T.C. SÜLEYMAN DEMİREL UNIVERSITY GRADUATE SCHOOL …

T.C.

SÜLEYMAN DEMİREL UNIVERSITY

GRADUATE SCHOOL OF NATURAL AND APPLIED

SCIENCES

RADIATION SHIELDING PROPERTIES FOR SOME

COMPOSITE MATERIALS

Abdlhamed Faaq Abdlhamed ABDLHAMED

Supervisor

Prof. Dr. İskender AKKURT

Co-Supervisor

Asst. Prof. Dr. Zeynep PARLAR

THE DEGREE OF MASTER OF SCIENCE

DEPARTMENT OF PHYSICS

ISPARTA - 2018

i

TABLE OF CONTENTS

Page

TABLE OF CONTENTS .............................................................................................. i

ABSTRACT ................................................................................................................ iii ÖZET........................................................................................................................... iv

ACKNOWLEDGEMENTS ......................................................................................... v LIST OF FIGURES .................................................................................................... vi LIST OF TABLES .................................................................................................... viii 1. INTERODUCTION ................................................................................................ 1

1.1.Radiation and Properties ................................................................................... 1

1.2.The Types of Radiation ..................................................................................... 1 1.2.1. The Non-Ionizing Radiation ................................................................... 2

1.2.2. Ionizing Radiation .................................................................................. 2

1.2.2.1. Alpha (α) – Decay ........................................................................ 4 1.2.2.2. Beta (β)–Decay ............................................................................. 4 1.2.2.3. Neutron ......................................................................................... 6

1.2.2.4. Gamma ( )-Decay ........................................................................ 7

1.3. Gamma Interaction with Matter ....................................................................... 8 1.3.1. Compton Scattering ................................................................................ 8

1.3.2. Photoelectric Effect ................................................................................ 9 1.3.3. Pair Production ..................................................................................... 10

1.4. Radiation Measurements ................................................................................ 11 1.4.1. Gas-Filled Detectors ............................................................................. 12 1.4.2. Scintillation Detectors .......................................................................... 13

1.4.3. Semiconductor Detectors (Solid State Ionization Chambers) .............. 14 1.5.Biological Effects of Radiation ....................................................................... 15

1.5.1. Type of effects ...................................................................................... 15 1.5.1.1. Somatic Effects .......................................................................... 17

1.5.1.2. Genetic Effects ........................................................................... 18 1.6. Radiation Protection ....................................................................................... 19

1.6.1. Time ...................................................................................................... 19 1.6.2. Distance ................................................................................................ 19 1.6.3. Shielding ............................................................................................... 20

1.7. The Composite ............................................................................................... 21

1.7.1. Rubber .................................................................................................. 21 1.7.2. Jute ........................................................................................................ 23 1.7.3. Glass Fiber ............................................................................................ 24

2. THE LITERATURE BACKGROUND ................................................................ 26 3. MATERIALS AND METHODS .......................................................................... 31

3.1. Preparation Samples ....................................................................................... 31

3.2. Methods .......................................................................................................... 33

3.2.1. Gamma spectrometer system ............................................................... 33 3.2.2. NaI (TI) Detector .................................................................................. 34 3.2.3. Electronic units ..................................................................................... 35 3.2.4. Energy calibration ................................................................................ 35 3.2.5. Detection efficiency calibration ........................................................... 36

3.3. The Linear attenuation Coefficients Measurement ........................................ 37 4. RESULTS .............................................................................................................. 40 5. CONCLUSION ..................................................................................................... 52

ii

REFERENCES ........................................................................................................... 53

CURRICULUM VITA .............................................................................................. 57 ÖZGEÇMİŞ ............................................................................................................... 58

iii

ABSTRACT

M.Sc. Thesis

RADIATION SHIELDING PROPERTIES FOR SOME COMPOSITE

MATERIALS


Süleyman Demirel University

Graduate School of Natural and Applied Sciences

Department of Physics

Supervisor: Prof. Dr. İskender AKKURT

Co-Supervisor: Asst. Prof. Dr. Zeynep PARLAR

Radiation since its discovery has been one of the facts of life. Radiation has always

been existed in universe even it is due to the different sources. Radiation is a

important research field as it is used in a variety of different field and also it is

hazardous for helath. Therefore development of radiation shielding materials has

been recently a vital research field. Thus development of a composite materials to

sheild radiation and testing it agains gamma-ray are main goal of this thesis.

Composite materials have been developed in Istanbul Technical University and they

were tested against gamma rays from 60Co and 137Cs sources in Suleyman Demirel

University.

Keywords: Radiation, Shielding, Composite Materials, Radiation Protection.

2018, 58 pages

iv

ÖZET

Yüksek Lisans Tezi

BAZI KOMPOZİT MALZEMELERİN RADYASYONA KARŞI KORUMA

ÖZELLİKLERİ


Süleyman Demirel Üniversitesi

Fen Bilimleri Enstitüsü

Fizik Anabilim Dalı

Danışman: Prof. Dr. İskender AKKURT

II. Danışman: Yrd. Doç. Dr. Zeynep PARLAR

Radyasyon keşfinden bu yana hayatın gerçeklerinden biri olmuştur. Radyasyon

kaynağı gelişen teknoloji ile değişmekle birlikte doğada hep varolmuştur. Günlük

hayatta birçok alanda kullanılıyor olması yanında sağlık açısından zararlı etkilerinin

olması radyasyondan korunmayı önemli bir araştırma alanı yapmıştır. Bu nedenle

radyasyondan korunmak amaçlı olarak zırh malzemesi geliştirilmesi son zamanların

önemli konularından biridir. Bu yüzden zırh materyalinin (kompozit) özelliklerini ve

performansını değerlendirmek için bu çalışmanın temel amacıdır. İstanbul Teknik

Üniversitesinde geliştirilen zırh malzemeleri Suleyman Demirel Üniversitesinde 60Co

ve 137Cs gamma kaynakları ile test edilmiştir.

Anahtar Kelimeler: Radyasyon, Zırhlama, Kompozit Malzemeler, Radyasyondan

Korunma.

2018, 58 sayfa

v

ACKNOWLEDGEMENTS

I would like to thank God who helped me during all the steps of my study and gave

me the power and necessary patience to complete my thesis. I take this opportunity to

thank those people who were necessary for doing this work, and I am proud of them.

I am very grateful to my supervisor Prof. Dr. Iskender Akkurt for his expert

supervision of this work. His invaluable comments on the chapters of the thesis

enrich the whole work and for his words that always encourage me to work hard, and

I would like to thank him for all the recommendations during my study with him.

Also, I would like to thank my second supervisor, Asst. Assoc. Dr. Zeynep PARLAR

for her enthusiastic support of my project contributions and standing by my side in

the work period related to the experimental part of the composite work of my thesis.

With great thanks and respect.

I would like to thank Asst. Prof. Dr. Kadir GÜNOĞLU very much for everything he

did it for me, who helped me without fail during the study and measurements of

samples. My sincere greetings and appreciation.

And you, my dear family, you most of all deserve my gratitude. My dear mother,

brothers, and sisters; I give you my endless thanks for all your patience, support, and

love. I am lucky to have you. Yours affectionately.

Finally, I would like to thank all the friends who have stood and continue to support

me during my scientific career, you have all my love and appreciation.


ISPARTA, 2018

vi

LIST OF FIGURES

Page

Figure 1.1. The electromagnetic radiations .................................................................. 2 Figure 1.2. Ionizing processes ...................................................................................... 3

Figure 1.3. Damage caused by radiation in DNA ........................................................ 3 Figure 1.4. The alpha particle ...................................................................................... 4 Figure 1.5. Beta minus (β-) emission ........................................................................... 5 Figure 1.6. Beta plus (β+) emission .............................................................................. 5 Figure 1.7. Neutrons Radiation .................................................................................... 7

Figure 1.8. Gamma decay ............................................................................................ 8 Figure 1.9. The mechanism of Compton scattering ..................................................... 9 Figure 1.10. Schematic of the photoelectric absorption process ................................ 10 Figure 1.11. Schematic of the pair production process and annihilation ................... 11 Figure 1.12. Schame of Scintillation Counter ............................................................ 14

Figure 1.13. Radiation effects on the whole body system ......................................... 17 Figure 1.14. Comparison of penetration of various radiations................................... 21

Figure 1.15. Rubber used in the composite ................................................................ 23 Figure 1.16. Jute used in the composite ..................................................................... 24 Figure 1.17. Glass-Fiber used in the composite ......................................................... 25 Figure 3.1. Samples composite .................................................................................. 31

Figure 3.2. Cutting and compressing devices ............................................................ 32 Figure 3.3. Heating procedure .................................................................................... 32

Figure 3.4. Schematic view of gamma Spectrometer and electronic units ................ 33 Figure 3.5. Schematic view of NaI (Tl) detector ....................................................... 34 Figure 3.6. Photography of NaI (Tl) detector ............................................................ 34

Figure 3.7. The decay scheme of 137Cs ...................................................................... 35 Figure 3.8. The decay scheme of 60Co ....................................................................... 36

Figure 3.9. Energy spectrum and related fit ............................................................... 36

Figure 3.10. Detection efficiency of NaI (Tl) detector as a function of gamma-ray

energies for 0,5 cm distance to detector face ........................................ 37 Figure 3.11. I and Io are measured count in detector respectively with and without

the absorber of thickness x (cm) ........................................................... 38

Figure 3.12. Intensity of gamma rays with and without materials for 60Co and 137C

sources ................................................................................................... 39

Figure 4.1. The relationship between 𝝁 (cm-1) and energy (keV) for all composite

samples .................................................................................................... 43

Figure 4.2. The relationship between 𝝁 (cm-1) and energy (keV) for R100, R90-J10,

R80-J20, and R70-J30 ............................................................................. 44

Figure 4.3. The relationship between 𝝁 (cm-1) and energy (keV) for R100, R90-

GF10, R80-GF20, and R70-GF30 ........................................................... 44

Figure 4.4. The relationship between 𝝁 (cm-1) and energy (keV) for R100, R70-J10-

GF20, R70-J15-GF15, and R70-J20-GF10 ............................................. 45

Figure 4.5. The relationship between 𝝁 (cm-1) and jute rate ...................................... 45

Figure 4.6. The relationship between 𝝁 (cm-1) and glass-fiber rate ........................... 46

Figure 4.7. The 𝜇 (cm-1) as a function of energy and also jute and glass-fiber rates for

R100, R70-J10-GF20, R70-J15-GF15, and R70-J20-GF10 ................... 46 Figure 4.8. The relationship between mfp (cm) and energy (keV) for R100, R90-J10,

R80-J20, and R70-J30 ............................................................................. 47 Figure 4.9. The relationship between mfp (cm) and energy (keV) for R100, R90-

GF10, R80-GF20, and R70-GF30 ........................................................... 47

vii

Figure 4.10. The relationship between mfp (cm) and energy (keV) for R100, R70-J10

GF20, R70-J15-GF15, and R70-J20-GF10 ........................................... 48 Figure 4.11. The relationship between transmission rate (%) and thikness (cm) for

R100, R90-J10, R80-J20, and R70-J30 ................................................ 49

Figure 4.12. The relationship between transmission rate (%) and thikness (cm) for

R100, R90-GF10, R80-GF20, and R70-GF30 ...................................... 50 Figure 4.13. The relationship between transmission rate (%) and thickness (cm) for

R100, R70-J10-GF20, R70-J15-GF15, and R70-J20-GF10 ................. 51

viii

LIST OF TABLES

Page

Table 1.1. Effects ocur ............................................................................................... 16 Table 3.1. Materials composite rates ......................................................................... 33

1

1. INTRODUCTION

Radiation is one of many facts in our life. Natural radiation exists in our whole life.

Heat and light from nuclear reactions in the sun are fundamental to existence. In

environment the occurrence of radioactive materials is natural, carbon-14, potassium-

40 and polonium-210 are natural radioactive materials found in our bodies. Radiation

discovery developed many fields of life (Wrixon and Clark, 2004).

Each radiation type has different properties. Non-ionizing radiation may shake or

move the molecules. Ionizing radiation could break molecular bonds, causing

unpredictable chemical reactions. Ionizing radiation has not only energy waves but

particles. Humans can not see, taste, feel, hear, or smell the ionizing radiation. The

exposure to ionizing radiation comes from cosmic rays and some materials are

natural. Human exposure to natural radiation is in question for certain numbers of

mutations and maybe cancers. Exposure of radiation above natural limites is a cause

for concern since it may result in diseases (Karam and Stein, 2009).

1.1. Radiation and Properties

There are additionally different forms, which we use in everyday in life, like

microwaves, radio waves, radar and X-rays for cooking, communication, navigation,

medical examinations, respectively. Radiation also comes from Radioactive

materials. These materials occur naturally, also produced others artificially. We can

classify radiations according to the effects they produce in the matter. There are two

categories, ionizing and non-ionizing radiations. Ionizing radiation includes cosmic

rays, X-rays and the radiations from radioactive materials. Non-ionizing radiation

includes ultraviolet light, radiant heat, radio waves and microwaves. We can also

classify radiation in terms of its origin as natural or artificial radiation (Harish, 2011).

1.2. Types of Radiation

There are two kinds of radiation, non-ionizing radiation and ionizing radiation

(Canadian nuclear safetly, 2012).

2

1.2.1. Non-Ionizing Radiation

Non-ionizing radiation is the radiation that does not have enough energy for ionizing

atoms or molecules. The electromagnetic (EM) spectrum includes all kinds of EM

radiation from the lowest to the highest energy, wavelength and frequency. EM

radiation is an energy that travels with light speed and spreads. The visible light that

originates from lamps in our homes and the radio waves originates from microwave

or radio stations are some examples of EM radiation that all we are using in our

homes. Figure 1.1 shows the whole electromagnetic spectrum including non-ionizing

and ionizing types. We can seen from this figure that gamma rays and X-rays are

ionizing and all other types are non-ionizing radiation (Alsarray, 2016).

Figure 1.1. The electromagnetic radiations (Alsarray, 2016).

1.2.2. Ionizing Radiation

Ionizing radiation has sufficient energy to take out electrons from an atom. If the

energy of radiation is not high for ionizing an atom then it can excite atom by taking

out an electron up to higher orbit rather than kicking it out from the atom (Figure

1.2). The DNA could be damaged by the ionizing radiation. The hazardous on cell

can be different according to the absorbed radiation dose, When the body is exposed

to a high amount of ionizing radiation it can lead to a cancer (Harish, 2011).

3

Figure 1.2. Ionizing processes (Harish, 2011).

From Figure 1.3, we can see that the radiation can effect DNA by two ways, directly

or indirectly, and it shows the effect of ionizing radiation on DNA. In all cases this

can lead to dangerous health problems (Alsarray, 2016).

Figure 1.3. Damage caused by radiation in DNA

4

1.2.2.1. Alpha (α) – decay

Alpha particle is ionized helium atom and can be obtained from radioactive decay of

nucleus (Figure 1.4). It occurs in elements that have higher atomic mass (A), usually,

alphe decay occures with nucleus are heavy, like radon or uranium, and those ratio

between neutron and proton is very big.

The emissions consider are particulate, so it takes an amount of energy away from

nucleus. The helium nucleus He2+ ; 2n + 2p is a massive 4 AMU particle with a

charge 2+. Nuclear state equation displays that the atomic mass number A is

decreased by (-4), Z protons number is decreased by (-2), and N neutrons number

decreases by (–2). (Smith, 2010).

𝑋𝑛 → 𝑌𝑁−2𝑍−2𝐴−4 + 𝐻𝑒2

2+24

𝑍𝐴 (1.1)

The alpha particles can be stopped by the sheet of paper.

Figure 1.4. The alpha particle

1.2.2.2. Beta (β) - decay

There are three kinds of this decay: beta minus, beta positive and electron capture:

Beta minus decay (𝛃−)

When the neutron number is greater than proton number. The nucleus will be

radioactive in this case, we can solve this problem by converting more neutron to

proton to move closer to the line of stability by:

5

𝑛 → 𝑝+ + 𝛽− + �̅� (1.2)

For example, 60Co has too many neutrons and one of them is converted to a proton,

an electron is created in the nucleus that will be emitted from the nucleus. (Figure

1.5) shows this process schematically. The penetration ability of beta particles is

greater than alpha particles. by a few centimeters of plastic or human body tissue.

Figure 1.5. Beta minus (β-) emission

Beta plus decay (β+)

When the proton number is greater than the neutron number in the nucleus which

will be radioactive, we can solve this problem by converting more protons to

neutrons:

𝑝+ → 𝑛 + 𝛽+ + 𝑣 (1.3)

Occures and through this process positron (positive electron) will be created in the

nucleus, then these positron will be emitted (Figure 1.6).

Figure 1.6. Beta plus (β+) emission

6

Electron capture

In the third case which is electron capture, proton will capture the closest electron

and will convert to neutron:

𝑝+ + 𝑒− → 𝑛 + 𝜈 (1.4)

In electron capture, a parent nucleus may capture one of its own electrons and emit a

photon and a neutrino. This photon may be an x-ray or gamma-ray photon. The x-ray

wave is emitted as one of the electrons in the higher orbitals fills the gap that left by

the electron was captured by nucleus (Mehmet, 2016).

𝐴𝑡 + 𝑒−10 → 𝑃𝑜84

218 85218

1.2.2.3. Neutron

Neutron radiation consists of a free neutron and it can ordinarily be emitted as a

result of induced or spontaneous nuclear fission. The neutron has no electric charge,

it can travel hundreds meters or thousand meters in the air. However, its

effectiveness can be stopped if it is exposed to hydrogen-rich substances. Because

mostly the interaction of neutron with the material is (n, p) reaction and thus proton

reach materials, for example concrete or water are applicable materials to shield

neutron. (Figure 1.7) shows the processes of neutron emission, There are kinds of

neutrons producing from the ionization process that can make matters radioactive or

substances, these procedures called the effective neutrons, which are a primary

method of fabricating radioactive sources used in the many fields of academic,

health, and industrial area and its several applications.

As we mentioned, the neutrons have high ability to travel perhaps thousands of

meters in the air and moderate distance (a few meters in the solids) because they

have high energy making them more penetration to the materials.

7

Figure 1.7. Neutrons Radiation

1.2.2.4. Gamma ( )-decay

Gamma radiations, also called as gamma rays, are electromagnetic radiation have

high frequency and therefore high energy with very short of wavelength (~10-3 A.U.

to 1 A.U.). Hence, they have no electric charges and can not be deflected by

magnetic and electric fields (Figure 1.8). Gamma rays are ionizing radiations and are

thus cause of biologically hazardous. They are produced of the decay of high energy

states of (highly unstable) of atomic nuclei. They can also be created in other

process. Also this ray can produced from naturally occurring radioactive isotopes,

and secondary radiations produced from atmospheric interactions with particles of

cosmic rays. Gamma-rays produced by the number of astronomical process in which

very high energy of electrons are produced that in turns cause secondary gamma

rays, inverse Compton scattering and Synchrotron radiation. Typically, gamma-rays

have frequencies above ten exahertz (or >1019 Hz), therefore have the energies above

100 KeV and wavelengths less than 10 pico-meters (less than diameter of an atom).

Gamma-rays from radioactive decays are defined as gamma rays no matter what

their energy. Gamma decay commonly produces energies of a few hundred of KeV

and almost less than 10 MeV (Gorden, 2011).

8

Figure 1.8. Gamma decay

1.3. Gamma Rays Interaction with Matter

Three main photon interactions mechanisms, they are: photon scattering (elastic or

inelastic), photoelectric effects, and pair production. In this section, each of

those processes is shortly delineate beside its effects from a radiation attenu-

ation purpose of view. When gamma-rays are passing through the matter, the

probability of absorption in a thin layer is proportional to the layer thickness. This

can lead to an exponential of intensity decrease with the thickness, which is

formulated in following equation:

I (x) = Ioe−μ𝑥 (1.5)

Where, 𝜇 = 𝑛. 𝜎 is the coefficient of absorption, measured in cm−1, I is the atoms

number per cm3 in the material, 𝜎 is the absorption cross section in cm2 and x is the

material thickness in cm (Angüner, 2008).

1.3.1. Compton Scattering

Compton scattering is a direct interaction of the incident gamma photon with a

loosely bound (outer shell) electrons in the absorbing material, transferring only a

part of the incident gamma ray energy to the kinetic energy of the recoil electron.

The interaction results are that the incoming photon is treated in energy and deflected

from its original direction and an electron called as a recoil electron is generated.

This energy can be different from zero (θ=0) to a significantly large fraction of the

9

original energy of the gamma ray (θ=π), depending on scattering angle. (Figure 1.9)

shows a schematic of Compton scattering process.

Figure 1.9. The mechanism of Compton scattering

1.3.2. Photoelectric Effect

In the photoelectric absorption, an incident energy photon is completely absorbed by

an atom and one of the electrons is released, creating an ion and free electron. Then,

an energetic electron known as a photoelectron is removed from one shell of the

electron shells with a kinetic energy given by incident photon gamma energy Eγ (hν)

(Figure 1.10). Minus binding energy of electron in its origin shell (Eb). For that case

to occur photoelectric effect, the incident photon energy must be greater than the

binding energy of the inner shell electron. The photoelectric absorption process is

shown schematically in the diagram below. The energy of the ejected photoelectron

can be expressed as follows:

𝐸𝑒− = ℎ𝜈 − 𝐸𝑏 (1.6)

Where Ee- represents the kinetic energy of the photoelectron, and Eb is the binding

energy of the photoelectron's original shell. When an electron in one of the bound

electron shells is ejected from the atom by an external excitation x-ray or gamma

photon, a vacancy is created. A free electron can then be captured from the medium

or a rearrangement of electrons from outer shell orbits of the atom then fills this

vacancy leading to the emission of characteristic X-ray and this in turn, produces a

hole in the L or M shell (Peshawa, 2015).

10

Figure 1.10. Schematic of the photoelectric absorption process

1.3.3. Pair Production

Pair production is the third significant process for gamma-ray photons interactions in

matter. This interaction occurs at the points which are closes to the nuclei of the

absorbing material due to the high electric field. This process occurs within the

Coulomb field of the nucleus resulting in the conversion of gamma ray into an

electron-positron pair. For this quantum mechanical effect, gamma ray should be

carried an energy at least equivalent to the combined rest mass of two particles

making 1022 keV in all. Therefore, minimum γ-ray energy of 1.022 MeV is required

for any incident photon to under go this process. Any excess energy carried in by the

photon above 1.022 MeV is transferred into kinetic energy which shares by the

electron_positron pair. Total kinetic energy of the electron_positron pair is given by

following equation:

𝐸𝑒− + 𝐸𝑒+ = ℎ𝜈 − 2𝑚0𝑐2 (1.7)

After electron and positron pair is created, they typically travel a few millimeters in

the material before losing their energies in the absorbing medium. As the positron

slows down, it can combine with an atomic electron in the surrounding material and

subsequently annihilate to form two photons, called annihilation photons, each with

energies of about moc2 = 0.511 MeV which are emitted back-to-back (to conserve

linear momentum). Generally, the annihilation of pair of an electron-positron does

not occur until the energy of the posıtron reduced to near thermal energy, releasing

two energies of 511 keV. Figure 1.11 shows pair production process schematically.

11

Figure 1.11. Schematic of the pair production process and annihilation

1.4. Radiation Measurements

The radiation has been measured by the scientists, they used several idioms depend

on if they are discussing radiation is coming from radioactive sources, the dose of

radiation which absorbed by a person, or the biological risk that affects on a person's

health. Most of the scientists in the international community measure radiation by

using the System International (SI), which is a uniform system of weights and

measures that developed from the metric system. In the United States, however, the

traditional system of measurement is still used. They used different measurement

units depend on which aspect of radiation is being measured. The amount of

radiation emitted, by a radioactive material measured using the conventional unit

Curie that came from the name of famed scientist Marie Curie (CI), or the (SI) unit

Becquerel (Bq). The radiation dose absorbed by a person is measured by using (rad)

or gray (Gy). The biological risk of exposure to radiation measured by using REM or

Sievert (https://emergency.cdc.gov/radiation/measurement.asp).

The measurement unit of emitted radiation is Bq in the SI unit or Ci in the

conventional unit. When the nucleus has many particles, too much energy, or too

much mass will emit radiation to get stability. The nucleus disintegrates, in an

attempt to reach a stable case. As the nucleus disintegrates, energy is emitted as

radiations. The Ci or Bq is used to express the number of disintegrations of

radioactive atoms in the radioactive materials over a time period. For example, one

12

Ci = 37 billion (37 × 109) disintegrations per 1s. The Ci is being substituted by Bq.

Since the one Bq is equal to one disintegration per second, one Ci = 37 billion (37 ×

109) Bq. Ci or Bq could be used to assign to the amount of the radioactive materials

released into the environment.

When a person is exposed to the radiation, the energy is deposited in tissues of the

body. The amount of energy deposited per unit weight of the human tissue is called

the absorbed dose. The Absorbed dose is measured by using the conventional unit is

rad or the SI unit is Gy. The rad, which stands for radiation absorbed dose, was the

conventional unit of measurement, but it has been changed by the Gy where (Gy =

100 rad) (Knoll, 2010).

There are various types of radiation detectors such as: gas-filled detectors,

scintillation detectors, semiconductor detectors. Radiation detectors and detection

systems are also classified according to their physical form (gas, liquid, and solid

materials). Some of them may provide different types of information about radiation,

such as its energy, intensity, and/or the type of radiation which is being measured. In

general the methods of detection are based on the process of ionization or excitation

of atoms in the detector by the passage of a charged particle (Peshawa, 2015).

1.4.1. Gas-Filled Detectors

It is known that people cannot feel, smell, see or test ionizing radiation and therefore

it was necessary to rely entirely on special devices detection and measurement of

radiation. Several detector types take benefit of the ionizing effect of radiation on

gases. It can be collected the ion pairs separately. When a potential gradient is

applied between the two electrodes in a gas-filled ion chamber, the positively

charged molecules will move to the cathode and the negative ions (electrons) will

move quickly to the anode, thereby generating a measurable pulse. These pulses can

be easily measured by the associated devices as integrated current. There are three

main kinds of Gas-field detectors:

1- Ionization Chamber

2- Proportional Counter

3- Geiger-Mueller Counter

13

All are based on the collection of the electrical charge from the detector as a result of

the collection of the charges made in the detector chamber (Flakus, 1982).

1.4.2. Scintillation Detectors

The scintillation counter is a useful tool when counting gamma rays. The efficiency

is high and it has proper solving power. An important feature of this detector is that

the energy of the emitted photon is proportional to the incoming radiation. It gives

informations about the type and of radiation energy coming from this situation.

This process is also considered as one method available is the most useful for the

spectroscopy and detection of a variety of radiations widely. These are properties of

the ideal scintillation material that should have it:

1. Should change the kinetic energy of charged particles into detectable light with a

large scintillation qualification.

2. The light produced should be proportional to the deposited energy over a wide

range as possible.

3. The medium should be transparent to the wavelength of emission for good light

summation.

4. The decay time of the resulted scintillation must be short so that quick signal

pulses may be created.

5. The materials should be of good optical quality and subject to industry in sizes

large enough to be of usefulness.

6. The indicator of refraction should be close to that of glass (-1.5) to allow efficient

coupling of the scintillation light to a photomultiplier tube or other light sensor

(Knoll, 2010).

The energy of the radiation coming into the conduction band with the excited

electrons makes the return to the valence band at least once and then emits a photon.

This photon falls into the photocathode, causing a photoelectric phenomenon, which

results in a photocathode electron sputtering. The output pulses come to fruition as

these electrons, which are broken, are amplified and accelerated by the dyads found

in a photomultiplier tube (Figure 1.12).

14

Figure 1.12. Schame of Scintillation Counter

At the present time, the most commonly detectors used are sodium iodide (NaI), and

High Purity Germanium detector (HPGe). Both of these spectrometers have their

advantages and disadvantages. Sodium iodide scintillator NaI (Tl) is the most

frequently used because it has a high efficiency for detecting gamma radiation and

does not require cooling, while HPGe detectors provide superior energy resolution.

Scintillation detectors designed for use in different applications are divided into three

groups:

Inorganic scintillators (NaI (Tl), CsI (Tl), CsI (Na), LaBr3 (Ce) etc.).

Organic scintillators (Anthracene crystal, POPOP liquid, Stilbene Crystal etc.).

Plastic scintillators (NE102A, BC400, BC406, etc.).

1.4.3. Semiconductor Detectors

The primary ionization should be collected to make the direct measurements of the

nuclear radiation energy. Condensed, the phases have higher densities than gases

and so provide more efficient stopping of the radiation per unit of length. However,

the metals allow rapid recombination of the electron-positive ion pairs and

insulators inhibit the collection of charge. Therefore, only semiconductors are

extensively used for detection of radiation. Metals and insulators like concrete are

15

used extensively in the shielding of radiation, also some transparent inorganic

crystals have a special sensitivity for radiation.

The most common semiconductors used to construct "solid state ionization

chambers" are silicon and germanium. These materials should be extraordinarily

pure to look at the primary ionization (~ 105 electrons). The germanium devices

should be cooled for reducing the thermal noise to look at the outcoming

signals. The properties of small-scale devices based on the groups III and V

materials, e.g. GaAs, have been studied but no large- scale applications have been

made. The size and shape of the available semiconductors have grown with time but

are still severely limited by production techniques and the availability of high purity

materials.

Early solid state device relied on observing the ionization in semiconductors. The

early devices were impractical due to the requirement of extremely pure materials.

Modern devices are based on the junction diodes. These diodes have a rectifying

junction that only allows one direction of the flow current. Incident radiations

creates ionization inside the bulk of the diode, and creates a pulse of current in the

opposite direction to the normal current flow through a diode that is straightforward

for detection (Knoll, 2010).

1.5. Biological Effects of Radiation

One quality of the ionizing radiation on body of human is that the energy absorbed

is low but the biological effects are reality. For example, after receiving dose 10 Gy

from lethal, the temperature of body only will increase by 0.02o C, but the dose can

lead to death of all the exposed entities. The second quality is the latent radiation

biological effects. Acute biological effects may occur within hours to several days,

while the long-term the effects appears usually several years after exposure.

1.5.1. Type of effects

According to the subjects which the effects will occur, we can classify the biological

effects of radiation to two types: Somatic effects and Genetic effects, see (Table1).

16

Table 1.1. Effects occur

Characteristic of

effects Occurring time Object Effects on organs

Deterministic

Effects

Acute Effects

Somatic Effects

Skin damage

Damage of

reproductive system

Damage of blood

forming system

Damage of digestive

system

Damage of central

nervous system

Latent Effects

Cataract

Damage of

immunization system

Stochastic Effects Cancer

Genetic Effects Heredity effects

The deposition of energy by ionization could be a random method. Even at terribly

doses there is some probability that enough energy could also be deposited into a

vital volume among a cell to lead to cellular changes or cell death. But because

of the exceptional ability of cells to repair harm, enzymatic, and repair mechanisms

would lead in several instances to the proper DNA repair and also

the cell can survive with none modification to its perform or genetic structure.

Another result is likely to be a boom a mutation. The cell will survive but with

modulation in the DNA sequence for the cell’s genome. The mutated cells have the

ability for reproduction and thus perpetuate the mutation. If the mutated cell is a

somatic cell, the mutation could be a result of a malignant tumor. If the mutated cells

are germ cells, it could lead to a hereditary effect. These are stochastic effects and

their results cancer or hereditary effects. (Figure 1.13) shows both ways by which

radiation can affect the whole system of the body (Domenech, 2016).

17

Figure 1.13. Radiation effects on the whole body system

1.5.1.1. Somatic Effects

The somatic effects of radiation can be classified as Cataract and Cancer Induction as

detailed below.

A. Cataract Induction

The eye lens are different from other organs in that injured or dead cells are not

removed. Single dose of various of hundred rem have chosen opacities that overlap

with vision among a year. Larger dose is required when the doses are fractionated

over a period of a many years, and the cataract appear various years after the last

exposure. BEIR Report (1990) (a report stated by specific committees of National

Research Council) noted that the cataract induction may not be a trouble for the

doses currently permitted radiation workers (National Research Council (US), 1980).

18

B. Cancer Induction

Cancers arise in an exceeding sort of tissues and organs that are thought to be the

basic somatic effect of the low and medium exposure to radiation. Organs and tissues

vary greatly in their disposition to cancer inducement by radiation. Inducement of

leukemia by the radiation features because of the natural rarity of the disease

(National Research Council (US), 1980).

1.5.1.2. Genetic Effects

A mutation can be defined as an inheritable issue in the genetic materials inside the

chromosomes. Generally, mutations can be classified into two types, dominant and

recessive. The effects of dominant mutations can usually be seen in the first and

subsequent generations while the effects of recessive mutations can not be seen until

a child receives a similarly changed gene for that trait from his parents. This may not

be recorded for many generations or it may never be. Mutations can be a harmful

cause when it ranges from undetectable to fatal. In this part, the mutational effects

mean only those inheritable conditions which are usually hard enough to require

medical care in a person's lifetime at some time.

ICRP estimated the induced radiation probability which hard hereditary effects in all

workers population of a radiation to be 6Χ10−5 per rem. UNSCEAR report concluded

that about ¼ of the affected descendants could be grandchildren and children. This

information could be gathered with the average annual dose about 10 millirem to

estimate the genetic risks of our research and workers in the field of laboratory

medicine radiation. Assuming that 11 years of exposure before to conception (The

average age of fathers at the time of birth in the US is 29, while the average age of

the mother is less than 29 yeas. The minimum age for occupational exposure to the

ionizing radiation is 18 years), and the chance of a serious birth trouble in all

descendants of the workers due to previous occupational exposure is less than one in

a hundred thousand due to previous the occupational exposure 11 0.01 6 10−5 ∼

7 10−6. This hazard is very low, and many experts state the risk to be nonexistent.

However, it must not be surprising for such a small dose. For perspective, the

radiation dose due to single round trip air flight may be as high as 20 millirem,

19

substantially higher than the average annual dose of research and workers are

working in laboratory medicine (National Research Council (US), 1980).

1.6. Radiation Protection

The organisms are continuously exposed to radiation in daily life and may be

biologically hazarded if the exposure is direct and for a long time without shield

prtection, Therefore, there are three main ways to protect against radiation impacts

(Time, Distance and Shielding), It should be noted, that the best way to protect

against radiation that all these factors work at same time.

1.6.1. Time

It is the first and simplest rule to protect from radiation. Applications using

radioactive sources should be completed as soon as possible, according to the

Permissible Maximum Dose "ALARA: As Low As Reasonably Achievable"

principle (Silva, and et al., 2008). The dose absorbed by the radiation-exposed

substance varies with time. The exposure speed (exposure in unit time) at a distance r

from a source with A activity is given by the following formula. where Γ is a

constant connected to the source.

△ 𝑋 = Γ 𝐴

𝑟2 Δ𝑡 (1.8)

1.6.2. Distance

The free pathways of particulate radiation such as alpha and beta radiation, even in

air, are rather short. Neutron and gamma radiation, however, take longer to travel

than others, but their intensity decreases with distance. For this reason, one of the

effective ways of avoiding damage to a radiation source is to keep it as far away

from the source as possible. The amount of radiation that is exposed varies

depending on the distance according to inverse square law. To confirm this, it is

enough to assume that the rays emerging from the source at the same time form

spherical wavefronts at a distance r after a time t. Thus, the intensity of the rays on

20

the surface of any spherical wave ceiling is inversely proportional to the intensity of

the source Io and the area of the sphere surface.

𝐼 ∝ 𝐼𝑜

4𝜋𝑟2 (1.9)

Here, I is a radiation intensity is at a distance r. Also, the dose rate D2 at a distance d2

is smaller than that at distances d1 from the dose rate D1 at another distance d1

(d1 <d2).

𝐷2 = (𝑑1

𝑑2)

2

𝐷1 (1.10)

1.6.3. Shielding

Shielding is a barrier between the radiation source and the system intended to protect

it from radiation to destroy or minimize the effects of radiation. The main purpose of

the armor is to allow the emitted radiation to interact with the atoms of the armor

material to lose all or part of their energy. The type and energy of the radiation is

important in the selection of the shielding material to be used for this purpose.

Because the mechanisms of interaction of nuclear radiation with matter vary

depending on the load, mass and energy they possess.

When the penetration capabilities of radiation into matter are compared, it turns out

that the most dangerous types of radiation are gamma and neutron. Neutrons are

unloaded; gamma rays are both uncharged and massless. Figure 1.14 shows the

penetration of different types of radiation in different materials.

21

Figure 1.14. Comparison of penetration of various radiations

1.7. The Composite Materials

The composite can define as a combination of different materials in the composition,

where the particular constituents keep their separate identities. These separate

constituents work together to give the needed mechanical strength or the hardness of

the parts in the composite. Composite materials are composed of two or more several

phases. In the recent years, a rapid growth happened in the exhaustion of fiber

reinforced polymer, yielding an individual composition of high performance, large

versatility and processing features at favorable costs by permutation and composition

of different fibers and polymers, a wide range of composites, having individual

properties to applications as alternatives to ordinary materials like woods, metals,

etc. have been produced (Azim, 2017).

In this study of radiation shilding were prodused the composite contain three

materials are: Rubber as the main material, Jute and Glass-Fiber.

1.7.1. Rubber

Rubber can be defined as a material that may be compressed or stretched when the

pressure is removed and will back quickly to its original state, without permanent

perversion. The elastomers or elastic polymers are the primary chemical building

blocks of rubber, These are large chain like molecules, which when cured

22

(vulcanized) form chemical crosslinks between the chains of polymer. The ability of

the chains to stretch and spring back is a measure of the rubber’s elasticity and

resilience under pressure. The first common elastomer was Polyisoprene is made

from natural rubber. In general, rubber materials are identified by their low elasticity

modulus. They are resistant to water, alkalis and weak acids. Rubber is also a good

electrical insulator and can be used as a bonding factor. There are two types of

rubber are natural and synthetic. Natural rubber is produced from the latex (milky

juice) of the Hevea Brasiliensis tree. It is, therefore, a renewable resource unlike

synthetic rubber, which is produced from petrochemicals.

Natural rubber is harvested as latex, a natural source of isoprene, named

caoutchouc or India rubber which is dried out for trade processing. This form of

natural rubber dried out is the chemical Polyisoprene. Although it has some

elasticities, Polyisoprene is mixed with chemicals and vulcanized to make the

completed product. Natural rubber is both elastic and viscous making it a model

polymer for the effective and static engineering applications.

Synthetic rubber is made from oil by-manufactures using either solution or

emulsion polymerisation mechanisms Figure 1.15. Applying polymer chemistry

techniques allow greater opportunities for customisation of mechanical properties

and increased resistance to temperature, chemicals, and solvents than is possible with

natural rubber. Synthetic rubber properties are far wider than those of natural rubber.

There are up to 20 different classes of synthetic rubber giving a far wider choice of

rubbers to meet the wanted mechanical properties, temperature, and chemical

resistance, required for a specific application, synthetic rubber may have the

following advantages over natural rubber like more resistant to oil, certain chemicals

and oxygen, better aging and weathering resistance, resilience over a wider range of

temperature (Awagon, 1996).

The rubber is used in the composite which prepared from recycled footwear comes

from the base of the bottom of the sports shoes that the athletes are using it (Yıldız,

F., et al., 2017).

23

Figure 1.15. Rubber used in the composite (Lee, M. J., & Rahimifard, S., 2012)

1.7.2. Jute

Jute is produced from the genus Corchorus plants Figure 1.16, which has about 100

kinds. This is one of the cheapest natural fibers, with highest manufacture volume.

Bangladesh, India, and China supply are the best situations for the growth of the jute.

Ray and co-workers investigated widely alkali cure jute fiber reinforced with vinyl

ester resin. In their studies, they considered the mechanical, thermal, dynamic, and

impact fatigue behavior compared with that of untreated jute fiber–vinyl ester

composites. Longer alkali treatment ejected hemicelluloses and improved the

crystallinity, enabling better dispersion of the fiber. The impact, thermal, dynamic

and mechanical properties were excellent owing to the alkali treatments, contain

treatment time, concentration and situations. The enhancements in tensile strength

above 50%, in bending strength about 30%, and in impact strength 90% were spotted

in composites and are comparative to values achieved for pure Biopol sheets.

Degradation studies displaied that after 150 days of compost burial more than 50%

weight loss of jute/Biopol composites happens.

24

Figure 1.16. Jute used in the composite

The effects of hybridization on the tensile properties of jute–cotton woven fabric

reinforced polyester composites were studied as a function of the fiber contents,

orientation, and roving texture. It was spotted that tensile properties along the

direction of jute roving alignment increase steadily with fiber content above to 50%

and then show a slope to decrease. The tensile strength of composites with 50% fiber

content in parallel to the jute roving is about 220% higher than pure polyester resin.

Jute fiber reinforced PP composite was evaluated regarding the effect of matrix

preparation, the effect of gamma radiation, the influence of interfacial adhesion on

creep and dynamic mechanical attitude, the effect of silane coupling agent, and the

effect of natural rubber. The properties were studied of jute/plastic composites,

including the thermal stability, crystallinity, modulation, trans-esterification,

weathering, durability, fiber orientation on frictional and wear attitude, eco-design of

automotive components, and alkylation. Polyester resin used as a matrix for jute fiber

reinforced composite and the relationship between water absorption and dielectric

attitude, the elastic properties, notched strength and fracture criteria, impact damage

characterization, weathering, and thermal attitude, and effect of silane treatment was

examined (Faruk, 2012).

1.7.3. Glass Fiber

Glass Fibers are among the most multilateral industrial materials known nowadays.

They are readily prepared from the raw materials, which available provide in

virtually unlimited. They display useful bulk properties like transparency, hardness,

25

resistance to any chemical attacks, stability, and inertness, as well as wished for fiber

properties like flexibility, strength, and stiffness. Glass-fibers are used in the

preparation of structural composites, printed circuit boards and a wide range of

special objective products. Glass melt is made by fusing (co-melting) silica with

minerals, which contain the oxides necessary to form a given composition. The

molten mass is quickly cooled to prevent crystallization and created into glass-fibers

by a fashion also known as fiberization (Figure 1.17) (Wallenberger, et al., 2001).

Figure 1.17. Glass-Fiber used in the composite

Glass fiber preparation is the conversion of different raw materials (predominantly

borosilicates) with high-temperature into a homogeneous melt, followed by the

fabrication of this melt into glass fibers. The two main types of glass fiber products,

textile, and wool, are manufactured by similar processes. Glass fiber production may

be classified into three stapes: raw materials handling, glass melting, and wool glass

fiber forming and finishing, this last stape being slightly different for textile and

wool glass fiber preparation (Industry, W. F. I. M., 1983).

26

2. THE LITERATURE BACKGROUND

There are many studies and scientific experiments on radiation shielding because it is

a very important subject to protect against the risk of radiation and its impact on

human health. In this section, a summary of some works and experiments related to

radiation protection and reduction of risks to human life is presented.

I. Akkurt and his team, (2004), the linear attenuation coefficients and total mass

attenuation coefficients of gamma rays for barite, marble, and limra were obtained.

The results show in the high energy region that attenuation begins to increase.

Increasing the attenuation coefficient of photons, while increasing the photon energy,

results in successive collisions due to Compton scattering. The calculations also

show that the linear attenuation coefficients increased with increasing the density of

materials, and mass attenuation coefficients remain constant.

C. Basyigit and other, (2005), Investigated the photon attenuation coefficients for

marble. The total linear attenuation coefficients μ (cm−1) have been obtained using

the XCOM program at the energies of 1 keV to 1 GeV for six various types of

natural marbles. The results show that the linear attenuation coefficients μ depend on

the photon energy, and even at low energy slight fluctuation has been spotted

between the results of calculations, there are no significant differences at higher

energies. The linear attenuation coefficients μ decrease clearly at low energy Eγ

0.1 MeV, decrease clearly in the middle 0.1 MeV Eγ 10 MeV and slightly

increase in high energy Eγ 10 MeV.

M. A. Mosiewicki and other, (2007), natural rubber NR effect on Wood Reinforced

Tannin Composites was investigated, natural rubber latex was added to the

composite materials formulated from a quebracho tannin adhesive crosslinked with

hexamethylenetetramine and wood flour as a filler. The impact properties presented a

similar trend, with the largest change happening between 0 to 5% natural rubber in

the matrix formulation.

Ibrahim Turkmen and other, (2008), they have calculated radiation attenuation

coefficients in Portland cements were mixed with silica fume, blast furnace slag and

natural zeolite, the radiation attenuation coefficients expressed as mass attenuation

27

coefficients for Portland cement, zeolite, blast furnace slag, silica fume and their

mixed kinds in function of the Photon energy over the energy range of 1 keV and 2

MeV. It was spotted that the different percentages of constituents in cement and that

cement mixed with different additives such as zeolite, silica fume and blast furnace

slag, lead to important variations in total mass attenuation coefficients. The elemental

samples compositions were resolved using a wavelength dispersive XRF

spectrometer. The calculated values of total mass attenuation coefficients were

looked on the basis of different percentages of cement constituents and cement

mixed with different additives.

I. Akkurt and other, (2009a), investigated photon attenuation coefficient for pumice,

where it was measured the linear attenuation coefficients of two different kinds of

Golcuk-Isparta pumice which are alternative building materials were measured. The

results were compared with calculation and some other materials where the results

showed that the linear attenuation coefficients are higher for compressed pumice than

for natural pumice.

M.E. Medhat, (2009b), calculated gamma ray attenuation coefficient of some

building materials in Egypt, calculated by XCOM Program, the mass attenuation

coefficients l/q for building material samples were measured for five energies which

are 59.5, 356.5, 661.6, 1173.2 and 1332.5 keV, the materials were studied here

represent walls (three types of red clay bricks) and roofs (cement, gypsum and

concrete) representative of those used in Egypt. Bricks were covered in two sides by

the cement, which is used in inner walls, may protect from radiation about 46–67%

more than brick itself. Concrete was covered by two gypsum layers and cement,

which is used for ceilings, the protection efficiency increased about 6% more than

concrete itself.

I. Akkurt and his team, (2010a), measured the radiation shielding of concrete

containing zeolite, the μ for concrete which contain zeolite as an aggregate

concentrations differently 0%, 10%, 30% and 50%, the results were compared with

the calculations. The linear attenuation coefficient measured with four blocks of

concrete were decreased with increasing zeolite concentration. It was concluded that

the addition of zeolite as an aggregate in concrete was not an alternative option to be

used for the purposes of radiation shielding.

28

I. Akkurt and other, (2012a), studied photon attenuation coefficients of concrete

including marble aggregates. The linear attenuation coefficients for the six different

types of concretes were measured at photon energies of 662, 1173 and 1332 keV and

calculated at photon energies of 1 keV–100 GeV. The measurements linear

attenuation coefficients were decreased linearly with increasing photon energy for all

kinds of concrete. It is also clear, that decreasing of linear attenuation coefficients

with the increasing of the photon energies.

K. H. Mahdi, (2012b), studied and calculated gamma ray attenuation coefficients for

different composites. The linear attenuation coefficient was displayed as a function

of the applied energy. It was clear that the linear attenuation coefficient sharply

decreased with the increase of the photon energy in the range between 0.1 to 1 MeV

for all composites. Such a behavior could be ascribed to the photoelectric and

compton scattering which were the main predominant interactions in this region.

This sharp decrease was considered to be an indication that the μ was very sensitive

for the photon energy.

Akkurt and his team, (2013a), investigated and measured the radiation shielding

properties of iron doped into clay samples at 662 keV. The measurements were

performed using gamma spectrometer system which contains NaI(Tl) and 16k MCA

analyzer.

Mohammed M. Al-Humaiqani and other, (2013b), examined gamma radiation

shielding properties of high strength high performance concretes prepared with

different types of normal and heavy aggregates. It was found that the compressive

strength of the normal concrete had almost no effect on the attenuation of gamma

rays. The Linear and mass attenuation coefficients were calculated and compared

with the past researchs and a good agreement was found. However, the HPCs density

considerably affected the attenuation of gamma rays. With the increase in the

density, the attenuation coefficients increased linearly. This endorsed that the

relationship between HPC density and the gamma attenuation coefficients was linear.

Rezaul K Khan and other, (2014a), comparative experimental studies on the physico-

mechanical properties of jute caddies reinforced polyester and polypropylene

composite, non-woven jute caddies (JC, jute wastage) reinforced unsaturated

29

polyester resin (UPR) and polypropylene (PP)-based at random oriented

discontinuous fibre composites with fibre loading 40-65% were fabricated by

compression molding, the composites were irradiated with gamma rays 60Co of dose

varied from 2.5 kGy to 12.5 kGy. Tensile and flexural properties of the composites

were found to be improved significantly after irradiation. TS and BS of JC/UPR

composites were increased 29.86 and 14.60% respectively at 7.5 kGy, while for

JC/PP composites the increments were 21.69 and 7.78% respectively at 5.0 kGy.

Sardar M Shauddin and others, (2014b), studied the effects of fiber inclusion and γ

radiation on physico-mechanical properties of jute caddies reinforced waste

polyethylene composite, jute mill wastage which is known as jute caddies (JC)

reinforced waste polyethylene (WPE)-based low cost at random oriented

discontinuous fibrous composite was fabricated by using traditional hand layup

method. Fourier Transform Infrared Spectroscopy (FT-IR) was used for investigation

the chemical composition of both raw jute and jute caddies. Jute caddies content in

the composite was varied from 20 to 45% where, 32% JC enriched composite

showed that the best performance in mechanical tests, both types of composites were

irradiated with gamma rays of dose varied from 2.5kGy to 12.5 kGy, where

composites irradiated with 5 kGy dose leads to the best results.

TP Sathishkumar and his team, (2014c), studied glass fiber-reinforced polymer

composites, glass fiber reinforced polymeric (GFRP) composites were the most

commonly used in the preparation of composite materials. Ultimate tensile strength

and flexural strength of the fiber glass polyester composite was increased with

increasing in the fiber glass Vf of fiber weight fractions. The elastic strain of the

composite was increased with the fiber glass Vf over 0.25, and then subsequently

decreased with further increase in fiber glass Vf. The Young’s modulus of elasticity

of the composite increased with the fiber glass Vf.

Raghavendra Supreeth B.S and his team, (2015), studied the effects of gamma

irradiation on the mechanical properties of natural fibers reinforced hybrid

composites, using of natural fibers as reinforcement in polymeric composites for

technical applications has been a research subject of scientist. These test samples

were exposed to different doses of gamma radiation in the range of 1-20 kGy. Then,

mechanical properties of hybrid composites were tested by using computerized UTM

30

machine as per the ASTM standards. The hybrid composites were irradiated with 5

kGy radiation dosage lead to the best results.

Esraa Alsarray, (2016a), investigated radiation shielding properties of some

composite materials. The measurement showed that the barite was an important

materials to shield against gamma rays. With the increasing barite rate in composite

material, the linear attenuation coefficients were also increased.

M. A. Çakıroğlu and others, (2016b), investigated radiation shielding properties of

polypropylene fiber reinforced shotcrete. For this purpose, measured linear

attenuation coefficient for shotcrete reinforced with polypropylene fiber, produced

using the dry mixing process. Measured the linear attenuation coefficients for dry

mixed shotcrete at photon energies of 1173, 1332 and 662 keV. The measurements

were carried out by using gamma spectrometer containing NaI (Tl) detector and a

Multichannel Analyzer (MCA), the linear attenuation coefficients increased with the

increasing fiber percentage for all studied gamma energies.

31

3. MATERIALS AND METHODS

3.1. Preparation Samples

In the study of linear attenuation coefficient μ (cm-1), Rubber as a main material, Jute

and Glass-Fibre (R-J-G/F) as additions to composite in (Figure 3.1) were used to

produce ten samples with different rates. The average sizes of these composites about

(L8.5, W5.5, Th2) cm, the average thickness about 2cm. All preparation and

production of samples have been done according to the details given in (Table 3.1).

The composite materials were manufactured in Istanbul Technique University's

laboratory ( Parlar, Z., et al., 2015).

At first, the procedure was to cut Rubber by the cutting device, also cutting Jute and

Glass-fiber to small pieces, then, the materials were mixed with hands, after that, we

put the mixed materials in a mold for a few minutes under the compressing device.

The oven was heated about 150°C while the mold inside. Finally, we took the

components out of the oven, see (Figure 3.2 and 3.3). and cutting them according to

the sizes that mentioned above.

.

Figure 3.1. Samples composite

32

Figure 3.2. Cutting and compressing devices

Figure 3.3. Heating procedure

33

Table 3.1. Materials composite rates

No Sample Rubber (%) Jute (%) Glass-

Fibre (%)

Thickness

(cm)

1 R100 100 - - 2

2 R90-J10 90 10 - 2,1

3 R80-J20 80 20 - 2,1

4 R70-J30 70 30 - 2,2

5 R90-GF10 90 - 10 1,9

6 R80-GF20 80 - 20 2,1

7 R70-GF30 70 - 30 2

8 R70-J15-GF15 70 15 15 2,3

9 R70-J10-GF20 70 10 20 2,1

10 R70-J20-GF10 70 20 10 2,1

3.2. Methods

3.2.1. Gamma ray spectrometer system

Gamma spectrometer system is applied to know if a material is radioactive,

radioactive, or the source of radioactivity. Also it can identif gamma rays energy of

the samples. It consists of a detector which is NaI(Tl), counting electronic system

(spectroscopy amplifier, high voltage, Multichannel Analyses MCA) and a personal

computer PC (where software was installed is MAESTRO 32) for recording data. In

order to reduce the background level of the system, the detector is cvered by shield

prepared from lead about 6 cm on all side (Figure 3.4).

Figure 3.4. Schematic view of gamma Spectrometer and electronic units

34

3.2.2. NaI (TI) Detector

The scintillation detectors are consist of scintillator material and photomultiplier

pipe connected to the materials. It can cause ionization or excitation if the ionizing

radiation interacts with scintillation materials. Scintillation light emitted by the

phosphor was collected by photomultiplier tubes, it can change to the voltage pulse

as can be shown from Figure 3.5. The obtained amplitude pulse is proportional to the

energy of the radiation. This detector can be used to count and energy discrimination.

Figure 3.5. Schematic view of NaI (Tl) detector

Iodine has high atomic number in crystal of NaI, also can be obtained the high

detection efficiency. In order to activate, usually a small amount of thallium is added

to crystal, and this new structure is named Nal (TI). In (Figure 3.6) 3" x 3" NaI (Tl)

detector is seen.

Figure 3.6. Photography of NaI (Tl) detector

35

3.2.3. Electronic units

The high voltage unit in the spectrometer is used to provide a high potential. The

preamplifier could modify voltage signal which is presenting from the detector

related with the energy is stored. It can be the pulses. the pulses divide into signals

and detect the height of pulse.

The Multi-Channel Analyzer (MCA) analyses the output signals from the amplifier.

Subsequently, each analyzed signal is converted to digital form and stocked,

according to its amplitude, in a memory channel. Each channel is characterized by its

energy and picas which can be got by the accumulation of signals.

3.2.4. Energy calibration

In order for introducing the response of radiation to the detector should be calibrated

before using a gamma-ray spectrometer. That may be done by using known the

radiation sources. The using of the gamma-ray sources which are 60Co and 137Cs, the

energy calibration of NaI (Tl) detector has been achieved. From the Figures 3.7 and

3.8, it can be shown that the decay scheme of the two our sources were used. The

energies 662, 1173 and 1332 keV of gamma rays are released from those sources can

be seen in these figures respectively.

Figure 3.7. The decay scheme of 137Cs

36

Figure 3.8. The decay scheme of 60Co

In the upper part of Figure 3.9 has been shown that the energy spectrum from the

sources which are used in our system related to the channel number. In the lower

section of figure 3.9, the related of fit energy against the channel number has shown.

Figure 3.9. Energy spectrum and related fit

3.2.5. Detection efficiency calibration

The detector performance is crucial in radiation measurement for the reason that in

the detector material, and before detection is made possible, the radiation can travel

37

large distances between interactions. The detectors efficiency is not reached 100%.

The gamma spectrometer performance depends precisely on the detection efficiency

knowledge. The latter, and as depicted in Figure 3.10, is a measure of radiation

percentage detected by a given detector from the overall yield emitted by the source.

The detection efficiency can change the volume and shape of the detector material,

absorption cross-section in the material, attenuation layers in front of the detector,

distance and position from the source to the detector. (Akkurt, 2014).

Figure 3.10. Detection efficiency of NaI (Tl) detector as a function of gamma-ray

energies for 0,5 cm distance to detector face (Akkurt, 2014).

3.3. The Linear attenuation Coefficients Measurement

Prepared composite samples have been tested in front of gamma-ray by using gamma

spectrometer in the laboratories of Suleyman Demirel University. The measurements

have been done with the energies 662, 1173 and 1332keV of gamma-rays. The

gamma-rays for those energies from 137Cs and 60Co will be obtained. For each

sample, the linear attenuation coefficients are measured using the relationship

between incoming and passing through material.

38

According to the initial intensity of gamma-ray Io passed through zero thickness of

the material is known, the linear attenuation coefficients may be obtained by the

following equation:

𝐼 = 𝐼0𝑒−𝜇𝓍 (3.1)

Where, I and Io are counted in the detector respectively, with and without the

materials (Figure 3.11). Plotting ln (Io/I) versus x would give straight line and could

be acquired from the slope value.

Figure 3.11. I and Io are measured count in detector respectively with and without

the absorber of thickness x (cm)

The intensity of incoming (Io) and passing through (I) gamma rays is obtained from

energy spectrum. It is a number of count under the spectrum peak. In (Figure 3.12)

displayed the intensity of the gamma rays with and without absorber materials for

137Cs and 60Co. From this figure could be seen that the dot line is (Io), the attenuated

gamma-ray spectrum, and full line is (I) the attenuated gamma-ray spectrum.

39

Figure 3.12. Intensity of gamma rays with and without materials for 60Co and 137C

sources

The half-value layer (HVL), which is the thickness of the transmitted density up to

half of the first density, defines the effectiveness of gamma ray shielding and the

tenth-valur layer (TVL) is the thicknesses of an absorber that will reduce the γ-

radiation to tenth of its intensity. Those are obtained as:

2lnHVL ,

10lnTVL (3.2)

The mean free path (cm) is defined as: the average distance between consecutive

interactions of gamma-ray photons with matter measures its probability of interaction

and it is given as:

𝑚𝑓𝑝 = 1

𝜇 (3.3)

40

4. RESULTS AND DISCUSSIONS

In the present work, the linear attenuation coefficients (µ) cm-1 were measured for

Rubber, Jute and Glass-Fiber (R-J-G/F) composite with different rates which is used

as absorber by using gamma spectrometer system and by using life radioactive

sources 137Cs and 60Co for the energies (662, 1173, and 1332) keV respectively.

The results are shown in Figure 4.1 where the linear attenuation coefficient µ (cm-1)

plotted as a function of gamma energy. It can be seen from this figüre that the highest

value of linear attenuation coefficient at low energy 662 keV of pure rubber in

sample R100 has been obtained while it decreases with increasing ratio of jute in

R90-J10, R80-J20, and R70-J30 respectively. It also decreases with increasing ratio

of glass-fiber in R90-GF10, R80-GF20, and R70-GF30 in composite materials

respectively.

Figure 4.2 shows the linear attenuation coefficient as a fuction of gamma energy for

R100, R90-J10, R80-J20, and R70-J30. It can be seen that the linear attenuation

cofficient decreases with increasing the energy.


R100, R90-GF10, R80-GF20, and R70-GF30. It is also clear from this figüre that the

linear attenuation cofficient decreases with increasing the gamma ray energy.


R100, R70-J10-GF20, R70-J15-GF15, and R70-J20-GF10. It is proved ones again

from this figüre that the linear attenuation cofficient decreases with increasing the

energy.

In Figur 4.5 shows that the linear attenuation coefficient as a fuction of gamma ray

energies and jute rate for all samples. It can be seen that the linear attenuation

cofficient decreases with gamma ray energies and the ratio of jute in the composite.

In Figur 4.6 shows that the linear attenuation coefficient as a fuction of gamma ray

energies and glass-fiber rate for all samples. It can be seen that the linear attenuation

cofficient decreases with gamma ray energies and increasing the ratio of glass-fiber

in the composite.

41

Figur 4.7 shows that the linear attenuation coefficient as a fuction of gamma ray

energies and jute and glass-fiber rates in samples. The energies were 662, 1173 and

1332 keV while rates were 0%, 10%, 15%, 20% for jute and 0%, 10%, 15%, 20% for

glass-fiber. As can be seen from this figure that the linear attenuation coefficient

decreases with both increasing energies and the ratio of jute and glass-fiber.

As the mfp is an important parameter in radiation dosimeter, the mfp results have

been obtained. Figure 4.8 shows the mean free path mfp (cm) as a function of energy

to the samples R100, R90-J10, R80-J20 and R70-J30. It can be seen from these

fugures that, the mean free path of R100 is 11.94 cm, 15.31 cm and 16.37 cm for the

energy of 662, 1173 and 1332 keV respectively. The mean free path of R90-J10 is

12.64 cm at the energy 662 keV, it is increases to 16.74 cm with increasing the

energy to 1173 keV, and it also increases to 17.89 cm with increasing the energy to

1332 keV. The mean free path of R80-J20 is 13.18 cm at the energy 662 keV, it is

increases to 18.71 cm with increasing the energy to 1173 keV, and it also increases to

20.45 cm with increasing the energy to 1332 keV. The mean free path of R70-J30 is

14.25 cm at the energy 662 keV, it is increases to 20.19 cm with increasing the


1332 keV.

Figure 4.9 shows the mean free path mfp (cm) as a function of energy to the samples

R100, R90-GF10, R80-GF20, and R70-GF30. From these results, it can be

concluded that the mean free path of R100 is 11.94 cm at the energy 662 keV, it is


16.37 cm with increasing the energy to 1332 keV. The mean free path of R90-GF10

is 13.3 cm at the energy 662 keV, it is increases to 16.2 cm with increasing the


1332 keV. The mean free path of R80-GF20 is 14.2 cm at the energy 662 keV, it is


18.7 cm with increasing the energy to 1332 keV. The mean free path of R70-GF30 is

15.4 cm at the energy 662 keV, it is increases to 19 cm with increasing the energy to

1173 keV, and it also increases to 20.1 cm with increasing the energy to 1332 keV.

42

Figure 4.10 shows the mean free path mfp (cm) as a function of energy to the

samples R100, R70-J10-GF20, R70-J15-GF15, and R70-J20-GF10. From these

results, it can be seen that the mean free path of R100 is 11.94 cm at the energy 662

keV, it is increases to 15.31 cm with increasing the energy to 1173 keV, and it also

increases to 16.37 cm with increasing the energy to 1332 keV. The mean free path of

R70-J10-GF20 is 13.1 cm at the energy 662 keV, it is increases to 17.1 cm with

increasing the energy to 1173 keV, and it also increases to 18 cm with increasing the

energy to 1332 keV. The mean free path of R70-J15-GF15 is 13.7 cm at the energy

662 keV, it is increases to 19.7 cm with increasing the energy to 1173 keV, and it

also increases to 20.6 cm with increasing the energy to 1332 keV. The mean free

path of R70-J20-GF10 is 14.3 cm at the energy 662 keV, it is increases to 20.9 cm

with increasing the energy to 1173 keV, and it also increases to 22.5 cm with

increasing the energy to 1332 keV. Those all shown that the mfp increases with the

increasing gamma-ray energies.

Figure 4.11 shows the gamma-ray transmission rate (%) through materials as a

function of thickness (cm) to the four samples R100, R90-J10, R80-J20 and R70-J30

at three energies 662, 1173, 1332 keV. From these results, It is clear that while a

smaller thickness of R100 is required to stop 𝛾-rays, the larger thickness is required

for R90-J10, R80-J20 and R70-J30 respectively. The HVL and TVL values for those

materials are also indicated in this figure. The same results can be said for those

values. Also, it can be seen that the HVL and TVL increase with increasing the

energy.


function of thickness (cm) to the four samples R100, R90-GF10, R80-GF20, and

R70-GF30 at three energies 662, 1173, 1337 keV. From these results, It is clear that

while a smaller thickness of R100 is required to stop gamma-rays, a larger thickness

is required for R90-GF10, R80-GF20, and R70-GF30 respectively. The HVL and

TVL values for those materials are also indicated in this figure. Also, it can be seen

that the HVL and TVL increase with the increasing of energy.

43


function of thickness (cm) to the four samples R100, R70-J10-GF20, R70-J15-GF15,

and R70-J20-GF10 at three energies 662, 1173, 1332 keV. From these results, It is

clear that while a smaller thickness of R100 is required to stop gamma-rays, the

larger thickness is required for R70-J10-GF20, R70-J15-GF15, and R70-J20-GF10

respectively. The HVL and TVL values for those materials are also indicated in this

figure. The same results can be said for those values. Also, it can be seen that the

HVL and TVL increase with the increasing of energy.

Figure 4.1. The relationship between 𝝁 (cm-1) and energy (keV) for all composite

samples

44

Figure 4.2. The relationship between 𝝁 (cm-1) and energy (keV) for R100, R90-J10,

R80-J20, and R70-J30

Figure 4.3. The relationship between 𝝁 (cm-1) and energy (keV) for R100, R90-

GF10, R80-GF20, and R70-GF30

45

Figure 4.4. The relationship between 𝝁 (cm-1) and energy (keV) for R100, R70-J10-

GF20, R70-J15-GF15, and R70-J20-GF10

Figure 4.5. The relationship between 𝝁 (cm-1) and jute rate

46

Figure 4.6. The relationship between 𝝁 (cm-1) and glass-fiber rate

Figure 4.7. The 𝝁 (cm-1) as a function of energy and also jute and glass-fiber rates for

R100, R70-J10-GF20, R70-J15-GF15, and R70-J20-GF10

47

Figure 4.8. The relationship between mfp (cm) and energy (keV) for R100, R90-J10,

R80-J20, and R70-J30

Figure 4.9. The relationship between mfp (cm) and energy (keV) for R100, R90-

GF10, R80-GF20, and R70-GF30

48

Figure 4.10. The relationship between mfp (cm) and energy (keV) for R100, R70-J10

GF20, R70-J15-GF15, and R70-J20-GF10

49


R100, R90-J10, R80-J20, and R70-J30

50


R100, R90-GF10, R80-GF20, and R70-GF30

51

Figure 4.13. The relationship between transmission rate (%) and thickness (cm) for

R100, R70-J10-GF20, R70-J15-GF15, and R70-J20-GF10

52

5. CONCLUSION

As the radiation is very important phenemone in our life it is used in a variety of

different fields. On the other hand its higher dose is hazardous for human health and

it must be protected. Shielding is the main method to be protected from radiation and

thus the development of shielding materials becomes popular for researhers. In the

recent years, there have been many studies to find new composite materials used in

the radiation shielding. For this purposes composite materials consist of the Rubber,

as the main material, Jute and Glass-Fiber for radiation shielding have been done and

ten samples have been produced. These produced materials have been tested against

gamma-ray of 662, 1173, 1332 keV energies. The measurement has shown that the

jute and glass-fiber are not important materials to shield gamma rays when we have

added them with rubber. With the increasing jute rate and glass-fiber in the

composite material, the linear attenuation coefficient is decreased. This leads that

rubber without jute and glass-fiber is more useful for shielding against gamma-rays.

Further researches should be carried out to find out more suitable shielding materials.

53

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57

CURRICULUM VITAE

Name and Surname : Abdlhamed Faaq Abdlhamed ABDLHAMED

Birth date: Iraq- Anbar, 1988

Sex: Male

Marital status: Single

Languages: English and Turkish (A2)

E-mail : [email protected]

Educational Status

High School: AL-Anbar, ANNA School, 2006

Bachelor's degree: University of Anbar, College of Education for Pure Sciences

Department: Physics, 2010

Professional Experience :

Worked in schools of Iraqi Sunni Endowment Diwan as a lecturer between

2010 – 2011.

Teacher of Physics in a Secondary School in The Ministry of Education from

2011 to till now.

Publications :

Abdlhamed, A., Günoğlu, K., Parlar, Z., Akkurt, İ. (2017). “Gamma-Ray Shielding

Properties of Some Composite Material”. The 4th International Conference on

Computational and Experimental Science and Engineering (ICCESEN-2017)

Antalya – Turkey.

mailto:[email protected]

58

ÖZGEÇMİŞ

Adı Soyadı: Abdlhamed Faaq Abdlhamed ABDLHAMED

Doğum yeri - tarihi: Anbar – Irak, 1988

Cinsiyeti: Erkek

Medeni durumu: Bekar

Yabancı dili: İngilizce ve Türkçe (A2)

E-posta: [email protected]

Eğitim Durumu

Yüksekokul: AL-Anbar, ANNA Okulu, 2006

Üniversite: Anbar Üniversitesi, Eğitim Fakültesi

Bölüm: Fizik, 2010

Mesleki Tecrübe

Sünnî vakıf okullarında 2010 - 2011 staj öğretmenliği yaptım.

Millî Eğitim Bakanlığı'nda 2011 yılından bu yana bir orta öğretim okulunda

fizik öğretmeni olarak çalışmaktayım.

Yayınlar

Abdlhamed, A., Günoğlu, K., Parlar, Z., Akkurt, İ. (2017). “Gamma-Ray Shielding

Properties of Some Composite Material”. The 4th International Conference on

Computational and Experimental Science and Engineering (ICCESEN-2017)

Antalya – Turkey.

mailto:[email protected]

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