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Glass Materials in Nuclear Technology for Gamma Ray and Neutron Radiation Shielding: a Review

January 2020

Authors:

Aly Saeed

Egyptian Russian University

Raed El Shazly

Al-Azhar University

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The review deals the investigation of glass doped with elements exhibiting high cross section for shielding against the neutrons and gamma ray radiation. The attenuation ability of pure gamma ray, fast neutrons, slow neutrons, and total gamma rays is reviewed and discussed. A short literature review on the glass materials for nuclear radiation shielding is also given.

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Nonlinear Optics, Quantum Optics, Vol. 53, pp. 107–159

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107

Glass Materials in Nuclear Technology for

Gamma Ray and Neutron Radiation Shielding:

a Review

Aly SAeed1, R. M. el ShAzly1, y. h. elbAShAR2,*, A. M. Abou el-AzM1,

M. N. h. CoMSAN3, M. M. el-okR1 ANd W. A. kANSouh3

1Physics Department, Faculty of Science, Al-Azhar University, Cairo, Egypt

2Egypt Nanotechnology Center (EGNC), Cairo University, Giza, Egypt

3Nuclear Research Center, Egyptian Atomic Energy Authority, Egypt

Received: September 1, 2019. Accepted: March 2, 2020.

The review deals the investigation of glass doped with elements exhibit-

ing high cross section for shielding against the neutrons and gamma ray

radiation. The attenuation ability of pure gamma ray, fast neutrons, slow

neutrons, and total gamma rays is reviewed and discussed. A short litera-

ture review on the glass materials for nuclear radiation shielding is also

given.

Keywords: Glass shielding, gamma ray shielding, nuclear technology, neutron

shielding, nuclear spectroscopy

1. INTRODUCTION

Life on earth developed under avpermanent exposure to radiation, which is a

part of our daily life. So the radiation and the life are inseparable. In recent

decades of the twentieth century a huge revolution took place in the use of

nuclear technology in many different applications and the use of ionizing

radiation in various ﬁelds of industrial, agricultural, medical, and many other

areas. Thus the exposure to ionizing radiation is no longer limited to natural

sources but the artiﬁcial sources become a major part of this exposure. Expo-

Corresponding author’s e-mail: y_elbashar@yahoo.com

108 Aly SAeed et al.

sure to ionizing radiation consists mainly of two distinct components, namely,

internal and external exposure. The exposure causes deformation of the

atomic structure of the living tissue, leading to its damage. The protection

from exposure to the nuclear radiation was an important subject considered

largely in the ﬁeld of nuclear physics. The general principles for radiation

protection are time, distance, and shielding. Shielding is generally preferred

due to its efﬁciency in intrinsically safe working conditions, whereas reliance

on distance and time of exposure involves continuous administrative control

over workers [1]. The type and amount of shielding required depends on the

type of radiation, the activity of radiation source and the dose rate that is

acceptable outside the shielding material. However, there are other factors for

the choice of shielding material such as their cost and weight. An effective

radiation shielding will cause a large energy loss in a small penetration depth

without emission of more hazardous radiation. At the same time irradiation

side effects on its physical properties should be small. A number of experi-

mental and theoretical studies were performed on radiation shielding, which

has wide, different application areas with different materials (e.g. concrete,

semi- conductor, polymer, Lipowitz alloy, colemanite, etc.) [2].

Nowadays, some of the promising materials are glassy systems in the

regard of radiation shielding because they can be transparent to visible light

and their properties can be modiﬁed by changing their composition and the

preparation technique. Several glasses were used for nuclear engineering

applications because they accomplish the double task of allowing visibility

while absorbing radiations like gamma-ray and neutrons. A good shielding

glass should exhibit a high absorption cross-section and at the same time the

irradiation effects on its mechanical, electrical and optical properties should

be small. In the last decades, many experimental and theoretical studies, deal-

ing with the investigation of the attenuation and penetration of gamma ray

and neutrons in different types of glassy shielding were carried out [3].

1.1 Literature survey

C.A. Stone et al. [1994], studied glass containing a high concentration of 6Li

and its use for the thermal neutron shielding. The glass composition was the

following: 31 mol% of 6Li2O and 69 mol% of SiO2. Studies were performed

on the second formulation that contained as much as 37 mol% of 6Li2O and

59 mol% of SiO2, with 4 mol% Al2O3 added to prevent crystallization at

such high 6Li2O concentration. The lithium silicate glasses can be formed

into variety of shapes using conventional glass fabrication techniques [4].

Kulwant Singh et al. [2002], investigated the mass attenuation coefﬁ-

cients of the bismuth borate glass at 356, 662, 1173, and 1332 keV photon

energies using a narrow beam transmission method. The calculated values by

(WinXCom software version 3.1) were compared with the obtained experi-

mental data. Appreciable variations were observed in these coefﬁcients

caused by the change in the chemical composition of glasses. On the other

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 109

hand, the coefﬁcients were then used to determine effective atomic numbers

of glass samples, which were found to be constant with bismuth concentra-

tion and energy. The obtained results have shown that the bismuth based glass

can be used as radiation shielding glass [5].

Harvinder Singh et al. [2003] determined experimentally the values of

the mass attenuation coefﬁcients (µm), mean free path (mfp), effective atomic

numbers (Zeff) and effective electron density (Ne,eff) of xZnO.2xPbO.(1-3x)

B2O3 (where x=0.1–0.26) glasses using the photon energies of 511, 662,

1173, and 1332 keV and compared them with the theoretical data. The spe-

ciﬁc volume of glasses was derived from density measurements and studied

as a function of composition. It was pointed out that these glasses are poten-

tially interesting for application in radiation shielding [6].

Narveer Singh et al. [2004] studied the linear and mass attenuation coef-

ﬁcients of two different glass systems: lead borate and bismuth lead borate

glasses, using narrow beam transmission method at 662 keV. The obtained

values were used to calculate the half value layer (HVL) parameters. These

parameters were also calculated theoretically for some standard radiation

shielding concretes at the same energy. The effect of replacing lead by bis-

muth was analyzed in terms of density, molar volume, and mass attenuation.

The two studied glassy systems were found to be promising as gamma-ray

shielding materials. Their radiation shielding properties were also better than

those of investigated concretes, with an added advantage of being transparent

in visible light. Their high values of mass attenuation coefﬁcients and low

values of HVL in comparison to concretes indicated that the volume required

is less than for the concretes [7].

Kulwant Singh et al. [2005] used the transmission method to measure the

mass attenuation coefﬁcients of the system xCaO-(0.3 -x) SrO-0.7B2O3 at

photon energies 511, 662, 1173, and 1332 keV. They used the obtained data

to calculate the effective atomic numbers, and effective electron density. The

mass attenuation coefﬁcients, at a given photon energy, varied slowly with

composition. This was due to the almost constant electron density, and the

fact that the attenuation is determined essentially by the Compton scattering.

The results were compared with the theoretical calculation, carried out by

WinXCom computer program. The obtained data should be helpful in poten-

tial applications in gamma ray shielding, for practical applications, e.g. in

connection with the radioactive waste disposal [8].

Narveer Singh et al. [2006] investigated the PbO.BaO.B2O3 glass sys-

tem in terms of molar mass, mass attenuation coefﬁcients and half value layer

parameters by using gamma-ray at 511, 662 and 1274 keV photon energies.

Gamma-ray attenuation coefﬁcients of the prepared glass samples were com-

pared with the tabulations based upon the results of XCOM computer soft-

ware. Good agreement was observed between the experimental and the

theoretical data. The results had uncertainty less than 3%. The radiation

shielding properties of the glass system were compared with some standard

110 Aly SAeed et al.

radiation shielding concretes and showed that these glasses are better than

ordinary concrete in terms of the volume required for shield design, with

added advantage of being transparent in visible light [9].

Kulwant Singh et al. [2006] determined experimentally the mass attenu-

ation coefﬁcients for lead and bismuth borate ﬂy-ash glasses at 81, 356, 511,

662, 1173, and 1332 keV photon energies, using a narrow-beam transmission

method. The coefﬁcients of glasses were then used to determine their interac-

tion cross sections, the photon mean free path, effective atomic numbers, and

the electron densities. Results indicated that these ﬂy-ash glasses exhibit an

interesting potential for application in low-energy gamma ray shielding. This

was due to their large mass attenuation coefﬁcients and low values of HVL in

comparison to concretes. These authors observed also that the bismuth glasses

have higher density and mass attenuation coefﬁcients than lead-containing

ones. It has shown that bismuth could substitute lead to improve the radia-

tion-shielding properties of glasses. It is apparent that the ﬂy-ash glasses can

be advantageously used as the radiation-shielding materials [10].

Sukhpal Singh et al. [2008] measured the mass attenuation coefﬁcient of

barium–borate–ﬂy-ash glasses for γ-ray photon energies of 356, 662, 1173

and 1332 keV using a narrow beam transmission geometry. The theoretical

values were obtained by using the ‘mixture rule’ and the ‘XCOM’ computer

software. The photon beam was highly collimated and overall scattering

acceptance angle was less than 3%. The uncertainty in results was less than

3%. The so obtained coefﬁcients were then used to calculate values of mean

free path, effective atomic number and electron density. Good agreements

were observed between the experimental and the theoretical values. From the

obtained results it was reported that from the shielding point of view the

barium–borate–ﬂy-ash glasses exhibit better shieldings to gamma radiation

than the standard radiation shielding concretes and the ordinary barium–

borate glasses [11].

K.J. Singh et al. [2008] determined experimentally the gamma-ray atten-

uation coefﬁcients using a narrow beam transmission method. It was done for

the xPbO(1-x)SiO2 (x = 0.45–0.70) glass system at 662, 1173 and 1332 keV

photon energies. These values were also obtained theoretically using the

‘mixture rule’ and the ‘XCOM’ computer software. The results were then

used to calculate the half value layer parameters. Gamma-ray shielding prop-

erties of PbO–SiO2 glass samples were compared with ordinary concretes.

The molar volume, FTIR and acoustic investigations were employed to study

the structural properties of the prepared glass system. The PbO–SiO2 glasses

are potential candidates for gamma ray shielding. Lower values of HVL in

PbO– SiO2 glasses as compared to that of concrete suggest smaller size

requirements with the advantage of being transparent visible[12].

S.R. Manohara et al. [2009] calculated the effective atomic number

(Zeff), the effective electron density (Ne,eff), and the energy dependence

(ED), at photon energies from 1 keV to 1 GeV for CaO–SrO–B2O3, PbO–

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 111

B2O3, Bi2O3–B2O3, and PbO–Bi2O3–B2O3 glasses with potential applica-

tions as gamma ray shielding materials. For medium-Z glasses, Zeff is almost

constant and equal to the mean atomic number in a wide energy range, typi-

cally 0.3 < E < 4 MeV. In such cases, Compton scattering is the dominating

photon interaction mechanism. In contrast, for high-Zeff glasses there is no

energy region where Compton scattering is truly dominating. The results

were compared with concrete and other standard shielding materials and

showed an additional advantage of being transparent in visible. The single-

valued effective atomic number calculated by XMuDat is approximately

valid at low energies where photoelectric absorption is dominating. Heavy-

metal oxide glasses containing PbO and/or Bi2O3 are promising gamma ray

shielding materials due to their high effective atomic number and a strong

absorption of gamma ray. [13].

K. Kirdsiri et al. [2009] studied the mass attenuation coefﬁcients, total

interaction cross-sections and effective atomic numbers of lead borate glass

system with different compositions at 662 keV gamma ray energy. The results

were in fair agreement with the theoretical values, calculated by the WinX-

Com program. The mass attenuation coefﬁcients were increasing with the

increase PbO content. This is due to the increase of effective atomic number

of glass samples, which increase the probability of photoelectric absorption

in glass. However, the Compton scattering gives dominant contribution to the

total mass attenuation coefﬁcients for the studied glass samples. The shield-

ing properties of studied glass samples are better than the ordinary concrete

shieldings and commercial window glass with added advantage of transpar-

ency in visible region. These results indicated a good potential glasses for use

as radiation shielding materials [14].

R.S. Kaundal et al. [2010] prepared glasses of the system PbO–SrO–

B2O3 with the value of molar ratio R (=PbO/B2O3) in the region 0.14≤R≤2.0

using the melt quenching technique, in order to evaluate gamma-ray shield-

ing properties. Mass attenuation coefﬁcients have been calculated with the

XCOM computer program. The longitudinal velocities of ultrasonic waves

were measured in glass samples at room temperature using the pulse echo

technique. The results indicated that with the increase of R-value, stability of

glass network decreases. The decrease of glass network indicates the increase

in the number of boron atoms with non- bridging oxygen at the expense of

decrease of tetrahedral borate units. This feature may lead to open glass struc-

ture with lesser rigidity of the glass samples. DSC studies have been under-

taken to measure the glass transition temperature and to get an idea about the

stability of the glass network with increasing R value [15].

J. Kaewkhao et al. [2010] determined the mass attenuation coefﬁcients

and some other shielding parameters of borate glass matrices containing with

Bi2O3 and BaO at 662 keV, and compared with PbO in the same glass struc-

ture. The theoretical values were calculated by WinXCom software and com-

pared with the experimental data. The results found that the mass attenuation

112 Aly SAeed et al.

coefﬁcients increased with the increasing of Bi2O3, BaO and PbO concentra-

tion. Moreover the HVL of glass samples were also better than for ordinary

concretes and commercial window glass. These results reﬂected that the Bi-

based glass can replace Pb in radiation shielding material. In the case of Ba

compound, it may be used at appropriate energy such as X-rays or lower

energy of gamma rays [16].

P. Limkitjaroenporn et al. [2011] studied the lead sodium borate glasses

with compositions xPbO-20Na2O-(80-x)B2O3 (where x=5, 10, 15, 20, 25,

30, 35, 40, 45, 50 and 55 mol%). The samples were prepared using the melt-

quenching method and investigated with respect to their optical, physical,

structural and gamma-rays shielding properties. The densities of these glass

samples were increasing with the increase of PbO concentration. The FTIR

spectra and molar volumes indicated that PbO acts differently in this glass

structures over their compositions. Gamma-ray shielding properties, param-

eters such as mass attenuation coefﬁcients, effective atomic number and half

value layer were increasing with the increasing PbO concentration. The

results showed a relative difference of less than 1% between experimental

and theoretical values. Moreover, the half value layers of the glass systems

were compared with some standard radiation shielding materials and they

exhibited better shielding properties than barite and ferrite concretes at 15

and 25 mol% of PbO, respectively [17].

K. Kirdsiri et al. [2011] investigated the radiation shielding and optical

properties of xBi2O3-(100-x)SiO2, xPbO-(100-x)SiO2 and xBaO-(100- x)

SiO2 glass systems (where 30 ≤ x ≤ 70 is the composition by weight%). Mass

attenuation coefﬁcients of glasses at 662 keV were improved by increasing

the Bi2O3 and PbO content, which raised the photoelectric absorption in the

glass matrices. The results indicated that photon is strongly attenuated in

Bi2O3 and PbO containing glasses, more than in BaO containing glass. The

results from the optical absorption spectra showed an edge that was not

sharply deﬁned; clearly indicating the amorphous nature of glass samples. It

was observed that the cutoff wavelength for Bi2O3 containing glass was lon-

ger than for PbO and BaO containing glasses [18].

H. A. Saudi et al. [2011] have prepared a glass system with chemical

formula xBi2O3–(30-x) CdO–10B2O3–20Fe2O3-40P2O5 (0≤ x ≤30)

weight % to be used as a radiation shield. The mass attenuation coefﬁcient

and half value layer of the glass system to gamma ray were measured

experimentally and compared with those calculated using the WinXCom

program. A database of effective mass removal cross-sections for fast neu-

trons was also introduced in this work. The obtained results were correlated

with the structural characteristics of these glasses obtained from their IR

spectra. The inﬂuence of gamma and neutrons irradiation on structural

properties was also studied [19].

Natthakridta CHANTHIMA et al. [2011] investigated experimentally

and theoretically the mass attenuation coefﬁcients and partial interaction of

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 113

xRmOn:(100-x)SiO2 glass system (where RmOn are Bi2O3, PbO and BaO,

with 30≤ x ≤70 % by weight) at the photon energy of 662 keV. The theo-

retical values of total and partial interaction were obtained by the WinX-

Com software. The glass system was prepared by the melt-quenching

method and its radiation shielding properties were measured. Mass attenu-

ation coefﬁcients of glasses increased with the increasing Bi2O3 and PbO

content. On the contrary, there was no signiﬁcant change in mass attenua-

tion coefﬁcient when the fraction of BaO increased. These results indicated

that photons were more attenuated in Bi2O3 and PbO. The HVL and effec-

tive atomic number results indicated that Bi can replace Pb at this energy as

a gamma-ray shielding material. For Bi2O3 and BaO, HVL was better than

for the ordinary concrete and commercial windows. It meant that Bi2O3

and BaO glasses can be used to shield gamma rays. It was suggested that

Bi2O3 and BaO glass will open new possibility for lead-free radiation pro-

tecting glasses with non-toxicity [20].

J. Kaewkhao [2011] have synthesized the Bi2O3-SiO2 glass system

by using the melt-quenching method. The radiation shielding properties

of the glass samples at various levels of bismuth content were measured

at 662 keV. Comparisons were made with theoretical values calculated by

using the WinXCom. The experimental values were generally in good

agreement with the theoretical ones. Furthermore, a comparison was

made with a lead-borate glass system with the same level of the additive.

The radiation shielding properties were found to improve with increasing

Bi2O3 concentration. The different values of Compton scattering yielded

a higher mass attenuation coefﬁcients for the bismuth-silicate than for the

bismuth-borate glass, respectively. These results reﬂect a potential useful-

ness of bismuth-based glasses as new materials for lead-free radiation-

shielding glasses [21].

Suparat TUSCHAROEN et al. [2011] have prepared glasses with com-

position xBaO.(80-x)B2O3.20Fly-ash (x = 45, 50, 55, 60, 65 and 70 wt.%)

using the melt-quenching method. Mass attenuation coefﬁcients of the fabri-

cated glass system at 662 keV were determined using the narrow beam trans-

mission method. The results were consistent with theoretical calculations

done using the WinXCom software. The mass attenuation coefﬁcient for the

given photon energy was found to vary slowly with composition but to

increase with the increasing fraction of BaO. The glass samples having 45-60

wt.% of BaO content showed the value of mean free path shorter than in the

case of ordinary hematite-serpentine, ilmentite-limonite and basalt-magnetite

concretes. The values were smaller than in the case of all conventional con-

cretes with higher than 65 wt.% content. The mean free path results showed

that the better shielding properties were achieved at higher BaO concentra-

tions. The density measurements were reported as a function of the BaO frac-

tion. The data could be useful for potential applications of ﬂy-ash in

developing of radiation shielding glasses [22].

114 Aly SAeed et al.

P. Limkitjaroenporn et al. [2011] have determined the mass attenuation

coefﬁcients of glass systems xPbO: 20Na2O: (80-x) B2O3 (x = 5, 10, 15, 20,

25, 30, 35, 40, 45, 50 and 55% mol) at 662 keV photon energy using the

gamma rays transmission method. The theoretical values of mass attenuation

coefﬁcients were calculated by the WinXCom program. The coefﬁcients

were then used to determine the effective atomic numbers of the glass sam-

ples. All shielding parameters increased with the increasing PbO concentra-

tion. [23].

Krit Won-in et al. [2011] have prepared colorless lead free glass sam-

ples from 40 wt% local quartz sand and various concentration of BaCO3

(20-40 wt%) as the main composition in order to study their radiation

shielding properties. The gamma ray attenuation characteristics were stud-

ied for the photon energy of 662 keV. Density and refractive index were

also determined. They found that changes of prepared glass compositions

can produce better material in terms of both clarity and radiation shielding

properties. The attenuation coefﬁcients were found to increase linearly with

the increasing BaCO3 content. Both the density and refractive index were

found also to increase. The measured linear and mass attenuation coefﬁ-

cients of the glass sample containing 40-wt % BaCO3 were 0.234 cm-1 and

0.0726cm2g-1, respectively. The lead-free high refractive index glass, that

is one of the ecofriendly environmental materials, can be used as the gamma

radiation shielding glass [24].

S. R. Manohara et al. [2011] introduced a geometric progression ﬁtting

formula for calculating the gamma ray exposure buildup factors for three

heavy metal oxide (HMO) glass systems: PbO-Bi2O3-B2O3, PbO-B2O3,

and Bi2O3-B2O3. Computations were done by interpolation method using

the geometric progression ﬁtting formula and ANSI/ANS- 6.4.3 library for

the energy range from 15 keV to 15 MeV, up to penetration depths of 40 mfp

(mean free path). The buildup factors were studied as function of incident

photon energy and penetration depth. These factors for HMO glasses cannot

be found in any standard database. They are useful for practical calculations

in gamma ray shield designs. They help also to determine and control the

thickness of the used shielding material [25].

S. Kaewjaeng et al. [2012] studied the effect of BaO on optical, physical

and radiation shielding properties of SiO2-15B2O3-2Al2O3- 10CaO-

23Na2O glasses. Glasses in composition (50-x) SiO2-15B2O3- 2Al2O3-

10CaO -23Na2O-xBaO, where x = 0, 5, 10, 15 and 20 % mol were fabricated

by the melt-quenching technique. The results have shown that the density and

molar volume of glasses were found to increase with increasing BaO concen-

tration. The optical spectra were measured by UV-VIS spectrophotometer

and showed a shift to longer wavelength. Gamma-ray shielding properties at

662 keV of SiO2-B2O3-Al2O3-CaO-Na2O-BaO glasses were studied. The

experimental results showed a good agreement with the theoretical values

which calculated by WinXCom program. Moreover, the comparison of ordi-

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 115

nary concrete and commercial window in term of half value layer was dis-

cussed and showed better radiation shielding properties with the advantage of

being transparent in visible light [26].

N. Chanthima et al. [2012] calculated the mass attenuation coefﬁcient,

effective atomic number, effective electron density and HVL of xRmOn:(1-x)

SiO2 glass system (where RmOn are Bi2O3, PbO and BaO, with 0.3 ≤ x ≤ 0.7

as fraction of weight) by theoretical approach using WinXCom program in

the energy region from 1 keV to 100 GeV. In addition, the HVL of these glass

samples was compared with some standard shielding concretes. They

observed that the value of these parameters change with energy and composi-

tion of the silicate glasses. Better shielding properties of glass samples were

obtained as compared with some standard shielding concretes. These results

indicate that these glasses can be used as radiation shielding materials [27].

Sandeep Kaur et al. [2013] evaluated the gamma ray shielding properties

of PbO-Al2O3-B2O3 and PbO-Al2O3-SiO2 glass systems in terms of mass

attenuation coefﬁcient, half value layer, mean free path and effective atomic

number parameters. Structural information of both the glass systems has been

obtained by using density, XRD, DSC and ultrasonic measurements. Gamma

ray shielding properties of the investigated glass systems have been com-

pared with standard nuclear radiation shielding concretes. PbO-B2O3-Al2O3

and PbO-SiO2-Al2O3 glass systems can be the potential candidates for

gamma-ray shielding applications. Higher mass attenuation coefﬁcients and

lower HVL values of the investigated glass samples for most of the composi-

tion range show that most of the prepared glass samples are better than exist-

ing concretes. They have estimated that these glasses have smaller volume

requirements than the others as gamma ray shielding materials in nuclear

reactors [28].

2. RADIATION SHIELDING MATERIALS

2.1. Introduction

Radiation refers to the whole spectra of electromagnetic waves as well as

atomic and subatomic particles that were discovered [29]. The main sources

of radiation can be categorized as given in Table 1.

Radiation can be divided into two main categories: ionizing and non-ionizing

one. Non-ionizing radiation is electromagnetic radiation with wavelength of

about 10 nm or longer, this part of electromagnetic spectra includes “Radio-wave,

Micro-wave, Infrared, Visible light and Ultraviolet”. The ionizing radiation refers

to the radiation, which can remove an electron from atoms or molecules of the

medium, it traverse. The ionizing radiation includes the rest of electromagnetic

spectra with wavelength shorter than 10 nm such as X-rays and γ-rays as well as

atomic and subatomic particles such as α-particle, β-particle, positron, proton,

neutron, heavy ions, and mesons. The ionizing radiations are commonly classi-

116 Aly SAeed et al.

ﬁed into two principal types. The ﬁrst is the directly ionizing radiations that

include radiations of energetic particles carrying an electric charge, such as beta,

alpha particles, protons, and heavy ions. Another type of ionizing radiation are

the indirectly ionizing, such as neutrons and gamma rays, which are not charged

and cause ionization by an indirect way [31].

Directly ionizing radiation interacts very strongly with shielding media,

therefore they can be easily stopped. A very thin barrier like sheet of paper

can stop alpha particles so that alpha particles are usually not hazardous out-

side the body, but they cause damage if ingested. By contrast, indirectly ion-

izing radiation may be more penetrating and the shielding required may be

quite massive and expensive, gamma rays and neutrons are very penetrating

and can pass through thick barriers [32]. Several feet of concrete or lead

would be needed to stop highly energetic gamma rays. On the other hands,

light elements such as hydrogen and parafﬁn are needed to moderate a fast

neutron and high capture cross section elements such as boron and cadmium

are necessary to absorb thermal neutrons (Fig 1).

For these reasons nowadays much attention has been paid for shielding

against neutrons and gamma ray because they are the highly penetrating

through matter. Any medium can stop gamma ray and neutrons is very sufﬁ-

ciently to stop any other type of nuclear radiation [32].

FIGURE 1

Penetrating power of ionizing radiation [32].

TABLE 1

Main sources of radiation [30].

Category Source Type of radiation

Natural Radioactivity (decaying process) α, β- particles and γ-rays

Cosmic Rays Neutrons, protons, electrons, photons, etc

Artiﬁcial Nuclear Reactors Neutrons, γ-rays and residual radioactivity

Accelerators Electrons, protons, X-rays, neutrons, etc

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 117

2.2. Radiation Hazards and Principles of Protection

People are exposed to low levels of radiation every day from background

radiation and from artiﬁcial sources. Low levels of radiation do not cause any

detectable harm. Exposure to higher levels of radiation, however, can affect

the body in a number of harmful ways, and these damaging effects may not

be seen for many years, due to the increased susceptibility of the human body

to damage from ionizing radiation, the National Council on Radiation Protec-

tion and Measurement (NCRP) recommended maximum radiation dose lim-

its allowable annually (Table 2) [33-34].

The severity of the effects depends on the amount of radiation absorbed by

the body, the type of radiation, how the radiation is taken into the body

(breathed in, eaten, or absorbed through cuts in the skin), and the length of

time a person is exposed [36].

Exposure to above-normal levels of radiation can completely kill or dam-

age living tissue, it can also increase the risk of developing cancer much later

in life. On the other hand, the exposure to very large doses of radiation can

lead to the death (Table 3) [36].

With increasing use of radioactive sources human exposure to ionizing

radiation has increased and the health hazards are increased, so practical

TABLE 2

Dose limit recommendations (N.C.R.P) [35] in Sieverts (Sv).

Type of exposure group Maximum permissible absorbed

dose equivalent

Whole Body:

– Prospective annual limit

– Reprospective annual units

– Long term accumulation to N years of age

¾Skin

¾Hands

¾Forearms

¾Other organs

¾Fertile women

50 mSv any one year

100-150 mSv, any one year

500 (N-18) mSv

150 mSv, any one year

750 mSv, any one year

300 mSv, any one year

150 mSv, any one year

5 mSv is gestation period

Public or occasionally exposed individual

¾Individual or occasional

¾Students

5 mSv, any one year

1 mSv, any one year

Population as a whole 1.750 mSv, any one year

Emergency

– life saving

•Whole body

•Hands, forearms

– less urgent

•Whole body

•Hands, forearms

1000 mSv

2000 mSv

250 mSv

1000 mSv, total

118 Aly SAeed et al.

methods of protection were developed. In essence, the three effective factors

of protection from ionizing radiation are [37]:

i) Distance from radiation source

Distance is a very effective protection factor and often the least expensive

way to reduce radiation exposure. As one moves away from a point source of

radiation, the amount of radiation at a given distance from the source is

inversely proportional to the square of the distance (inverse square law) [37].

I2 = I1[d1/ d2]2

I2 = intensity at a distance (d2) from a point source

I1 = intensity at distance (d1) from the same point source

Exposure time

Reducing the exposure time is a very practical method of radiation protec-

tion. Since the amount of exposure occurs as a function of its duration.

Shorter time means less absorption of radiation [37].

Dose rate × exposure time = total dose

ii) Shielding materials

Along with understanding of the characteristics and potential beneﬁts of dif-

ferent types of radiation came awareness of their potential harm, thus from

the need for radiation protection, the radiation shielding design and analysis

born. Radiation shielding serves a number of functions, foremost among

these is reducing the radiation exposure to persons near radiation sources,

shielding used for this purpose is named biological shielding. Shields are also

used in some nuclear reactors to reduce the intensity of γ-rays incident on the

reactor vessel, which protects the vessel from excessive heating due to γ-ray

TABLE 3

Biological effects of acute dose [36].

Acute dose Probable effect

0-250 mSv No obvious injury

250-500 mSv Possible blood changes but no serious injury

500-1000 mSv Blood-cell changes, some injury, no disability

1000-2000 mSv Injury and disability certain

2000-4000mSv Injury and disability certain, death possible

4000 mSv 50 percent fatal

6000 or more mSv Probably 100 percent fatal

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 119

absorption and reduces radiation damage due to neutrons; these shields are

named thermal shields. Sometimes shields are used to protect delicate elec-

tronic apparatus that otherwise would not function properly in a radiation

shield such apparatus shields are used, for example, in some types of military

equipment [38].

2.3. Radiation shielding materials

Variety of materials can be used for radiation shielding, but to choose an

appropriate type of shielding material, the type of radiation that is being

shielded, the energy of the radiation, and the level of dose reduction are in

need to be considered. In choosing a shielding material, the ﬁrst consider-

ation must be effectiveness. If dealing with external radiation protection, the

most important consideration is personnel protection. An effective shield will

cause a large energy loss in a relatively small penetration distance without

emission of more hazardous radiation. However, other factors may also inﬂu-

ence the choice of shielding materials such as, cost, weight of the materials as

well as how much space is available for the materials. The effectiveness of the

shielding material is determined by the interactions between the incident

radiation and the atoms of the absorbing medium. The interactions, which

take place, depend mainly upon the type of radiation, the energy of the radia-

tion, and the density of the absorbing medium. Among all the types of radia-

tion, gamma ray and neutron radiation shielding materials will be discussed

in some details [39].

2.3.1. Gamma-Ray Shielding Materials

Gamma radiation is the most difﬁcult to shield against it. Therefore pres-

ents the biggest problem in the nuclear plant. The penetrating power of the

gamma is due, in part, to the fact that it has neither charge no mass, there-

fore, it does not interact as frequently as do the other types of radiation. The

most probable of gamma interaction are those near the nucleus or interac-

tions with the electrons around the nucleus. For this reason, more gamma

interactions occur in dense materials that have many electrons [40]. The

most common form of effective gamma rays shielding materials are: lead,

steel, concrete …etc. A few details of common gamma rays shielding mate-

rials are given below.

Lead: lead is a common shielding material used for gamma radiation. The

properties of lead that make it excellent shielding materials are its density,

high atomic number, high level of stability, ease of fabrication, high degree of

ﬂexibility in application and its availability. [41]

Concrete: concrete is widely used as a shielding for gamma rays because

it is cheap and adaptable for any construction design. Concrete often used in

the construction of large volume shields because of its low cost and most

commonly used as the outer constituent of a shield with its own activity

shielded by an inner layer of steel, lead or other materials. [42]

120 Aly SAeed et al.

Iron: It is also a common gamma-ray shielding material and is often used

in situations where the size or conﬁguration of the shield would make its

construction from lead alone too expensive. In such circumstances, an outer

layer of steel with inner lead lining is often an effective compromise [43].

Tungsten: For small shields or collimators, tungsten is an attractive mate-

rial. It has been shown that tungsten can be effectively and economically

utilized in large-scale applications and in shielding conditions that require a

high degree of radiation attenuation in a limited thickness. It possesses supe-

rior shielding factors compared to lead while removing the accompanying

toxicity hazard and mixed waste processing costs. However, tungsten is over

thirty times more expensive than lead; therefore, it is used sparingly and is

almost never used for massive shields [44].

Depleted uranium: DU is an excellent gamma radiation-shielding mate-

rial; it can provide equivalent gamma shielding with less thickness than other

materials such as lead and stainless steel. In shipping container applications,

DU has been used to provide gamma shielding for spent fuel casks. Another

advantageous of DU is that it can provide structural strength similar to stain-

less steel, although this characteristic of DU has not yet been recognized by

regulatory agencies [45].

2.3.2. Neutron shielding materials

Shielding against neutrons is based on three successive steps

1. Slow down fast neutron by scattering.

2. Absorb the slow neutrons.

3. Absorb the secondary produced gamma rays

The effective materials in slowing down neutrons are the light elements, par-

ticularly hydrogen. Many hydrogenous materials, such as water and parafﬁn,

make efﬁcient neutron shields. However, water shields have the disadvantage

of needing maintenance; also, evaporation can lead to a potentially dangerous

loss of shielding while parafﬁn is ﬂammable. At higher energies (10 MeV),

the cross section for interaction with hydrogen (1 barn) is not as effective in

slowing down neutrons. If the energy of the fast neutron is sufﬁciently high,

inelastic scattering with nuclei can take place in which the recoil nucleus is

elevated to one of its excited states during the collision. The nucleus quickly

de-excites, emitting a gamma ray, and the neutron loses a greater fraction of

its energy than it would in an equivalent elastic collision [46]. Inelastic scat-

tering and the subsequent gamma ray play an important role in the shielding

of high-energy neutrons. Materials with good inelastic scattering properties,

such as iron, are used to offset this decrease in cross section with increased

neutron energy. Lead, which is generally used to provide shielding against

gammas, has another advantage of shielding against fast neutrons because of

its high cross section for inelastic scattering when it is uniformly distributed

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 121

throughout a hydrogenous material. These materials can cause a large change

in neutron energy after collision for high-energy neutrons but have little effect

on the neutrons at lower energy, below 0.1 MeV [46].

Thermal neutrons with energies of 0.025 eV or less are absorbed with high

effectively by thin layers of boron or cadmium. Boron is often used in the

form of boron carbide (B4C) or boron-loaded solutions. One commonly used

material is Boral, a mixture of boron carbide and aluminum, which is avail-

able in sheets of varying thickness. Cadmium has the disadvantage of emit-

ting high-energy gamma rays after neutron capture, which may necessitate

additional gamma ray shielding. Cadmium is more effective than boron for

absorbing thermal neutrons, whereas boron is more effective for absorbing

epithermal neutrons (energy range 0.1 eV to 10 eV) [46].

2.3.3. Current Neutron-Gamma Radiation Shielding Materials

An effective radiation shield consists of a combination of hydrogenous or

other low-atomic mass number materials to moderate neutrons, as well as

materials with high capture cross section for thermal neutron, and heavy ele-

ments to absorb gamma rays (Table 4).

Examples of hybrid shielding materials are polyethylene and lead, con-

crete containing scrap iron, and more exotic materials such as lithium hydride.

Shields that are more effective can sometimes be obtained by adding boron,

lithium, or lead to polyethylene [47].

2.4. Novel shielding materials

Some studies already carried out to seek for novel shielding materials with

both the neutron and gamma ray shielding ability. A number of them are

brieﬂy presented below.

Special glasses have been developed which accomplish the double task

of allowing visibility while absorbing gamma ray and neutrons, thus pro-

TABLE 4

Material properties for gamma and neutron absorption [38].

Material

Density

(g/cm3) Advantages Disadvantages

Lead 11.350 High gamma attenuation Poor neutron absorption

Cadmium 8.642 Good neutron absorption High mass

Concrete 2.320 Inexpensive, high neutron moderation High volume

Boron 2.31 Good neutron absorption Poor gamma absorption

Carbon 2.250 good neutron moderation, Low mass Poor gamma attenuation

Water 1.0 Inexpensive, high neutron moderation Vapor unless pressurized

Lithium 0.533 low mass, High neutron absorption Poor gamma absorption

122 Aly SAeed et al.

tecting the observer. Glasses rich in heavy elements such as lead, bismuth

and barium are an excellent absorber for gamma rays, on the other hand

glasses containing boron and cadmium are used to absorb slow neutrons..

In industry, a variety of neutron-gamma shielding materials have been

developed and supplied by industry. Some examples are given here., boro-

nated rubber and other shielding rubber & polyethylene composite (e.g.,

lead-rubber & plastics composite) are produced in the Boron Rubbers India

Company. The composite neutron-gamma guard mixtures which have been

proved the attributes/properties for potential applications in management of

solid, liquid and sludge radioactive wastes, as well as mixed grout mixtures.

In some applications, given the additional need for weight reduction and

practical considerations, there is a need for multifunctional materials which

could perform structural or other roles while providing good radiation shield-

ing capability. Characterization of this neutron-gamma shielding material

and the methods to enhance strength and ductility of this integration structure

is being studied [38].

3. GLASSY MATERIALS

3.1. Introduction

Glass is an amorphous material that can be prepared by melting and quench-

ing method. The glass was used from ancient in many applications like

cups, windows, bottles etc. Nowadays the glass ﬁnds enhanced interest due

to its wide range applications such as lasers, optical recording, optical ﬁl-

ters, smart windows and radiation shielding, and detection. These types of

application are very important in many space of modern technology [48].

3.2. Glass Deﬁnition

The glass can be deﬁne as an amorphous materials completely lacking in long

range order, periodic structure and exhibiting a region of glass transforma-tion

behavior, so that any materials organic, or inorganic, or metallic formed by any

technique, which exhibits glass transformation behavior, is a glass [48].

3.3. Volume- temperature diagram

The relation between crystal, liquid and glass can be explained by means of

V-T curve, which reﬂects the behavior observed for substance, which upon

cooling from a liquid state either changes into a crystal or moves into the

glassy state depending upon the rate of cooling.

As shown in Figure 2 regardless of whether melts had originated from a

crystal line or vitreous solid; above the melting point (Tm) they are identical

and show the identical decrease of volume with temperature along AB. If the

rate of cooling is slow, crystallization occurs if i) There is a sufﬁciently large

number of nuclei forming in unit time (nucleation frequency) and ii) The

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 123

linear growth rate of these nuclei are large enough. The volume is decreased

sharply along BC, upon further cooling of the crystal so formed shrink along

the crystal line to point D. If the rate of cooling is sufﬁciently rapid, a super-

cooled liquid is obtained. As temperature decreases, the structure of the liquid

continues to rearrange, but no abrupt decrease in volume observed, due to a

discontinuous structural rearrangement occurs. As cooling continues, the vis-

cosity of the system is rapidly increases. At sufﬁcient low temperature, the

molecular groups can’t rearrange themselves fast enough to reach the volume

characteristic of that temperature and the frozen liquid is now glass. This is

called glass transformation region and occurs over a range of temperature not

a ﬁxed point. Therefore, every glass has its characteristic temperature (Tg) in

which only below it the material is glass. If the temperature of the glass is

held constant at T, which is a little below Tg, the volume G continue to

decrease slowly. Eventually it reaches the level G’ on the dotted line, which

is a smooth continuation of contraction graph BE of the supercooled liquid.

The properties of glass show a change with time in the vicinity of Tg. This

process by which the glass reaches a more stable condition is known as stabi-

lization. Above Tg no such time – dependence properties is observed. Because

FIGURE 2

Relationship between glassy, liquid, and solid states [47].

124 Aly SAeed et al.

of the existence of stabilization effects, the properties of glass depend to a

certain extent on the rate at which it has been cooled particularly through the

transformation range [48].

3.4. Structural theories of glass formation

The earliest glasses used by man where found in nature and provide to be

useful to early man it is not surprising that the desire to produce glasses at

will developed thousands year ago. Many theories were development to

explain the glass formations, which usually are grouped under the heading of

structure theories of glass formation.

3.4.1. Goldschmidt observations

The earliest and simplest theory of glass formation was based on the observa-

tions by Goldschmidt that glasses of the general formula RnOm from most

easily where the ionic radius of the cation R to the oxygen ion lies in the

range 0.2 to 0.4 since radius ratios in this range tend to produce cations sur-

rounded by four oxygen ions in the form of tetrahedral (Fig. 3) [48].

3.4.2. Zachariasen’s Rules

In 1932 W.H. Zachariasen published a paper “The Atomic Arrangement in

Glass”, which described the ability of various oxides to form glass and had

great impact on material scientists of the period. Zachariasen’s work placed

the understanding of glass structure and its relationship to composition on a

chemical basis for the ﬁrst time.

Zachariasen considered the relative glass forming ability of oxides and con-

cluded that the ultimate condition for glass formation is that a substance can

form extended three-dimensional networks lacking periodicity but with energy

content comparable with that of the corresponding crystal network. From these

conditions, he derived four rules for oxide structure that allow selection of those

oxides that tend to form glasses. The rules are the following:

1. An oxygen atom is linked to not more than two glass-forming atoms.

2. The coordination number of the glass-forming atoms is small.

FIGURE 3

Tetrahedral form for glass [47].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 125

3. The oxygen polyhedra share corners with each other, not edges or faces.

4. The polyhedra are linked in the three dimensional network.

From these considerations, Zachariasen concluded that the following oxides

should be glass formers:

B2O3, SiO2, GeO2, P2O5, As2O3, Sb2O3, V2O5, P2O3, Sb2O5, As2-

O5, Nb2O5, and Ta2O5.

At the time of Zachariasen’s research, only the ﬁrst ﬁve oxides above [in

bold] had been made into glasses, but since then glasses of two other oxides

[underlined] in the list have been prepared. Five others on his list have never

been successfully prepared as glasses [49].

3.4.3. Stanworth classiﬁcation

Stanworth suggested three more criteria on the glass-forming tendency of

oxides. These three criteria are:

1. The cation valance must be three or greater

2. The tendency of glass formation increases with decreasing cation size.

3. The electronegativity should be between 1.5 and 2.1 on Pauling’s scale.

Accordingly, Stanworth divided all oxides into four groups according to the

glass forming ability. They are:

i. Strong glass formers, which form glasses by their own when the melts

of their oxides are cooled. e.g., SiO2, GeO2, P2O5, and B2O3

ii. Intermediate glass formers which form glasses by their own but only on

rapid cooling, e.g., As2O3, Sb2O3 etc

iii. Conditioned glass formers, which will not form glasses by their own but

each will do so when melted with suitable quantity of non-forming

oxides, e.g. TeO2, SeO2, MoO3, WO3, Bi2O3, Al2O3 and V2O5.

iv. Non-glass forming oxides, e.g. oxides of Tin and Chromium.

The main glass-forming oxides are SiO2, B2O3, GeO2, and P2O5 all of

which come from a certain area of the periodic table. They are oxides of ele-

ments with intermediate electro-negativity. These elements are not sufﬁ-

ciently electropositive to form ionic structures, but also are not sufﬁciently

electronegative to form covalently bonded molecular structure. Instead,

bonding is usually a mixture of ionic and covalent, and the structures are best

regarded as three-dimensional polymeric structures. Oxides of other elements

around this group in the periodic table also show a tendency for glass forma-

tion. Oxides like Na2O, K2O, PbO and CaO which when added in a small

quantities to the glass network forming oxides produce drastic changes in the

properties (like melting point, electrical conductivity etc.) of the glass form-

ing oxides. Such oxides also modify the network structure of the glass and

hence they are termed as network modiﬁers. [49]

126 Aly SAeed et al.

3.5. Methods of Glass Preparation

Different techniques have been used to prepare amorphous/ glassy materials

in various forms like bulk, sheet, powder, thin ﬁlms, etc (Fig. 4). The follow-

ing techniques are used to prepare amorphous/glassy materials [50-51].

1. Melt quenching

2. Sol gel technique

3. Thermal evaporation,

4. Sputtering,

5. Glow discharge decomposition,

6. Chemical vapor deposition,

7. Electrolytic decomposition,

8. Chemical reaction,

9. Reaction amorphization,

10. Neutron Irradiation etc,

Among the above, melt quench technique is the simple and widely used for

the preparation of glassy materials because of the following reasons,

i. Glass preparation and handling are very easy

ii. Speedy preparations of innumerable compositions of the glasses are pos-

sible.

iii. Ability to produce wide variety of new oxide glasses

iv. Bulk glasses can be prepared and comparisons can be made with pulver-

ized glasses

v. Simultaneously, both amorphous and crystalline nature can be obtained

in the same melt, etc.

Hence, melt quench technique is used to prepare different compositions of

SAT. SPT and SVT glassy compounds

3.5.1. Melt– quench technique

Glass is prepared by cooling the molten liquid form of the compound sufﬁ-

ciently quickly (Fig. 4(b)). The cooling must be sufﬁciently fast to preclude

crystal nucleation and growth. The crystallization rate of an under cooled

liquid depends on the rate of crystal nucleation and on the speed with which

the crystal-liquid interface moves. These are strongly dependent on the

reduced temperature Tr =Tc-Tm and the under cooling ∆Tr = (Tm-T)/Tm.

High values of viscosity reduce the crystallization front velocity and hence,

the crystallization rate. Cooling rate is a critical factor in determining glass

formation. Generally, the quenching rate is from 105 to 109 KS-1 depends

upon the preparation method and type of materials.

Various types of materials can be prepared in amorphous/glassy form

using the melt-quench technique. Table 5 gives various techniques used for

creating different cooling rates to produce various types of glasses [53-54].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 127

3.6. Properties of Glass

Glass is a solid that belongs to the disordered state of matter. The most

common names for it are non-crystalline solids or amorphous solids or sol-

ids with liquid like structure. From macroscopic point of view, glass is hard,

FIGURE 4

Schematic presentations of different methods for the preparation of glasses: a) Slow cooling, b)

quenching, c) R F. sputtering, d) thermal evaporation [51].

TABLE 5

Various techniques and different cooling rates [53].

Technique Cooling rate (KS-1)*

Sputtering, Evaporation

Melt spinning, extraction

Chill-block, Splat-cooling

Liquid Quenching

Air Quenching

Annealing

Large telescope mirror

Optical glass

Ordinary glass

≥109

106-108

105

102-103

1-10

10-5

3x10-4

10-2-10-4

*Kelvin/second

128 Aly SAeed et al.

difﬁcult to deform, transparent because there are no irregularities to deviate

any light ray passing through it and easy to break (brittle substance). In

addition, it is isotropic, except if internal stress or a texture is present.

Under microscope, it looks like a continuous and homogeneous material; in

large scale, it might be inhomogeneous. When compared with crystal, like

crystal glass must be made of an extended three-dimensional lattice but the

diffuse character of x-ray scattering shows that, this lattice neither symmet-

ric nor periodic unlike crystal (Fig. 5). The density, the mechanical proper-

ties and the thermal properties of glasses are similar to those of the

corresponding crystal. However, unlike crystal, glasses do not have sharp,

well-deﬁned melting point and don’t cleave in preferred directions. Practi-

cally, glasses are more preferred as host matrices than crystalline materials

because of their high transparency, ease of mass production and shaping,

chemical durability, and dispersibility of many additives at high concentra-

tion [54-55].

3.7. Different Types of Glass

There are different types of glasses such as silicate glasses, alkali silicate

glasses, boric oxide glasses, alkali borate glasses, alkali borosilicate glasses

phosphate glasses and so on. Since the present study deals with borate glasses,

the structure and properties of vitreous B2O3 and borate glasses are discussed

in the following sections.

3.7.1. Borate Glass

In B2O3 glass, the oxygen coordination around each B is only three, and

hence the basic structural unit is a BO3 triangle with B is slightly above the

FIGURE 5

Ordered crystalline form (a) and, (b) a random network glassy form of the same oxide [55].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 129

plane of three oxygens. All the three oxygen are bridging between neighbour-

ing triangles. Several authors believe that an important constituent of vitreous

B2O3 is the boroxyl group (Fig. 6).

This group is a planar ring consists of six-members, alternatively boron

and oxygen atoms. These groups are linked together in a three-dimensional

network by boron-oxygen-boron bonds as shown in Figure 6. The presence of

boroxol groups in B2O3 glass was conﬁrmed from X-ray diffraction and

various spectroscopic studies. From the X-ray diffraction measurements, the

boron-oxygen and oxygen-oxygen distances in the boroxol groups were esti-

mated as 1.37Ao, 2.40 Ao respectively [57].

3.7.2. Alkali orate glass

B2O3 is a glass fobrmer and the boron ion is trivalent positive. The introduc-

tion of oxygen from a modiﬁer oxide into boric oxide glass brings about one

of the two possibilities.

a. Creates non-bridging oxygen, and the oxygen co-ordination around the

boron remains three as shown by the following reaction.

FIGURE 6

Conﬁguration of two boroxyl groups linked by a bridging oxygen •-boron; O-oxygen [57].

130 Aly SAeed et al.

b. Converts boron from 3-coordination state to 4-coordination state as

shown by the reaction.

In the RO: group, the oxygen is fully bridging and one negative charge each

from oxygen satisﬁes the three positive charges on the boron ion. After the

conversion from BO3 to Bo4, all the oxygen remain bridging, the extra nega-

tive charges on the [BO3 group is satisﬁed by an alkali M+ ion in the vicinity].

An Increase in the Bo4 tetrahedral in the glass structure increases the connec-

tivity of the network, and hence ﬂow related properties decrease and viscosity

increases. The addition of alkali oxide to glassy B2O3 causes the gradual

change in the co-ordination number of boron from three to four [58-61].

3.8. Applications of Glass in Nuclear Science

Interest in glass is due to its diverse applications, has been observed from

time immemorial. Glass as a material is comparatively relatively easy to pro-

duce as well as for synthesis in a wide range of composition. Chemical com-

position of the glass plays an important role in determining properties of

glass. Glass has many applications in laser eyewear, laser safety, nuclear

shielding for gamma ray and neutrons, water treatment, antibacterial, and

many applications [62-92].

3.8.1. Glass as a shielding material

Nowadays nuclear radiation has been used in a wide space of modern tech-

nology like medical, agriculture, drilling and exploration of oils….etc. Expo-

sure radiation doses may harm various types of cell at varying effects and due

to no lower safety radiation limit. International committee on radiation pro-

tection (ICRP) had developed the principle for utilization of radiation with

safety precaution by using “As Low As Reasonable Achievable ‘ ALARM’ “.

Radiation workers needs a high quality or suitable shielding materials for

reduction of unnecessary radiation dose to working personal Many studies

conducted on various materials that serve the purpose of radiation shielding

like normal and heavy concrete, foam and rubber...etc, and hence they are

promising materials in this regard. Glass materials are possible alternatives

with two advantages brought by their transparency to visible light and their

properties can be modiﬁed by using composition and preparation techniques.

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 131

Several glasses developed which can use as a shield for different nuclear

application because they accomplish the double task of allowing visibility

and excellent absorber for nuclear radiation like gamma ray and neutrons. A

good shielding glass should have high value of interaction cross-section and

the effect of irradiation in mechanicals, electrical, thermal, and optical prop-

erties should be small [93-94].

3.8.2. Glass as nuclear waste disposal materials

With the increased potential of the nuclear power and the associated repro-

cessing of the fuel, the disposal of radioactive wastes will become an increas-

ingly important consideration. Many nuclear wastes are currently sorted as

liquid solutions, which impose very stringent demands on containment mate-

rials. Glasses are whoever showing considerable promise as a means of stor-

ing nuclear waste. The stability of glasses considered to their possible use in

the disposal of radioactive waste. Many of the technique that has been devel-

oped to improve the reliability of structural glasses will be applicable to

glasses intended for radioactive waste disposal [95-96].

3.8.3. Glass as radiation detectors

The most common glass scintillators are cerium-activated lithium or boron

silicates. Since both lithium and boron have large neutron cross-sections,

glass detectors are particularly well suited to the detection of thermal (slow)

neutrons. Lithium is more widely used than boron since it has a greater energy

release on capturing a neutron and therefore greater light output. Glass scin-

tillators are however sensitive to electrons and γ-rays as well (pulse height

discrimination can be used for particle identiﬁcation). Being very robust, they

are also well suited to harsh environmental conditions. Their response time is

≈10 ns, their light output is however low, typically ≈30% of that of anthracene

[98].

4. INTERACTION OF RADIATION WITH MATTER AND

DETECTION TECHNIQUES

4.1. Introduction

The detection, characterization, and effects of radiation are almost entirely

dependent upon their interaction with matter. The ﬂows of charged particles

“direct ionizing radiation” such as alpha particles, beta particles and electrons,

are strongly interact with matter and cause dense ionization in the medium,

because through coulomb interaction with matter it directly causes ionization

and excitation of atoms so they easily stopped in matter. To shield from alpha

particles, a thin sheet of paper is generally sufﬁcient. For beta particles having,

those energies commonly encountered, a few millimeters of aluminum, or a

132 Aly SAeed et al.

similar material is sufﬁcient. Indirect ionizing radiation (neutrons, γ-rays) are

radiation of particles or photons, which have no charge and during interaction

with matter can transfer energy to charged particles, nuclei, and atom electrons

due to electromagnetic or nuclear interaction. The high penetration of indirect

ionizing radiation makes them very difﬁcult to stop [98].

In this chapter, we will focus our debate on indirectly ionizing radiation in

terms of the interaction mechanisms and detection methods.

4.2. Gamma Rays

Gamma radiation is electromagnetic radiation of high frequency and there-

fore high energy. Gamma rays are ionizing radiation and are thus biologically

hazardous [99].

4.2.1. Gamma Ray Sources

Natural sources of gamma rays on Earth include gamma decay from natu-

rally occurring radioisotopes such as potassium-40, and as a secondary

radiation from various atmospheric interactions with cosmic ray particles.

Some rare terrestrial natural sources, which produce gamma rays that are

not of a nuclear origin, are lightning strikes and terrestrial gamma-ray

ﬂashes, which produce high-energy emissions from natural high-energy

voltages. Gamma rays are produced by a number of astronomical processes

in which very high-energy electrons are produced, such electrons produce

secondary gamma rays by the mechanisms of bremsstrahlung, inverse Comp-

ton scattering and synchrotron radiation. Large fractions of such astronom-

ical gamma rays are screened by Earth’s atmosphere and must be detected

by spacecraft. Notable artiﬁcial sources of gamma rays include ﬁssion such

as occurs in nuclear reactors, and high energy physics experiments, such as

neutral and nuclear fusion [100].

4.2.2. Interaction of Gamma Rays with matter

Knowledge of gamma-ray interactions is important to the nondestructive

essayist in order to understand gamma-ray detection and attenuation. The

interaction of gamma rays with matter is markedly different from that of

charged particles such as α- or β-particles. The difference is apparent in the

much greater penetrating power of γ-rays and in the absorption laws. Gamma

rays, which are electromagnetic radiations, show a characteristic exponential

absorption in matter, and have no deﬁnite range [101].

Gamma rays have many modes of interaction with matter which can

divided into three categories from point of view the importance of attenuating

gamma rays

i. Primary importance

1. Photoelectric effect

2. Compton scattering

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 133

ii. Pair production

1. Secondary importance

2. Coherent (Rayleigh ) electron scattering

3. Annihilation radiation

4. Fluorescence radiation

5. Bremsstrahlung

iii. Negligible

1. Thomson scattering from the nucleus

2. Potential scattering

3. Crystal scattering

4. Nuclear interaction

a. Nuclear photoeffect

b. Nuclear scattering

5. Radiative corrections to lower order processes

The ﬁrst category of interactions is the most frequent and therefore we will

discuss them below in some details [102].

i) Photoelectric effect

In this effect, an incident photon ejects tightly bound electrons from an atom

or a molecule (Fig. 7), and the energy of the ejected electron is equal to the

difference between the incident photon energy and the energy of that electron

in its atom [102].

The energy carried away by the emitted electron can be found by subtract-

ing the binding energy from the incident photon energy, that is

EEE

=−

γ (1)

FIGURE 7

Photoelectric effect interaction [100].

134 Aly SAeed et al.

where Eb is the binding energy of the electron, Eγ is the incident photon

energy and Ee is the emitted electron energy.

In this reaction, the conservation of the energy is possible through impart-

ing some momentum to the remainder of the atom. The possibility of the

photoelectric absorption tends to increases rapidly with both the decrease in

the photon energy and the increase in the atomic number of the irradiated

material.

In the case of the heaviest elements, the photoelectric absorption is rather

insigniﬁcant for gamma ray energies greater than one MeV. Also in the pho-

toelectric effect, the photoelectrons are mainly responsible for the ionization

produced by the low energy photons [71].

ii) Compton scattering

In this interaction, the photon, with incident energy Eγ, scatters from an

electron normally regarded as being free. The result is a photon of lower

energy

′

γ, and an electron recoiling with an amount of kinetic energy T

which depends on the scattering angle. The process is illustrated schemati-

cally in Figure 8 Energy conservation gives for the kinetic energy of the

electron as [103]:

TE Eb

=−

′

γ (2)

From energy and momentum conservation one can derive the ratio of scat-

tered (

′

γ) to incident photon energy (Eγ)

′=+−

γγ

εθ

11(cos ) (3)

where, θγ is the scattering angle of the photon with respect to its original

direction. For backscattering (θγ = π) the energy transfer to the electron EKin

reaches a maximum value

Kin

max=

 (4)

The Compton Effect occurs primarily in the absorption of high γ-ray energy

and low atomic numbers. The cross section for this process, given by the

Klein-Nishina formula, can be approximated at high energies by

σε

=⋅

ln (5)

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 135

In Compton scattering only a fraction of the photon energy is transferred to

the electron. Therefore, one deﬁnes an energy scattering cross section [103].

σσ

cs c

=′ (6)

In addition, energy absorption cross-section

σσσσ

ca ccsc

Kin

=− = (7)

iii) Pair production

A gamma ray with energy of at least 1.022 MeV can create an electron-posi-

tron pair when it is under inﬂuence of the strong electromagnetic ﬁeld near a

nucleus (Fig. 9). In this interaction, the nucleus receives a very small amount

of recoil energy to conserve momentum, but the nucleus is otherwise

unchanged and the gamma ray disappears.

This interaction has a threshold of 1.022 MeV because that is the minimum

energy required to create the electron and positron. If the gamma ray energy

exceeds 1.022 MeV, the excess energy shared between the electron and posi-

tron as kinetic energy. This interaction process is relatively unimportant for

nuclear material assay because most important gamma-ray signatures are

below 1.022 MeV. The electron and positron from pair production rapidly

slowed down in the absorber. After losing its kinetic energy, the positron com-

FIGURE 8

The Compton effect [101].

136 Aly SAeed et al.

bines with an electron in an annihilation process, which releases two gamma

rays with energies of 1.022 MeV. These lower energy gamma rays may interact

further with the absorbing material or may escape. In a gamma-ray detector,

this interaction often gives three peaks for a high-energy gamma ray [103].

The probability of the three processes described above for barium is illus-

trated in Figure 10.

4.2.3. Attenuation of Gamma Rays

The attenuation of gamma ray intensity is due primarily to combinations of the

photoelectric effect, the Compton scattering effect, and pair production. In each

FIGURE 9

Pair production interaction [103].

FIGURE 10

The relative importance of the three major types of gamma-ray interaction with Barium (Xcom

program version 3.1).

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 137

case, the photon energy is absorbing by the energy transferred to kinetic energy

of an electron or an electron-positron pair. The photoelectric effect and pair pro-

duction are processes in which the gamma ray’s energy is mostly absorbed in the

material. The Compton Effect involves both scattering and absorption. Cross sec-

tion of an absorbing material of thickness x with an intensity of radiation I0 inci-

dent perpendicularly to the face of the material (Fig. 11), as the radiation is

absorbed in the material, the intensity will be I1=I(x) as it enters a thin slice of the

material of thickness Dx located a distance x, from the entrance face. Assuming

that the material absorbs some of the radiation, the intensity of the radiation as it

leaves the slice is I2 = I(x + Dx) and will be less. The change in intensity DI will

be( I1-I2) and will be a decrease (negative) that is proportional to the thickness of

the thin slice, Dx, and the intensity of the radiation, I(x), so that

∆∆

IX Ix x() ()

=−µ

(8)

The constant of the proportionality µ is called the linear attenuation coefﬁ-

cient; its value is dependent on the gamma ray photon energy. Equation (4-8)

is the standard relationship for a change in a quantity that is proportional to

that quantity and is the basis for the typical exponential relationship. Equa-

tion (4-8) can be rearranged as

∆∆

Ix=−µ (9)

and rewritten as a differential equation

∆Ix

IX dx

()

=−µ (10)

FIGURE 11

Cross section of absorbing material of thickness x with incident radiation intensity I0 [102].

138 Aly SAeed et al.

By applying the boundary conditions, I(0) = I0 when x’=0, the methods of

integral calculus yields the familiar solution

Ix Ie x

()=−

µ (11)

The functional dependence of I on the absorber thickness x is implied implic-

itly. If the natural logarithm is taken on both sides of equation (4-11), then

ln() ln()IIx

=−

0µ (12)

This equation is in the form of an equation for a straight line, y = a + bx,

where yIaIb== =−ln(),ln( ), ,

0µ and x = x. If ln(I) is plotted as a function

of x, then the results should be a straight line whose slope is negative and has

the value µ linear attenuation coefﬁcient (in cm-1).

The total linear attenuation coefﬁcient µ represents the probability of pho-

ton interaction per unit path length. The quantity µ is the sum of the probabil-

ity for each of the three processes, that is.

µ = t + s + π

where t is the linear attenuation coefﬁcient for the photoelectric effect, and s

is that for the Compton Effect, and π is that for pair production. The coefﬁ-

cient µ is a function of the energy of the gamma rays and the absorbing sub-

stance.

Another useful concept is the half-value thickness, X½, which is the value

of the absorber thickness that reduces the intensity to the half.

when xX

II==

120

IX IIe

1200

()

−

µln()

µ=ln

()

X (13)

The way for comparing the absorption of photons in different material is

to use the tenth value thickness (T.V.T.). The (T.V.L) is the thickness,

which reduces the initial intensity to one-tenth of its value; it is related to

the HVL by:

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 139

xXxx

110 1212

10 10

232== =

ln() ln()

ln() .

µ (14)

Several values of TVT for different materials and photon energies are listed

in Table 6 [104].

4.3. Neutrons

Neutron is a neutral subatomic particle that is a constituent of every atomic

nucleus, except ordinary hydrogen atom, it has no electrical charge and rest

mass equal to 1.67493x10-27 kg, neutrons and protons commonly called

together nucleons where they account for 99.9 percent of atom’s mass. The

English physicist James Chadwick discovered the neutron in 1932 and within

a few years of this discovery, many of studies were conducted on the new

particle [105].

4.3.1. Neutron sources

Since the discovery of the neutron, studies have shown that neutrons can be

produced in a variety of ways. Neutron sources and some of their properties

are discuss below in some details [106-108]:

I (α,n) reactions - These sources are often prepared bymaking an intimate

mixture of α-emitter and ﬁnely divided powder ofthe target substance (Fig.

12). 226Ra, 2l0Po, and 239Pu are common as α-emitters used in the past.

Recently, 241Am and 238Pu sources have been used. B, Be, Li, Na, and F are

used as target materials (Table 7) [106-108].

TABLE 6

Photon tenth value thickness in cm for Al, Fe, Pb and concrete [104].

Energy (MeV) Aluminum Iron Lead Concrete

0.05 21 1.6 0.25 18

0.1 50 8.1 0.37 49

0.2 70 20 2.03 77

0.5 101 35 13 111

1 139 49 29 153

1.5 170 60 39 188

2 198 69 44 218

3 241 81 48 265

4 274 88 48 306

5 300 93 47 338

140 Aly SAeed et al.

Beryllium is most commonly used since it gives the highest yield. These

sources emit neutrons, which have a spectrum of energies, up to about 10 MeV.

II Photoneutron reactions - Some nuclei will emit neutrons when exposed to

high-energy-rays (Fig. 12) The target substances are thus limited mainly to

beryllium and deuterium, which have low neutron binding energies (1.67and

2.23 MeV, respectively). These sources emit nearly monoenergetic neutrons,

generally below about 1MeV (Table 8) [106-108].

III Accelerator sources - Nuclear reactions caused by high-speed charged

particles impinging on a suitable target yield neutrons. Types of reactions

used are (α, n), (α, 2n), (p,n), (γ,n), and (d, n). Accelerator sources are useful

for producing monoenergetic neutrons over a wide energy range. Since these

machines produce very high-energy charged particles, many new target sub-

TABLE 7

(α, n) neutron sources [106].

Source Half life

Average neutron

energy (MeV)

Yield

n/s Ci

226Ra + Be 1600 yrs 5 1.7 × 107

226Ra + B 1600 yrs 3 6.8 × 106

222Rn + Be 3.8 d 5 1.5 × 107

210Po + Be 138 d 4 3 × 106

210Po + B 138 d 2.5 9 × 105

210Po + F 138 d 1.4 4 × 105

210Po + Li 138 d 0.42 9 × 104

239Pu + Be 24,000 yrs 4 106

TABLE 8

Photoneutron sources [106].

Source Have-Life Energy(MeV) Neutron Yield (per GBq)

Neutron Gamma

24Na/Be 15 h 0.967 2.754 3.4x104

24Na/D 15 h 0.263 2.754 3.3x104

28Al/Be 2.24 min 0.101 1.779 3.3x103

124Sb/Be 60.2 d 0.023 1.691 2.2x104

88Y/D 107 d .0152, 0.949 1.836, 2.734 2.3x104

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 141

stances can be used. Proton and deuteron accelerators of 10 MeV can produce

monoenergetic neutrons at all energies up to 27 MeV [108-110].

IV Reactors - The ﬁssion process in reactors produces neutrons with a spec-

trum of energy, extending up to approximately 11 MeV [110].

V Spontaneous ﬁssion sources - The neutron spectra from these sources are

similar to that produced by ﬁssion in reactors. These sources are desirable as

calibration sources since they have ﬁssion like spectrum. 252Cf is at present

the most notable of these types of sources [110].

VI Fusion neutrons - When light elements are fused to form a heavier

nucleus, neutrons are released. Two reactions of interest in fusion research

are[110]:

245

14 1

HH He

HHeHen Mev

+→ +

+→+

(. )

4.3.2. Classiﬁcation of Neutrons by Energy

Neutrons are classiﬁed according to their energy because the type of reaction

that a neutron undergoes depends very strongly on its energy. Neutrons are

classiﬁed according to the following energy scheme [111-112]:

I. Slow neutrons: These are neutrons with energies less than 1kev, and

divided into three groups as follows

FIGURE 12

Two types of neutron sources [109].

142 Aly SAeed et al.

a. Cold neutron: The neutron with an average energy less than 0.025

eV, they produced by device which depends on the coherent scattering

of slow neutron.

b. Thermal neutron: Thermal neutrons are those, which are in thermal

equilibrium with molecules or atoms of the surrounding medium. The

energy distribution of this group of neutron can be represented by

Maxwell–Boltzmann distribution. The thermal neutrons have energy

of about 0.025 eV at room temperature. One may use 0.4 eV as the

upper limit of this energy range.

c. Epithermal neutron: The neutrons, which are cover energy range

from 0.4 eV to 1 keV.

II. Intermediate energy neutron: These are the neutrons of energies

between slow and fast neutrons, and which cover energy range from

1kev to 500 keV.

III. Fast neutrons: These are neutrons having energies from 0.5 MeV to 20

MeV, and are produced by many types of nuclear reactions.

I V. Relativistic neutrons: This range includes neutrons of energy greater

than 20 MeV.

4.3.3. Neutrons Interactions

The study of neutrons interactions with matter still forms a large part of the

present experimental work in the ﬁeld of physics. There are a number of pro-

cesses, which a neutron can enter into while passing through matter. The

particular effect, which occurs, depends upon the properties of the substances

and the energy of the neutron.

The interactions of neutrons with nuclei are divided into two categories:

scattering and absorption.

I. Scattering reactions

In scattering interactions, the neutron interacts with a nucleus, but both par-

ticles reappear after the reaction. The importance of scattering becomes

greater, however, because the neutron can transfer an appreciable amount of

energy in one collision. The secondary radiations in this case are recoil nuclei,

which have picked up a detectable amount of energy from neutron collisions.

At each scattering site, the neutron loses energy and is thereby moderated or

slowed to lower energy. A scattering collision is indicated as an (n, n) reaction

or as:

+→+

The two categories of scattering reactions, elastic and inelastic scattering are

described in the following paragraph [1].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 143

I1. Elastic scattering

In an elastic scattering reaction between a neutrons and a target nucleus

(Fig. 13), there is no energy transferred into nuclear excitation. Momentum

and kinetic energy of the “system” are conserved although there is usually

some transfer of kinetic energy from the neutron to the target nucleus. The

target nucleus gains the amount of kinetic energy that the neutron loses.

Elastic scattering of neutrons by nuclei can occur in two ways. The

more unusual of the two interactions is the absorption of the neutron,

forming a compound nucleus, followed by there-emission of a neutron in

such a way that the total kinetic energy is conserved and the nucleus

returns to its ground state. This is known as resonance elastic scattering

and does the neutron possess very dependent upon the initial kinetic

energy. Due to formation of the compound nucleus, it is also referred to

as compound elastic scattering. The second, more usual method is termed

potential elastic scattering and can be understood by visualizing the neu-

trons and nuclei to be much like billiard balls with impenetr able surfaces.

Potential scattering takes place with incident neutrons that have energy of

up to about 1 MeV. In potential scattering, the neutron does not actually

touch the nucleus and a compound nucleus is not formed. Instead, the

neutron is scattered by the short-range nuclear forces when it approaches

close enough to the nucleus [101,113].

FIGURE 13

Elastic collision, momentum and energy are conserved [111].

144 Aly SAeed et al.

I2. Inelastic scattering

When a fast neutron undergoes inelastic scattering (Fig. 14), it is ﬁrst cap-

tured by the target nucleus to form an excited state of the compound nucleus;

a neutron of lower kinetic energy is then emitted, leaving the target nucleus

in an excited state. In other words, in an inelastic scattering collision, some

(or all) of the kinetic energy of the neutron is converted into excitation energy

of the target nucleus.

This excess energy is subsequently emitted as one or more photons of

gamma radiation, called inelastic-scattering gamma rays. The total energy of

these gamma rays is equal to the excess energy of the excited state of the

target nucleus. For elements of moderate and high mass number, the mini-

mum excitation energy, i.e., the energy of the lowest excited state of the target

nucleus above the ground state is usually from 0.1 to 1 MeV. Hence, only

neutrons with energy exceeding this amount can be inelastically scattered

because of nuclear excitation. With decreasing mass number of the nucleus,

there is a general tendency for the excitation energy to increase, so that the

neutrons must have higher energies if they are to undergo inelastic scattering

[101,113].

II Absorption reactions

Most absorption reactions result in the loss of a neutron coupled with the

production of a charged particle or gamma ray. When the product nucleus is

radioactive, additional radiations emitted later. Radiative capture, particle

FIGURE 14

Inelastic Scattering neutrons, the target nucleus, and the total gamma energy emitted is equal to

the initial kinetic energy of the incident neutron [111].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 145

ejection, and ﬁssion are all categorized as absorption reactions and are brieﬂy

described below [112,113].

II1. Radiative capture

In a radiative capture, the incident neutron enters the target nucleus forming

a compound nucleus.

AXn Z

AX+→

The compound nucleus then decays to its ground state by gamma emission

and described by (n, γ) reaction, this process most common occurs with slow

neutrons. The compound nucleus resulting from this process has almost exci-

tation energy of the order of 8 MeV [114].

II2. Particle ejection

In a particle, ejection reaction the incident particle enters the target nucleus

forming a compound nucleus. The newly formed compound nucleus has been

excited to a high enough energy level to cause it to eject a new particle while

the incident neutron remains in the nucleus. After the new particle is ejected,

the remaining nucleus may or may not exist in an excited state depending

upon the mass-energy balance of the reaction [115]. Table 9 lists some exam-

ples of particle eject reactions.

II3. Fission

One of the most important interactions that neutrons can cause is ﬁssion, in

which the nucleus of a heavy atom splits into two smaller nuclei known as

ﬁssion fragments and emits large amounts of energy and additional free neu-

trons. Hence, ﬁssion products include free neutrons, photons in the form

gamma rays, and other fragments such as beta particles and alpha particles.

This reaction occurs with certain nuclei of high atomic and mass numbers,

and can be described by (n, f) reaction. The ﬁssion process can become a

chain reaction producing large amount of neutrons, which become source to

TABLE 9

Particle eject reactions [114].

Reaction Reaction Name

111

+→ +

−

(n, p) Reaction

nX Y

+→ +

−

α(n, α) Reaction

2nX

+→ +

−

(n, 2n)Reaction

146 Aly SAeed et al.

a nuclear reactor. The neutron energy required to cause ﬁssion depends on the

nuclear properties of the target. Thus 235U, for example, has a large cross

section for thermal neutron, whereas 238U undergoes ﬁssion only if irradi-

ated by fast neutrons [116].

4.3.4. Neutron thermalization

Since neutrons are neutral particles, they cannot lose energy via ionization;

hence, they are slowed down by collisions with nuclei. Fission neutrons are

produced at an average energy of 2 MeV with energy extending to 11 MeV

and immediately begin to slow down as the result of numerous scattering

reactions with a variety of target nuclei. After a number of collisions, the

speed of a neutron is reduced to such an extent that it has approximately the

same average kinetic energy of the atoms or molecules of the medium in

which the neutron is undergoing elastic scattering. This energy, which is a

small fraction of an electron volt at ordinary temperature (0.025 eV at 20.5

OC), is frequently referred to as the thermal energy, since it depends upon the

temperature of the medium. Neutrons whose energies have been reduced to

values in this region (<1eV) are designated thermal neutrons. The process of

reducing the energy of a neutron to the thermal region by elastic scattering is

referred to as thermalization, slowing down, or moderation. The material

used for the purpose of thermalizing neutrons is called a moderator; a good

moderator reduces the speed of neutrons in a small number of collisions, but

does not absorb them largely [46].

4.3.5. Neutron Attenuation

Although a neutron beam has a relatively long range, there are always some

neutrons absorbed in the matter. The characteristic quantity for this absorp-

tion is called attenuation coefﬁcient and is denoted by (µ). For polyenergetic

neutrons, there is an exponential decrease in their intensity variation during

the passage of neutrons through the matter, such that

IIe

=−µ (15)

The attenuation coefﬁcient has the unit cm-1 and in case of neutron physics,

it is also called macroscopic cross-section ∑. The physical meaning of the

macroscopic cross-section is the probability of neutron interaction per one

centimeter. The quantity µ, divided by the density ρ of the target material is

called mass attenuation coefﬁcient and is equal to

ρσ= N

A (16)

where:

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 147

NA = Avogadro’s number,

M = the atomic or molar mass,

s = microscopic cross-section.

The macroscopic cross-section can be used for deﬁning a quantity called

reaction rate RR. Its meaning gives the number of reactions per unit time.

RR V

=⋅⋅Σφ

(17)

where

φ = neutron ﬂux.

V = target volume.

The microscopic cross-section can be deﬁned as the area within which the

number of nuclei – neutron reaction taking place. The unit of the microscopic

cross-section is cm2, but usually barn is used instead (1 barn=10-24 cm2).

The relationship between microscopic and macroscopic cross-section corre-

sponds to the physical meaning:

Σ= ⋅

(18)

where:

n = number of nuclei per unit volume.

ρ = the density of the material in gm/cm3.

The dependence of neutron microscopic cross-section on energy of the neu-

tron usually consists of three parts as in Figure 15.

The ﬁrst part (E~1 eV) is called “1/v region”, because the dependence

approximately is proportional to 1/v, where v is the neutron velocity. In this

region

σσ σEE

()

(19)

where

s0= microscopic cross-section at the velocity v0 = 2200 m/s.

E0 = energy of the neutron at the velocity v0 =2200 m/s.

The second region (E from ~ 1 eV to ~ 103 eV) is characterized by

extremely large changes of the microscopic cross-section and is called “reso-

nance region”. In the third region, neutrons have high energy more than ~ 103

eV. In this region, the microscopic cross-section is approaching the geometri-

cal cross-section of the target nucleus [46].

148 Aly SAeed et al.

4.4. Detection Techniques

It is impossible to smell, see, taste, or sense ionizing radiation. Humans have

no senses for α, β, γ rays, and neutrons; therefore, one has to develop detec-

tors to replace the missing capability to see, smell, taste,or sense ionizing

radiation. The necessity for the measurement of radiation exposures origi-

nates from the fact that this type of radiation has to be surveyed, controlled,

and limited. Humans also have to be protected against unexpected exposures.

4.4.1. Gamma rays Detection

Photons of gamma rays must ﬁrst produce charged particles in an interaction

process, which are then normally detected via processes of ionization, excita-

tion, and scintillation. The interactions of photons are fundamentally differ-

ent from those of charged particles since in a photon interaction process the

photon is either completely absorbed (photoelectric effect, pair production)

or relatively large angles (Compton Effect). The kinds of detectors commonly

used can be categorized as [117]:

i. Gas-Filled Detectors

a. Ionization Chambers

b. Geiger Counters

c. Proportional Counters

ii. Solid State Detectors

a. Scintillation Detectors

FIGURE 15

The dependence of the neutron microscopic cross-section on its energy [46].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 149

1. NaI(Tl) detectors

2. Plastic scintillators

b. Semiconductor Detectors

1. Ge(Li) detector

2. Si(Li) detector

iii. Personal Dosimeters

a. The Pocket Ion Chamber

b. The Film Badge

c. The Thermoluminescent Dosimeter

4.4.2. Neutrons Detection

All neutrons detection relies on observing a neutron-induced nuclear

reaction. Neutrons have mass but no electrical charge. Because of this,

they cannot directly produce ionization in a detector, and therefore cannot

be directly detected. This means that neutron detectors must rely upon a

conversion process where an incident neutron interacts with a nucleus to

produce a secondary charged particle. These charged particles are then

directly detected and from them the presence of neutrons is deduced.

Neutrons are generally detected through nuclear reactions that result in

prompt energetic charged particles such as protons, alpha particles, and so

on. Virtually every type of neutron detector involves the combination of a

target material designed to carry out this conversion together with one of

theconventional radiation detectors. Different techniques are employed

for neutron detection in different energy regions, for the reason that the

cross section for neutron interactions in most materials is a strong func-

tion of neutron energy [118].

1. Fast neutron detection

As we go higher in the neutron energy range, the probability that a neutron

will come in contact with the reaction components mentioned in previous

section of this review, becomes smaller. In this case, a material that will slow

down (or “moderate”) the fast neutrons is needed in order for our detector to

be of useful efﬁciency. Hydrogen is often used for this purpose (of slowing

down neutrons), and fast neutrons undergo elastic scattering while they are

being slowed down. The hydrogenous material used to slow down neutrons is

called a moderator and usually is several centimeters thickness and surrounds

the detector. However, one must be extremely cautious about the moderator

thickness due to the fact that neutrons may sometimes be completely stopped

inside the moderating material and not penetrate the detector’s active volume.

For neutrons in the intermediate energy range (keV) a couple centimeters

moderator thickness is required, but for higher ranges (MeV), it should be a

lot more (close to tens of centimeters) when the moderator material is poly-

ethylene or parafﬁn [119].

150 Aly SAeed et al.

2. Slow neutron detection

Traditional slow neutron detectors mostly utilize gas-based detectors or scin-

tillators and employ a converter material with large total cross section. In

these detectors, neutrons are converted to charged particles (typically a-parti-

cles) which can ionize the detector material and thus indirectly indicate that a

neutron interaction occurred [120].

a. BF3 proportional counter

b. Lithium scintillator

c. 3He proportional counter

d. Fission counters

We will mention some of the basic characteristics of the used detector in our

work.

4.5. Scintillation counters

Scintillation was the ﬁrst method used to detect ionizing radiation. When

radiation loses energy in a luminescent material, called a scintillator or

phosphor, it causes electronic transitions to excited states in the material.

The excited states decay by emitting photons, which can be observed and

related quantitatively to the action of the radiation. Scintillators employed

for radiation detection are usually surrounded by reﬂecting surfaces to trap

as much light as possible. The light is fed into a photomultiplier tube for

generation of an electrical signal. There a photosensitive cathode converts a

fraction of the photons into photoelectrons, which are accelerated through

an electric ﬁeld toward another electrode, called a dynode. In striking the

dynode, each electron ejects a number of secondary electrons, giving rise to

electron multiplication. These secondary electrons are then accelerated

through a number of additional dynode stages achieving electron multipli-

cations in the range 107–1010. The magnitude of the ﬁnal signal is propor-

tional to the scintillator lightoutput, which, under the right conditions, is

proportional to the energy loss that produced the scintillation. Good scintil-

lator materials should have a number of characteristics. They should efﬁ-

ciently convert the energy deposited by a charged particle or photon into

detectable light. The efﬁciency of a scintillator is deﬁned as the fraction of

the energy deposited that is converted into visible light. The highest efﬁ-

ciency, about 13%, is obtained with sodium iodide. A good scintillator

should also have a linear energy response; that is, the constant of propor-

tionality between the light yield and the energy deposited should be inde-

pendent of the particle or photon energy. The luminescence should be rapid,

so that pulses are generated quickly and high-count rates can be resolved.

The scintillator should also be transparent to its own emitted light. Finally,

it should have good optical quality for coupling to a light pipe or photomul-

tiplier tube. The choice of a particular scintillation detector represents

abalancing of these factors for a given applications [121-122].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 151

There are different types of scintillators that are classiﬁed into inorganic

and organic-based scintillators.

4.5.1. Inorganic scintillators

Inorganic scintillator crystals are made with small amounts of activator

impurities to increase the ﬂuorescence efﬁciency and to produce photons

in the visible region. As shown in ﬁgure (4.10), the crystal is character-

ized by valence and conduction bands. The activator provides electron

energy levels in the forbidden gap of the pure crystal. When a charged

particle interacts with the crystal, it promotes electrons from the valence

band into the conduction band, leaving behind positively charged holes. A

hole can drift to an activator site and ionize it. An electron can then drop

into the ionized site and form an excited neutral impurity complex, which

then decays with the emission of a visible photon. Because the photon

energies are less than the width of the forbidden gap, the crystal does not

absorb them.

The alkali halides are good scintillators. In addition to its efﬁcient light

yield, sodium iodide doped with thallium [NaI (Tl)] is almost linear in its

energy response. It can be machined into a variety of sizes and shapes.

Disadvantages are that it is hygroscopic and somewhat fragile. NaI has

become a standard scintillator material for gamma-ray spectroscopy.

CsI(Na), CsI(Tl), and LiI(Eu) are examples of other inorganic scintilla-

tors. Silver-activated zinc sulﬁde is also commonly used. It is available

only as a polycrystalline powder, from which thin ﬁlms and screens can

be made. The use of ZnS, therefore, is limited primarily to the detection

of heavy charged particles. (Rutherford used ZnS detectors in his alpha

particle scattering experiments.) Glass scintillators are also widely used

[121-122].

FIGURE 16

Energy-level diagram for activated crystal scintillator [121].

152 Aly SAeed et al.

i. Sodium iodide – NaI(Tl)

This is the most commonly used scintillator material (Fig. 17), the excellent

light yield is the most notable property of NaI(Tl), and its response is nearly

linear over the signiﬁcant energy range. It has come to be accepted as the

standard scintillation material for gamma-ray spectroscopy.

Photons striking a sodium iodide (NaI) crystal, which contains 0.5 mole

percent of thallium iodide (Tl) as an activator, cause the emission of a short

ﬂash of light in the wavelength range of 3300-5000 A (in the ultraviolet

region). A photomultiply tube, which gives a pulse corresponding to the light

intensity, detects the light ﬂashes; a multi-channel counter measures these

pulses. The output pulses from a scintillation counter are proportional to the

energy of the radiation. Electronic devices have been built not only to detect

the pulses, but also to measure the pulse heights. The measurements enable

us to plot the intensity (number of pulses) versus energy (pulse height), yield-

ing a spectrum of the source [123].

4.5.3. Organic scintillators

Fluorescence in organic materials results from transitions in individual

molecules. Incident radiation causes electronic excitations of molecules

into discrete states, from which they decay by photon emission. Since the

process is molecular, the same ﬂuorescence can occur with the organic

scintillator in the solid, liquid, or vapor state. Fluorescence in an inorganic

scintillator, on the other hand, depends on the existence of a regular crystal-

line lattice. Organic scintillators are available in a variety of forms. Anthra-

cene and stilbene are the most common organic crystalline scintillators,

anthracene having the highest efﬁciency of any organic material. Organic

scintillators can be polymerized into plastics. Liquid scintillators (e.g.,

xylene, toluene) are often used and are practical when large volumes are

required. Radioactive samples can be dissolved or suspended in them for

FIGURE 17

NaI (TI) crystal detection technique [123].

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 153

high-efﬁciency counting. Liquid scintillators are especially suited for mea-

suring soft beta rays, such as those from 14C or 3H. High-Z elements (e.g.,

leador tin) are sometimes added to organic scintillator materials to achieve

greater photoelectric conversion, but usually at the cost of decreased efﬁ-

ciency. Compared with inorganic scintillators, organic materials have much

faster response, but generally yield less light. Because of their low-Z con-

stituents, there are little or no photoelectric peaks in gamma-ray pulse-

height spectra without the addition of high-Z elements. Organic scintillators

are generally most useful for measuring alpha and beta rays and for detect-

ing fast neutrons through the recoil protons produced [124-126].

i. Stilbene crystal

The stilbene crystal (C14 H12 ) is the most commonly used organic scintilla-

tor, it has a density, ρ=1.16gm.cm-3, and contains 4.658 x 1022 hydrogen

atoms and 5.435 x 1022 carbon atoms.cm-3. The chemical composition of a

stilbene crystal shows that it consists of sufﬁciently light nuclei. Therefore, in

gamma quanta detection, the photo effect takes place only in the region of

low energies (Eγ<0.1 MeV). The effect of pair production becomes appre-

ciable in the region of gamma quanta of energies Eγ =1.5 MeV. However, the

contribution of this effect is not signiﬁcant, because the pair production

cross-section at Eγ = 10 MeV is about 20 % to of the total cross-section.

Therefore, in a wide range of energies, the Compton effect is the predominant

process for gamma quantum interaction with the stilbene scintillator. The

electrons, generated as a result of the Compton scattering possess a continu-

ous energy spectrum in the energy range from zero to a certain maximum

value, Emax. For gamma quanta of energy Eγ the value of Emaxcan be calcu-

lated as below:

max

=−

−

10 511 2

(20)

where Emax and Eγ are measured in MeV.

The shape of recoil electron energy distributions strongly depends on the

recorded gamma-quantum energy Eγ and on the scintillation dimensions

because of the leakage of high-energy electrons beyond the scintillator. The

fast neutrons on passing through the scintillator may experience elastic and

inelastic scattering by the nuclei of the elements constituting the scintillator.

However, inelastic scattering occurs on heavy nuclei and the energy imparted

by the neutrons to these nuclei is not high. Moreover, heavy nuclei do not

generate light pulses in the scintillator. For this reason, a contribution of neu-

tron inelastic scattering is usually neglected for the registration by a scintilla-

tion counter. The fast neutrons can transfer to the nucleus the highest possible

154 Aly SAeed et al.

energy on elastic scattering by protons. In this case, at a head-on collision, the

neutron transfers its total energy to the proton (EP= En). However, in fact,

this shape is idealized, since there are a number of distortion factors, which

will distort the pulse shape. For stilbene crystals, three time-constituents of

decay have been discovered,fast component with effective decay time t0=

6x10-9sec., independent on the nature of the particle emanating the scintilla-

tion. The other two components have a decay time, which varies from about

0.35 to 10µsec. In scintillation excitations by protons, these two components

are approximately, twice as intensive as for scintillation produced by elec-

trons. The main advantages of the stilbene crystal are:

a. Absence of long-lived neutron activation products in the scintillator;

b. Improved response time;

c. Relatively high resolution compared with the NaI (Tl) crystal;

d. Relatively simple method for unfolding the measured pulse amplitude

spectrum compared to the complex gamma-ray spectrum by a singleNaI

(Tl) scintillation spectrometer, which is a quite difﬁcult task. While the

disadvantage of the stilbene scintillator is that, it is affected by a sharp

increase or decrease of temperature and mechanical impacts. [127-128].

4.5.4. Helium-3 proportional counters

Neutron detection by 3He (Fig. 18) is base on the reaction

The cross section for this reaction is quite high for thermal neutrons (4500

barn at 0.025eV). Proportional counters ﬁlled with 3He are widely used,

especially in time-of-ﬂight measurements. The efﬁciency of the counter can

be increased by increasing the pressure. The 3He spectrometer is still a useful

detector for measuring the neutron spectrum in the energy range 0.1 to 1

MeV [128].

3276He nHHkeV+→ ++

FIGURE 18

3He slow neutron detector.

GlASS foR GAMMA ANd NeutRoN RAdiAtioN ShieldiNG 155

5. CONCLUSIONS

In the present work, show a review of the Fast and slow neutrons, total gamma

rays, and pure gamma-ray shielding properties. The Barium is a good attenu-

ator for fast neutrons, total gamma rays and pure gamma rays. Boron is an

efﬁcient attenuator for slow neutrons. The investigated glassy barriers were

found to be a good absorber for gamma rays under transparent conditions. On

the other hand some disadvantages were observed with neutrons such as the

required thickness for attenuate both fast and slow neutrons.

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(2011).

A multi-function glass shield for neutrons and gamma rays of boron- and bismuth-reinforced silicate glass

Article

Full-text available

Oct 2024

A successful attempt to produce a multi-function glass shield for attenuating neutrons and gamma rays by reinforcing a silicate glass network with boron and bismuth has been accomplished. A composition of 20SiO2-80Na2O (BSiBi0) was proposed to be used as a host glass network and prepared using the melt/annealing techniques. The low concentration of SiO2 in BSiBi0 was not sufficient to form a stable glass network. Then, the proposed BSiBi0 was modified with 10, 20, 30, and 40 mol% of each of B2O3 and Bi2O3 (BSiBi1, BSiBi2, BSiBi3, and BSiBi4) simultaneously. The structural effects of adding B³⁺ and Bi³⁺ were studied through X-ray diffraction, density, and FTIR, which all showed enhancement of glass forming ability, a former role of Bi³⁺ ions, and crowded the glass network by BO4 units. The derived structural parameters -

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molar volume, mean silicon – silicon separation, mean boron – boron separation, oxygen packing density, packing density, and number of bridging/non-bridging oxygen -

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were extensively discussed to explore the impact of B³⁺ and Bi³⁺ on the formed network. The richness of the proposed host glass network by B³⁺ and Bi³⁺ enhanced its thermal stability. The obtained elastic properties by ultrasonic measurements reflect the increase of the glass rigidity with increasing concentrations of B³⁺ and Bi³⁺ ions. The obtained glasses have high visible light transparency and almost complete UV absorption. The measured shielding parameters against two types of neutron energies (total slow and slow) and a wide range of gamma rays’ energies showed a significant improvement in the shielding efficiency of the considered glasses. The total slow neutrons, slow neutrons, and gamma rays’ attenuation abilities were improved by 22.9, 135.5, and 73.8 -

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199.5%. High thermal stability, elasticity, visible light transparency, and neutrons and gamma rays’ attenuation performance features give the produced glasses, especially BSiBi4 glass, preference as shielding materials in nuclear fields.

Experimental and theoretical study on the physico-mechanical characteristics and radiation shielding capability of hardened alumina sludge waste-cement pastes containing MnFe2O4-nanoparticles

Article

Full-text available

Oct 2024

This research investigates the impact of incorporating low-cost MnFe2O4 spinel nanoparticles (MF-NPs) at varying concentrations (0.5, 1, and 2 mass%) on the mechanical and physical properties of blended cement pastes. These pastes were produced by replacing different proportions (5, 10, and 15 mass%) of ordinary Portland cement (OPC) with activated alumina sludge waste (AAS), to promote sustainability. Also, the research examined the gamma radiation shielding effectiveness of certain hardened composites against a ¹³⁷Cs gamma radiation source with an energy of 661.64 keV using a NaI (Tl) detector (Oxford Model) with 3″ × 3″, amplifier and 16 k multi-channel analyzer. The linear attenuation coefficient (LAC) of all the studded samples were calculated theoretically using a Monte Carlo code MCNP-5 code. The gamma radiation shielding properties were analyzed in depth using a Monte Carlo code MCNP-5 simulation model. The theoretical and experimental results for LAC were found to be in complete agreement. Phy-X/PSD software was applied to estimate the mass attenuation coefficient (MAC) for gamma radiation at various energies, as well as the effective atomic number (Zeff), mean free path (MFP), half-value layer (HVL), and tenth-value layer (TVL). The findings demonstrated that the addition of 0.5% MnFe2O4 nanoparticles (MF-NPs) to blended cement pastes exhibited the best physical and mechanical characteristics, as well as the most effective gamma radiation shielding.

Recycling lead-acid batteries in polymeric materials to enhance their efficiency as gamma ray shielding materials

Article

Apr 2025
J ALLOY COMPD

Efficient red-NIR laser of Er/Yb, Er/Nd, and Er/Ce co-doped in oxyfluorophosphate glass

Article

Full-text available

Mar 2025

Being one of the most important tools in industrial and medical fields, the red-NIR laser has caught a lot of attention in research studies to improve its emission efficiency and the stability of its lasing materials, both thermally and mechanically. In this regard, the emission intensity of the red-NIR laser generated from Er³⁺ ions doped glass network was enhanced by adding Yb³⁺, Nd³⁺, and Ce³⁺ ions. A host glass network of 44P2O5–15ZnO–10Pb3O4–15NaF–15MgF2–1Er2O3 was proposed as a red-NIR lasing material and was reinforced by 0.5 and 1 mol% of Yb³⁺, Nd³⁺, or Ce³⁺ ions (labeled as PZPbNMEr³⁺:RE³⁺). The structural variations resulting from compositional changes, which were examined by XRD, density, FTIR, and Raman measurements, showed an increase in the glass network tightness, leading to better thermal stability and elasticity of the considered PZPbNMEr³⁺:RE³⁺ glasses, indicating their suitability as lasing materials. Optically, all the distinctive optical absorption bands of the Er³⁺, Yb³⁺, Nd³⁺, and Ce³⁺ ions appeared in the spectral region of 200–2500 nm. Additionally, a significant impact of the induced strain or crystal field of the added RE³⁺ ions on the optical properties was observed. A successful excitation was conducted by the green wavelength of 525 nm to the host glass PZPbNMEr³⁺, generating five wavelengths, one in the red region at 631 nm and four in the NIR at 748, 801, 1034, and 1527 nm. A significant enhancement in the intensity and redshift of the formed bands were observed with the inclusion of Yb³⁺, Nd³⁺, and Ce³⁺ ions, and new NIR bands were generated. Additionally, an energy transfer between Er³⁺/Yb³⁺ and Er³⁺/Nd³⁺ was observed. Accordingly, the considered PZPbNMEr³⁺:RE³⁺ glasses had highly efficient photoluminescence in the red-NIR region alongside their high thermal stability and good elasticity, making them a suitable choice for laser media.

Production, Analysis, and Assessment of Gamma-Ray Shielding Performance of Gd2O3-Doped ZnO-B2O3 Glasses using MCNP6 Simulations

Article

Mar 2025
RADIAT PHYS CHEM

Adjusting the cool white light emission in borate glass by controlling Ce3+/Cr3+ energy transfer for white light-emitting diodes and outdoor lighting

Article

Mar 2025
PHYSICA B

Reinforcing silicate glass with waste lead and cadmium batteries to be used as shielding materials for gamma rays and thermal neutrons

Article

Apr 2025

Silicate Glass Reinforced by Bi and B as Efficient Protective Materials Against Gamma Rays and Neutrons

Article

Jan 2025

The growing role of nuclear radiation in various domains necessitated a continuous development in enhancing the efficiency of radiation shielding materials. Generally, a shield that has the ability to attenuate thermal and fast neutrons and gamma rays simultaneously is the most effective in various applications. Therefore, in this article, a host glass of sodium silicate (labeled as BSiBi0) was reinforced by 10, 20, 30, and 40 mol% of each of Bi and B ions simultaneously (BSiBi1, BSiBi2, BSiBi3, and BSiBi4, respectively) to produce an integrated shield capable of attenuating thermal and fast neutrons and gamma rays. To check the attenuation efficiency of the proposed glasses, fast neutrons removal cross-sections S_F, thermal neutrons total macroscopic cross-sections S_Th, total mass attenuation coefficients µ_m, linear attenuation coefficients µ, mean free paths MFP, half-value layers HVL, effective atomic numbers (Z_eff), effective electron densities (N_eff), and buildup factors for energy absorption EABF and exposure EBF were studied extensively. Compared to the host glass (BSiBi0), the attenuation of gamma rays in the BSiBi4 glass (40 mol% of each of Bi2O3 and B2O3) improved by 3.9%-92.5% in the studied energy range. While, its ability to attenuate the fast neutrons improved by 86.0%, and thermal neutrons by 1402.4%. Hence, the BSiBi4 glass was nominated as an efficient shield in relevant applications due to its high attenuation ability for both neutrons and gamma rays.

An insight into advanced glass systems for radiation shielding applications: A review on different modifiers and heavy metal oxides-based glasses

Article

Nov 2024

Ionizing radiation from natural and many synthetic sources is a remarkable tool in many scientific, production, quality control, food preservation, medical, security, and other technological processes. The need to protect humans (public and personnel), gadgets, the environment, and animals from the harmful effects of radiation, while maintaining and expanding the scope of application has made radiation protection an important topic to discuss. Among the methods and materials available for radiation control, shielding and the use of glass shields are the most effective and attractive methods and materials, respectively. In this report, the basic parameters for measuring shielding competences, basic shielding materials and their shortcomings, and glass shields are discussed. Five categories of glasses, namely, borate, germanate, silicate, phosphate, and tellurites, with important shielding attributes, are reviewed. The role of chemical composition, density, and mean atomic number as gamma shielding delineating factors was emphasized. The weaknesses and comparable advantages of each glass system were presented as well. The review concludes by looking at the trend and future of glass shields and research in radiation technology. The data and analysis presented in this review provides scientists and radiation protection technologist on the impact of certain chemical oxides on shielding efficacies of different glass systems.

Experimental study of gamma-ray attenuation capability of B2O3-ZnO-Na2O-Fe2O3 glass system

Article

Full-text available

Aug 2024

In the present work, a glass system with developed composition consisting of B2O3, ZnO, Na2O and Fe2O3 samples has been investigated. Glass samples were prepared using the melt quenching method and the density of the system was measured using Archimedes’ principle. Spectroscopic analysis using a gamma source and a high-purity germanium detector at four energies of 0.0595, 0.6617, 1.173, and 1.333 MeV emitted from Am-241, Cs-137, and Co-60 were used to determine the attenuation parameters of present glass composites. The sample containing 45 B2O3 + 10 Na2O + 40 ZnO + 5 Fe2O3 (coded BNZF-4) had the highest mass attenuation coefficient (MAC) value at all the energies discussed compared to the other composites. Whoever, the BNZF-1 sample had the lowest value at all ranges of energies. The transmission factors (TF, %) of the manufactured samples were calculated, at 0.0595 MeV (TF, %) values are 32.6429 and 6.4612 for samples BNZF-1 and BNZF-4, respectively. The statistical results demonstrated significantly better to increase the ZnO concentration in the sample, where the percentage of zinc oxide inside the prepared glass samples has the following direction BNZF -4 > BNZF -3 > BNZF -2 > BNZF -1. The significance of this study is that transparent, environmentally harmless glass composites with relatively high density have been prepared that can be used as shielding materials against gamma rays, especially at low energies.

Nuclear and Particle Physics

Book

May 2006

B. R. Martin

Introduction to Glass Science and Technology

Book

Oct 2020

James E Shelby

Presenting the fundamental topics in glass science and technology, this concise introduction includes glass formation, crystallization, and phase separation. Glass structure models, with emphasis on the oxygen balance method, are presented in detail. Several chapters discuss the viscosity, density, thermal expansion, and mechanical properties of glasses as well as their optical and magnetic behavior and the diffusion of ions, atoms, and molecules and their effect on electrical conductivity, chemical durability, and other related behavior. In addition to the effects of atomic structure on the properties of glasses, the effects of phase separation, crystallization, and water content, which are neglected in most texts, are discussed extensively. Glass technology is addressed in chapters dealing with the raw materials for producing glasses, batch calculations, and the melting and fining processes. The compositions, properties, and production of commercial glasses are also presented. A chapter is devoted to the use of thermal analysis in the study of glasses, including their crystallization behavior. This expanded, third edition, includes new chapters on doped vitreous silica and the, often overlooked, role of halides on glass formation and properties. In addition, solutions to all of the exercises at the ends of chapters are included for the first time in this edition. This introductory text is ideal for undergraduates in materials science, ceramics, or inorganic chemistry. It will also be useful to the graduate student, engineer, or scientist seeking basic knowledge of the formation, properties, and production of glass in support of their work.

Measurement & Detection of Radiation

Book

Jul 2021

Permissible Dose: A History of Radiation Protection in the Twentieth Century

Book

Jan 2000

How much radiation is too much? This book examines the evolution, over more than a hundred years, of radiation protection standards and efforts to ensure radiation safety for nuclear workers and for the general public. The risks of radiation—caused by fallout from nuclear bomb testing, exposure from medical or manufacturing procedures, effluents from nuclear power, or radioactivity from other sources—have aroused more sustained controversy and public fear than any other comparable industrial or environmental hazard. The book clarifies the entire radiation debate, showing that permissible dose levels are a key to the principles and practices that have prevailed in the field of radiation protection since the 1930s, and to their highly charged political and scientific history as well.

Fundamentals of Nuclear Science and Engineering

Book

Nov 2016

Radiation Protection in the Health Sciences

Book

Dec 1995

Atoms, Radiation, and Radiation Protection

Book

May 2007

James E. Turner

The influence of nickel ions on the electric behaviour of P 2 O 5 –ZnO–Na 2 O glassy system

Article

Jul 2019
J Microw Power Electromagn Energ

P2O5–ZnO–Na2O glasses doped with various concentrations of NiO in the range from 0 to 6 mol % are prepared. The measurements of the electric properties of the samples are studied at different temperatures and frequencies. It is found that the dielectric constant decreases with increasing frequency as well as the electric modulus shows relaxation behaviour. The effect of different NiO concentrations and their chemical states in the glass network appears in the analysis of the activation energy, where Ni²⁺ ions have octahedral and tetrahedral states. From the frequency exponent result, the present samples exhibit correlated barrier hopping (CBH) conduction mechanism. The results show that some samples have great dielectric properties.

Molecular spectroscopic study of borate glass doped neodymium oxide

Article

Jan 2017

This work presents the molecular spectroscopic analysis of glass systems with a chemical composition wt% (x-42) B2O3. (100-x) Na2O.40ZnO.2Nd2O3 (where x=100, 95, 90, 85,80and75) and which were prepared using the melt quenching method. The X-ray diffraction analysis confirmed the amorphous nature of the prepared glass. The effects of Na2O content on the glass system influences the estimated values with changing of the boron network of BO3 and the BO4. The most active IR peaks were observed in the region from 400 cm⁻¹ to 1600 cm⁻¹. The FT-IR spectra were analyzed to determine and differentiate the various vibrational modes by applying a deconvolution method to them.

The Properties of Optical Glass

Book

Jan 1998

This book, entitled The Properties of Optical Glass, is one of a series reporting on research and development activities on products and processes conducted by the Schott Group. The scientifically founded development of new products and technical processes has traditionally been of vital importance at Schott and has always been per formed on a scale determined by the prospects for application of our special glasses. The scale has increased enormously since the reconstruction of the Schott Glaswerke in Mainz. The range of expert knowledge required for that could never have been supplied by Schott alone. It is also a tradition in our company to cultivate collaboration with customers, universities, and research institutes. Publications in numerous technical journals, which since 1969 we have edited to a regular timeplan as Forschungsberichte - 'research reports' - formed the basis of this cooperation. They contain up-to-date information on the most var ious topics for the expert but are not suited as survey material for those whose standpoint is more remote. This is the point where we would like to place our series, to stimulate the exchange of thoughts, so that we can consider from different points of view the possibilities offered by those incredibly versatile materials, glass and glass ceramics. We would like to show scientists and engineers, interested customers, and friends and employees of our firm the knowledge that has been won through our research and development at Schott in cooperation with the users of our materials.

Glass Materials in Nuclear Technology for Gamma Ray and Neutron Radiation Shielding: a Review

Abstract

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