i 
 
ASSESMENT OF THE SHIELDING INTEGRITY OF Co – 60                   
GAMMA- RAY SCANNER AT AFLAO BORDER, GHANA. 
  
 
 
This thesis is submitted to the: 
 
 
 
DEPARTMENT OF NUCLEAR SAFETY AND SECURITY 
SCHOOL OF ALLIED AND NUCLEAR SCIENCES 
UNIVERSITY OF GHANA, LEGON 
 
 
By 
 
 
EDWIN CAPACITY AGBEMAFO 
(10506731) 
Bsc. (U.C.C, Cape Coast), 2008 
 
 
In partial fulfilment of the requirements for the Award of the degree of  
 
 
MASTEROF PHILOSOPHY 
 
IN 
 
 RADIATION PROTECTION. 
 
 
 
JULY, 2016 
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DECLARATION 
This thesis is the result of research work undertaken by Edwin Capacity Agbemafo of 
the Department of Nuclear Safety and Security, University of Ghana, Legon, under the 
supervision of Dr. J. Owusu Banahene and Professor J.J Fletcher. 
 
 
 
 
Sign ..........................................   ………………………… 
(Edwin Capacity Agbemafo)     Date 
Student  
 
 
 
 
Sign ..........................................   ………………………….. 
(Dr. J. Owusu-Banahene)     Date 
Principal Supervisor 
 
 
 
 
Sign ..........................................   ………………………….. 
Date 
Co-Supervisor 
 
 
 
 
 
 
(Professor J.J Fletcher) 
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DEDICATION 
This work is dedicated to my lovely wife Precious Mawunya Ahortor and my daughter 
Eyram Debrascha Ardeen-Capacity. 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
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ACKNOWLEGMENT 
I would like to express my profound gratitude to the Almighty God who gave me the 
ability to complete this work. 
I am very grateful to my supervisory team; Dr. J. Owusu-Banahene and Prof. J.J 
Feltcher for their invaluable support, advice and encouragement especially during my 
research and thesis writing. 
Special thanks go to my Head of Department, Dr. Augutine Faanu, also to Lecturers 
and Course mates at the Department of Nuclear Safety and Security. 
My appreciation also go to Mr. Lawluvi Henry and Mr. Akrobortu Emmanuel all of 
Nuclear Regulatory Authority of Ghana, for their motivations and encouragements 
during my stay on campus.  
My heartfelt appreciation goes to my siblings Messrs Kwaku Kpodo, Emmanuel 
Kpodo, Yawo Amable Kpodo, Christian Agbemafo, Felicia Agbemafo and Justine 
Agbemafo - who, as always, have supported and encouraged my education. 
 
Finally, I would like to thank the management and staff, especially Samuel Mensah and 
Mr. Carl Lokko of NICK TC Scan Ltd, Aflao Border, for granting me the permission to 
carry out this research work on their facility. 
 
 
 
 
 
 
 
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LIST OF ABBREVIATIONS 
 
ALARA-   As Low As Reasonably Achievable 
BSS-   Basic Safety Standards 
CCTV-  Closed Circuit Television 
Gy-   Gray 
HVL-   Half Value Layer 
IAEA-   International Atomic Energy Agency 
ICRP-   International Commission on Radiological Protection,  
IDR-   Instantaneous Dose Rate 
IPEM-  Institute of Physics and Engineering in Medicine 
MeV-   Mega-electronvolt 
mSv   milli-sievert 
NCRP-  National Council on Radiation Protection and Measurements 
POI-   Point Of Interest 
SAD-   Source Axis Distance 
SRS-   Safety Report Series 
TADR-  Time Averaged Dose-equivalent Rate 
TBq-   Tera-Becquerel 
TVL-   Tenth Value Layer 
USA-   United State of America 
 
 
 
 
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TABLE OF CONTENTS 
DECLARATION ............................................................................................................. ii 
DEDICATION ................................................................................................................ iii 
ACKNOWLEGMENT ................................................................................................... iv 
TABLE OF CONTENTS ................................................................................................ vi 
LIST OF TABLES .......................................................................................................... ix 
LIST OF FIGURES ......................................................................................................... x 
ABSTRACT .................................................................................................................... xi 
CHAPTER ONE .............................................................................................................. 1 
1.0. Introduction ................................................................................................... 1 
1.1.       Background .................................................................................................... 1 
1.2 Statement of the Problem .............................................................................. 4 
1.3.      Objectives ....................................................................................................... 6 
1.4.       Relevance of the Study ................................................................................... 6 
1.5.       Scope .............................................................................................................. 7 
1.6.       Structure of the Thesis ................................................................................... 7 
CHAPTER TWO ............................................................................................................. 8 
LITERATURE REVIEW ................................................................................................ 8 
2.0. Introduction ................................................................................................... 8 
2.1. A brief history of radiation shielding ............................................................ 8 
2.2 Destination Inspection. .................................................................................. 9 
2.3. Scanning Technology .................................................................................. 10 
2.4.        Radioactive Decay Scheme of Co-60 .......................................................... 12 
2.5. Shielding materials ...................................................................................... 13 
2.5.1. Concrete .................................................................................................... 13 
2.5.2.  Lead .................................................................................................. 14 
2.5.3. Steel .................................................................................................. 14 
2.5.4. Polyethylene and Paraffin ................................................................. 14 
2.5.5. Earth.................................................................................................. 15 
2.6.1. Good geometry shielding situation ................................................... 17 
2.6.2. Point isotropic source ....................................................................... 17 
2.7. Shielding for X-ray ...................................................................................... 20 
2.8.       The classification of work areas .................................................................. 25 
2.9.       Occupancy Factors ...................................................................................... 26 
2.11.     Assessment of Occupational Exposure........................................................ 27 
2.12.0   Calculation method ...................................................................................... 28 
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2.12.1 Primary barrier shielding. ................................................................. 28 
2.12.2 Secondary barrier shielding. ............................................................. 30 
2.13.0    Average instantaneous dose rate calculations. .................................. 32 
2.13.1 Instantaneous dose rate due to the primary radiation (  ) .............. 32 
2.13.2 Dose rate from the scattered radiation. ............................................. 33 
2.13.3 Dose from the leakage radiation ....................................................... 33 
2.14.0 Time Averaged Dose-Equivalent Rates ........................................... 34 
2.15.0  Calculation of required barrier thickness for X – ray of energy 6 MeV.
 ………………………………………………………………………..35 
    2.15.1 Primary barrier shielding. ................................................................. 35 
2.16.0    Secondary barrier shielding ............................................................... 36 
     2.16.1 Scattered radiation ............................................................................ 36 
MATERIALS AND METHODS ................................................................................... 37 
3.0 Introduction ........................................................................................................ 37 
3.1   Materials ........................................................................................................... 37 
3.2 Technical specification of the facility. ............................................................... 38 
3.3 Area of Survey ............................................................................................. 38 
3.4 Method ................................................................................................................ 39 
3.5.0 Instruments used ...................................................................................... 40 
3.5.1 Dose rate meter..................................................................................... 40 
3.6. Shielding barrier calculations ...................................................................... 41 
3.6.1. Calculation of the workload for the gamma ray source. ................... 41 
3.6.2. Primary Barrier Shielding for the Gamma Source ........................... 42 
3.6.3.0 Secondary barrier shielding for the 1.25 MeV Gamma-ray scanner. 42 
3.7. Average instantaneous Dose Rate Calculation for the primary Gamma ray 
radiation .................................................................................................................... 43 
3.8.       Instantaneous dose rate from leakage gamma ray radiation. ....................... 44 
3.9. Instantaneous dose rate from scattered gamma ray radiation. ..................... 44 
3.10.0    Barrier thickness consideration for X-ray radiation of energy 6MeV 45 
3.10.1 Barrier thickness for primary X-ray beam of energy 6MeV ............ 45 
3.10.2 Barrier thickness for leakage radiation for 6MeV X-ray radiation. . 46 
CHAPTER FOUR .......................................................................................................... 47 
RESULTS AND DISCUSSION .................................................................................... 47 
4.0       Introduction .................................................................................................. 47 
4.1       Primary barrier thickness for the gamma ray scanner. ................................. 47 
4.2.0    Secondary barrier thickness for the Co-60 scanner ...................................... 48 
4.2. 1.  Leakage Barrier Thickness for the Co-60 Gamma-ray Scanner ................... 49 
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4.3     Scattered Barrier Thickness for the Co-60 Gamma-ray Scanner .................... 53 
4.4.   Dose rates for the secondary barrier thickness of the gamma ray scanner. ..... 55 
4.5.   Comparison between measured and calculated dose rate for the gamma-ray 
scanner. ..................................................................................................................... 57 
4.6.   Barrier thickness calculation for X-ray of energy 6MeV ................................ 58 
4.7.  Scatter and leakage radiation barrier thickness barrier for 6MeV X-ray source.
 .................................................................................................................................. 59 
4.9.  Comparison of commissioning shielding requirement with the shielding 
requirement at the present state. ............................................................................... 62 
CHAPTER FIVE ........................................................................................................... 66 
5.1 Conclusions. ................................................................................................ 66 
5.2 Recommendation ......................................................................................... 66 
REFERENCES. ............................................................................................................. 68 
APPENDICES ............................................................................................................... 74 
Appendix A............................................................................................................... 74 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
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LIST OF TABLES 
Table 2. 1: Shielding Materials and Their Densities. Source: IAEA Safety series ....... 15 
 
Table 4. 1:  Calculated and measured primary barrier thickness for the gamma ray .... 47 
Table 4. 2: Leakage barrier thickness compared with designed specification ............... 49 
Table 4. 3: Scatter Barrier thickness compared with designed specifications. .............. 53 
Table 4. 4: Calculated dose rates for the secondary barrier thickness for the gamma ray 
scanner. .......................................................................................................................... 55 
Table 4. 5: Comparison between measured and calculated dose rates for the gamma-ray 
scanner. .......................................................................................................................... 57 
Table 4. 6:  Calculated primary barrier thickness for X-ray of energy 6 MeV. ............. 58 
Table 4. 7: Calculated scattered radiation barrier thickness barrier for 6MeV X-ray 
source. ............................................................................................................................ 59 
 
Table A. 1: Tenth Value Layers (TVL) for Co-60 and some selected Energies for 
concrete of density 2350kgm-3 ....................................................................................... 74 
 
Table B. 1: First TVL (mm concrete) for Container scatter radiation at some selected 
scatter angles. ................................................................................................................. 75 
 
Table C. 1: Summary of recommended/legal effective dose limits and designed dose 
limits .............................................................................................................................. 76 
 
Table D. 1: Scatter fraction of dose α, at 1 m, for 400 cm2 incidence beam. ................ 77 
Table E. 1: Occupancy  factor for some selected areas ................................................. 78 
 
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LIST OF FIGURES 
Figure 2. 1: Typical Non - Intrusive System Image (GAMMA RAY) .......................... 11 
Figure 2. 2: Radioactive decay scheme for Co-60 ......................................................... 12 
Figure 2.3: Point isotropic source shielding configuration ............................................ 18 
Figure 3. 1: Layout of the NICK TC Gamma ray scan facility. .................................... 39 
Figure 3. 2: Rad eye G-10 gamma survey meter ........................................................... 40 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
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ABSTRACT 
This study examines the current state of the shielding integrity of the  38.7 TBq Co-60 
gamma ray scanner with an average energy of 1.25 MeV operated by NICK TC Scan 
Limited, which has been in use for destination inspection at Aflao Border of Ghana, for 
the past six years, (2010-2016). The facility uses a high energy ionizing radiation in its 
operation; therefore continuous adequacy of the installed biological shielding is critical 
to the protection and safety of the workers and the general public. The workload of the 
facility has increased since its commissioning, requiring the review of the status of the 
installed shielding. Theoretical calculations for dose rates and barrier thicknesses based 
on tenth – value- layer (TVL) concept and NCRP 151, 2005 recommendations, were 
done around the scanning facility using the current operational data. The results were 
then compared with the measured dose rates and the shielding thickness constituted 
during the commissioning stage, and international standards. Calculated dose rate at 
commissioning state ranges from 0.6µSv/hr to 2.4 µSv/hr with an average dose rate of 
1. 43µSv/hr and that of the current operational state ranges from 1.1 µSv/hr to 2.6 
µSv/hr with an average dose rate of 1.54µSv/hr, indicating an increase of 7.9%. Even 
though the dose rates were all below the recommended dose limit of 20µSvh-1 by 
NCRP, there has been an increase in dose to the staff and the general public. It has been 
observed that, the workload has increased three-fold from the commissioning stage to 
current operational state over the past six years. The assessment done on the installed 
shielding using the current operational data indicates that the shielding is inadequate in 
providing protection for the general public and the workers against X-ray radiation 
source of energy of at least 6MeV, and therefore the facility in its current state cannot 
be used to house a linear accelerator of energy up to 10MeV. 
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CHAPTER ONE  
INTRODUCTION 
1.0. Overview 
This chapter gives a brief explanation of the research work carried out.  
It touches on the background, the problem statement, the objectives of the research, 
relevance and the scope of the research work. 
1.1.  Background 
Non-intrusive inspection of cargo containers has become a key issue in recent 
years to parry terrorist activities and to easily verify the shipping content at 
customs in order to curb contraband goods. Such systems are mainly based on X- 
and gamma-ray radiography which produces inside images of the cargos. 
There is a worldwide need for effective scanning of cargo containers in order to 
detect possible contraband goods such as illicit drugs, explosives, nuclear materials 
or weapons, as well as verification of declared manifests. The manual inspection of 
large containers is not practical because of the time constraints and the high labour 
requirements for unpacking and repacking the cargo content. Therefore, there is a 
need for nonintrusive scanning systems. 
Most non-intrusive cargo screening (Industrial radiography) systems are based on 
the use of a radiation (X- or gamma-rays). Those systems can provide high-
resolution intensity images of the cargo contents and are well suited for detecting 
metal-based objects such as weapons. The images obtained from the systems are 
easy to interpret due to the high contrast shapes obtained from the scanning. The 
most used technologies are based on X and gamma-rays systems. 
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Industrial radiography is an established practice that provides benefits concurrent 
with radiation risks. The radiation protection objective, therefore, is to keep the 
risks As Low as Reasonably Achievable (ALARA) while maximizing benefits. 
The purpose of shielding is to protect workers as well as the general public from 
the harmful effects of ionizing radiation. Therefore, adequate shielding design is 
essential during planning and building of such scanner facilities. (NCRP 47, 2002) 
Effective shielding for the scanner units remains a crucial consideration when 
installing scanner facilities. This is to reduce primary and scatter radiation, hence, 
limiting radiation exposure to staff and the general public. 
The radiation shields and design of the scanner facility describe how the required 
facility structure should be determined. The radiation shields of the rooms adjacent 
to the scanning equipment are usually thick to be able to attenuate high energy 
radiation associated with scanning equipment. 
It is important to ensure that the shielding design is installed properly since 
retrofitting  made after the room has been completed are very expensive. 
When designing the radiation shield, the workload of the facility, the intended use 
and the occupancy of the room adjacent to the scanner facility are taken into 
consideration. 
It is often impractical to make an overall experimental determination of the 
adequacy of the shielding adequacy prior to the completion of the building and 
installation of the radiation source. However, shielding voids may be detected by 
use of suitable portable gamma-ray detectors. 
Poor design of scanner shielding can result in unnecessary exposures to the staff 
and the general public. These radiation exposures to the ionizing the radiation can 
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result in two main detrimental health effects, these are Deterministic and Stochastic 
effects. 
Deterministic effects occur at relatively large doses and are called deterministic 
because they are certain to occur, if the dose exceeds a threshold level. 
Deterministic effects are the result of various processes, mainly cell death and 
delayed cell division, caused by exposure to large radiation doses. 
Severity of a particular deterministic effect in an exposed individual increases with 
the dose above the threshold for the occurrence of the effect. 
Radiation exposure can also induce delayed effects such as malignancies and 
hereditary effects, which are expressed after a latency period and may be 
epidemiologically detectable in a population. This cancer induction, referred to as 
stochastic effects, usually takes place over the entire range of doses without a 
threshold level with the probability of occurrence of cancer being higher for higher 
doses and severity of cancer that may result from irradiation is independent of 
dose. 
Stochastic effects may ensue if an irradiated cell is modified rather than killed, 
since modified cells may develop into a cancer after a prolonged delay. 
If a cell damaged by radiation exposure is a germ cell whose function is to transmit 
genetic information to progeny, it is conceivable that hereditary effects of various 
types may develop in the descendants of the exposed individual. 
The broad aim of optimization is to ensure that, the magnitude of the individual 
doses, the number of people exposed, and the likelihood of incurring exposures 
where these are not certain to be received, are all kept As Low As Reasonably 
Acceptable (ALARA), economic and social factors being taken into account. 
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The process of optimising protection is essentially source-related and should first 
be applied at the design stage of any project. This is done by ensuring that the 
installations and radiation sources are provided the best practical protection and 
safety measures possible. 
For the purpose of implementing the principle of optimization and taking into 
account the exposures from different sources, the design of the shielding shall be 
based on dose limit. 
The Basic Safety Series (BSS) defines dose limits the value of the effective dose or 
the equivalent dose to individuals in an exposure situations that is not to be 
exceeded. 
When planning for the construction of gamma and X-ray scanner facilities, the 
dose limits are the doses in the controlled and supervised areas for which the 
facility is designed. (BSS 2014). 
It is important that for effective radiological protection system, the principles of 
distance, time and shielding are all employed to ensure that all radiation doses are 
kept below acceptable dose limits. 
Scanning rooms should also comply with both national and international 
regulations that deal with shielding requirement to render an installation safe from 
the radiation protection view point. The design also takes into consideration the 
structural building codes. 
 
1.2 Statement of the Problem 
Effective shielding for scanner units remains a crucial consideration when 
installing or remodelling containerized scanning facilities. This is to reduce 
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primary and scatter radiation, hence, limiting exposure to staff and the general 
public. 
The facility at the Aflao border uses a 3.8 TBq (103.4 Ci) Co-60 source. The 
enormous radiation energy from this source during beam-on state needs to be 
shielded in order to prevent undue exposure of the workers and the general public, 
therefore continuity and the integrity of the installed shielding of the facility are 
very important. 
Since commissioning of the scanner units, no work has been done to re-evaluate 
the shielding designs. Workload of the facility may have increased with time, 
which may result in increased scan per day; number of working days per week 
which will require that the number of tenth value layers be increased. 
In addition, concrete under radiation exposure, heats up and this drives the water 
out of its internal structure, resulting in swelling, cracking and spalling. 
So as the facility ages, its shielding can develop cracks and voids due to the 
exposure to the gamma-rays. 
It is advisable that during this deteriorating phase, an on-site concrete testing 
service be carried out at regular intervals. 
It is against this backdrop that there is the need to conduct research on this facility 
to determine the current protection status of the installed barriers. The proposed 
research would undertake radiological assessment on the shielding to determined 
wether the installed shielding still provides the needed protection of the staff and 
the general public. 
 
 
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1.3. Objectives 
In installing shield for radiation facility, it is advisable that the shielding design 
stage include consideration of possible future needs for new equipment, higher 
radiation energies and increased workloads.  
The primary objectives of this research are: 
 Assess the shielding integrity of the Co-60 scanning facility at Aflao Border 
by performing shielding calculations using current operational data of the 
facility, and to determine whether or not there has been any deterioration in 
the shielding since commissioning. 
 Verify whether the shielding provides the needed protection for the staff 
and the general public in its current operational states. 
 Analyse the suitability of the current shielding to house a higher energy up 
to 10 MeV X-ray scanners. 
 Make appropriate recommendations to address any radiological health and 
safety problems identified. 
 
1.4.  Relevance of the Study 
Cargo scanning facilities use ionizing radiation. It is therefore important that steps 
are taken to prevent overexposure to these radiations which usually results in 
radiation hazard, and this requires adequate shielding and periodic evaluation of 
the shielding adequacy of the scanning facilities. 
The unit has been in operation for six years till now (2010 – 2016), the operational 
states of the facility has changed with significant variation in workload and with 
new structures built around the facility. 
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This study is relevant because re-evaluation of the shielding integrity of both 
theoretical and verification by measurement will provide the basis for determining 
the current status of protection and safety of staff and members of the public. 
This work will also serve as a reference data for the Nuclear Regulatory Authority 
in enhancing the regulatory process. 
 
1.5.  Scope 
This study is focused on assessment shielding integrity of 38.1TBq Co-60 gamma 
ray destination inspection facility operated by NICK TC Scan Limited at the Aflao 
border town of Ghana.  Dose rates would be measured at critical areas around the 
facility. These rates would be compared with international values, and 
recommendations be made if there is any adverse findings in the operation of the 
facility. 
 
1.6.  Structure of the Thesis 
This thesis work covers five chapters. Chapter One deals with the background to 
the study, problem statement, objectives of the study, relevance and justification of 
the research and scope of the work.  In Chapter Two, a review of related literature 
in radiation is discussed. Chapter Three contains Research Materials and Methods 
used to assess the shielding integrity of the Co-60 gamma ray scanner. Chapter 
Four presents results and discussions on the findings from the research. Chapter 
Five presents Conclusions of the study and Recommendations from the findings of 
the research work. 
The reference section provides the related literature citations used. 
 
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CHAPTER TWO 
LITERATURE REVIEW 
2.0. Introduction 
This chapter presents a detailed overview of the work done in this research area in 
published literature. It discusses the history of shielding, decay scheme and 
properties of gamma rays, key considerations in structural design of gamma ray 
scanners, theory of shielding for gamma and X-ray scanners. 
 
2.1. A brief history of radiation shielding 
The hazards of x rays were recognized after some months of Roentgen’s 1895 
discovery, but dose limitation by time, distance, and shielding was at the discretion 
of the individual practitioner until about 1913. Only then were there organized 
professional efforts to establish guides for radiation protection, and not until about 
1925 were there instruments available to quantify radiation exposure. In his survey 
of organization for radiation protection, Taylor (1979) begins with British and 
German efforts at establishing guidance for X-ray shielding. In 1913, the German 
Radiological Society on X-Ray Protection Measures issued recommendations that 
2 mm of lead shielding was needed, regardless of generator voltage, workload, or 
filtration. In Britain, the Roentgen Society addressed radiation protection, stressing 
operator protection, the need for beam collimation, and the importance of scattered 
X rays. No explicit recommendations on shielding requirements were issued.  
In 1921, the British X-Ray and Radium Protection Committee issued broad 
guidelines, both physical and administrative, on radiation protection in X-ray 
facilities. For diagnostic examinations, 2 mm of lead screening was recommended 
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for the operator, as well as gloves with effectively 0.5 mm of lead shielding. For 
superficial therapy (up to 100 kV X- rays), 2 mm of lead shielding was 
recommended.  
For deep therapy (in excess of 100 kV X- rays) 3 mm of lead shielding was 
recommended. Again, filtration and workload were not addressed. (Mutscheller, 
1925) introduced important concepts in x-ray shielding. He expressed the erythema 
dose, ED, quantitatively in terms of the beam current i (mA), exposure time t 
(min), and source-to-receiver distance r (m), namely,  
ED = 0.00368
  
  
        2.1 
independent of X-ray energy. Years later, as observed by Taylor (1981), unit 
erythema dose was equated to an exposure X of about 600 R. Thus, in modern 
terms, 
X  = 2.2 
  
  
         2.2 
 
2.2 Destination Inspection. 
Destination Inspection (DI) which falls under medico-legal exposure by virtue of 
the incorporation of radiological imaging without medical indication is a service 
whereby goods and import declaration are verified before exportation and also on 
arrival in the importing country. 
The benefits of destination inspections include but not limited to: 
  Avoidance  of the unpacking and repacking of goods  
  Inspection is completed quickly, leading to faster clearance  
  Reduction in congestion at ports, harbours and country borders  
  Increased in Customs revenues  
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  Detection of discrepancies between goods and declarations   
  Reduced smuggling and fraud  
  Improved security  
  Increased compliance.(COTECNA, 2008) 
 
2.3. Scanning Technology 
Radiological imaging, which is carried out in destination inspection technology 
applies to non-destructive methods of inspecting and identifying goods in 
transportation system. It exploits a radiation source (X-ray, gamma rays or 
neutrons) to penetrate a cargo container and produce an image of its contents. The 
images are typically detecting the transmitted radiation on the other side of the 
cargo by different techniques: fluoroscopy or transmission which shows the object 
in superimposition. Computed tomography (CT) normally serves fore luggage 
control as it digital images without superimposition; and scatter radiations are used 
to identify chemical components. These images are assessed and matched with 
entry declaration invoices, bills of entry and packing list to determine whether they 
appear to reasonably represent the declared goods. Irregular or unexpected object 
may warrant further scanning and inspection to determine whether the objects are 
dangerous. Implied in this process is the need for image interpretation, which is 
typically performed by humans. Cargo scanning imagery may attain different 
levels of accuracy, depending on the type and specification of the selected 
technology. High penetration equipment, for instance, can both identify the nature 
of a cargo and allow quantity checks. (SGS, 2006). (Figure 2.1). 
 
 
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Figure 2. 1: Typical Non - Intrusive System Image (GAMMA RAY) 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
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2.4.  Radioactive Decay Scheme of Co-60 
Co-60 is a radionuclide produced in a nuclear reactor by neutron activation of Co-
59. 
Co -60 is a beta emitter with maximum energy of 0.31MeV. Its half-life is 5.27 
years and decay into an excited state of Ni-60. The excited nucleus of Ni-60 
undergoes two photon emission in cascade, with energies 1.17 MeV and 1.33MeV. 
The average energy of the two cascades gamma photons is 1.25MeV. (Figure 2.2) 
 
Figure 2. 2: Radioactive decay scheme for Co-60 
 
 
 
 
 
 
 
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2.5. Shielding materials 
Radiation shielding is based on the principle of attenuation, which is the ability to 
reduce a wave’s or ray’s effect by blocking or bouncing particles through a barrier 
material. Charged particles may be attenuated by losing energy due to interactions 
with electrons in the barrier, while X-ray and gamma radiation are attenuated 
through either photoemission, scattering, or pair production. Neutrons can be made 
less harmful through a combination of elastic and inelastic scattering, and most 
neutron barriers are constructed with materials that encourage these processes. 
The materials typically used for shielding purposes are ordinary concrete, heavy 
concrete, lead, steel, polyethylene, paraffin, earth and wood. 
 
2.5.1. Concrete 
Concrete has many advantages and is commonly used. It gives good X-ray and 
neutron shielding (concrete contains hydrogen), good structural strength and is 
relatively inexpensive. The required thickness to attenuate the incident photons to 
one tenth, the tenth-value layer (TVL), in concrete is much larger than for neutrons 
and therefore an adequate shielding for photons is more than enough for 
photoneutrons. For new buildings to house radiation scanner facilities, concrete 
will usually be the material of choice since it is the least expensive. However, if 
space is at a premium it may be necessary to use a higher density building material. 
(IAEA SRS 47, 2006) 
 
 
 
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2.5.2.  Lead 
Lead has a very high density and is a very good X- and gamma ray shielding 
material, but it will need support by a second material to handle the neutrons, 
which it is nearly transparent to, albeit that the high-energy neutrons get attenuated 
by inelastic scattering. Lead is also toxic and more expensive than concrete. 
 
2.5.3. Steel 
Steel is also expensive compared to concrete, but is not toxic. It is more efficient 
than concrete as an X-ray and gamma ray shielding material, but less efficient than 
lead. It is also nearly transparent to neutrons, but is a good structural material for 
shielding the neutron. 
 
2.5.4. Polyethylene and Paraffin 
These two materials are considered together since they are very similar. Paraffin, 
sometimes called paraffin wax, has the same percentage of hydrogen (14.3%) as 
polyethylene and is less expensive. However, it has lower density and is flammable 
so it is usually avoided in any permanent barriers. Polyethylene is perhaps the best 
neutron shielding material available, but it is relatively expensive. It is available 
both pure and loaded with varying percentages of boron to increase the thermal 
neutron capture. It is usually used for neutron shielding, and where space is at a 
premium (e.g., in doors) around ducts or in ceiling shields where insufficient height 
is available for less costly methods. It is easy to fabricate and relatively strong. 
Standard Borated polyethylene (BPE) contains 5 % boron by weight.                     
(Fadlalla, 2010). 
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2.5.5. Earth 
Earth is also commonly used as a shielding material by placing the accelerator 
room partially or entirely underground. Earth is not a well-defined material and its 
density can vary considerably. While earth composition is quite variable, it is not 
too different elementally from concrete and it is sufficient to consider it as 
equivalent to concrete with a density of 1.5 g cm–3. A common problem with a 
room that is below is that, there may be a diagonal path up through mostly earth 
where the shielding is insufficient. Sand that is in contact with the earth will 
normally be damp enough that it can be treated the same as earth. Dry sand, 
however, such as bagged sand, needs some hydrogenous material beyond it in 
order to be an adequate neutron shield. (Fadlalla, 2010). 
 
Table 2. 1: Shielding Materials and Their Densities. Source: IAEA Safety series  
report no. 47, 2006. 
 
Building 
material 
Density 
(kg·m3) 
Comment 
Concrete 2350 Will vary with mineral content 
Barytes 
concrete 
3400 - 3500 
Most commonly used for dense concrete but 
expensive 
Iron-ore with 
Ferrosilicon 
4000 - 5400 
Range of densities which depend on proportions 
of ore mixture to sand 
Ledite 
3844 and 
4613 
Pre-moulded high density interlocking blocks 
Clay bricks 1600 
May be used for installations up to 500 kV with 
supplementary lead or 
Breeze blocks 1100 - 1400 Steel shielding 
Earth fill  1600 May be useful if bunker is below ground level 
Steel  
7900 
Normally used as a supplementary shielding on 
an existing treatment room 
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2.6. Shielding of Gamma Radiation 
When gamma radiation is incident on a finite thickness of material, there exists 
some probability that the radiation will interact in the material and later be 
attenuated. 
Thus, any of the common gamma interaction processes may result in secondary 
photons that have a finite probability of reaching the dose point. The extent to 
which such secondary photons add to the fluence or dose at the dose point is 
usually described through the use of an appropriate build-up factor. 
Build-up factors may refer to various quantities of interest, such as photon fluence, 
photon energy fluence, exposure, or dose, and the values among all are somewhat 
different. Much of the available build-up data relates to determination of exposure 
or kerma in a small air volume envisioned to be located within the shielding 
medium of interest. 
These data are also suitable for evaluation of dose to water or other low-Z material 
of interest. 
The dose build-up factor is a dimensionless quantity that represents the ratio of 
total dose (including the dose from secondary photons) at the dose point to primary 
photon dose at the same point. The primary photon dose naturally comes from 
original photons that have penetrated the shielding material without interacting. 
Magnitudes of build-up factors vary widely, ranging from a minimum of 1.0 to 
very large values, depending on source and shield characteristics. 
 
 
 
 
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2.6.1. Good geometry shielding situation 
When a narrow parallel beam of photons passes through a relatively thin shield, 
and if the dose point is many beam diameters away from the exit surface of the 
shield, we have a situation referred to in photon shielding as good geometry. This 
means, simply, that virtually all of the photons arriving at the dose point will be 
primary photons, and the dose, D, or dose rate, at a point of interest outside the 
shield, is related to the unshielded dose, Do, or dose rate, at the point by 
      
          2.10 
where: 
  is the linear attenuation coefficient for the photons of the energy of interest in the 
shield material, and x is the shield linear thickness. 
2.6.2. Point isotropic source 
The most popular source geometry involved in many calculations is the point 
isotropic source. While no real source is a true point, many sources are sufficiently 
small in dimensions that they can be treated mathematically as point sources. In 
practice, if the distance from source to dose point exceeds about three times the 
maximum source dimension and self-attenuation within the source volume is not a 
concern, the errors resulting from treating the source as a point will not exceed a 
few percentages. The assumption that the source is isotropic means that radiation 
of concern is emitted uniformly in all directions throughout 4π geometry. Consider 
a source of monoenergetic gamma radiation that emits S gamma rays per second 
and that is situated at a distance r (cm) from the dose point, and a shield of 
thickness T (cm) through which the gamma radiation passes before reaching the 
dose point. 
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where 
D is the dose point 
S is the radiation source 
2.6.2.1. Unshielded dose rate 
The unshielded dose rate at the dose point is given by 
 ̇ = 
   
 
 
    
         2.11 
where E is the photon energy, MeV, 
 
  
 is the mass energy absorption coefficient for the material at the dose point, cm2 
g-1 
4π   is the beam area 
k is a collective constant to convert energy fluence rate to dose rate; if the dose rate 
is in gray/hour, k will have a value of 5.76 x 10-7. 
 
2.6. 2.2      Shielded primary photon dose rate 
The primary photon dose rate is attenuated exponentially, and the dose rate from 
primary photons, taking account of the shield, is given by 
 ̇ = 
   
 
 
    
    
        2.12 
Figure 2.3: Point isotropic source shielding configuration 
T 
S D 
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where µ is the linear attenuation coefficient for the photons in the shield material. 
This expression does not account for the build-up of secondary radiation and will 
generally underestimate the true dose rate, especially for thick shields and when the 
dose point is close to the shield surface. 
2.6.2.3 Shielded dose rate accounting for build-up 
The added effect of the build-up is taken into account by incorporating a point 
isotropic source dose build-up factor, B, gives 
 ̇ = 
   
 
 
     
    
        2.13 
The magnitude of the build-up factor depends on the photon energy, the shield 
material and thickness, the source and shield geometry, and the distance from the 
shield surface to the dose point. In most cases, dose build-up factors for point 
isotropic sources have been determined under the assumption that both the source 
and the dose point reside within an infinite volume of the shield material. As a 
consequence, shielded doses evaluated using such build-up factors tend to be 
conservative for most practical situations in which the dose point is outside the 
shield and not subject to backscattering from shield material behind the dose point. 
(Shultis & Faw 1996).When one wishes to determine the shield thickness to yield a 
specific dose rate, equation (2.13) cannot be solved explicitly for T because the 
value of B depends on T. Solutions can be obtained by making educated guesses 
for the value of T, looking up the corresponding values of B, and solving for the 
dose rates; results can be plotted, and the correct value of T determined for the 
desired value of dose rate. Alternatively, we can use an analytical form of the 
build-up factor that can be incorporated into equation (2.13) and, through an 
iterative process using a computer or calculator, and solve for the desired 
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thickness. There are a number of algebraic expressions that have been used to 
represent B. 
Among the most popular is an expression referred to as Taylor’s form of the build-
up factor, given by 
B =    
     + (1-    
          2.14 
Where  A1 and A2µ and β are constants for a given energy and shield material. 
 
2.7. Shielding for X-ray 
Shielding for X-ray equipment is considered under the two categories; 
 source shielding and  
 structural shielding.  
Source shielding is usually provided by the manufacturer of the X-ray tube 
housing. 
For non-medical x-ray equipment, the tube housing and other parts of the X-ray 
apparatus, such as the transformer, are shielded to reduce the leakage X-ray 
radiation through the housing to acceptable levels. 
Protective tube housing designed to limit the amount of leakage radiation are 
provided for all X-ray machines. 
Leakage radiation, as used in these specifications for tube housings, means all 
radiation coming from the tube housing except for the useful beam. 
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Structural shielding for an X-ray facility provides protection from the useful or 
primary x-ray beam, from leakage radiation and from scatter radiation. It encloses 
both the x-ray equipment and the object being irradiated. 
The amount of scatter radiation depends on the X-ray field size, energy of the 
useful beam, the effective atomic number of the scattering media and the angle 
between the incoming useful beam and the direction of scatter. 
A key design parameter is the facility workload (W): 
                       2.15 
where W is the weekly workload, usually given in mA-min per week;  
E is the tube current multiplied by the exposure time per view, usually given in        
mA s;  
Nv is the number of views per cargo irradiated;  
Np is the number of cargo per week and  
k is a conversion factor (1 min divided by 60 s). 
Another key design parameter is the use factor U for a wall (or floor or ceiling) n. 
The wall may be protecting any occupied area such as a control room, office or 
waiting room.  
 
 
 
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The use factor is given by: 
    
 
  
         2.16 
where, N is the number of views for which the primary X-ray beam is directed 
toward wall n. 
The structural shielding requirements for a given X-ray facility are determined by 
the following: 
 the maximum tube potential, in kilovolts-peak (kVp), at which the X- ray 
tube is operated. 
 the maximum beam current, in mA, at which the X- ray system is operated 
 the workload (W), which is a measure, in suitable units (usually mA-min 
per week), of the amount of use of the X-ray system 
 the use factor (U), which is the fraction of the workload during which the 
useful beam is pointed in the direction of interest 
 the occupancy factor (T), which is the factor by which the workload should 
be multiplied to correct for the degree or type of occupancy of the area to 
be protected 
 the maximum permissible dose equivalent rate (P) to a person for 
controlled and non-controlled areas (typical absorbed dose limits are 1 
mGy for a controlled area in one week and 0.1 mGy for a non-controlled 
area in one week) 
 type of shielding material. 
 the distance (d) from the source to the location being protected. 
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With these considerations included, the value of the primary beam transmission 
factor B in mGy per mA-min at one metre is given by: 
    
   
   
         2.17 
Shielding of the x-ray facility must be constructed so that protection is not 
impaired by joints; by openings for ducts, pipes and so on, that pass through the 
barriers; or by conduits, service boxes and so on, embedded in the barriers. The 
shielding should cover not only the back of the service boxes, but also the sides, or 
be extended sufficiently to offer equivalent protection. Conduits that pass through 
barriers should have sufficient bends to reduce the radiation to the required level. 
Observation windows must have shielding equivalent to that required for the 
partition (barrier) or door in which they are located. 
Radiation scanning facilities may require door interlocks, warning lights, closed 
circuit television or means for audible (e.g. voice or buzzer) and visual 
communication between anyone who may be in the facility and the operator. 
Protective barriers are of two types: 
 Primary protective barriers, which are sufficient to attenuate the primary 
(useful) beam to the required level 
 Secondary protective barriers, which are sufficient to attenuate leakage, 
scattered and stray radiation to the required level. 
In a design of secondary protective barrier, the required thickness to protect against 
each component is calculated separately. If the required thicknesses are about the 
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same, an additional HVL is added to the greatest calculated thickness. If the 
greatest difference between the calculated thicknesses is one TVL or more, the 
thickest of the calculated values will suffice. 
Scattered radiation intensity depends on scattering angle, energy of the useful 
beam, field size or scattering area, and subject composition. 
When designing secondary protective barriers, the following assumptions are 
made: 
 When X-rays are produced at 500 kV or less, the energy of the scattered 
radiation is equal to the energy of the useful beam. 
 After being scattered, the X-ray energy spectrum for beams generated at 
voltages greater than 500 kV are degraded to that of a 500 kV beam, and 
the absorbed dose rate at 1 m and 90 degrees from the scatterer is 0.1% of 
that in the useful beam at the point of scattering. 
The transmission relationship for scattered radiation is written in terms of the 
scattering transmission factor (Bs) with units of mGym
2 (mA-min)–1: 
  =  
     
     
 
 
 
   
    
        2.18 
where P is the maximum weekly absorbed dose rate (mSv), 
dscat is the distance from the X- ray tube’s target and the scanner, 
dsec is the distance from the scatterer to the point of interest that the secondary 
barriers are meant to shield, 
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α  is the ratio of scattered radiation to incident radiation,  
F is a factor accounting for the fact that X ray output increases with voltage. 
Smaller values of B require thicker shields. 
2.8. The classification of work areas 
Approach to exercising control of the source of exposure and over the workers who 
are occupationally exposed has been recommended by the ICRP (1990). It involves 
formally designated proximal localities to the source into controlled and supervised 
areas (ICRP 1990). A controlled area is one in which access may be restricted 
during periods of irradiation, requiring workers to follow well-established 
procedures and practices aimed specifically at strictly controlling radiation 
exposures. In identifying a controlled area, it may be useful to make use of existing 
physical boundaries, such as the external walls of the facility and/or room walls. 
A supervised area shall be designated in any area not already designated as a 
controlled area, but where occupational exposure conditions need to be kept under 
review even though specific protection measures and safety provisions are not 
normally needed. 
The ICRP 1990 recommend that the classification of work areas are best based on 
operational experience and judgement, taking into account both expected exposure 
levels and of the likely variations in these exposures. Designation of areas may be 
defined in terms of the dose rate at the boundary, with the aim of ensuring that 
anyone outside a designated area will not need to be regarded as occupationally 
exposed. Thus, doses received outside the designated areas should be low enough 
to ensure that in normal conditions, the level of protection for those who work in 
the facility will be less than the dose constraint of 1 mSv per year. 
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Values of dose rate based on a fraction of the relevant dose limit have often been 
used in the past for defining the boundaries of controlled areas. However, careful 
consideration should be taken of the length of time for which the dose rate remains 
at or above the defined level and the risk from potential exposures. Periodic review 
of the exposure situation may necessitate the need for extra protective measures or 
changes to the boundary of a work area. 
There should be signs at entrances to controlled area to indicate to staff, especially 
to maintenance staff, the special procedures currently applied in the area and that a 
radiation source is likely to be present (ICRP 1999). 
 
2.9. Occupancy Factors 
Evaluation of strategically shielded scanner facility must certainly involve an 
estimation of occupancy for all adjacent areas. The occupancy factor for each 
selected monitoring site should represent the average time that the maximally 
exposed individual is present when the radiation is present. 
NCRP No. 147 (2004) provides a list of suggested occupancy factors that are to be 
used as a guide in calculations when accurate occupancy data are not available. 
 
2.10 Use Factor 
The use factor describes the different beam orientation used for treatment when 
calculating the required barrier thickness for each beam orientation. If conventional 
treatment techniques are to be used, NCRP Report 49 suggests a use factor of 1 for 
the floor with the beam pointing vertically down and0.25 for each wall and ceiling 
if specific values are not available. These use factors may depend on the particular 
use of the facility and also on the energy used. For example, a facility performing a 
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large number of total body irradiations may have a use factor greater than 0.25 for 
one wall, and lower for other walls. (NCRP, 2004) 
 
2.11. Assessment of Occupational Exposure 
During the implementation or evaluation of a radiation protection plan there should 
be tracking of changes to plan outcome indicators enabling the opportunity for 
feedback. This routinely occurs for occupational exposure in the form of individual 
monitoring or in the form of a radiation surveying. 
In the characterization of occupation exposure, the main indicators to be examined 
are the level of the collective dose and the distribution of individual doses (IAEA 
2002). The assessment of collective dose from occupation exposure is usually 
based on the recorded dose from individual monitoring. Such information provides 
confidence and may provide data of use in reviewing the radiation protection 
program. Workers exposed to radiation from sources not related to, or required by 
their work shall receive the same level of protection as if they were members of the 
public. 
A formal individual dose assessment would normally be performed should 
monitoring indicate that the corresponding annual effective dose exceeds 1mSv, 
and should certainly be conducted in total annual effective doses estimated to be 
above 5 mSv (IAEA 1999). 
In the case of shielding design, goals are used in the evaluation of the effectiveness 
of barrier construction for the required level of protection. Ideally one should 
perform a radiation survey when the source is of greatest activity, sampling nearby 
controlled and uncontrolled areas. In addition to providing information on the 
control of occupational exposure, a program of individual monitoring may be 
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helpful in confirming the classification of work areas. For example, when 
combined with data on the frequency of the jobs performed, their duration, 
measured dose rates and the number of workers exposed. In practice, it is possible 
to achieve an accuracy of about 10% at the 95% confidence level for measurements 
of radiation fields in good laboratory conditions. The NCRP recommend the setting 
of a weekly shielding design goal for a area at an equivalent dose value of 0.1 mSv 
per week and 0.02 mSv per week for an uncontrolled area. 
Measurements should be made at a number of occupied distances from the source, 
including outside the facility (e.g. on the roof or an adjacent footpath). The 
measurements should be taken from the source to the nearest likely approach of the 
sensitive organs of a person to the barrier (Podgorsak, 2000). The NCRP 
recommend a measurement point of no closer than 30 cm to a wall. For the floor 
below, not higher than 1.7 m from that floor level and, for ceiling transmission, a 
distance of at least0.5 m above the floor of the room above. (Podgorsak, 2006) 
If, as a result of a radiation survey, supplementary shielding is added to the 
protective barriers, the survey should be repeated to evaluate the adequacy of the 
shielding after the modification. 
 
2.12.0 Calculation method 
For radiation safety purposes, the barriers thicknesses must be designed to 
attenuate the primary, leakage and scatter radiations. 
2.12.1 Primary barrier shielding. 
The primary barrier is the shielding used for the attenuation of the primary beam. It 
is givens as, (McGinley, 2002) 
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   = 
         
   
        2.19 
P = the design limit for a public area which is 0.02mSvhr-1 
d  = the source axis to the point behind barrier 
SAD  = source axis distance 
W = workload 
U = use factor 
T = occupancy factor 
The workload W, the average absorbed dose of radiation produced by a source in a 
specified time (mostly in a week) and expressed as; 
 
  =                     2.20 
Where 
       is dose per scan 
   is number of scan per day 
   is number of scan days per week. 
The required number (n) of tenth layer value (TVL) of concrete needed to shield 
the primary radiation from the source is determined by the equation. 
n = - (               2.21 
therefore the thickness(   of concrete required to attenuate the primary beam to the 
desired level is: 
   =                                                  2.22 
 
 
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2.12.2 Secondary barrier shielding. 
2.12.2.1. Scattered radiation 
The secondary barrier must shield against both the leakage and scatter radiations. 
As described in section 2.2.3, at megavoltage energies the dominant photon 
interaction is the Compton effect. In the Compton interaction, the photon is not 
absorbed, rather it is scattered at an angle and with reduced energy. Both the 
leakage and scatter radiations undergo Compton scattering. 
The leakage beam scatters multiple times, and thus loses significant energy, 
while traversing the dense material (typically lead) in the head of the radiation 
generator. The scatter radiation, by definition, is scattered at least once before 
reaching the secondary barrier. 
Two main sources of scatter radiation are possible:  
 scatter radiation produced by scattering of the primary beam within the 
cargo,  
 scatter radiation produced by scattering of the primary and cargo scatter 
radiations with the walls and fixtures of the scanning tunnel.  
A secondary barrier with thickness adequate to attenuate scatter radiation from the 
primary beam will generally be sufficient for cargo scatter radiations, since 
multiple scattering interactions reduce photon energy and penetrability 
significantly. 
Barrier thicknesses are calculated for the leakage and patient-scatter beams 
independently due to the difference in energy between them. Leakage radiation, 
having undergone multiple scattering within the head of the scanning machine, is 
typically of a much lower energy than the cargo-scatter radiation, which is 
scattered less often before reaching the walls. 
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The barrier transmission (  ) required to shield against scattered radiation from the 
container is given by the equation (NCRP 151, 2005), (McGinley, 2002) 
  =  
     
     
 
 
 
   
    
       2.23 
 
Where 
α is the scatter fraction or the fraction of the primary beam absorbed dose that 
scatters at a particular angle.  
     is the distance from the primary radiation source to the scattering material, 
usually taken as the SAD of the machine. 
     is the distance from the scattering point to the point of interest (POI). 
F is the incident field area (in cm2) on the scatterer. 
The number (n) of TVL required to shield the scattered radiation from the 
container is given by, (McGinley, 2002) 
n =      
 
  
          2.24 
The required thickness,      for the attenuation of the scattered radiation is 
   =                                               )   2.25 
2.12.2.2. Leakage radiation shielding 
Based on NCRP 151, 2005, the required transmission barrier is given by the 
equation 
   = 
       
 
  
        2.26 
P is the design limit for a public area which is 0.02mSvhr-1 
    is the source axis to the point of interest behind barrier 
W = workload 
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T = occupancy factor 
 
The number (n) of TVL required shielding the leakage radiation from the scanner 
head is given by 
n =     
 
  
          2.27 
The required thickness,      for the attenuation of the leakage radiation is 
   =                                                .  2.28 
 
2.13.0 Average instantaneous dose rate calculations. 
In the IAEA SRS 47 report, the instantaneous dose rate (IDR) is defined as the 
direct reading of a survey meter that gives the dose per hour, averaged over a 
minute. 
 
2.13.1 Instantaneous dose rate due to the primary radiation (  ) 
     = 
       
  
       2.29 
Where 
    is the maximum dose rate output at 1.0m is 0.8Gy/min. 
    is the primary barrier transmitted factor given by the relation 
   = 
 
  
      
        2.30 
Where    is the thickness of the primary barrier 
 
 
 
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2.13.2 Dose rate from the scattered radiation. 
The dose from the scattered radiation is given by, 
 
     = 
       
 
   
    
    
     
       2.31 
Where    = 
 
  
       
      2.32 
 
α is the scatter fraction or the fraction of the primary beam absorbed 
dose that scatters at a particular angle.  
     is the distance from the primary radiation source to the scattering 
material. Usually taken as the SAD of the machine. 
     is the distance from the scattering point to the point of interest (POI). 
F is the incident field area (in cm2) on the scatterer. (NCRP 151, 2005) 
2.13.3 Dose from the leakage radiation 
The dose from the leakage radiation is given by 
 
     = 
       
 
   
    
    
     
            2.33 
 
Where    = 
 
  
        
            2.34 
α is the scatter fraction or the fraction of the primary beam absorbed 
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dose that scatters at a particular angle.  
     is the distance from the primary radiation source to the scattering material. 
Usually taken as the SAD of the machine. 
     is the distance from the scattering point to the point of interest (POI). 
F is the incident field area (in cm2) on the scatterer. 
 
Total dose rate for leakage and scatter radiation is defined by the equation; 
 
             +             2.35 
 
2.14.0 Time Averaged Dose-Equivalent Rates 
When designing radiation shielding barriers it is usual to assume that the workload 
will be evenly distributed throughout the year. Therefore, it is reasonable to design 
a barrier to meet a weekly value equal to one-fiftieth of the annual shielding design 
goal (NCRP, 2004). However, further scaling the shielding design goal to shorter 
intervals is not appropriate and may be incompatible with the ALARA principle. It 
is more useful if the workload and use factor are considered together with the IDR 
when evaluating the adequacy of a barrier. 
For this purpose, the concept of time averaged dose equivalent rate (TADR) is 
used. 
The TADR is the barrier attenuated dose-equivalent rate averaged over a specified 
time or period of operation. TADR is proportional to IDR, and depends on values 
of W and U. There are two periods of operation of particular interest to radiation 
protection, the week and the hour. (NCRP 151, 2005) 
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The maximum dose in-any-hour (TADRh) due to leakage and scattered gamma 
radiation is given by: 
       
   
         2.36 
Where    is workload in one hour and the rest of the symbols have their usual 
meaning as stated earlier in the text. 
 
2.15.0  Calculation of required barrier thickness for X – ray of energy 6 MeV. 
2.15.1 Primary barrier shielding. 
The primary barrier is the shielding used for the attenuation of the primary beam. 
It is given as    
   
   
 (McGinley, 2002) 
P is the design limit for a public area which is 0.02mSvhr-1 
d  = the source axis to the point behind barrier 
SAD  = source axis distance 
W = workload 
U = use factor 
T = occupancy factor 
The workload W, the average absorbed dose of radiation produced by a source in a 
specified time (mostly in a week) and expressed; 
n = - (             (2.31) 
therefore the thickness(     of concrete required to attenuate the primary beam to 
the desired level is: 
   =                                                 
 
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2.16.0 Secondary barrier shielding 
2.16.1 Scattered radiation 
The barrier transmission (  ) required to shield against scattered radiation from the 
container/cargo is given by the equation (NCRP 151. 2005) 
  =  
     
     
 
 
 
   
    
       2.33 
Where 
α is the scatter fraction or the fraction of the primary beam absorbed 
dose that scatters at a particular angle.  
     is the distance from the primary radiation source to the scatteringmaterial. 
     is the distance from the scattering point to the point of interest (POI). 
F is the incident field area (in cm2) on the scatterer. 
The number (n) of TVL required to shield the scattered radiation from the 
container is given equation   n = - (            
  
The required thickness,      for the attenuation of the scattered radiation is 
   =                                               
 
 
 
 
 
 
 
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 CHAPTER THREE 
MATERIALS AND METHODS 
3.0 Introduction 
This section describes the method and the materials used in the assessment of the 
scanner. 
The layout of the facility, distance measurements, dose measurements at critical 
locations around the facility, primary and secondary barrier thickness calculations 
are describe in this section. 
 
3.1   Materials 
The gamma ray scanner facility at Aflao is one of the scanning facilities operated 
by NICK TC Scan limited of Ghana. 
The facility was built in October, 2010. The primary shield is made of concrete of 
density 2350 kgm-3. Its thickness was measured to be 500 mm.  
The secondary shield is also made of concrete of thickness 430 mm. The scanning 
tunnel is 30 m long 
The scanning method is such that the scanned container or cargo remains still 
whilst the scanning unit moves at a speed of 12m/min. The scanning time for each 
cargo is 90 seconds. 
Safety systems installed at the facility helps determine whether anyone is in the 
scanning tunnel area or not during scanning.  
A motion sensor at both ends of the tunnel triggers the retraction of the source into 
its shield when it detects an intrusion during beam-on state.  
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CCTV cameras are also installed in the tunnel for monitoring the tunnel before and 
during the scanning process.  
Other safety systems at the facility include; safety confirmation devices, an alarm, 
radiation symbol on the door, and traffic marshals at both ends of the tunnel. 
 
3.2 Technical specification of the facility. 
Manufacturer: Beijing Huahixing Science Tech-Development Co. Ltd. 
Model: TC-SCAN FMG 
Serial Number: HFMG01405032 
Scan speed: 6-18m/min 
Power rating 30KVA 
Target W x H: 3m x 4.6m 
Date of manufacture: January, 2006 
Radionuclide: Co-60 
Half-life: 5.7 years 
Initial activity: 3.8 TBq (103.4 Ci)  
Dimension of the tunnel: L x B: 30m x 18 m 
 
3.3 Area of Survey 
Figure 3.1 shows the outline of the NICK TC Scan facility at the Aflao Border. The 
areas where the survey was carried out are the Manager’s Office, the Secretary’s 
Office, Control Room, Check-in, Reception, Washrooms and the Marshal post. 
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Figure 3. 1: Layout of the NICK TC Gamma ray scan facility. 
 
3.4 Method 
Measurements of distances were taken using surveyor’s metre, from the source 
axis and the scatterer. 
The following points of interest were considered 
 Source axis to the point of interest for the primary barrier shielding. 
 Source axis to the scatterer for the scattered radiation shielding.. 
 Container to the point of interest for scattered radiation shielding. 
 
 
 
 
 
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3.5.0 Instruments used 
3.5.1 Dose rate meter 
The survey meter used in this work is a Rad-eye G10 which was manufactured in 
Germany by Thermo electron Scientific Inc. 
It is a portable radiation detection and measurement instrument used to check 
personnel, equipment and facilities for radioactive contamination or to measure 
external or ambient ionizing radiation fields. The instruments are designed to be 
handheld, are battery powered and of low mass to allow easy manipulation. Other 
features include an easily readable display, in counts or radiation dose, and an 
audible indication of the count rate. These features, among others makes this 
detector one of the most used detectors. 
The  Rad-eye G10 gamma survey meter (Fig 3.2) is specifically designed for 
gamma survey from background up to personal safety levels and used by first 
responders, in the nuclear power industry, radiography and for radiation protection 
purposes. 
 
Figure 3. 2: Rad eye G-10 gamma survey meter 
 
 
 
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Technical specifications of the rad-eye G10 survey meter 
Model               Rad-Eye G10 
Serial number   46506/76  F-Nr0738 
Country of Origin   Germany 
Detector   Energy compensated GM tube 
Measuring range  0.5 µSv/h -100 mSv/h 
Energy range   50keV- 3MeV 
 
3.6. Shielding barrier calculations 
Two types of radiation reach the walls of the room and must be accounted for in 
shielding design calculations are; 
 Primary beam (either attenuated through the container or unattenuated) and 
 Secondary radiation arising from leakage through the shielded head of the 
radiation generator and from scattered radiation produced by the interaction 
of the primary beam within the container. Since the energy of the primary 
and secondary beams differ significantly, they must be considered 
independently and require separate shielding design calculations. 
 
3.6.1. Calculation of the workload for the gamma ray source. 
The workload was calculated using equation (2.20) 
 
 
 
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3.6.2. Primary Barrier Shielding for the Gamma Source 
The barrier attenuation factor need to shield against primary radiation was 
calculated using equation (2.19). 
Since there are no structures around the space behind the primary barrier, the only 
group of people affected by transmitted beam outside the barrier is the public. 
The number of tenth-value layers of concrete required to shield against primary 
radiation was also determined using the equation (2.21) 
The thickness, tp, of concrete required to shield against the primary radiation is 
given by tp = number of TVLs  x  TVL in concrete for 1.25 MeV primary γ 
concrete. 
The TVL for concrete (density 2350 kgm-3) for gamma radiation is 218 mm. 
 
3.6.3.0 Secondary barrier shielding for the 1.25 MeV Gamma-ray Scanner. 
3.6.3.1. Leakage Gamma Radiation 
The areas outside the shielding where people who are affected by leakage radiation 
are located are the Manager’s office, the secretary’s office, control room, check-in, 
the reception, the washrooms and the marshal post. 
The required attenuation to shield against leakage radiation from the scanner head 
for the above named areas was calculated using equation (2.26). 
The number of tenth value layer of needed to shield against the leakage gamma 
radiation was calculated using equation (2.21) 
The thickness of required to shield against leakage radiation was calculated using 
equation (2.22). 
 
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3.6.3.2 Scattered gamma radiation 
The areas outside the shielding where people who are affected by scattered 
radiation are located are the manager’s office, the secretary’s office, control room, 
check-in, the reception, the washrooms and the marshal post. 
The needed barrier thickness to effectively shield the scattered gamma ray from the 
container was determined using equation (2.23). 
The number of tenth value layer of needed to shield against the scattered gamma 
radiation was calculated using equation (2.21) 
The thickness of required to shield against scattered radiation was calculated using 
equation (2.22). ie 
   =                                                            ) 
Where the TVL for 1.25 MeV in concrete of density 2350 kgm-3 for scattered 
radiation is 151 mm. 
 
3.7. Average instantaneous Dose Rate Calculation for the primary Gamma 
ray radiation 
Primary barrier transmission factor required was calculated using equation 2.30i.e. 
   = 
 
  
      
 
Where te is thickness of the primary barrier employed which was measured to be 
500 mm. 
TVL for the gamma ray for the concrete used is 218 mm. 
 
 
 
 
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The dose rate for the primary radiation was determined using equation (2.29) 
 
   = 
       
  
 
Where 
DRm is the maximum dose rate output at 1.0 m is 0.8Gy/min. 
 
3.8. Instantaneous dose rate from leakage gamma ray radiation. 
The expected transmission factor was determined using 
    = 
 
  
      
 
Where the barrier thickness used for the leakage is 430 mm, and the TVL for the 
concrete used for the leakage radiation is 218 mm. 
 
The dose rate from the leakage radiation was calculated using equation 
     = 
       
 
   
    
    
     
  
  
3.9. Instantaneous dose rate from scattered gamma ray radiation. 
The needed scatter transmission factor from the container was determined using the 
equation 
    = 
 
  
       
 
Where te is given as 430 mm; and the TVL for the scatter radiation for the concrete 
used is 151 mm. 
 
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The dose rate due to the scatterer was calculated using equation 2.31 
ie 
     = 
       
 
   
    
    
     
  
Where 
    = 
 
  
       
 
 
3.10.0 Barrier thickness consideration for X-ray radiation of energy 6MeV 
3.10.1 Barrier thickness for primary X-ray beam of energy 6MeV 
 
The barrier needed factor required to shield against primary X-ray was determined 
using equation (2.17) 
   = 
   
   
 
P is the design limit for a public area which is 0.02mSv/hr 
d  = the source axis to the point behind barrier 
SAD  = source axis distance 
W = workload (Gy/wk) 
U = use factor 
T = occupancy factor 
The tenth value layer was also calculated using equation 
n = - (       
where     is attenuation barrier determined. 
The needed shielding thickness was calculated using equation 
tp = number of  TVLs  x  TVL in concrete for 6MeV primary X- ray radiation. 
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The TVL for concrete (density 2350 kgm-3) for x-ray radiation is 343 mm. 
 
3.10.2 Barrier thickness for leakage radiation for 6MeV X-ray radiation. 
The areas outside the shielding where people who are affected by leakage radiation 
are located are the manager’s office, the secretary’s office, control room, check-in, 
the reception, the washrooms and the marshal post. 
The barrier transmission, B, needed to attenuate the radiation to the scatterer 
acceptable level was calculated using the equation 
 
  =  
     
     
 
 
 
   
    
 
where 
P is the maximum weekly absorbed dose rate (mGy), 
dscat is the distance from the x ray tube’s target and the scanner, 
dsec is the distance from the scatterer to the point of interest that the secondary 
barriers are meant to shield, 
α is the ratio of scattered radiation to incident radiation,  
F is a factor accounting for the fact that x ray output increases with voltage. 
Smaller values of B require thicker shields. 
 
 
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CHAPTER FOUR 
RESULTS AND DISCUSSION 
4.0 Introduction 
The results obtained from the research are presented and discussed in this section. 
The discussions are in the context of the international standards found in the 
reviewed literature. 
 
4.1 Primary barrier thickness for the gamma ray scanner. 
Table 4. 1:  Calculated and measured primary barrier thickness for the gamma ray  
scanner 
 
The group of people that are affected by transmitted beam through the primary 
barrier is the public, since there are no structures behind the barrier.  
The nearest distance of approach behind the primary barrier is 15.8 m, and the  
calculated barrier thickness needed to protect the occupant in this area is 329.43 
mm of concrete. The installed barrier thickness for this purpose was measured to 
be 500.00 mm, an indication of over-shielding. This consideration might be given 
to make provision for increase in workload and or for higher energy source. And 
also for the purpose of radiation protection, it is more desirable to over-shield a 
radiation source than to under shield it. This gives an added protection against the 
source. 
POI    P 
(Sv/wk) 
 
d/m 
 
W 
 
U 
 
Be 
 
nTVLs 
Calculated 
thickness 
(mm) 
Designed 
Thickness 
(mm) 
Behind 
the 
primary 
barrier 
 
2.0E-5 
 
15.8 
 
0.06 
 
0.25 
 
3.08E-2 
 
1.51 
 
329.3 
 
500.00 
         
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4.2.0 Secondary barrier thickness for the Co-60 scanner 
The radiation transmitted by the secondary barrier (leakage and scatter) would 
affect people located at the following places, Manager's Office, Secretary’s Office, 
Washroom, Control Room, Check In Office, Reception, Waiting Area, Marshall 
Post and the corridor. 
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 4.2. 1.  Leakage Barrier Thickness for the Co-60 Gamma-ray Scanner 
 Table 4. 2: Leakage barrier thickness compared with designed specification 
 
 
 
 
 
 
 
 
 
 
POI 
    P 
 (Sv/hr) 
 
d/m 
 
T 
 
Bl 
 
nTVLs 
Calculated  
Thickness 
(mm) 
Designed 
thickness 
(mm) 
Manager's Office 1.00E-04 8.27 1.00 6.33E-03 2.20 479.25 430 
Secretary’s Office 1.00E-04 8.60 1.00 6.85E-03 2.16 471.85 430 
Washroom 1.00E-04 8.00 1.00 5.93E-03 2.23 485.54 430 
Control Room  1.00E-04 12.50 1.00 1.45E-02 1.84 401.03 430 
Check In Office 1.00E-04 13.20 1.00 1.61E-02 1.79 390.72 430 
Reception 2.00E-05 14.69 1.00 4.00E-03 2.40 398.94 430 
Waiting Area 2.00E-05 16.75 0.25 2.08E-02 1.68 366.74 430 
Marshall Post 1.00E-04 11.85 1.00 1.30E-02 1.89 411.15 430 
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The average secondary shielding employed by the structural shielding engineers 
against leakage gamma radiation was 430 mm of concrete. The result for the 
calculated leakage barrier thickness for the gamma-ray scanner based on the current 
operational data is tabulated in Table 4.2. 
From the table, the Manager’s Office requires a concrete shield thickness of 479.25 
mm against the leakage gamma radiation. This is against the 430 mm of concrete 
shield installed by the shielding engineers. There is therefore a shielding deficit of 
49.25 mm of concrete. In order for the occupant of the office to receive adequate 
protection from leakage radiation, 49.25 mm of concrete needs to be added to the 
current barrier thickness. 
 
The Secretary’s office requires a shielding of thickness 471.85 mm for him to be 
adequately protected against the leakage radiation. Comparing this value with the 
installed shield shows a deficit of 41.85 mm, which must be added to the current 
thickness in order that the secretary is well protected. 
 
To effectively shield the washroom from any leakage radiation from the scanner 
shielding head requires a wall thickness of was 485.54 mm. Comparing this thickness 
with the concrete shield employed during the construction of the facility shows a 
shielding difference of 55.54 mm which must be added to 430 mm employed during 
the construction of the facility. 
 
The control room is at a distance of 12.50 m from the gamma ray scanner source axis 
and would require a leakage shielding thickness of 401.03 mm. Comparing this with 
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shielding employed the expert during the construction of the facility shows an over 
shielding thus giving the occupants of the control room an added protection. 
 
The calculated thickness expected to adequately shield the check-in office against the 
leakage gamma radiation is 390.72 mm. There is an excess shielding of 39.28 mm, an 
indication that the occupants of check-in office are well protected against leakage 
radiation from the source. 
The reception is at distance of 14.69 m from the gamma ray source-axis requires a 
concrete shielding barrier thickness of 398.94 mm. The installed shielding thickness is 
430 mm, showing an extra concrete shield thickness of 31.06 mm. 
 
The corridor which is 12.70 mm from the gamma ray source-axis requires a concrete 
shield of 419.15 mm. Comparing this with the designed thickness shows that occupant 
of this area are adequately protected from the leakage radiation. This reason for this 
might be due to the fact that the designed shield needs to protect both far and near 
occupants. 
 
According to the leakage shielding calculations against the gamma ray, the marshal 
post requires an effective concrete thickness of 411.15 mm. Comparing this with that 
employed by the shielding experts shows an excess shielding of 18.85 mm concrete. 
This means that for a worker occupying the office is adequately protected from the 
leakage radiation from the source a shielding, therefore does not require any 
additional concrete to be added to the current shield. 
 
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The waiting area is a public access area where the customers or members of the public 
who come to the facility sit. A member of the public located at this area who is at a 
distance of 16.75 m requires a concrete shielding thickness of 366.74 mm in order to 
receive adequate protection against the leakage radiation. The average secondary 
barrier concrete thickness employed by the structural shielding engineers was 430 
mm.  
 
The average concrete shielding thickness required to adequately protect all the various 
points of interest likely to be affected by the leakage radiation based on the current 
operational data was calculated to be 428.70 mm. This value is approximately equal 
to the secondary leakage barrier shield thickness of 430 mm installed during the 
construction of the facility. Therefore, the shielding at its current state provides the 
needed protection against leakage radiation from the source head. 
 
 
 
 
 
 
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4.3 Scattered Barrier Thickness for the Co-60 Gamma-ray Scanner 
Table 4. 3: Scatter Barrier thickness compared with designed specifications. 
      
Thickness 
POI P d/m T Bs nTVLs Calculated Designed 
 (Sv/hr)     (mm) (mm) 
Manager's Office 1.00E-04 8.27 1.00 6.96E-03 2.16 325.78 430 
Secretary’s Office 1.00E-04 8.60 1.00 7.53E-03 2.12 320.64 430 
Washroom 1.00E-04 8.00 1.00 6.51E-03 2.19 330.13 430 
Control Room  1.00E-04 12.50 1.00 1.59E-02 1.80 271.60 430 
Check In Office 1.00E-04 13.20 1.00 1.77E-02 1.75 264.45 430 
Reception 2.00E-05 14.69 1.00 4.39E-03 2.36 355.97 430 
Waiting Area 2.00E-05 16.75 0.25 2.28E-02 1.64 247.84 430 
Marshall Post 1.00E-04 11.85 1.00 1.43E-02 1.85 278.60 430 
Corridor 2.00E-05 12.70 0.25 1.31E-02 1.88 284.15 430 
 
In order that the Manager's Office would be adequately protected against scattered 
gamma radiation from the container, the calculated concrete shielding thickness of 
325.78 mm is required. However, the average thickness employed as secondary 
barrier shielding to contain both leakage and scattered radiation is 430 mm, which 
shows an extra concrete shield of 104.22 mm. Based on this, the manager can be said 
to be well protected from the scattered radiation from the container. 
The secretary’s office needs a concrete shield of 320.64 mm for effective protection 
against radiation beam from the scatterer (container) which is at a distance of 8.60 m 
away. Comparing this calculated shield with that installed for this purpose shows an 
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extra shielding thickness of l09.36 mm concrete which gives an added protection to 
secretary. 
From the control room, a concrete of thickness 271.60 mm is require for an effective 
protection against the gamma  ray radiation scattered by the container. Here, there is a 
difference of extra 158.40 mm compared with the installed shield, indicating an over 
shielding for a worker occupying the control room to receive the needed protection 
against scattered radiation. This can be explained by the fact that the shielding barrier 
was for both the leakage and scatter radiation. 
The calculated shield thickness to protect the washroom against the gamma radiation 
scattered by the container was 330.13 mm. Therefore, any person vising the 
washroom during beam-on state receives the needed protection against scattered 
radiation. 
 
From the container to the check-in office is 2.25 m and requires a barrier thickness of 
264.45 mm to adequately protect its occupant from the scattered beam from the 
container. Comparing with the employed concrete of 430 mm shows a difference of 
165.55 mm. This is due to the fact that leakage radiation is more penetrating than 
scattered radiation and that the structural shielding engineers might have considered 
leakage as more important than the scattered radiation in constituting the average 
secondary barrier shielding, since this shield is to protect the occupants from both 
leakage and scattered radiation. 
The scattered barrier thickness required to adequately protect the receptionist against 
the scattered radiation by the container was calculated as 355.97 mm concrete. This 
shows a shielding difference of 74.03 mm compared with that of 430 mm employed 
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by the shielding engineers, giving the receptionist an added protection against the 
scattered radiation. 
The waiting room was located at a distance of 16.75 m from the scatterer. The public 
in this area requires a concrete shielding thickness of 330.13 mm for them to receive 
adequate protection from the gamma radiation scattered by the container. Therefore 
additional shield of 99.87 mm was constituted when the shielding experts constructed 
the facility. The 330.13 mm is purposely meant to contain only the scattered radiation 
but in installing the secondary barrier shielding, consideration was given to leakage as 
well as scattered radiation and this perhaps explains why an average secondary barrier 
shielding of 430 mm was put in place to cater for both radiations since leakage is 
more important than scattered radiation and a thicker shield must be preferred to a 
smaller one in such a circumstance. 
 
4.4. Dose rates for the secondary barrier thickness of the gamma ray scanner. 
Table 4. 4: Calculated dose rates for the secondary barrier thickness for the gamma 
ray scanner. 
POI d/m IDRs 
(µSvh-1) 
IDRl 
(µSvh-1) 
IDRT 
(µSvh-1) 
TADRh  
(µSvh-1) 
Manager's Office 8.27 5.44E-05 4.48E-04 5.03E-04 2.62E-06 
Secretary’s Office 8.60 5.03E-05 4.15E-04 4.65E-04 2.42E-06 
Washroom 8.00 5.82E-5 4.79E-04 5.37E-04 2.80E-06 
Control Room  12.50 2.38E-05 1.96E-04 2.20E-04 1.15E-06 
Check In Office 13.20 2.14E-05 1.76E-04 1.97E-04 1.03E-06 
Reception 14.69 1.72E-05 1.42E-04 1.59E-04 8.30E-07 
Waiting Area 16.75 1.33E-05 1.09E-04 1.23E-04 6.38E-07 
Marshall Post 11.85 2.65E-05 2.18E-04 2.45E-04 1.28E-06 
Corridor 12.70 2.31E-05 1.90E-04 2.13E-04 1.11E-06 
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Based on the concrete shielding design thickness of 430 mm, the calculated dose rate 
at the manager’s office was 2.62 µSvh-1. When compared with the recommended limit 
of 20 µSvh-1 as specified in NCRP (2005) shows that the secondary shielding is 
adequate to prevent the manager from unnecessary dose from the radiation beam. 
 
From the calculated dose rate for the position of the secretary’s office, it can be stated 
that the secretary received far less dose from leakage and scattered beam than the 
stipulated minimum dose. 
The dose rate outside the secondary shielding to the control room based on the 430 
mm concrete shield thickness of the secondary barrier was calculated to be                        
1.15 µSvh-1, which falls below the 20 µSvh-1 as specified by NCRP (2005).  
This means that the occupant of this room does not need any additional shielding to 
be well protected. 
Based on the concrete shielding design thickness of 430 mm, the calculated dose rate 
at the reception was 0.83µSvh-1. When compared with the recommended limit of 20 
µSvh-1 as specified in NCRP (2005) shows that the secondary shielding is adequate to 
prevent the occupants of the reception from unnecessary dose from the radiation 
beam. 
The waiting area receives a dose  of 0.63 µSvh-1.  
This is below the maximum recommended limit of 20 µSvh-1 as specified in NCRP 
(2005) shows that the secondary shielding is adequate to prevent the persons found at 
this area during the beam-on state from unnecessary dose from the radiation beam. 
 
 
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4.5. Comparison between measured and calculated dose rate for the gamma-ray 
scanner. 
Table 4. 5: Comparison between measured and calculated dose rates for the gamma-
ray scanner. 
POI d/m Calculated Measured Deviation 
Manager's Office 8.27 2.62 1.77 0.85 
Secretary’s Office 8.60 2.42 1.34 1.08 
Washroom 8.00 2.80 2.48 0.32 
Control Room  12.50 1.15 1.00 0.15 
Check In Office 13.20 1.03 0.93 0.10 
Reception 14.69 0.83 0.72 0.11 
Waiting Area 16.75 0.63 0.61 0.02 
Marshall Post 11.85 1.28 0.78 0.50 
Corridor 12.70 1.11 1.00 0.11 
 
From Table 4.5, it can be seen that there is a small deviation between each pair of 
values, nonetheless, they both fell below the limit of 20 µSvh-1 as specified in NRCP 
151, 2005.  
The secondary barrier has not deteriorated much enough to pose any radiation hazard 
to staff and the general public. 
The calculated dose rate for the primary barrier was 0.62 µSvh-1  whilst that recorded 
by the dose rate meter was 0.93 µSvh-1. These two values are all less than the 
specified maximum limit of 20 µSvh-1, therefore the primary barrier still provides the 
necessary protection for the public against the gamma ray radiation from the cobalt -
60 source. 
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4.6. Barrier thickness calculation for X-ray of energy 6MeV 
Table 4. 6:  Calculated primary barrier thickness for X-ray of energy 6 MeV. 
 
POI 
   P 
(Sv/hr) 
 
d/m 
 
U 
 
T 
 
Be 
 
nTVLs 
Calculated 
thickness 
(mm) 
Designed 
Thickness 
(mm) 
         
Behind 
the 
primary 
barrier 
 
2.0E-5 
 
15.8 
 
0.06 
 
0.25 
 
3.08E-2 
 
1.51 
 
421.29 
 
500.00 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
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4.7. Scatter and leakage radiation barrier thickness barrier for 6MeV X-ray 
source. 
 
Table 4. 7: Calculated scattered radiation barrier thickness barrier for 6MeV X-ray 
source. 
 
            Thickness 
POI P  d/m T Bs nTVLs Calculated Designed  
  (Sv/hr)         (mm) (mm) 
Manager's Office 1.00E-04 8.27 1.00 6.96E-03 2.16 368.92 430 
Secretary’s Office 1.00E-04 8.60 1.00 7.53E-03 2.12 363.11 430 
Washroom 1.00E-04 8.00 1.00 6.51E-03 2.19 373.85 430 
Control Room  1.00E-04 12.50 1.00 1.59E-02 1.80 307.57 430 
Check In Office 1.00E-04 13.20 1.00 1.77E-02 1.75 299.48 430 
Reception 2.00E-05 14.69 1.00 4.39E-03 2.36 403.11 430 
Waiting Area 2.00E-05 16.75 0.25 2.28E-02 1.64 280.67 430 
Marshall Post 1.00E-04 11.85 1.00 1.43E-02 1.85 315.50 430 
Corridor 2.00E-05 12.70 0.25 1.31E-02 1.88 321.78 430 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
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Table 4. 8:  Calculated leakage radiation barrier thickness barrier for 6MeV X-ray 
source. 
 
 
The values in the Tables 4.6, 4.7 and 4.8 are calculated for 6MeV X-ray using the 
current work load of1080Gy/wk and other current operational data for the selected 
point of interest around the facility. The results show that even though, the primary 
shield can provide the necessary protection for the public, the existing wall thickness 
for the secondary radiation attenuation is not enough to give the needed protection for 
the main occupant of the facility. 
The average secondary barrier required is 672.69 mm (ie the average of the averages 
of the leakage and scattered values), and the installed barrier thickness is 430 mm. 
This means that an additional concrete of 242 mm thickness is required to enable the 
current shield to provide the needed protection against radiation from an X-ray of 
energy of at the least 6MeV.  
 
POI 
   P 
(Sv/hr) 
 
d/m 
 
T 
 
Bl 
 
nTVLs 
Calculated 
Thickness  
(mm) 
Designed 
Thickness 
(mm) 
Manager's Office 1.00E-04 8.27 1.00 4.75E-03 2.32 648.22 430   
Secretary’s Office 1.00E-04 8.60 1.00 5.14E-03 2.29 638.73 430   
Washroom 1.00E-04 8.00 1.00 4.44E-03 2.35 656.26 430   
Control Room  1.00E-04 12.50 1.00 1.09E-02 1.96 548.11 430   
Check In Office 1.00E-04 13.20 1.00 1.21E-02 1.92 534.90 430   
Reception 2.00E-05 14.69 1.00 3.00E-03 2.52 704.00 430   
Waiting Area 2.00E-05 16.75 0.25 8.57E-04 3.07 855.66 430   
Marshall Post 1.00E-04 11.85 1.00 9.75E-03 2.01 561.05 430   
Corridor 2.00E-05 12.70 0.25 5.60E-04 3.25 907.25 430   
          
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Therefore, the shielding at its current state cannot be used house X –ray energy 
6MeV. 
 
 
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  4.9. Comparison of commissioning shielding requirement with the shielding requirement at the present state. 
  Table 4. 9: Calculated values leakage barrier thickness for the Gamma ray scanner at commissioning stage. 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
POI 
 
P 
(Sv/hr) 
 
d/m 
 
T 
 
Bl 
 
nTVLs 
Calculated 
thickness 
(mm) 
Designed  
thickness  
(mm) 
Manager's Office 1.00E-04 8.27 1.00 1.90E-02 1.72 375.24 430 
Secretary’s Office 1.00E-04 8.60 1.00 2.05E-02 1.69 367.83 430 
Washroom 1.00E-04 8.00 1.00 1.78E-02 1.75 381.53 430 
Control Room  1.00E-04 12.50 1.00 4.34E-02 1.36 297.02 430 
Check In Office 1.00E-04 13.20 1.00 4.84E-02 1.32 286.70 430 
Reception 2.00E-05 14.69 1.00 1.20E-02 1.92 418.83 430 
Waiting Area 2.00E-05 16.75 0.25 6.23E-02 1.21 262.73 430 
Marshall Post 1.00E-04 11.85 1.00 3.90E-02 1.41 307.13 430 
Corridor 2.00E-05 12.70 0.25 3.58E-02 1.45 315.14 430 
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Table 4. 10: Calculated values for scattered barrier thickness for the Gamma ray scanner at commissioning stage 
 
 
 
 
        POI  
P 
(Sv/hr) 
 
d/m 
 
T 
 
Bs 
 
nTVLS 
Calculated 
Thickness 
(mm) 
designed 
Thickness 
(mm) 
Manager's Office 1.00E-04 8.27 1.00 2.09E-02 1.68 253.73 430   
Secretary’s Office 1.00E-04 8.60 1.00 2.26E-02 1.65 248.60 430   
Washroom 1.00E-04 8.00 1.00 1.95E-02 1.71 258.08 430   
Control Room  1.00E-04 12.50 1.00 4.77E-02 1.32 199.55 430   
Check In Office 1.00E-04 13.20 1.00 5.32E-02 1.27 192.40 430   
Reception 2.00E-05 14.69 1.00 1.32E-02 1.88 283.92 430   
Waiting Area 2.00E-05 16.75 0.25 6.85E-02 1.16 175.80 430   
Marshall Post 1.00E-04 11.85 1.00 4.29E-02 1.37 206.55 430   
Corridor 2.00E-05 12.70 0.25 3.94E-02 1.40 212.10 430   
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The facility started operation in 2010 with the average workload of 360Gy/week, ie 
20 scans per day.  
From the Tables 4.9 and 4.10 the barrier thickness needed to adequately protect the 
workers and the general public is approximately 290 mm. But the structural engineers 
have installed a thickness of 430 mm, an excess concrete of 140 mm. It can be seen 
that there was an over-shielding of the source during its early stage of operation. 
However, in about six years from commissioning, the workload has increased by 
almost three-fold, (from 360Gy/wk to 1080Gy/wk). This requires that the half-value 
layer of the concrete be increased in order to provide maximum protection of people 
around the facility. 
 
Table 4. 11: Calculated dose rate at commissioning stage. 
POI d/m IDRs 
µSv/hr 
IDRl 
µSv/hr 
IDRT 
µSv/hr 
TADRh 
µSv/hr 
Manager's Office 8.27 1.81E-05 4.48E-04 4.67E-04 2.43E-06 
Secretary’s Office 8.60 1.68E-05 4.15E-04 4.31E-04 2.25E-06 
Washroom 8.00 1.94E-05 4.79E-04 4.99E-04 2.60E-06 
Control Room  12.50 7.94E-06 1.96E-04 2.04E-04 1.06E-06 
Check In Office 13.20 7.12E-06 1.76E-04 1.83E-04 9.54E-07 
Reception 14.69 5.75E-06 1.42E-04 1.48E-04 7.70E-07 
Waiting Area 16.75 4.42E-06 1.09E-04 1.14E-04 5.92E-07 
Marshall Post 11.85 8.83E-06 2.18E-04 2.27E-04 1.18E-06 
Corridor 12.70 7.69E-06 1.90E-04 1.98E-04 1.03E-06 
 
 
 
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Table 4. 12: Comparison of calculated dose rate at commissioning stage with the 
calculated dose rate the current stage. 
POI 
Commissioning 
stage (µSv/hr) 
Current state 
(µSv/hr) 
Manager's Office 2.43 2.62 
Secretary’s Office 2.25 2.42 
Washroom 2.60 2.80 
Control Room  1.06 1.15 
Check In Office 0.95 1.03 
Reception 0.77 0.83 
Waiting Area 0.59 0.64 
Marshall Post 1.18 1.28 
Corridor 1.03 1.11 
 
Table 4.12 compares the dose rate at commissioning stage and the current state. 
Comparing each pair of entries, it can be seen that there has been increment in dose 
rate over the past six years. This is mainly due to increase in the workload as a result 
of increase in the average number of scan per day. 
However, these increases in dose rates to the individuals still fall below the limit 
20µSv/hr. (NCRP 151, 2005), but due to the cumulative nature of radiation doses, it is 
advisable that management of the facility takes steps to optimise dose to staff as well 
as the general public by  keeping an optimal scan per day. 
 
 
 
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CHAPTER FIVE 
5.1 Conclusions. 
The assessment of the shielding integrity was carried out using commissioning data 
and the current operational data. Dose rate were measured beyond the primary and 
secondary barriers. Both the average measured and calculated dose rate for primary 
barrier were below the 0.02 mSvhr-1 limit specified by NRCP 151, 2005. This shows 
that the shielding is adequate for the protection of the workers. 
The secondary shielding is also enough for the persons at the premises of the Co-60 
scanner, since the measured dose rates were below the specified limit of 0.02 mSvhr-1 
It was however, observed that due to increase in workload over the past six years, 
there has been increase in dose to the staff and the general public.   
The assessment done on the install shielding using the current operational data 
indicates that the shielding is inadequate in providing protection for the general public 
and the workers against X-ray radiation source of energy 6MeV. 
 
5.2 Recommendation 
As the result of the research carried on the facility, the following recommendations 
are made: 
To Management of the NICK scan facility, 
 Optimise dose by reducing the weekly workload by keeping an optimal 
number of scans per week, preferably 800Gy/wk. 
 Ensure that door to the tunnel is always closed during scanning time. 
 Provision of local rules and procedures at the workplace to be followed by all 
staff and it should be supervised by a qualified Radiation Safety Officer (RSO) 
at the facility.  
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To regulatory authority –Nuclear Regulatory Authority, Ghana. 
 Encourage all practices no matter the level of exposure to develop an ALARA 
culture for the facility. 
 Ensure regular shielding integrity checks on all practices within the country, at 
least once in every two years.  
 Enforce the need for the facility to conduct random checks in aspects of 
workplace monitoring. 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
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APPENDICES 
Appendix A 
 
Table A. 1: Tenth Value Layers (TVL) for Co-60 and some selected Energies for 
concrete of density 2350kgm-3 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
Radiation Source/ 
Energy 
 
Co-60 
 
                             X-ray 
1.25 MeV 6MeV 10MeV 15 MeV 
Primary gamma ray / 
x - ray 
218 343 389 432 
Leakage gamma/ 
x-ray 
218 279 305 330 
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Appendix B  
 
Table B. 1: First TVL (mm concrete) for Container scatter radiation at some selected 
scatter angles. 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
 
Scattered angle /o 
 
Co-60 
 
                             X-ray 
1.25 MeV 6MeV 10MeV 15 MeV 
15 218 343 389 432 
30 213 279 305 330 
45 197 229 23 237 
60 189 205 209 299 
90 151 171 173 174 
135 128 144 144 145 
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Appendix C 
Table C. 1: Summary of recommended/legal effective dose limits and designed dose 
limits 
Dose limit IAEA USA United Kingdom 
Occupational 
exposure dose 
limit 
20 mSv per year 
averaged over 5 
consecutive years 
and 50 mSv in any 
single year 
Implied annual limit 
of 10 mSv,  
cumulative 
dose of age × 10 mSv, 
and 50 mSv in any 
single year  
20 mSv in a year or 
100 mSv in 5 consecutive 
years and 50 mSv in any 
single year 
Design limit for 
occupational 
exposure 
 
 Fraction of 10 mSv 
annually. 
6 mSv in a year  
IDR is 7.5 μSv·h–1 
Public dose limit  
 
 
1 mSv in a year Infrequently, 5 mSv 
annually, and 
continually 
1 mSv 
annually 
Design limit for 
public area 
 1 mSv annually 
20 μSv in any hour  
0.3 mSv in a year  
IDR is <7.5 μSv·h–1  
TADR is <0.5 μSv·h–1 
TADR2000 <0.15 μSv·h–1 
 
 
 
 
 
 
 
 
 
 
 
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Appendix D 
Table D. 1: Scatter fraction of dose α, at 1 m, for 400 cm2 incidence beam. 
 
 
 
 
 
 
 
 
 
 
 
 
 
Angle/o 
 
Co-60 
6 MeV 10 MeV 18 MeV 24MeV 
Max a α at 1.5cm Max α  α at 2.5 cm Max α  αat 2.5 cm Maxα α at 2.5 cm 
10 1.1E–2 1.68E–2 1.04E–2 1.69E-2 1.66E–2 2.43E-2 1.42 E–2 2.74E–2 1.78E–2 
20 8.0E–2 1.15E–2 6.73E –3 1.03E–2  5.79E–3 1.172E-2 5.39E–3 1.27E–2 6.32E–3 
30 6.0E–3 5.36E–3 2.77E-3 6.73E–3  3.18E–3 7.13E–3 2.53E–3 7.21E–3 2.74E–3 
45 3.7E–3 2.97E–3 1.39 E–3 3.25E–3  1.35E–3 3.05E–3 8.64E-4 3.06E–3 8.30E–4 
60 2.2E–3 1.74E–3 8.24E–4 1.84E–3  7.46E–4 1.42E–3 4.24E–4 1.37E–3 3.86E–4 
90 9.1E–4 7.27E–4 4.26E–4 7.14E–4 3.81E–4 3.75E–4 1.89E–4 3.53E–4 1.74E–4 
135 5.4E–4  4.88E–4 3.00E–4 3.70E–4 3.02E–4 2.59E–4 1.24E–4 2.33E–4 1.20E–4 
150 1.5E–4 3.28E-4 3.16E–4 2.87E-4  2.74E–4 2.26E–4 1.20E-4 2.12E–4 1.13E–4 
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Appendix E 
Table E. 1: Occupancy  factor for some selected areas 
Type of area NCRP   IPEM 
Offices, Reception areas, Shops, Control room 
Children’s play areas, Staff rooms 
 
1 
 
1 
Corridors 1/4 0.2 
Toilets, bathrooms, Outside areas with seating 1/16 0.1 
Stairways, unattended waiting rooms,  1/16 0.05 
 
 
 
 
 
 
 
 
 
 
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