Fit(R =0.9946)

0,05

H3BO3

0,04 0,03

MgB2

0,02 0,01 0,00 0

Hydrogen(%)

Figure 4. Total number of produced radioisotopes per primary neutrons.

Turgay Korkut & Abdulhalik Karabulut

Radiation protection is an important issue about protecting people and the environment from the harmful effects of ionizing radiation, which includes both particle radiation and high energy electromagnetic radiation. Radiation technologies widely used in industry and medicine. The most important disadvantage of radiation is causing microscopic damage to living tissue, resulting in skin burns and radiation sickness at high exposures and statistically elevated risks of cancer, tumors and genetic damage at low exposures. In this paper, we aimed to obtain neutron transmissions and simulate radioisotope products for some boron compounds. Our findings may be useful for nuclear processes, nuclear science laboratories, neutron experiments, radiation therapy hospitals especially BNCT rooms, etc…

References 1.

Hong, T., Stock, C., Cabrera, I., Broholm, C., Qiu, Y., Leao, J.B., Poulton, S.J., Copley, J.R.D. 2010, Phys. Rev. B 82 (18), 184424. 2. Mat’as, S., Dudzik, E. Feyerherm, R., Gerischer, S., Klemke, S., Prokes, K., Orendacova,A. 2010, Phys. Rev. B 82 (18), 184427. 3. Korkut, T., Karabulut, A., Budak, G., Korkut, H. 2010, J Radioanal Nucl Chem 286, 61. 4. Adib, M., Habib, N., Fathaalla, M. 2007, Annals of Nuclear Energy 33 (7), 627. 5. Borella, A., Gunsing, F., Moxon, M., Schillebeeckx, P., Siegler, P. 2007, Phys. Rev. C 76 (1), 014605. 6. Sato, H., Takada, O., Satoh, S., Kamiyama, T., Kiyangi, Y. 2009, Nucl Instrum Methods Phys Res A 623(1), 597. 7. Korkut, T., Ün, A., Demir, F., Karabulut, A., Budak, G., Şahin, R., Oltulu, M. 2010, Ann of Nucl Energy 37, 996. 8. Leinweber, M., Barry, D.P., Burke, J.A., Drindak, N.J., Danon, Y., Block, R.C., Francis,N.C., Moretti, B.E. 2010, Nuclear Science and Engineering 165 (2), 2000. 9. Mac Gillivray, H. T., Dodd, R. J. 1982, Astrophysics and Space Science 86(2), 437. 10. Ferrari, A., Sala, P.R., Fasso, A., Ranft, J. 2005, FLUKA: a multiparticle transport code.CERN 2005-10 INFN/TC 05/11, SLACR-773. 11. Korkut, T., Korkut, H., Karabulut, A., Budak, G. 2011, Ann Nucl Energy 38 (1) 56. 12. Yushkov, A., Ambrosio, M., Aramo, C., D'Urso, D., Valore, L., Guarino, F. Phys. Rev. D 81 (12) 123004. Canberra Data Sheet http://www.canberra.ru/html/products/healthphysics_rms/adm600/NPSeries.pdf 241

Am-Be Radioactive Material Safety Data Sheet http://www.stuarthunt.com/pdfs/Americium_241Beryllium.pdf

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Nuclear Science and Technology, 2012: 33-39 ISBN: 978-81-7895-546-9 Editor: Turgay Korkut

4. A new neutron absorber material: Oil loaded paraffin wax Bünyamin Aygün and Gökhan Budak*

Faculty of Science, Department of Physics, Atatürk University, 25040, Erzurum, Turkey

Abstract. Paraffin wax is produced from petroleum by de-waxing light lubricating oil stocks. It is one of the finest neutron moderators. Paraffin oil is refined from petroleum and is relatively cheap to produce. In this paper, oil concentration values of paraffin wax samples were changed. Neutron dose equivalent transmissions were measured using BF3 neutron detector and 4.5 MeV 241Am-Be neutron source for seven paraffin wax samples including different oil percentages. Also neutron macroscopic cross sections and half value layer values were calculated. As a result, neutron shielding properties of paraffin wax increased with increasing oil concentration. This is the first study about neutron shielding capabilities of paraffin wax having different oil quantity.

Introduction In chemistry, paraffin is the common name for the alkane hydrocarbons with the general formula CnH2n+2. The solid forms of paraffin, called paraffin wax, are from the heaviest molecules from C20H42 to C40H82. Because of the high hydrogen content, paraffin is frequently used for neutron shielding. Paraffin oil (liquid paraffin, white oil, kerosen) is colorless transparent oily liquid. Density of oil is approximately 0.86-0.91g/cm3. Major components of paraffin oil are C16 - C20 n-alkanes (CnH2n). Correspondence/Reprint request: Dr. Bünyamin Aygün, Faculty of Science, Department of Physics, Atatürk University, 25040, Erzurum, Turkey. E-mail: baygun25@hotmail.com; *gbudak@atauni.edu.tr

Bünyamin Aygün & Gökhan Budak

Neutron shielding studies have been performed widely. Hayashi et al. did neutron transport calculations of zirconium borohydride and zirconium hydride [1]. Yarar and Bayülken studied on neutron shielding efficiency and radioactivity of concrete shields containing colemanite [2]. There are several important studies relevant to paraffin wax as a neutron moderator. Akaho et al. determined neutron transmission parameters of several samples which have different hydrogen concentrations together with paraffin wax [3]. Nieminen et al. studied on a Monte Carlo simulation associated with neutron shielding using lead and paraffin [4]. Uddin et al. used paraffin wax to shield 241Am-Be neutron source [5]. Józefowicz et al. used paraffin and iron filters in neutron dosimetry study [6]. Hashemi-Nezhad et al. used paraffin and Am-Be neutron source for slow neutron flux measurements [7]. Seguchi et al. determined fast neutron irradiation effects on paraffin [8]. Minsky et al. and Korkut et al. used paraffin wax to shield neutron source [9,10]. Fast neutron removal cross section values of paraffin wax was measured by Desdin and Ceballos [11]. Fuga was determined removal cross sections of some materials for14, 6 MeV neutrons [12]. El Abd was measured thermal neutron cross sections for tungsten and molybdenum by using (n,γ) nuclear reactions [13]. Paraffin wax includes a petroleum product called paraffin-oil or liquid paraffin. There isn’t any study associated with neutron moderating properties of paraffin wax consisting of different paraffin-oil concentration up to now. The aim of this work is to measure neutron macroscopic cross sections and to show half value layers of seven paraffin wax samples having different oil percentages.

Experiments Seven different paraffin wax samples were exposed 4.5 MeV neutrons. Samples have 11x8x2 cm3 dimensions. Oil percentages and code numbers of specimens are shown in Table 1. Table 1. Oil percentages and code numbers of specimens.

Use of paraffin wax against neutrons

Physical specifications of oil-loaded paraffin wax samples are shown in Table 2. We used 241Am-Be neutron source which emits 4.5 MeV neutron particles. Physical form of 241Am-Be neutron source is compacted mixture of americium oxide with beryllium metal. Alpha particles sending from 241Am have approximately 5.5 keV maximum energy [14]. Table 2. Physical specifications of samples.

In this paper, a NP-100B neutron detector was used. It provides to detect slow and fast neutrons. Tissue equivalent dose rates of the neutron field can be measured by it. The detectors contain a proportional counter which produces pulses resulting from neutron interactions within it. The probes contain components to moderate and attenuate neutrons. Due to the fact that neutrons have no charge, they can only be detected indirectly through nuclear reactions that create charged particles. The NP100B detector uses 10B as the conversion target [15]. Our neutron source was shielded using collimator box included paraffin and lead according to 241Am-Be source data sheet. To avoid interactions between neutrons with air molecules in collimator box, a shield rod used. At the time of measurements samples were fixed in collimator box as adjacent neutron source. Detector was fixed combined with the exit of collimator box. It is satisfied that neutrons passed from samples arrived to detector probe. The Canberra NP100B series neutron detector was positioned front of circular hole where is surface of collimator box. First, the background measurements were

Bünyamin Aygün & Gökhan Budak

done 100 times. And then samples were put between collimator and detector. By this means 100 counts were done for each sample. Measurement results were read on 606M model transportable rate meter and system PC via RADACS software. Experimental details are illustrated in Fig 1.

Figure 1. Experimental design.

Theoretical The macroscopic cross section is a measure of neutron shielding capability of samples. Its symbol is Σ and unit of Σ is cm-1. The microscopic quantity that characterizes the neutron interaction with matter is called the cross-section with symbol σ. This describes the effective cross-sectional area to neutrons represented by each nucleus of the moderating material. The units are usually the barn where 1 barn is equivalent to 10-24 cm-2. The relationship between Σ and σ is: Σ = Nσ

(1)

where N is the number of nuclei per unit volume. Total cross-section value is obtained using by the various types of interactions of neutrons:

ΣTOTAL = Σ SCATTER + Σ CAPTURE + Σ FISSION + .....

(2)

Eq-3 is used to find total absorption. H = H 0e ΣTOTAL x

(3)

where H0 and H are some known value of equivalent dose rates respectively without sample between source-detector and there is a sample between source-detector situations. x is material thickness. Equivalent dose rate was

Use of paraffin wax against neutrons

used instead of beam intensity because of our equivalent dose rate measurements. Σ was determined for seven samples using by Eq-4. ⎛ H ⎞ ⎟⎟ ln⎜⎜ H Σ = ⎝ 0 ⎠ cm −1 x

(4)

The thickness of any given material where 50% of the incident equivalent dose rate has been attenuated is known as the half-value layer (HVL). The HVL is expressed in units of thickness (cm). We calculated HVL values for our specimens. HVL =

(5)

ln 2 0.693 = Σ Σ

Results and discussion In this paper, we determined neutron dose equivalent transmissions and macroscopic cross sections for seven paraffin samples having different oil concentrations. Results of measurements are shown in Fig.2. As can be seen 0.84 Macroscopic Cross Section Transmission

0.17

0.76

0.72

Transmission

Macroscopic Cross Section

0.80

0.13

0.68

0.09 0.00

4.00

8.00

12.00

0.64 16.00

Oil(%)

Figure 2. Transmissions and macroscopic cross sections as a function of oil percentages.

Bünyamin Aygün & Gökhan Budak

Fig.2, dose equivalent transmissions decrease with increasing oil percentage (from P1 to P7) and besides macroscopic cross section value increases with going from P1 to P7. It can be said that when oil content in paraffin is increase, dose of exposing detector decreases. Also HVL (half value layer) values calculated using Eq.-5. HVL and Σ values can be seen in Table 3. As can be seen Table.3, HVL values decrease with increasing oil percentage. We found an equation between dose transmission and oil percentage. T −1 = 1,4551526 +

0,17493342 P

where T is equivalent dose rate transmission

(6) ⎛H ⎞ ⎜ H ⎟ 0 ⎠ ⎝

and P is oil percentage.

Obtained experimental results and Eq-6 obviously show that sample encoded P7 is a finer fast neutron absorber than others. Table 3. HVL values of samples.

Conclusions Paraffin wax is one of the most usable neutron moderators. We investigated fast neutron dose equivalent transmission, macroscopic cross section and HVL values of seven different paraffin wax samples. Each of these samples has different oil percentages. If oil is added in paraffin wax, neutron moderation properties of paraffin wax includes because of its high hydrogen content. According to our experimental results and suggested equation, oil concentration in paraffin wax has positive impress on neutron absorption capability.

Use of paraffin wax against neutrons

Acknowledgements This work was supported by The Turkish Scientific and Technological Research Council of Turkey (TUBITAK) (Project No. 107T199), Atat端rk University (Project No. BAP 2008/86) and Mercan Kimya Corporation. The authors wish to thank to TUBITAK, Atat端rk University and Mercan Kimya Co.

References 1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13. 14. 15.

Hayashi, T., Tobita, K., Nakamori, Y., Orimo, S. 2009, Journal of Nuclear Materials, 386, 119. Yarar, Y., Bay端lken, A. 1994, Journal of Nuclear Materials, 212, 1720. Akaho, E.H.K., Jonah, S.A., Nyarko, B.J.B., Osae, S., Maakuu, B.T., SerforArmah, Y., Kyre A.W.K. 2002, Appl. Radiat. Isotopes 57, 831. Nieminen, M., Torsti, J.J., Valtonen, E. 1979, Phys. Scr., 20, 29. Uddin, M.S., Zaman, M.R., Hossain, S.M., Spahn, I., Sudar, S., Qaim, S.M. 2010, Appl. Radiat.Isotopes, 68-9, 1656. J坦zefowicz, K., Golnik, N., Zeilczynski, M. 1992, Radiat Prot Dosimetry, 44 (1-4), 139. Hashemi-Nezhad, S.R., Brandt, R., Westmeier, W., Westmeier, H., Wan, J.S., Vater, P. 2001, Radiat. Meas., 34 (1-6), 319. Seguchi, T., Hayakawa, N., Tamura, N., Katsumura, Y., Tabata, Y. 1991, Radiation Phys. Chem. 37(1), 141. Minsky, D.M., Valda, A.A., Kreiner, A.J., Green, S., Wojnecki, C., Ghani, Z. 2009, Appl. Radiat. Isotopes, 67, 179. Korkut, T., Karabulut, A., Budak, G., Korkut, H. 2010, J Radioanal Nucl Chem, 286, 61. Desdin, L., Ceballos, C. 2000, J Radioanal Nucl Chem, 243(3), 835. Fuga, P. 1990, J Radioanal Nucl Chem, 149(2), 287. El Abd, A. 2010, J Radioanal Nucl Chem., 284(2), 321. 241 Am-Be Radioactive Material Safety Data Sheet http://www.stuarthunt.com/pdfs/Americium_241Beryllium.pdf Canberra Data Sheet http://www.canberra.ru/html/products/healthphysics_rms/adm600/NPSeries.pdf

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5. Gamma and neutron shielding characteristics of concretes containing different colemanite proportions Osman Gencel 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. Osman Gencel, Department of Civil Engineering, Faculty of Engineering Bartin University, Bartin, Turkey. E-mail: osmangencel@gmail.com

Osman Gencel

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.

Radiation shielding properties of colemanite loaded concrete

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/cm3; 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 196-1 (1994) and EN 197-2 (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/m3. 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.

Osman Gencel

in Gencel et al. [7]. Slabs of 12Ă&#x2014;12Ă&#x2014;2 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 I0 (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 = I0 e -ax

(1)

where x is the thickness of the slab in cm.

Figure 1. Geometry setup.

Radiation shielding properties of colemanite loaded concrete

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 Âľ. 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 ÎŁR 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.

Osman Gencel

The particle source is a point source in air that is situated 20 cm away from the slab (target) along â&#x20AC;&#x201C;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 slabâ&#x20AC;&#x2122;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/cm3) is lower than that of the lime based aggregate (2.7 g/cm3). 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.

Radiation shielding properties of colemanite loaded concrete

Figure 3. Unit weights of concrete samples. Table 4. The measured and computed values of fast removal cross section (ÎŁR ) for neutrons.

Figure 4. Linear attenuation coefficients of samples for 30keV-30MeV energy range.

Osman Gencel

Table 4 lists the results of irradiation measurements and Monte Carlo simulations for fast removal cross sections of neutrons (ÎŁR; 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 ÎŁR values tend to decrease with density as mentioned above. Table 5. The measured* and computed values of linear attenuation coefficient (Âľ) for photons.

Radiation shielding properties of colemanite loaded concrete

Table 5 provides the results of Monte Carlo simulations for linear attenuation coefficient for photons (Âľ; in units of cm-1) in the energy range 30 keV â&#x20AC;&#x201C; 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 Kharita, M.H., Yousef, S., Al Nassar, M. 2009, Progr. Nucl. Energ., 51, 388. 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, 180. 4. Kharita, M.H., Takeyeddin, M., Al Nassar, M., Yousef, S. 2008, Progr. Nucl. Energ., 50, 33. 5. Demir, D. and Keles, G. 2006, Nucl. Instrum. Meth. B 245, 501. 6. Okuno, K. 2005, Radiat. Prot. Dosim., 115, 258. 7. Gencel, O., Brostow, W., Ozel, C., Filiz, M. 2010, Int. J. Phys. Sci. 5 (3), 216. 8. Gencel, O., Naziroglu, M., Celik, O., Yalman, K., Bayram, D. 2010, Biol. Trace. Elem. Res.,135, 253. 9. Cember, H. and Jhonson, T.E. 2008, Introduction to health physics, 4th 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-13709-M, Los Alamos National Laboratory, New Mexico. 12. Bozkurt, A., Yorulmaz, N., Kam, E., Karahan, G., Osmanlioglu, A.E. 2007, Radiat. Meas. 42, 1387. EN 196-1, (1994) Test Methods for Cement, CEN TC 51. EN 197-2, (2000) Conformity evaluation, CEN TC 30. 1. 2.

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Nuclear Science and Technology, 2012: 51-58 ISBN: 978-81-7895-546-9 Editor: Turgay Korkut

6. Estimation of neutron irradiation damages in Ni/n-GaAs Schottky contact layers via FLUKA Monte Carlo simulations 1

Hatun Korkut1, Turgay Korkut2, Hülya Doğan3 and Abdülmecit Türüt4

Department of Physics, Science and Art Faculty, Ağrı Ibrahim Çeçen University, Turkey; 2Department of Physics, Science and Art Faculty, Ağrı Ibrahim Çeçen University, Turkey; 3Department of Electronic Engineering Faculty, Sivas Cumhuriyet University, Turkey; 4Department of Physics, Science Faculty Erzurum Atatürk University, Turkey

Abstract. We estimated radiation effects of Ni/n-GaAs contact Schottky in neutron field by FLUKA Monte Carlo Code. Firstly, we determined elemental ingredients and layer thicknesses of diode by WDXRF spectroscopy technique. Then irradiation process was simulated in 1MeV neutron field for Ni/n-GaAs Schottky contact by FLUKA code. Results were interpreted for the purposes of Schottky diodes used in radiation applications.

Introduction Because of its successful electronic applications, Schottky contacts are used in a large field. It is an important issue that how Schottky diodes are affected by application conditions. Radiation application of Schottky contacts is a popular research field [1- 5]. FLUKA is able to simulate interactions and transports of many particles as hadrons, heavy ions and electromagnetic particles in a large perspective from Correspondence/Reprint request: Dr. Hatun Korkut, Department of Physics, Science and Art Faculty, Ağrı Ibrahim Çeçen University, Turkey. E-mail: hatunkorkut@hotmail.com

Hatun Korkut et al.

few keV (or thermal neutron) to high cosmic ray energies. The useful modern physics methods used in the development of the code have enhanced the usefulness of the code in different areas [6]. It is possible to predict lots of quantities from residual dose rates to activity of long-lived radionuclides by using FLUKA with an excellent accuracy [7]. Agosteo et al., studied direct interactions by irradiating a commercial PIN diode with thermal and fast neutron fields and compared experimental and FLUKA simulations for their solid state microdosimeter studies. Their study show that the consistency of the simulation results with the experimental data is satisfactory if the approximation made for the detector geometry is taken into account [8]. Neutron spectrometry was investigated with a recoil radiator-silicon detector device in another research. In the study the spectra of deposited energy were also calculated analytically via Monte Carlo simulations by FLUKA code. The effect of secondary charged particles produced by thermal and fast neutron interactions in the silicon diode was also investigated. FLUKA simulation of the deposited energies in the silicon diode for 2:7 MeV energetic neutrons and the results were compared to the experimental and analytical curves [9]. Wind investigated the energy response of RADFET for a wide spectrum of subatomic particles and photons from high energy photons, electrons and protons to neutrons by using FLUKA code [10]. Butterworth et al., estimated the radiation damage in the LHC cavities arising from beam gas collisions by FLUKA [11]. Korkut et al. recommended a new radiation shielding material by using FLUKA code [12]. Changing in electrical characteristics was studied depending on radiation effects on Schottky contacts in most studies. Interface spesific region has an effective role in the performance of Schottky contacts [13]. The change in Schottky diode characteristics depending on radiation environment is usually attributed to possible changes in metal-semiconductor interface region. But the origin of radiation effects in Schottky diode is not known in all its bearings. We used FLUKA simulation tool to see interactions in Schottky contact layers in the effect of 1 MeV energetic neutrons. In this stage the interaction of neutron irradiation with Ni/n-GaAs Schottky contact is discussed with respect to (dpa) displacements per atom and absorbed doses.

Experimental details In this paper, we used n-type GaAs wafer (Si-doped), (100) oriented with the free carrier concentration of 7.3x1015 cm-3 at room temperature conditions. In the chemical cleaning process the wafer was cleaned in trichloroethylene, acetone, and methanol for 3 minutes. After the chemical cleaning process, to remove the surface damages and undesirable impurities wafer was etched with

FLUKA simulations Ni/n-GaAs Schottky diodes

H2SO4:H2O2:H2O (5:1:1) for 1 minutes [14, 15]. For the fabricating of ohmic contact, a very small piece of indium was evaporated on the back side of the n-GaAs wafer and then the structure was annealed at 3000C for 3 minutes in N2 atmosphere to get low resistant ohmic contact. Ni evaporation was applied on the front face of the n-GaAs wafer as dots to get high quality Schottky contacts with the diameter of about 1.0 mm. Schottky contact layers are formed as seen in Fig.1. A wavelength-dispersive X-ray spectrometer (WDXRF, Rigaku ZSX-100e with Rhodium target X-Ray) was used to obtain elemental contents of diode. The sensitivity of WDXRF measurements is about ppm. As seen in Table.1, contents of diode were used in simulation process [16].

Figure 1. Layers and layer thicknesses of Ni/n-GaAs Schottky diode used in simulation process. Table 1. % Contents of Ni / n-GaAs Schottky Diode obtained from WDXRF results.

Hatun Korkut et al.

Simulation process Firstly thicknesses, densities and elemental contents (obtained from WDXRF measurements) of Schottky contact layers were entered FLUKA input file for primary 1 MeV neutron energies. Then simulation geometry has formed as seen in Fig 2. In this diagram Ni/ n-GaAs/In Schottky diode is in the center of simulation geometry. The Schottky contact is exposed to 1MeV neutrons along z axis. Schottky contact layers and their thicknesses are seen in Fig.1. As seen in Fig. 1, regions are labeled as below.

Figure 2. Simulation geometry of neutron irradiation processes in Ni/n-GaAs Schottky diode.

R1: Region 1; Black hole R2: Region 2; Vacuum R3: Region 3; Nickel, Schottky Metal Layer R4: Region 4; (Ni / n-GaAs), Interface Layer R5: Region 5; n-GaAs (Si Doped), Semiconductor Layer R6: Region 6; Indium, Ohmic Contact Metal Layer R7: Region 7; Vacuum We used NEW-DEFAults card to active some settings about simulation conditions (Inelastic form factor adjustments to Compton scattering were

FLUKA simulations Ni/n-GaAs Schottky diodes

activated. Particle transport and production threshold arrange as 10 MeV. Multiple scattering thresholds lowered to 20 MeV for secondary charged particles. Heavy particle bremsstrahlung activated with indicates photon production above 1 MeV- http://www.fluka.org/fluka.php?id=man_onl). USRBIN estimators were located the exits of each layers to calculate absorbed radiation doses and dpa values for each region. The program was run for 106 primaries for each irradiation processes. Dose values in each region (GeV/g/pr) are read in FLUKA output files. Uncertainties in simulation results are approximately %1.

Results and discussion Contact materials have precious place in the quality of Schottky contacts. Electrical conductivity of contact metal, carrier concrentation and band gap of semiconductor is impressive factors for electrical conduction mechanisms in Schottky diodes. Fabrication processes affect the chemical and physical properties of contacts as surface roughnesses and metal semiconductor interface transformations. The properties of surface and interface are directly interested in chemical cleaning and vacuum coating conditions. In this procedure, there are always been inadvertent effects. The all above reasons affect chemical ingredient rates and electrical dipole transformations of Schottky diodes. We assumed that these effects were in a minimum level in simulation process. Ni/ n-GaAs Schottky diode includes nickel as Schottky contact metal; indium as ohmic contact metal. n-type GaAs substrate is composed of gallium, arsenic and silicon. According to WDXRF results, there are four oxygenate compounds in interface oxide layer: Ga2O3, As2O3, NiO and SiO2 as shown in Table 1. We used the compositions, thicknesses and densities of Schottky contact layers in simulation process. WDXRF measurement sensitivity can affect the results of simulations. Small differences in content measurements may cause small differences in simulation results. Interface layer properties of Schottky contacts have an unneglectable importance in terms of conduction mechanism [13]. Fast neutrons produce a modification of the lattice structure and they affect the mechanical, electrical and other physical properties of irradiated materials. The incident particle kinetic energy may be transferred to the target atom in collision process. If the energy gets over a given transference threshold limit value, the target atom is displaced and then a stable vacancy pair and interstitial location (called Frankel pair) is created [18]. Neutrons cause non-ionizing effects in irradiated materials. Nonionizing energy can disturb the periodicity of the crystal. So, new deep states and trapping center formations are created into the material. The states can be changed by band

Hatun Korkut et al.

gap of the material [19]. If irradiated material is an electronic structure including transistors or diodes, all the components which compose the structure must be taken into consideration. Metals, basic semiconductor bulks, interface oxide regions, organic components etc. are keystones of electronic curcuit technology. Schottky diode, one of the most important elements of modern electronic world, is usually a multi-layered structure. Our sample (Ni/n-GaAs Schottky diode) has metal, semiconductor and interface oxide regions. We simulated 1 MeV neutron irradiation process for our sample. According to the simulation results, the interface layer absorbed maximum radiation doses as compared to the other layers. Absorbed doses by the each layer were shown in Fig.3. Dpa (displacements per atom) is a measure for damage amount for irradiated materials. Displacement damage can be produced by all the particles. There is a direct relation between the dpa value and the total numbers of defects (or Frenkel pairs). When x atom in the material has been displaced from its position, x quantitiy of dpa has been located in the lattice of the material [17]. Maximum absorbed doses caused the maximum quantity of dpa in interface layer as can be seen in Fig.3 and Fig.4 by the reason of 1 MeV energetic neutron irradiation. And so new interface dipole transformations were located in the interface layer. This result is consistent with experimental results and Tungâ&#x20AC;&#x2122;s theory [1, 13]. Akkurt et al., irradiated their Schottky diode with neutrons. They observed that electrical characteristics of diode changed with the effect of neutron field. Radiation damages were not only located in the interface layer, but also in the other layers as can be seen in Fig.4. There was a characteristic change in all the layers of electronic structure of Ni/n-GaAs Schottky diode due to neutron irradiation. Interface (0,060223 GeV/g/pr) Indium (0,001680 GeV/g/pr)

96%

0,62% 2,7% 0,34% GaAs

(0,000389 GeV/g/pr)

Nickel (0,000211 GeV/g/pr)

Figure 3. Absorbed doses for neutron irradiation by layers of Schottky diode.

FLUKA simulations Ni/n-GaAs Schottky diodes

Interface (9.4756E-18 dpa) Indium (1.0071E-19 dpa)

96%

0,97% 1% 1,9% GaAs (9.5683E-20 dpa) Nickel (1.8378E-19 dpa)

Figure 4. Dpa values for neutron irradiation by layers of Schottky diode.

Conclusion We fabricated Ni/n-GaAs Schottky contact in the form of Fig.1 and determined its elemental contents by WDXRF. FLUKA Monte Carlo Code was used to simulate 1 MeV energetic neutron irradiation process. Neutron field effect of dpa (displacements per atom) quantities and absorbed doses of the Ni/n-GaAs Schotttky contact layers were calculated by FLUKA for 1 MeV energetic neutrons. By way of addition, the implicit role of interface region in radiation field was emphasized via its maximum dose absorbtion and maximum dpa quantity property. These results enable us a foresight on neutron field induced electrical dipole changes in the interface layers. As a result, FLUKA simulations provide more efficient properties for Schottky contacts in different radiation conditions. To be able to estimate radiation effects on materials give us a proper way for choosing contact materials in the optional radiation applications. So, possible negative effects can be minimized under different radiation conditions.

Acknowledgments This investigation was supported by Ağrı İbrahim Çeçen University (BAP-F07). Experimental studies were done in Ataturk University Laboratories. The authors wish to thank to Ağrı İbrahim Çeçen University and Ataturk University.

References 1. 2.

Akkurt, I., Akyildirim, H., Özdemir, A.F., Aldemir, D.A. 2010, Rad Meas. 45 1381. Baranwal, V., Kumar, S., Pandey, A.C., Kanjilal, D. 2009, J Alloy Compd 480 962.

3. 4. 5. 6.

7. 8. 9. 10. 11. 12. 13. 14. 15. 16. 17. 18. 19.

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Çınar, K., Coşkun, C., Aydoğan, Ş., Asıl, H., Gür, E. 2010, Nucl Instrum Meth B 268 621. Güllü, Ö., Aydoğan, Ş., Şerifoğlu, K., Türüt, A. 2008, Nucl Instrum Methods Phys Res A 593 544. Uğurel, E., Aydoğan, Ş., Şerifoğlu, K., Türüt, A. 2008, Microelectron Eng 85 2299. Ballarini, F., Battistoni, G., Brugger, M., Campanella, M., Carboni, M., Cerutti, F., Empl, A., Fassò, A., Ferrari, A., Gadioli, E., Garzelli, M. V., Lantz, M., Mairani, A., Mostacci, A., Muraro, S., Ottolenghi, A., Patera, V., Pelliccioni, M., Pinsky, L., Ranft, J., Roesler, S.,Sala, P. R., Scannicchio, D., Smirnov, G., Sommerer, F., Trovati, S., Villari, R., Vlachoudis, V., Wilson, T., Zapp, N. 2007, Adv Space Res 40 1339. Brugger, M., Ferrari, A., Roesler., S., Ulrici, L. 2006, Nucl Instrum Methods Phys Res A 562 814. Agosteo, S., Fallica, P.G., Fazzi, A., Pola, A., Valvo, G., Zotto, P. 2005, Appl. Radiat. Isotopes 63 529. Agosteo, S., Birattari, C., D’Angeloa, G., Dal Corso, F., Para, A.F., Lippi, I., Pola, A., Zotto, P. 2003, Nucl Instrum Methods Phys Res A 515 589. Wind, M., Beck, P., Jaksic, A. 2009, IEEE T Nucl Sci 56-6 3387. Butterworth, A., Ferrari, A., Tsoulou, E., Vlachoudis V., Wijnands, T. 2005, Radiat Prot Dosim 116 (1–4) 521. Korkut, T., Korkut, H., Karabulut, A., Budak, G. 2011, Ann Nucl Energy 38 56. Tung, R.T.. 2001, Phys Rev B 64 205310. Sugawara, S., Saito, K. Yamauchi, Y., Shoji. M. 2001, Jpn. J. Appl. Phys. 40 6792. Wang, H.T., Chang, L.B., Cheng, Y.C., Lin, Y.K., Hsu. C.I.G. 1999, Cryst. Res. Technol. 34-8 1017. Doğan, H. 2006, Ph. D. Thesis, Department of Physics Atatürk University Graduate School of Natural and Applied Sciences. Vlachoudis, V., Smirnov, G., Ferrari, A. 27 Nov 2008, FLUKA Users Meeting. Vladimirov, P., Bouffard, S. 2008, R.Physique 9 303. Almaz, E., Stone, S., Blue, T.E., Heremans, J.P. 2010, Nucl Instrum Methods Phys Res A 622 200.

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Nuclear Science and Technology, 2012: 59-69 ISBN: 978-81-7895-546-9 Editor: Turgay Korkut

7. Measurement of mass attenuation coefficients by Si(Li), NaI(Tl) and Cd(Tl) detectors 1

Mustafa Recep Kaçal1, İbrahim Han2 and Ferdi Akman3

Giresun University, Faculty of Arts and Sciences, Department of Physics, 28100 Giresun, Turkey Ağrι İbrahim Çeçen University, Faculty of Sciences and Arts, Department of Physics, 04100 Ağrι Turkey; 3Bingöl University, Faculty of Arts and Sciences, Department of Physics, 12000 Bingöl, Turkey 2

Abstract. Mass attenuation coefficients of the some elements were determined for two photon energies using different three detectors. The samples were irradiated using 10 mCi Cd-109 and 100 mCi Am-241 radioactive point sources. The photons were separately counted by Si(Li), NaI(Tl) and Cd(Tl) detectors. It was observed that the mass attenuation coefficient decreases with increasing energy and increases with increasing atomic number. The measured values are compared with the theoretical ones calculated using WinXcom program.

Introduction This chapter concerns determination of photon attenuation or absorption properties of a material using the application of Lambert−Beer’s law with standard transmission method by adopting narrow beam geometry. The photon attenuation properties of a material can be evaluated by means of the linear attenuation coefficients (μ), mass attenuation coefficients (μ/ρ) and related Correspondence/Reprint request: Dr. Mustafa Recep Kaçal, Giresun University, Faculty of Arts and Sciences Department of Physics, 28100 Giresun, Turkey. E-mail: mustafakacal@hotmail.com

Mustafa Recep Kaçal et al.

parameters such as mean free path (mfp) and half value layer (x1/2) etc. In this method, these parameters are determined using from the spectrum obtained by sample and without sample each measurement. Photon spectra are recorded in the following order: firstly, source spectrum recorded with source but without sample and the incident spectrum (without attenuation) is obtained. The transmitted spectrum recorded with source and sample and the transmitted spectrum I (after attenuation) are obtained. In both the spectra the photo-peak had Gaussian distribution. Finally, by integrating the incident spectrum and the transmitted spectrum over selected width of the photo-peak, incident intensity I0 and transmitted intensity I are obtained. Generally, Si(Li), NaI(Tl) or Cd(Tl) detectors are used for count of photons. In this study for corporation all of these are used. When an x-ray beam passes through any matter, its intensity progressively reduces as a consequence of a complex series of interactions between x-ray photons and atoms of the attenuating medium. The linear attenuation coefficient (μ , cm-1) is defined as the probability of a radiation interacting with a material per unit path length [1]. It is related mass attenuation coefficient (μ/ρ gcm-2). The μ/ρ is a measure of the average number of interactions that occur between photons and matter mass per unit area. The accurate attenuation coefficient values of materials are a very essential parameter in nuclear and radiation physics, radiation dosimetry, radiography, spectrometry, crystallography, biological, medical, agricultural, environmental and industrial.

Historical background and current status of topic Since the mass attenuation coefficients are important in fundamental physics and many applied fields, the accurate values of mass attenuation coefficients for X- and γ- rays in several materials are essential for some fields such as, nuclear, radiation physics, radiation dosimetry, biological, medical, agricultural and industrial. Recently, there are a great number of experimental and theoretical investigations of mass attenuation coefficient. The mass attenuation coefficients for C2H4, CO2, N2, O2, CF4, Ne, H2S, HCl, Ar, Air, Mg, Al, SiO2, (C2H5)3PO4 materials have been determined by [2]. Hubbell presented tables of mass attenuation coefficients for 40 elements and 45 mixtures and compounds over the energy range from 1 keV to 20 MeV [3]. These tables were be reiterated with tabulation for all elements in the atomic range 1 ≤ Z ≤ 92 and 48 additional substances of dosimetric interest by the Hubbell and Seltzer [4] and Berger and Hubbell [5]. The photon attenuation coefficients in certain tissue equivalent compounds, perspex,

Mass attenuation coefficients

polyethylene, polycarbonate and teflon have been measured at energies 13.37, 17.44, 22.10, 32.06 and 44.23 keV [6]. Wang et al. measured systematically the mass attenuation coefficients in the range of X-ray energies between 1.486keV and 15.165 keV for SiH4 and; between 8.041 keV and 29.109 keV for Si [7]. Khanna et al. have measured γ-ray attenuation coefficients in some heavy metal oxide borate glasses at 662 keV [8]. Mass attenuation coefficients of 59.54 keV photons were measured in elements with atomic number ranging from low to high, and including lanthanides whose K-shell binding energies were close to those of incident photons [9]. Sing et al. determined attenuation coefficients for some dilute solutions (Li2SO4H2O, CuSO4.5H2O, NiSO4.6H2O, MgSO4.7H2O, NH4Cl) at 662 keV [10]. Orlic et al. published total photon mass attenuation coefficients for photon energies between 100 eV and 1000 MeV [11]. Measurements have been made to determine γ-rays attenuation coefficients very accurately by using an extremely narrow-collimated-beam transmission method and the effect of the sample thickness on the measured values of the mass attenuation coefficients (μ/ρ) cm2/g of perspex, bakelite, paraffin, Al, Cu, Pb and Hg have been investigated at three different γ-ray energies (59.54, 661.6 and 1332.5 keV) [12]. The transmissions of γ- rays at the energies, 81, 356, 511, 662, 835, 1274 and 1332 keV have been studied on the alloys brass, bronze, steel, aluminum-silicon and lead-antimony [13]. The mass attenuation coefficients for 22 high purity elemental materials (C, Al, Ti, V, Mn, Fe, Co, Ni, Cu, Zn, Zr, Nb, Mo, Rh, Pd, Ag, Cd, In, Ta, Pt, Au and Pb) were measured in the X-ray energy obtained by a variable-energy X-ray source range from 13 keV to 50 keV using a high purity germanium detector with thin (50 mm) Be window [14]. Angelonea et al. were measured the total absorption coefficients for some selected organic compounds relevant to health physics, Triaflol BN (C3H4O2)n, Triaflol TN (C12H18O7)n, Kapton (C44H20O10)n, and Melinex (C10H8N4O4)n in the X-ray energy range from 13 keV up to about 40 keV using a collimator, high purity germanium detector with thin Be window and variable energy X-ray source [15]. Söğüt et al. measured the total mass attenuation coefficients of Fe and Cu in various compounds using a Si(Li) detector with a resolution of 155 eV at 5,9 keV [16]. Turgut et al. were measured the total mass attenuation coefficients for the elements Co, Mn and Co2O3, compounds CoCl2.6H2O, CoSO4, CoSO4.7H2O, MnCO3, KMnO4, MnCl2.2H2O and MnCl2.4H2O at different energies between 4.508 and 11.210 keV using a secondary excitation method [17]. The μm values around the K-shell absorption edge of Nb, Zr and Mo as a parametric X-ray radiation (PXR) application of monochromatic hard X-ray radiation sources have been

Mustafa Recep Kaçal et al.

measured [18]. The gamma-ray attenuation coefficients for bismuth borate glasses were measured [19]. İçelli et al. were measured the mass attenuation coefficients for V2O3, VO2, VF3, NH4VO3, VF4, NiF2, NiCl2, NiCl2H2O, NiF24H2O, NiCl26H2O and Ni (ClO4)26H2O in the X-ray energy range from 15.746 to 40.930 keV using a Si (Li) detector [20]. İçelli et al. were measured to determine variation of the mass attenuation coefficients of H3BO3 according to percentage increasing concentration of H3BO3 by using an extremely narrow-collimated-beam transmission method in the energy range 15.746-40.930 keV with an X-ray transmission method [21]. The μm values of Ag in the 15–50 keV energy range with a level of uncertainty between 0.27% and 0.4% away from the K-edge were evaluated [22]. The X-ray linear attenuation coefficients was measured using characteristic X-rays with energies 32–66 keV produced by X-ray fluorescence using a secondary target system, and 140 keV γ-rays obtained from an unsealed Tc source for materials containing elements from hydrogen to calcium [23]. Measurements have been made to determine the mass attenuation coefficients of undoped ntype InSe, and Gd, Ho, Er doped n-InSe single crystals using a Si(Li) detector in the energy region 15.746-40.930 keV X-ray energies with energy dispersive X-ray fluorescence systems [24]. Chitralekha et al. have been measured mass attenuation coefficients for mono- and disaccharides at photon energies 5.947, 6.460 and 14.413 keV [25]. The μ m values for YBaCuO and BiPbSrCaCuO superconductors at 511keV, 661keV and 1274 keV energies and for MgB2 superconductor at some energy between 14.1keV and 29.7keV have been measured [26, 27]. The μ m values for BiPbSrCaCuO superconductor have been measured at different energies [28]. The X-rays attenuation coefficients for Cu, In and Se in elemental state and the semiconductor CuInSe2 were measured at 15 different energies from 11.9 to 37.3 keV by using the secondary excitation method [29]. Rettschlag et al. determined plutonium photon mass attenuation coefficients by using a collimated-beam transmission method in the energy range from 60 keV to 2615 keV [30]. The mass attenuation coefficients for cornea taken from keratitis patient and soft contact lens (-1.75, -3.75, -4 diopties), leiomyomata uteri and uterus were measured in the X-ray energy (5.9 keV) using a Si(Li) detector and Fe-55 annular source [31]. The mass attenuation coefficients were measured for various binary and ternary 3d transition metal alloys at different energies [32, 33, 34]. The total mass attenuation coefficients (for GaAs, GaAs (semi-insulating; S-I) GaAs:Si (N+), GaAs:Zn, InP:Fe, InP:Fe– As, InP:S and InP:Zn crystals were measured at various photon energies [35]. The mass attenuation coefficients (μm) for SiO2 {Quartz (1101), Quartz

Mass attenuation coefficients

(110 0) and Quartz (0 0 01)}, KAlSi3O8 {Orthoclase (010), Orthoclase (10 0)}, CaSO4.2H2O (gypsum), FeS2 (pyrite) and Mg2Si2O6 (pyroxene) natural minerals at 22.1, 25.0, 59.5 and 88.0 keV photon energies [36]. The total mass attenuation coefficients (μ/ρ) for pure Au and Au99Be1, Au88Ge12, Au95Zn5 alloys were measured at 59.5 and 88.0 keV photon energies [37]. Özdemir and Kurudirek determined total mass attenuation coefficients for 21 different compounds at 59.54 keV using a narrow beam good geometry setup [38]. The mass attenuation coefficients of the 59.54 keV radiation of Am241 point source in boron ores such as tincal, ulexite and colemanite were determined experimentally by a scintillation detector and theoretically [39]. Sharanabasappa et al. were measured mass attenuation coefficients for chromium and manganese compounds around absorption edge and for magnesium, nickel, copper, molybdenum and tantalum and three biological equivalent materials are compared with standard theoretical values [40, 41]. The mass attenuation coefficients, partial interactions and the effective atomic numbers (Zeff) of Bi2O3, PbO and BaO in xRmOn: (100-x) P2O5 (where x=30≤x≤70 (%by weight)) glass system have been investigated on the basis of the mixture rule at 662 keV [42]. Un et al. were determined the total mass attenuation coefficients, μm, for PbO, barite, colemanite, tincal and ulexite at 80.1, 302.9, 356.0, 661.7 and 1250.0 keV photon energies by using NaI(Tl) scintillation detector [43]. The photon attenuation coefficients of concrete includes barite in different rate were measured [44]. Kirdsiri et al. were measured photon attenuation coefficients of silicate glasses containing Bi2O3, PbO and BaO for comparison of radiation shielding and optical properties [45]. Gamma-ray attenuation coefficients of some building materials available in Egypt and photon attenuation parameters in different solid state track detectors in the energy range 1 keV–100 GeV were determined [46, 47]. In the present study, mass attenuation coefficients of V, Fe, Ni, Zn, Mo, Ag, Sb, Nd, Gd, Dy, Pt, Hg, Tl, Pb, Bi and U at 59.5 keV and 88 keV photon energies were measured by Si(Li), NaI(Tl) and Cd(Tl) detectors. The measured mass attenuation coefficients are compared with the theoretical values.

Detectors Semiconductor detector is fabricated from either elemental or compound single crystal materials, having a band gap of approximately 1 to 5 eV. The group IV elements silicon and germanium are by far the most widely-used semiconductors, although some semiconductors materials are finding use special applications as development work on them continues. Semiconductor

Mustafa Recep Kaçal et al.

detectors have a p-i-n diode structure, in which the intrinsic region is created by depletion of charge carriers when a reverse bias is applied across the diode. When photons interact within the depletion region, charge carriers (holes and electrons) are freed and are swept to their respective collecting electrode by the electric field. The resultant charge is integrated by a charge sensitive preamplifier and converted to a voltage pulse with amplitude proportional to the original photon energy. The advantage of semiconductor detectors is their excellent energy resolution; a disadvantage is that semiconductors usually have complex energy spectra showing the effect of charge carrier trapping and prominent K X-ray escape peaks. Trapping of either holes or electrons at sites of crystal defects and impurities, results in reduced efficiency of charge collection [48]. Sodium Iodide is an instrument for detecting X-rays and gamma rays. NaI is a kind of scintillation crystal with favorite properties, crystalline sodium iodide (NaI) converts an X-ray or gamma-ray photon into a pulse of visible light, the high Z of iodine in NaI gives good efficiency for gamma ray detection, and a small amount of Tl is added in order to activate the crystal, so the designation is usually NaI (Tl) for the crystal. The thallium activated sodium iodide scintillation detector produces energetic recoil electrons from photon interactions inside the crystal, which undergo inelastic collisions within the lattice to excite secondary electrons to the conduction band. Some of these electrons subsequently de-excite via the emission of photons in the visible region. This scintillation light is measured by a photomultiplier tube, where the photoelectrons are accelerated and multiplied through secondary emissions from a number of dynodes. The current pulse from the photomultiplier tube is passed through a charge sensitive preamplifier and fed to a shaping amplifier which is connected to a multi-channel pulse height analyzer (MCA) [49]. Cadmium Telluride (CdTe) detector has a combined high atomic numbers, with a good band gap energy for room temperature operation (1.5eV), and the probability of photoelectric effect in CdTe is 4-5 times higher than in Ge for typical gamma ray and X-ray irradiation on the cathode side, although CdTe detectors are efficient for low energy gamma rays. X-rays and gamma rays interact with CdTe atoms to create an average of one electron/hole pair for every 4.43 eV of energy lost in the CdTe. Depending on the energy of the incoming radiation, this energy loss is dominated by either the photoelectric effect or Compton scattering. The probability or efficiency of the detector to “stop” the incoming radiation and create electron/hole pairs increases with the thickness of CdTe [49]. The charge generates in CdTe, as an incident γ-ray or X-ray interacts with the semiconductor device and generates a number of

Mass attenuation coefficients

electron-hole pairs. These electron-hole pairs drift under the influence of an applied electric field and induce a charge Q on the electrodes, the induced charge Q is converted into a voltage pulse using a charge or current sensitive preamplifier, where ideally the output voltage is proportional to the initial deposited energy.

Experimental The experimental arrangement used in the present work is shown in Fig. 1. 59.5 keV and 88 keV Îł-rays of Cd-109 (10mCi) and Am-241 (100mCi) radioactive point sources were used to excite the targets, respectively. The intensities of fluorescent X-rays were measured using a high-resolution Si(Li), Cd(Tl) and NaI(Tl) detectors. The Si(Li), Cd(Tl) and NaI(Tl) detectors have a resolution of 160 eV at 5.90 keV, 1,5 keV at 122 keV and 8.5 keV at 662 keV, respectively. Spectroscopically pure foil samples of mass thickness ranging from 0.017 to 0.727 g/cm2 were used. The samples were placed in between the Îł-ray source and the detector. The detector absorbs a narrow beam of gamma rays passed through the sample. To minimize the effect of small angle scattering in the target transmitted gamma rays were further collimated. The spectra were obtained for a period 600s for each one of energy and elements. The measurements were repeated 5 times and took average of these measurements in order to minimize statistical error. A typical spectrum is given in Fig. 2.

Figure 1. Experimental setup.

Mustafa Recep Kaçal et al. 5

4.0x10

3.5x10

without absorber with absorber

3.0x10

Counts

2.5x10

2.0x10

1.5x10

1.0x10

5.0x10

0.0 57.5

58.0

58.5

59.0

59.5

60.0

60.5

Energy

Figure 2. The spectra of Am-241 radioactive point source without absorber and with absorber (Ti).

Mass attenuation coefficients for the elements at incident energies are determined by the transmission for collimated mono-energetic beam. If a material of thickness x (cm) is placed in the path of a beam of gamma or X-ray radiations, the intensity of the beam will be attenuated according to Beer–Lambert’s law: (1) where I0 and I are the unattenuated and attenuated photon intensity, respectively, μm(cm2/g) is the mass attenuation coefficient of the material and t(g/cm2) is sample mass thickness. Mathematical rearrangement of Eq. (1) yields the following equation for the mass attenuation coefficient: (2) The mass attenuation coefficients (μm) values for present elements obtained experimentally in the present investigation and a comparison with the theoretical values obtained by WinXcom program [14]. This program depends on applying the mixture rule to calculate the partial and total mass attenuation coefficients for all elements, compounds and mixtures at standard as well as selected energies.

Mass attenuation coefficients

Table 1. The experimental and theoretical values of mass attenuation coefficients Îźm (cm2/g).

Discussion The experimental and theoretical results for the mass attenuation coefficients (Îźm) are presented in Table 1. It is clearly seen from Table 1 that the mass attenuation coefficients depends on the photon energy. The mass attenuation coefficients of materials are decrease with increasing photon energy. As shown in Table 1, the experimental mass attenuation coefficients for almost all samples are agreement theoretical values of ones. The theoretical mass attenuation coefficients were calculated using XCOM program. The total experimental uncertainty of the measured mass attenuation coefficients depends on the uncertainties of the evaluation of peak area of I0 intensity without attenuation and I intensity after attenuation, mass thickness measurements and counting statistics. I0 intensity without attenuation and I intensity after attenuation for 59.54 keV and 88 keV

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photons emitted Cd-109 (10mCi) and Am-241 (100 mCi) radioactive point sources were counted by Si(Li), NaI(Tl) and Cd(Tl) detectors. Typical total uncertainty in the measured experimental mass attenuation coefficients is estimated to be 1–4%.

References 1. 2. 3. 4. 5. 6. 7. 8. 9. 10. 11. 12. 13. 14. 15. 16. 17. 18. 19. 20. 21.

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