iA CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR

IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR

FOR BORON NEUTRON CAPTURE THERAPY

Undergraduate Thesis

In Partial Fulfilment of the Requirements for the Degree of

Bachelor of Engineering in Nuclear Engineering

submitted by

NINA FAUZIAH

09/289119/TK/36010

presented to

DEPARTMENT OF PHYSICS ENGINEERING

FACULTY OF ENGINEERING

UNIVERSITAS GADJAH MADA

YOGYAKARTA

2013

ii

INTELLECTUAL PROPERTY STATEMENT

I, whom mentioned as follows:

Name : Nina Fauziah

NIM : 09/289119/TK/36010

Title of Thesis : A Conceptual Design of Neutron Collimator in the

Thermal Column of Kartini Research Reactor for Boron

Neutron Capture Therapy

certify that the thesis titled as mentioned above is my own original work in

accordance with the academic norms, and no portion of my thesis has been

copyrighted previously unless properly referenced.

If there is a breach, I will take full responsibility for any legal action that might be

caused.

Yogyakarta, July 22, 2013,

Who certifies the statement,

Nina Fauziah

NIM. 09/289119/TK/36010

iii

APPROVAL FORM

UNDERGRADUATE THESIS

A CONCEPTUAL DESIGN OF NEUTRON COLLIMATOR IN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR

FOR BORON NEUTRON CAPTURE THERAPY

by

Nina Fauziah09/289119/TK/36010

defended in front of the Board of Examiners on July 12, 2013

Board of Examiners

Chairman, Secretary,

Dr. Ir. Andang Widi Harto, M. T. Ir. Anung Muharini, M. T.

NIP. 196603041994031003 NIP. 196908011994122001

Chief Examiner, Co-Examiner,

Ir. Mondjo, M. Si. Prof. Ir. Yohannes Sardjono

NIP. 195308131981031001 NIP. 195906101981031002

Approved and certified to fulfill the requirements for graduation on July12,2013

Chairman of Department of Physics Engineering

Faculty of Engineering UGM

Prof. Ir. Sunarno, M. Eng., Ph. D.

NIP. 195511241983031001

iv

MINISTRY OF EDUCATION AND CULTUREUNIVERSITAS GADJAH MADAFACULTY OF ENGINEERING

DEPARTMENT OF PHYSICS ENGINEERING

FINAL PROJECT

Name : Nina Fauziah

NIM : 09/289119/TK/36010

Title of Thesis : A Conceptual Design of Neutron Collimator in the Thermal Column of Kartini Research Reactor for Boron NeutronCapture Therapy

Supervisor : Dr. Ir. Andang Widi Harto, M. T.

Co-Supervisor : Prof. Ir. Yohannes Sardjono

Problem : Boron Neutron Capture Therapy (BNCT) is a type of tumour therapy that uses neutron beam as radiation beam. A good therapy should destroy the tumour cells thoroughly without any significant side effect to the surrounding normal cells. For this reason, the IAEA recommends some criteria of the neutron beam used for BNCT purpose. Thus, a certain conceptual design of neutron collimator has to be made to fulfill the criteria.

Supervisor, Co-Supervisor,

Dr. Ir. Andang Widi Harto, M. T. Prof. Ir. Yohannes Sardjono

NIP. 196603041994031003 NIP. 195906101981031002

Chairman of Department of Physics Engineering

Faculty of Engineering UGM

Prof. Ir. Sunarno, M. Eng., Ph. D.

NIP. 195511241983031001

vDEDICATION

To my beloved parents, my mother Emma Siti Rochmah and my father Achmad

Damanhuri, for their religious guidance and affectionate care showed to me.To my

beloved elder sisters, Farida Apriyani, Dewi Damayanti, Nunung Nurul Falah, and

Fitrie Amelia, and my beloved elder brother, Guruh Agung Setiawan, for their

motivations and encouragements given to me.

vi

Indeed, within the heavens and earth are signs for the believers.

Al-Jathiyah (45) : 3

vii

ACKNOWLEDGEMENT

First and foremost, praises and thanks to the Allah S.W.T., the Almighty, for

the showers of blessings throughout my research work to complete the study

successfully. This thesis was produced with the assistance and guidance of the

following people to whom I would like to express my sincere gratitude.

1. My research advisors, Dr. Ir. Andang Widi Harto, M. T. and Prof. Ir.

Yohannes Sardjono, for giving me the opportunity to do research and

providing invaluable guidance throughout this research,

2. My examiners, Ir. Mondjo, M. Si. and Ir. Anung Muharini, M. T., for giving

me deeper lessons and understandings from the questions posed and the exact

answers told during the viva voce,

3. The Chairman of Department of Physics Engineering, UGM, Prof. Ir.

Sunarno, M. Eng., Ph. D.,

4. The Head of Pusat Teknologi Akselerator dan Proses Bahan Badan Tenaga

Nuklir Nasional (PTAPB-BATAN) Yogyakarta, Dr. Ir. Widi Setiawan, for

giving me the chance to do this final project work at BATAN,

5. The Head of Academic Affairs of Department of Physics Engineering, UGM,

Ferdiansjah, S. T., M. Eng. Sc., for the advices given to me in writing in

English,

6. All lecturers at Department of Physics Engineering, UGM, for all knowledge

shared,

7. All staffs of Department of Physics Engineering, UGM, for the kindness,

8. My best friends Anti, Dian, Dewa, and Dita, for all precious experiences we

have, and also for the supports given to me,

9. The greatest talented young poet I have ever met, Eckart Sulaksono, for

every-single-word in his poets which were very enjoyable even though I did

not understand it whatsoever,

10. My dear friends Manda, Sekar, Oksel, Desti, Sukma, Imel, Una, Laret,

Tukah, Rima, Indah, Vika, Binar, Dintan, Feni, Lina, Umi, Khusnul, Farkhad,

viii

Afwan, Aji, Nico, Ego, Ilham, Didik, Cecep, Gagad, Handoyo, Andik, Alief,

Irfan, Rizal, Baghir, Ario, Dio, Kamal, Helmi, Kamal, and all students of

Department of Physics Engineering, UGM, batch of 2009, for all

unforgettable togetherness,

11. All staff of Keluarga Mahasiswa Teknik Fisika, UGM, for the opportunity

given to me for being a part of them, and

12. Nourish, Helmi, Bemby, and Debi, for being ridiculous clowns in my sorrow.

Finally, my thanks go to all the people who have supported me to complete the final

project directly or indirectly.

Yogyakarta, July 22, 2013,

Writer

ix

TABLE OF CONTENTS

TITLE ............................................................................................................. i

INTELLECTUAL PROPERTY STATEMENT ................................................ ii

APPROVAL FORM ........................................................................................... iii

PROJECT FORM.............................................................................................. iv

DEDICATION ..................................................................................................... v

QUOTE ........................................................................................................... vi

ACKNOWLEDGEMENT ................................................................................. vii

TABLE OF CONTENTS.................................................................................... ix

LIST OF TABLES............................................................................................. xii

LIST OF FIGURES ......................................................................................... xiii

SYMBOLS AND ABBREVIATIONS.............................................................. xiv

ABSTRACT ...................................................................................................... xvi

INTISARI.. ....................................................................................................... xvii

I. INTRODUCTION ...................................................................................... 1

I.1. Background ......................................................................................... 1

I.2. Scope and Limitation........................................................................... 3

I.3. Objective ............................................................................................. 3

I.4. Advantages.......................................................................................... 4

II. LITERATURE REVIEW........................................................................... 5

II.1. Desired Neutron Beam Parameters ...................................................... 5

II.1.1. Epithermal Beam Intensity..................................................... 5

II.1.2. Incident Beam Quality ........................................................... 5

II.2. Neutron Source for BNCT ................................................................... 6

III. THEORETICAL BACKGROUND ........................................................... 9

III.1. Radiation Interactions with Matter ....................................................... 9

xIII.1.1. Neutron Interactions............................................................... 9

III.1.2. Gamma-ray Interactions....................................................... 12

III.2. The Monte Carlo Method and MCNP Program .................................. 14

III.2.1. Weight ................................................................................. 15

III.2.2. Particle Tracks ..................................................................... 16

III.2.3. Neutron Interactions............................................................. 16

III.2.4. Photon Interactions .............................................................. 17

IV. MATERIALS AND METHOD ................................................................ 18

IV.1. Materials ........................................................................................... 18

IV.2. Method of Study ................................................................................ 18

IV.2.1. Kartini Research Reactor Modelling .................................... 18

IV.2.2. Neutrons and Gamma Rays Recording ................................. 20

IV.2.3. Tally Selecting ..................................................................... 21

IV.2.4. Beam Criteria....................................................................... 26

IV.2.5. Collimator Conceptual Designing ........................................ 27

IV.3. Results Analysis ................................................................................ 30

V. RESULTS AND ANALYSIS.................................................................... 31

V.1. Reactor Criticality.............................................................................. 31

V.2. Collimator Conceptual Design ........................................................... 31

V.2.1. Collimator Wall ................................................................... 31

V.2.2. Moderator ............................................................................ 33

V.2.3. Filter .................................................................................... 36

V.2.4. Gamma-ray Shielding .......................................................... 38

V.2.5. Aperture............................................................................... 39

V.2.6. Environment Surrounding the Collimator............................. 40

VI. CONCLUSION AND RECOMMENDATION........................................ 42

VI.1. Conclusion ........................................................................................ 42

VI.2. Recommendation............................................................................... 43

xi

REFERENCE .................................................................................................... 45

APPENDICES.................................................................................................... 47

A. AN EXAMPLE OF MCNP5 INPUT CODES ......................................... 48

B. FIGURES OF REACTOR AND COLLIMATOR MODELS ................ 62

B.1. Reactor core model (top section)........................................................ 63

B.2. Reactor core model (side section). ..................................................... 64

B.3. Reactor core and collimator model (top section). ............................... 65

C. MEAN FREE PATH CALCULATIONS................................................. 66

xii

LIST OF TABLES

Table 1.1. Energies of the particles coming from neutron capture in 10B............... 2

Table 4.1. MCNP tally types .............................................................................. 21

Table 4.2. Beam parameters ............................................................................... 22

Table 4.3. Kerma coefficients for fast neutrons .................................................. 24

Table 4.4. Kerma coefficients for photons .......................................................... 26

Table 4.5. Beam criteria ..................................................................................... 27

Table 5.1. Comparison of moderator materials ................................................... 34

Table 5.2. Results of moderator (Al) thickness variations ................................... 35

Table 5.3. Results of -ray shielding (Bi) thickness variations ............................ 38

Table 5.4. Results of beam characteristics for different aperture diameter........... 40

Table 5.5 Results of beam characteristics for different aperture diameter of

graphite-surrounded collimator .......................................................... 40

xiii

LIST OF FIGURES

Figure 3.1. Random history of a neutron incident on a fissionable material slab ... 14

Figure 4.1. Core configuration ............................................................................. 19

Figure 5.1. Epithermal neutron flux for various thickness of wall (Ni) ................. 32

Figure 5.2. Scattering cross sections of 58Ni ......................................................... 33

Figure 5.3. Fast neutron component for various thickness of moderator (Al) ........ 35

Figure 5.4. Fast neutron component for various thickness of filter (60Ni) ............. 36

Figure 5.5. Thermal neutron component for various thickness of filter (60Ni)....... 36

Figure 5.6. Absorption cross sections of 60Ni ....................................................... 37

Figure 5.7. Gamma-ray component for various thickness of shielding (Bi)........... 38

Figure 5.8. Total cross sections of Bi ................................................................... 39

Figure 6.1. Collimator configuration .................................................................... 42

Figure 6.2. Collimator shielding configuration ..................................................... 43

xiv

SYMBOLS AND ABBREVIATIONS

Symbols

Symbol Quantity Unit

X Thickness cm

n Number of particle n

v Speed cm.s-1

A Area cm2

J Current n.cm-3.s-1

I Intensity n.cm-3.s-1

Flux n.cm-3.s-1N Atom density atom.cm-3

Microscopic cross section barn (10-24 cm2) Macroscopic cross section cm-1

Attenuation coefficient cm-1 Mean free path cm Mass density g.cm-3wf Weight fraction

M Atomic weight g.mole-1

Symbol Definition

Alpha Gamma6Li Lithium-66Li2CO3 Lithium (Lithium-6 enriched) carbonate7Li Lithium-710B Boron-1060Ni Nickel-60

Al Aluminum

AlF3 Aluminum fluoride

Al2O3 Aluminum oxide

B4C Boron carbide

Bi Bismuth

C Carbon

Cd Cadmium

xv

Symbol Definition

F Fluorine

H Hydrogen

Ni Nickel

O Oxygen

Pb Plumbum (lead)

PbF2 Lead fluoride

Abbreviations

Abbreviation Meaning

BATAN Badan Tenaga Nuklir Nasional

BNCT Boron Neutron Capture Therapy

IAEA International Atomic Energy Agency

ICRU International Commission onRadiation Units and Measurements

LET Linear Energy Transfer

MCNP Monte Carlo N-Particle

MCNP5 Monte Carlo N-Particle version 5

SAR Safety Analysis Report

xvi

A CONCEPTUAL DESIGN OF NEUTRON COLLIMATORIN THE THERMAL COLUMN OF KARTINI RESEARCH REACTOR

FOR BORON NEUTRON CAPTURE THERAPY

by

Nina Fauziah09/289119/TK/36010

Submitted to the Department of Physics EngineeringFaculty of Engineering Universitas Gadjah Mada on July 12, 2013

in partial fulfilment of the requirements for the Degree ofBachelor of Engineering in Nuclear Engineering

ABSTRACT

Studies were carried out to design a collimator which results in epithermal neutron beam for Boron Neutron Capture Therapy (BNCT) at the Kartini Research Reactor by means of Monte Carlo N-Particle (MCNP) codes. Reactor within 100 kW of thermal power was used as the neutron source. The design criteria were based on recommendation from the International Atomic Energy Agency (IAEA). All materials used were varied in size, according to the value of mean free path for each material. MCNP simulations indicated that by using 5 cm thick of Ni as collimator wall, 60 cm thick of Al as moderator, 15 cm thick of 60Ni as filter, 2 cm thick of Bi as -ray shielding, 3 cm thick of 6Li2CO3-polyethylene as beam delimiter, with 1 to 5 cm varied aperture size, epithermal neutron beam with maximum flux of 7.65 x 108 n.cm-2.s-1 could be produced. The beam has minimum fast neutron and -ray components of, respectively, 1.76 x 10-13 Gy.cm2.n-1 and 1.32 x 10-13 Gy.cm2.n-1, minimum thermal neutron per epithermal neutron ratio of 0.008, and maximum directionality of 0.73. It did not fully pass the IAEAs criteria, since the epithermal neutron flux was below the recommended value, 1.0 x 109 n.cm-2.s-1. Nonetheless, it was still usable with epithermal neutron flux exceeding 5.0 x 108

n.cm-2.s-1. When it was assumed that the graphite inside the thermal column was not discharged but only the part which was going to be replaced by the collimator, the performance of the collimator became better within the positive effect from the surrounding graphite that the beam resulted passed all criteria with epithermal neutron flux up to 1.68 x 109 n.cm-2.s-1.

Keywords: design, collimator, epithermal neutron beam, BNCT, MCNP, criteria

Supervisor : Dr. Ir. Andang Widi Harto, M. T.

Co-supervisor : Prof. Ir. Yohannes Sardjono

xvii

DESAIN KONSEPTUAL KOLIMATOR NETRON DI KOLOM TERMAL REAKTOR RISET KARTINI UNTUK BORON NEUTRON CAPTURE THERAPY

oleh

Nina Fauziah09/289119/TK/36010

Diajukan kepada Jurusan Teknik Fisika Fakultas TeknikUniversitas Gadjah Mada pada tanggal 12 Juli 2013

untuk memenuhi sebagian persyaratan untuk memperoleh derajatsarjana S-1 Program Studi Teknik Nuklir

INTISARI

Telah dilakukan penelitian tentang desain kolimator yang menghasilkan radiasi netron epitermal untuk Boron Neutron Capture Therapy (BNCT) di Reaktor Riset Kartini dengan menggunakan program Monte Carlo N-Particle(MCNP). Reaktor pada daya sebesar 100 kW digunakan sebagai sumber netron. Kriteria desain berdasar pada rekomendasi dari IAEA. Setiap material divariasikan ukurannya berdasarkan mean free path radiasi di dalam material tersebut. Simulasi MCNP menunjukkan bahwa dengan menggunakan 5 cm Ni sebagai dinding kolimator, 60 cm Al sebagai moderator, 15 cm 60Ni sebagai filter, 2 cm Bi sebagai perisai sinar-, 3 cm 6Li2CO3-polietilen sebagai penahan radiasinetron, pada variasi bukaan sebesar 1 sampai 5 cm, dihasilkan fluks netronepitermal maksimum sebesar 7,65 x 108 n.cm-2.s-1. Radiasi netron epitermal tersebut memiliki komponen netron cepat sebesar 1,76 x 10-13 Gy.cm2.n-1, komponen sinar- sebesar1,32 x 10-13 Gy.cm2.n-1, rasio netron termal per netron epitermal sebesar 0,008, dan direksionalitas maksimum sebesar 0,73. Hasil ini masih tidak memenuhi seluruh kriteria IAEA, karena fluks netron epitermal kurang dari 1,0 x 109 n.cm-2.s-1. Meski demikian, radiasi netron epitermal tersebut masih dapat digunakan karena fluksnya melebihi 5,0 x 108 n.cm-2.s-1. Pada saat diasumsikan bahwa bagian kolom termal yang tersisa di luar daerah kolimator tetap berisi grafit seperti semula, hasil keluaran kolimator menjadi lebih baik dengan fluks netron maksimum mencapai 1,68 x 109 n.cm-2.s-1.

Kata kunci: desain, kolimator, radiasi neutron epitermal, BNCT, MCNP, kriteria

Pembimbing Utama : Dr. Ir. Andang Widi Harto, M. T.

Pembimbing Pendamping : Prof. Ir. Yohannes Sardjono

1CHAPTER I

INTRODUCTION

I.1. Background

Cell is the basic structural and functional unit of all living organisms. In a

normal cell, the processes of cell division are controlled meanwhile in a tumour

cell, it no longer responds to the signals which control the growth and the death of

the cell. If the creation of abnormal cells happens rapidly, it is then known as

malignant tumour or cancer. Cancer cells can invade the adjoining parts of the

body and spreads to other organs,disrupting normal activities and causing serius

medical problems or even death.

Cancer is a leading cause of death worldwide and accounted for 7.6 million

deaths (around 13% of all deaths) in 2008. About 70% of all cancer deaths

occurred in low- and middle-income countries. In Indonesia, there were 136 males

and 109 females died of cancer for every 100,000 cancer cases in 2008. Deaths

from cancer worldwide are projected to continue to rise to over 13.1 million in

2030. [1]

These facts lead to a consideration that eradicating the tumour cells as soon

as possible is needed before it spreads to any nearby normal cells. There are

several kinds of treatment to cure the disease or considerably prolong life while

improving the patient's quality of life. Those treatments are, generally, sorted into

3 majors: surgery, radiotherapy, and systemic therapy [2].

Radiotherapy is a common cancer treatment that uses high doses of

radiation to destroy cancer cells and shrink tumours. X-rays, -rays, and charged particles are types of radiation used for cancer treatment. These radiations used in

high level of energy, thus they may cause ionizations in the surrounding normal

cells. Besides, those kinds of beam have been rarely effective since they were

found to have relatively low Linear Energy Transfer (LET) characteristics (53

keV.m-1 or less). [3,4,5]

2Boron Neutron Capture Therapy (BNCT) is another form of radiotherapy. In

BNCT, 10B and its carrier drug are administered to the patient. This carrier will

take these compounds to the location of the tumour cells, where 10B is supposed to

be accumulated. On the next step, the tumour area is to be irradiated by neutron

beam. There are two different neutron beams commonly used in BNCT: thermal

neutron beam for superficial tumours and epithermal neutron beam which may

penetrate to relatively deeper locations (8 cm to 10 cm depths). Theoretically, an

epithermal neutron becomes a thermal neutron when it reaches the tumour cells

after undergoes moderations by materials (especially water) contained in the

humans body along its path. Then, 10B in the tumour cells captures the thermal

neutron, resulting in a prompt nuclear reaction 10B(n,)7Li. The particles coming from the neutron capture by 10B have two possible energies that are reported in

Table 1.1. [6,7]

Table 1.1. Energies of the particles coming from neutron capture in 10B.

94% 6%

1.47 MeV 1.78 MeV7Li 0.84 MeV 1.01 MeV

0.48 MeV - Reference: [7]

Both -particle and the fission fragment 7Li have high LET characteristics(175 keV.m-1 and above) and short path lengths (approximately 4.5 to 10 m), hence the energy deposition is locally limited around the tumour cells. [4,5,8]

In Indonesia nowadays, three research reactors are available, all are operated

by the National Nuclear Energy Agency (BATAN). Those reactors are TRIGA

2000 reactor in Bandung, TRIGA MARK-II reactor in Yogyakarta, and

Multipurpose Research Reactor in Serpong. Of these three reactors exist, only

TRIGA reactors are planned to be added with a facility for BNCT purpose. Any

BNCT facility has not been established yet; feasibility study is still in its process,

indeed. In TRIGA MARK-II type research reactor in Yogyakarta, which has also

3been known as Kartini Research Reactor, the facility for BNCT is going to be

built for an advanced study which uses tumour-injected animals as the object. [6]

Kartini Research Reactor has an operational output thermal power of 100

kW. The thermal column of this reactor is planned to be implanted with a device

which is capable of narrowing the neutron beam, called as collimator. Thermal

column is selected since it is the most flexible part of the reactor which could be

modified. Due to the tendency of epithermal neutron beams usage for BNCT, the

collimator must contains materials needed to produce an epithermal neutron beam

which fulfill some particular characteristics. Thus, a proper collimator has to be

designed so that the output neutron beam reaches criteria recommended by the

International Atomic Energy Agency (IAEA).

I.2. Scope and Limitation

Here are the limitations of the study:

1. Kartini Research Reactor which operates steadily on 100 kW thermal power

is used as the neutron source,

2. The beam criteria are based on the IAEAs recommendations,

3. Simulations are conducted by using Monte Carlo N-Particle version 5

program,

4. Moderator varies in materials and thickness,

5. Wall, filter, and -ray shielding vary in thickness,6. Aperture varies in diameter.

I.3. Objective

The main purpose of this study is to make a conceptual collimator design for

BNCT that can be properly implanted in the thermal column of Kartini Research

Reactor and the output beam produced passes all criteria recommended by the

IAEA.

4I.4. Advantages

The advantages which may be gained as the implication of this study are:

1. Finding the design of the BNCT purpose-collimator which is proper to be

implanted in the thermal column of Kartini Research Reactor,

2. Being a reference for the next experiment about collimator design for

producing an epithermal neutron beam.

5CHAPTER II

LITERATURE REVIEWS

II.1. Desired Neutron Beam Parameters

Epithermal neutron beam entering tissue creates radiation field with a

maximum thermal flux at a depth 2 to 3 cm, which drops exponentially thereafter.

In contrast to the epithermal beam which shows a skin-sparing effect, the thermal

flux falls off exponentially from the surface. Thus, thermal neutron irradiations

have been used for tumour treatments in the skin. In general, however, the current

trend for treatment of patients is using epithermal neutron beams. [6]

The main collimator designing objective is to deliver an epithermal neutron

beam within a reasonable treatment time and to produce the desired thermal

neutrons at tumour depth with minimal other radiations present. The two principal

beam characteristic of interest are intensity and quality. Beam intensity will be the

main determinant of treatment time. Beam quality relates to the types, energies,

and relative intensities of all the radiations present. [6]

II.1.1. Epithermal Beam Intensity

For the purposes of reporting beam intensity, the common definition for an

epithermal energy range should be used, namely 0.5 eV to 10 keV. Current

experience shows that desirable minimum epithermal neutron beam intensity

would be 109 n.cm-2.s-1. Beam of 5 x 108n.cm-2.s-1 are usable, but result in rather

long irradiation times. Where there is a choice to be made, most practitioners

would rather have better quality rather than more intensity. [6]

II.1.2. Incident Beam Quality

Beam quality is determined by four parameters under free beam conditions.

They are discussed below in order of importance. [6]

61. The fast neutron component

In BNCT the energy range for fast neutrons is taken as > 10 keV. Fast

neutrons, which accompany the incident beam, have a number of undesirable

characteristics such as free radicals production. Therefore, it is one of the main

objectives of BNCT beam design to reduce the fast neutron component. In

existing facilities, the range of dose from this component is from 2.5 to 13 x 10-13

Gy.cm2 per epithermal neutron, meanwhile the target number should be 2 x 10-13

Gy.cm2 per epithermal neutron. [6]

2. The -ray component

It is desirable to remove -ray radiation from the beam. A target number for this should be 2 x 10-13 Gy.cm2 per epithermal neutron. The range in existing

facilities is from 1 to 13 x 10-13 Gy.cm2 per epithermal neutron. [6]

3. The ratio between the thermal flux and the epithermal flux

To reduce damage to the scalp, thermal neutrons in the incident beam

should be minimized. A target number for the ratio of thermal flux to epithermal

flux should be 0.05. [6]

4. The ratio between the total neutron current and the total neutron flux

This ratio provides a measure of the fraction of neutrons that are moving in

the forward beam direction. A high value is important for two reasons; to limit

divergence of the neutron beam (thereby, reduce undesired irradiation of other

tissues) and to permit flexibility in patient positioning along the beam central axis.

A target number for this ratio should be greater than 0.7. [6]

II.2. Neutron Source for BNCT

Several experiences in designing collimator for BNCT have been conducted

both based on the materials selection and the geometry optimizing. A collimator

at least consists of 5 components: collimator wall, moderator, filter, -rayshielding, and aperture. Hereby, explained each of those parts.

71. Collimator wall

Collimator wall should reflect neutrons back into the inner part of

collimator. Therefore, neutron reflecting type material is used. Suitable reflector

materials for this are those with high scattering cross section and high atomic

mass (resulting in little energy loss). They include Pb, Bi, PbF2. [6]

In his experiment, Marko Mauec (1998) found Ni outperformed other materials, Pb, Bi and PbF2, with the highest epithermal neutron flux as the result.

O. O. Gritzay et al. (2004) also made a collimator design with Kyviv Research

Reactor as the neutrons source. In their study they used Ni as collimator wall

layer. From this study they got that the epithermal neutron flux grew up as the Ni

layer became thicker up to 6.5 cm, then it started to fall off slowly. [9,10]

Walls that are used near the beam exit are beam delimiters and it should

absorb rather than reflect neutrons. This part is made of B4C or 6Li2CO3 dispersed

in polyethylene. Epithermal neutrons striking the wall of the collimator are

thermalized and captured. It should be noted that 10B emits a low energy capture

-ray (478 keV) but 6Li does not and its use is to be preferred in locations close to the patient. [6]

2. Moderator

Moderation of fast neutrons is best accomplished by low atomic mass

materials. Any moderator or filter materials chosen must not decompose in a high

radiation field, nor produce moisture. Any neutron activation products from the

materials should be short lived. Some suitable candidates that widely used are Al,

Al2O3, and AlF3. Combinations of Al followed by Al2O3 or AlF3 downstream are

very efficient because the O and F cross-sections fill in the valleys between the

energy resonance peaks of Al. [6]

3. Gamma-ray shielding

Materials such as Pb and Bi may be placed in the beam to reduce -rays originating from the reactor core, but these will nonetheless reduce neutron beam

8intensity. Bi is nearly as good as Pb for shielding -rays, while having a higher transmission of epithermal neutrons. [6]

4. Filter

The objective is to filter out all neutrons but the epithermal neutrons from

the reactor beam. For epithermal neutron beams, it is desirable to limit thermal

neutron contamination by filtering. Filter materials for thermal neutrons require

either elements with 6Li, 10B or Cd. Cd is most frequently used absorber due to the

reason that Cd is an effective (n,) converter. [6,8]

Not only thermal neutrons, but also fast neutrons are very necessary to

reduce. This can be done with natural or isotopically enriched materials, for which

an interference minimum in the total neutron cross section exists in epithermal

energy range. The total cross section of 60Ni isotope has the deep and wide

interference minimum in the energy range from several eV to 10 keV and

therefore this material is useful for BNCT purposes. [6]

5. Aperture

Aperture is a part of collimator which provides required cross section of the

beam. Because of its role in the collimator, it is often found to be located at the

end point of collimator. In this study, the collimator which is going to be built is

for trials with 1 to 2 cm sized tumour cell samples and tumour-injected animals as

the object. For the tumour-injected animals, the size of tumour cells would be

monitored. Once the tumour reaches the detectable size, it would be irradiated

immediately. Hence the minimum detectable size of tumour should be known.

James Michaelson (2003) used screening mammography to detect breast tumour.

According to the result of the study, it was found that the median size at which

breast tumours become operationally detectable by screening mammography was

approximately 7.5 mm, with relative efficiency of 50%. A higher relative

efficiency of 80% appeared for 10 mm tumour detection, and 100% for 30 mm

tumour detection. [11]

9CHAPTER III

THEORETICAL BACKGROUND

III.1. Radiation Interactions with Matter

The design of all nuclear systems such as reactors, radiation shields, and so

on, depends fundamentally on the way in which nuclear radiation interacts with

matter. In this section, these interactions are discussed for neutrons and -rays.

III.1.1. Neutron Interactions

It is important to recognise that, since neutrons are electrically neutral, they

are not affected by the electrons in an atom or by the positive charge of the

nucleus. As a consequence, neutrons pass through the atomic electron cloud and

interact directly with the nucleus. Neutrons may interact with nuclei in one or

more of the following ways. [4]

1. Scattering

Scatter is an important way for neutrons to lose kinetic energy. Neutron

scattering occurs when neutrons collide with the nuclei of atoms. Neutrons may

scatter from interaction with a nucleus either in an elastic or inelastic fashion. In

elastic scattering process, the neutron strikes the nucleus, which is almost always

in its ground state, the neutron reappears, and the nucleus is left in its ground

state. This interaction is abbreviated by the symbol (n,n). [12]

Inelastic scattering is identical to elastic scattering except that the nucleus is

left in an excited state. Inelastic scattering is denoted by the symbol (n,n). The

excited nucleus decays, by the emission of -rays. [4]

2. Absorption

Neutrons may enter the nucleus of an atom quite easily, as compared to the

particles since there is no coulomb or charge repulsion to overcome. Absorption

10

interaction may cause several kinds of reaction. In radiative capture which is

denoted by (n,), the neutron is captured by the nucleus, and one or more -rays, called capture -rays, are emitted. Another reaction is charged-particle reactions, which results in charged particle production, such as -particle and proton. Fission reaction can occur if neutrons collide with certain nuclei, causing the

nucleus to split apart. Fission reaction is the principal source of nuclear energy for

practical applications. [4]

The extent to which neutrons interact with nuclei is described in terms of

quantities known as cross sections. Suppose that a beam of monoenergetic

neutrons of area A impinges on a target of thickness X. If there are n neutrons per

cm3 in the beam and is the speed of the neutrons, then the quantity [4]

= n, (3.1)is called the intensity of the beam. One can think of the neutron flux in a reactor as

being comprised of many neutron beams traveling in various directions. Then, the

neutron flux becomes the scalar sum of these directional flux intensities (added as

numbers and not vectors), that is, = I1 + I2 + I3 +... Since the neutrons travel the distance cm in 1 second, all of the neutrons in the volume A in front of the target will hit the target in 1 second. Thus, n A = I A neutrons strike the entire target per second. The number that do collide are found to be proportional to the

beam intensity, to the atom density N of the target, and to the area and thickness

of the target. These observations can be summarized by the equation [4,12]

( ) = s , (3.2) where , the proportionality constant, is called the cross section. The factor N A Xin Equation (3.2) is the total number of nuclei in the target. The number of

collisions per second with a single nucleus is therefore just I. It follows that is equal to the number of collisions per second with one nucleus per unit intensity

of the beam or, in other words, the effective cross sectional area of the nucleus,

hence the term cross section. Each of the processes described by which neutrons

interact with nuclei is denoted by a characteristic cross section. Thus, elastic

11

scattering is described by the elastic scattering cross section, e; inelastic scattering by the inelastic scattering cross section, i; the (n,) reaction (radiative capture) by radiative capture cross section, ; and so on. The sum of the cross sections for all possible interactions is known as the total cross section and is

denoted by the symbol T; that is [4]

s = s + s + s + s + (3.3)The sum of the cross sections of all absorption reactions is known as the

absorption cross section and is denoted by a. Thus, [4]

s = s + s + s + sa + (3.4)The total scattering cross section is the sum of the elastic and inelastic scattering

cross section. Thus, [4]

s = s + s , (3.5)and [4]

s = s + s. (3.6)The product of the atom density N and cross section, as in Equation (3.2), occurs frequently in the equations of nuclear engineering; it is given the special

symbol and is called the macroscopic cross section. In particular, the product NT = T is called the macroscopic total cross section, N s = s is called the macroscopic scattering cross section, and so on. Since N and have units of cm-3

and cm2, respectively, has unit of cm-1. [4]

Let I(X) be the intensity of the neutrons that have not collided after

penetrating the distance X into the target. Then in traversing in additional distance

dX, the intensity of the uncollided beam is decreased by the number of neutrons

that have collided in the thin sheet of target having an area of 1 cm2 and the

thickness dX. From Equation 3.2, this decrease in intensity is given by [4]

() = s () = () . (3.7)This equation can be integrated with the result [4]

12

() = 0. (3.8)The intensity of the uncollided neutrons thus decreases exponentially with the

distance inside the target. [4]

When equation 3.7 is divided by I(X), the result is [4]

()() = .T dX is equal to the probability that neutron will interact in dX, and it may be concluded that T is the probability per unit path length that a neutron will undergo some sort of the collision as it moves about in a medium. The average

distance that a neutron moves between collisions is called the mean free path,

which is designated by the symbol (cm), [4]

= 1 . (3.10)

III.1.2. Gamma-ray Interactions

Although the term -ray is normally reserved for radiation emitted by nuclei and x-ray refers to radiation originating in transitions of atomic electrons, both

forms of radiation are called -rays in the present section. There is no fundamental difference between the two radiations, as they are both electromagnetic radiation.

Gamma-rays interact with matter in several ways. Ordinarily, however, only three

processes must be taken into account in nuclear engineering problems: the

photoelectric effect, pair production, and Compton effect. Alike neutrons, in -rayinteractions the term cross section is also used in the same way. [4]

1. The photoelectric effect

The photoelectric effect occurs when the electromagnetic radiation or

photon imparts all its energy to an orbital electron, the -ray disappears, and theelectron is ejected from the atom. The kinetic energy of the ejected photoelectron

is therefore equal to the energy of the photon less the binding energy of the

electron to the atom. If a -ray succeeds in ejecting an inner atomic electron, the

(3.9)

13

hole in the electronic structure is later filled by a transition of 1 of the outer

electrons into the vacant position, accompanied by the emission of x-rays

characteristic of the atom or by the ejection of an Auger electron. The

photoelectric cross section is denoted by the symbol pe. [4]

2. Pair production

In this process, the photon disappears and an electron pair, a positron and a

negatron, is created. Since the total rest-mass energy of the 2 electrons is 2 mc2 =

1.02 MeV, this effect does not occur unless the photon has at least this much

energy. Above this threshold, the cross section for a pair production, pp, increases steadily with increasing energy. The total kinetic energy of the negatron-

positron pair is equal to 1.02 MeV. Once formed, these electrons move about and

lose energy as a result of collisions with atoms in the surrounding medium. After

the positron has slowed down to very low energies, it combines with an electron,

the two particles disappear, and two photons are produced (annihilation radiation),

each having an energy of 0.511 MeV. [4]

3. The Compton effect

The Compton effect, or Compton scattering as it is sometimes called, is

simply the elastic scattering of a photon by an electron. An incident photon with

energy E is scattered through the angle and the struck electron recoils. Since the recoiling electron acquires some kinetic energy, the energy E' of the scattered

photon is less than E. This interaction is denoted by C. [4]

The total cross section per atom for -ray interaction is the sum of the cross sections for the photoelectric effect, pair production, and Compton scattering, [4]

s = s + s + s. (3.11)A macroscopic cross section can also be defined, like the macroscopic neutron

cross section, by multiplying T in by the atom density N. Such macroscopic -ray cross sections are called attenuation coefficients and are denoted by the symbol . Thus, [4]

14

m = s = m + m + m, (3.12)where is the total attenuation coefficient and pe, pp, and C are the attenuation coefficients for the three interaction processes. Like macroscopic cross sections

for neutrons, the various have units of cm-1. is equal to the probability per unit path that a -ray will have a collision in a medium and that [4]

= 1m ,is the mean free path of the -ray. If I0 is the intensity (-rays.cm-2.s-1) of the monoenergetic -ray beam striking a target of thickness X, then the intensity of the photons that penetrate the target without having a collision is [4]

() = 0 m . (3.14)

III.2. The Monte Carlo Method and MCNP Program

The Monte Carlo method can be used to duplicate theoretically a statistical

process (such as the interaction of nuclear particles with materials). The individual

probabilistic events that comprise a process are simulated sequentially. The

probability distributions governing these events are statistically sampled to

describe the total phenomenon. The statistical sampling process is based on the

selection of random numbers based on the physics rules and probabilities

governing the processes and materials involved. [14]

Figure 3.1. Random history of a neutron incident on a fissionable material slab.

(3.13)

fissionable material

incident neutron1

24

3 6

7

5

15

Figure 3.1 depicts a random history of a single neutron incident on a slab of

material that can undergo fission reaction. Numbers between 0 and 1 are selected

randomly to determine what and where interaction takes place In this particular

example, a neutron collision occurs at event 1. The neutron is scattered in the

direction shown. A photon is also produced and is temporarily stored (banked) for

later analysis. At event 2, fission occurs, resulting in the termination of the

incoming neutron and the birth of 2 outgoing neutrons and 1 photon. The neutron

and the photon are banked for later analysis. The first fission neutron is captured

at event 3 and terminated. The banked neutron is now retrieved and leaks out of

the slab at event 4. The fission-produced photon has a collision at event 5 and

leaks out at event 6. The remaining photon generated at event 1 is now followed

with a capture at event 7. This is a quite satisfying example of random phenomena

generated in the Monte Carlo method. As more and more such histories are

followed, the neutron and photon distributions become better known. [14]

III.2.1. Weight

If MCNP were used only to simulate exactly physical transport, then each

MCNP particle would represent one physical particle and would have unit weight.

For instance, each MCNP particle might represent a number w of particles emitted

from a source. This number w is the initial weight of the MCNP particle. The w

physical particles all would have different random walks, but one MCNP particle

representing these w physical particles will only have one random walk. The true

number of physical particles is preserved in MCNP in the sense of statistical

averages. Each MCNP particle result is multiplied by the weight so that the full

results of the w physical particles represented by each MCNP particle are

exhibited in the final results (tallies). This procedure allows users to normalize

their calculations to whatever source strength they desire, so that the expected

means will be independent of the number of source particles actually initiated in

the MCNP calculation. [14]

16

III.2.2. Particle Tracks

When a particle starts out from a source, a particle track is created. If that

track is split 2 for 1 at a splitting surface or collision, a second track is created and

there are now two tracks from the original source particle. Track length tallies use

the length of a track in a given cell to determine a quantity of interest, such as

fluence or energy deposition. Tracks crossing surfaces could also be used. [14]

III.2.3. Neutron Interactions

1. Scattering

The selection of an elastic collision is made with the probability [14]

ss + s =

ss s,

where el is the elastic scattering cross section, in is the inelastic cross section, ais the absorption cross section ((n,x) where x n that is, all neutron disappearing reactions), T is the total cross section (T = el + in + a). The selection of an inelastic collision is made with the remaining probability [14]

ss s.

2. Absorption

The terms absorption and capture are used interchangeably for non-fissile

nuclides, both meaning (n,0n). For fissile nuclides, absorption includes both

capture and fission reactions. [14]

In analog absorption, the particle is killed with probability a/T, where aand T are the absorption and total cross sections of the collision nuclide at the incoming neutron energy. The absorption cross section is specially defined for

MCNP as the sum of all (n,x) cross sections, where x is anything except neutrons.

Thus a is the sum of n,, n,, f, etc. Implicit absorption has a fraction of 1 -

(3.15)

(3.16)

17

a/T of the incident particle weight and energy is deposited in the collision cell corresponding to that portion of the particle that was absorbed. [14]

III.2.4. Photon Interactions

The physical processes treated are photoelectric effect, pair production, and

Compton scattering from free electrons. The photoelectric effect is regarded as an

absorption (without fluorescence). The total cross section t is regarded as the sum of three components [14]

s = s + s + s. (3.17)1. Photoelectric effect

This is treated as a pure absorption by capture with a corresponding

reduction in the photon weight, and hence does not result in the loss of a particle

history. Photoelectric happens with probability pe/T. [14]

2. Pair production

In a collision resulting in pair production [probability pp/(T pe)], either an electron-positron pair is created for further transport and the photon disappears,

or it is assumed that the kinetic energy weight (E 1.022) MeV of the electron-

positron pair produced is deposited as thermal energy at the point of collision,

with production of one photon of energy 0.511 MeV headed in one direction and

another photon of energy 0.511 MeV headed in the opposite direction. [14]

3. Compton scattering

The alternative to pair production is Compton scattering on a free electron,

with probability s/(T pe). This yields at once the energy weight (E E) deposited at the point of collision and the new direction of the scattered photon.

The energy deposited at the point of collision can then be used to make a

Compton recoil electron for further transport. [14]

18

CHAPTER IV

MATERIALS AND METHOD

IV.1. Materials

This study was a simulation-based experiment. Materials used are listed as

follows.

1. Computer

The computer used had specifications:

Processor : Intel Core i3 CPU 2.93 GHz

RAM : 2.00 GB

Operating System : 32-bit, Windows 7

2. Simulation Program

Monte Carlo N-Particle version 5 (MCNP5) was used for the simulations of

phenomena of interest. MCNP was a general-purpose Monte Carlo N-Particle

code that can be used for neutron, photon, electron, or coupled

neutron/photon/electron transports. Specific areas of application include, but were

not limited to, radiation protection and dosimetry, radiography, medical physics,

nuclear criticality safety, and also fission and fusion reactor design. MCNP5 was

the latest version of MCNP which included some additions of photonuclear

database, superimposed mesh tallies and time splitting ability. Meanwhile,

MCNP6 was still being developed.

IV.2. Method of Study

IV.2.1. Kartini Research Reactor Modelling

Kartini Research Reactor specifications are documented in the Safety

Analysis Report (SAR) of the reactor. It was needed to make a model of the

reactor since it would be used as the neutrons source.

19

An MCNP input le is divided into 3 main blocks (which are known as cards) so called cell cards, surface cards, and data cards. The rst two cards correspond to the geometry denition, while the data cards contain all the information related to the specication of the particle source, the denition of the materials, and the tallies. By using these codes, Kartini Research Reactor was

modelled, as the first step.

Figure 4.1. Core configuration.

Reference: [15]

A

CT

B1B6

B5 B2

B4 B3

9987

98839988

9994

9995 9996

IFE

C1

C2

C3

C10 C4

C9 C5

C8 C6C7

C11

C12 9892

9998

9981

9597

9598

9977 9976

99759983

9592

D18 D1

D17 D2

D16 D3

D15 D4

D14 D5

D13 D6

D12 D7

D11 D8

D10 D9

98819880

9877

9871

9352

9980

98789869

9594

9982

9879

9986

9985

9984

9997

9593

9873 9870

E24E1

E2

E23

E2 2

E21

E20

E3

E4

E5

E6

E7

E8

E9

E19

E18

E17

E16

E15

E10

E11

E14E13

E12

9350

9637

9979

9636

9887

9354

9596

9635

9978

9889

9639

9349

9872

9641

9595

9888 9353 9890

9891

9640

9885

9882

9886

F30G10F29

F28

F1F2

F27

F26

F25

F24

F23

F22

F21

F20

F19

F18

9535F17

F16 F15F14

F13

F12

F11

F10

F9

F8

F7

F6

F5

F4

F3

9891

G 1247

9876

9538

9540

9537

9536

9542

AmBe

9541

9539

G7

G1

G4

G3

G5

9543

G2666

2812G

9875

27922821

G

GG8

G9

2810G

2799G

PS

CR

CR

CR

20

Kartini Research Reactor is a TRIGA MARK-II research reactor type. It has

a maximum thermal power of 250 kW. The reactor was modelled by using

MCNP5 program with core configuration as depicted in Figure 4.1. Several other

parts of the reactor, whose existence were considered to affect to the reactor

criticality, were also modelled, such as the radial reflector, rotary specimen rack,

and piercing beam port. Moreover, the thermal column was also built since it

would become the point of interest; where the collimator would be built.

The desired thermal power for this study was 100 kW. According to the

Safety Analysis Report (SAR), for gaining 100 kW of thermal power the control

rods needed to be arranged in different axial positions. C5 control rod was

dragged to 100%, C9 to 65% and E1 to 55% of the active core height [15]. In this

step, criticality calculations were done and the neutron importance was restricted

only for those parts located in the inner side of radial and axial reflector. Thus,

neutrons that travelled out of this limit were not calculated or, considered as

leaking neutrons.

Some brief simulations were done to make sure that the criticality value was

approximately 1, and the thermal neutron flux in the Ring B was near (12.45+

0.23) x 1011 n.cm-2.s-1 [16]. Up to this point, it was not yet necessary to do a

copious number of iteration. So, in the KCODE card, using the default settings,

1,000 starting particles (or also called as history) with 130 total number of cycle

was merely enough. For neutron flux calculation tally card, F4:N, was used.

Deeper explanations about tally will be discussed later in Tally Selecting section.

IV.2.2. Neutrons and Gamma Rays Recording

Neutrons and -rays recording means that those neutrons and -rays which are released as the reaction stemmed from any interaction happens in the reactor

and then pass through a certain defined surface are written into a file, so that we

can use the surface as a new neutron source for the next further calculation. This

is a quite necessary method for reducing time consuming of the simulation.

21

Higher number of important cells would prolong the simulation time. By using

this method, for every modification done in the collimator design, we do not need

to include the reactor core in the calculation. We only use the new particle source.

Thus, the simulation time would be pretty much shortened.

In this study, it was very advantageous to record the one-directional particle

tracks that crossed the surface which separated the reactor and the thermal

column. The direction of the tracks must be from the reactor then entered the

thermal column. This part was done after one convinced with the reactor model

which had been made. Generally, the error and variance decrease as the larger

number of iterations taken. Thus a plenty number of histories per cycle were

needed. 107 histories per cycle were eventually used in each of 30 cycles. It took

about 3 to 4 days until the program finished the iteration process.

IV.2.3. Tally Selecting

In an MCNP input le, tallies are the information that a user wants to obtain by Monte Carlo calculation. Several tallies provided in MCNP5 are shown in

Table 4.1.

Table 4.1. MCNP tally types.

Mnemonic Tally Description Fn Units *Fn Units

F1:N or F1:P or F1:ECurrent integrated overa surface

particles MeV

F2:N or F2:P or F2:EFlux averaged over a surface

particles.cm-2 MeV.cm-2

F4:N or F4:P or F4:EFlux averaged over a cell

particles.cm-2 MeV.cm-2

F6:N or F6:N,P or F6:PEnergy deposition averaged over a cell

MeV.g-1 jerks.g-1

F8:P or F8:E or F8:P,EEnergy distribution of pulses created in a detector

pulses MeV

Reference: [14]

The abbreviation N, P, and E namely means neutron, photon, and electron.

22

Tallies were selected according to the parameters used in the beam criteria

suggested by the IAEA, as shown in Table 4.2 below.

Table 4.2.Beam parameters.

Parameter Nomenclature Epithermal beam intensity epi (n.cm-2.s-1) Fast neutron dose per epithermal neutron Df / epi (Gy.cm2.n-1)Gamma dose per epithermal neutron D / epi (Gy.cm2.n-1)Ratio between thermal flux and epithermal flux

th / epiRatio between neutron current and neutron flux

J / epi

Reference: [6]

By examining Table 4.2, it was found that the tallies needed were neutron flux,

neutron dose rate, -ray dose rate, and neutron current. The tallies exploited for this work are F4:N for the calculation of neutron ux and dose rate averaged over a cell, F4:P for the calculation of photon dose rate averaged over a cell, and also

F1:N for the calculation of neutron current integrated over a surface. F4 can be

replaced, indeed, by F2, but it leads to a more complex code since we have to trim

the surface and use the desired one.

F4 tally was used for 3 aims. Meanwhile, in fact, in MCNP each tally can

only be used for one aim. In other words, having two F4:N for flux and dose

calculation, and an F4:P in the same input file is not allowed. One needs to put

one or two digits of additional number between F and n (the tally number) to

make a difference for each tally. In this study, for instance, F4:N was used for

neutron flux calculation, F14:N for fast neutron dose rate calculation, and F24:P

for photon dose rate calculation.

Normalization was clearly needed since the output unit from each MCNP

tally did not match the unit used by the IAEA. First of all, fission rate needed for

generating 100 kW thermal power was calculated as follows.

23

(15 ) 1 1 1.62 13 1 2 = .121 15 .Therefore, to produce 100 kW of thermal power, one needs 3.121 x 1015 fissions

per second. By using this fission rate, normalization factor for each tally were

calculated as follows.

1. Neutron flux and dose rate (F4:N and F14:N)

For an average of 2.42 neutrons per fission [4], the normalization factor is

.121 15 2.42 = 7. 15 .This result was used both for neutron flux (F4:N) and neutron dose rate (F14:N)

calculations.

2. Gamma dose rate (F24:P)

For 1 -ray per fission [4], the normalization factor is

.121 15 1 = .121 15 .3. Neutron current, F1:N

For an average of 2,42 neutrons per fission [4], the normalization factor is

.121 15 2.42 = 7. 15 .It needs to be divided with the area which is prependicular to the neutron current.

In this study, the multiplication factor for F1:N tally was varied due to its

dependence on the size of collimator aperture. The maximum aperture diameter

used was 5 cm, meanwhile the minimum was 1 cm. For 3 cm aperture diameter,

the normalization factor for F1:N was

(7. 15) (1. ) = 1.68 15 . .

24

Energy classifications for neutrons should be included in the input file for

flux calculation, so each of thermal, epithermal, and fast neutron fluxes appeared

in the output file. MCNP needed the upper limit of neutron energy for the energy

bins. The energy limits of 5 x 10-7, 10-2, and 20 MeV were used. Those values,

respectively, denote the upper limit of thermal, epithermal, and fast neutron

energy spectrums. The total neutron flux would appear automatically.

Furthermore, an important step in the dosimetry evaluation was to relate the

radiation passing through a unit volume of a material (fluence) to the energy

release (kerma) in the material. Therefore, the latest fluence-to-kerma conversion

coefficients or kerma coefficients used in Dosimetry System 2002 (DS02) from

ICRU Report 63 were taken into account of neutron and photon doses. The kerma

coefficients for neutrons and photons in air were used. Since it was fast neutron

and -ray dose rate needed, the kerma coefficients for neutrons used were only those with energy higher than 10-2 MeV (the lower energy limit of fast neutron),

meanwhile kerma coefficients for photon were all used. Respectively, Table 4.3

and 4.4 shows the kerma coefficients for fast neutrons and photons.

Table 4.3. Kerma coefficients for fast neutrons.

Neutron Energy (Mev)

Kerma Coefficient (Gy.cm2)

Neutron Energy (Mev)

Kerma Coefficient (Gy.cm2)

1.10 E-2 1.09 E-12 1.55 E-1 8.86 E-122.00 E-2 1.88 E-12 1.65 E-1 9.19 E-123.60 E-2 3.11 E-12 1.75 E-1 9.51 E-126.30 E-2 4.82 E-12 1.85 E-1 9.83 E-128.20 E-2 5.86 E-12 1.95 E-1 1.01 E-118.60 E-2 6.05 E-12 2.10 E-1 1.06 E-119.00 E-2 6.24 E-12 2.30 E-1 1.11 E-119.40 E-2 6.44 E-12 2.50 E-1 1.16 E-119.80 E-2 6.62 E-12 2.70 E-1 1.21 E-111.05 E-1 6.92 E-12 2.90 E-1 1.27 E-111.10 E-1 7.35 E-12 3.10 E-1 1.31 E-111.25 E-1 7.76 E-12 3.30 E-1 1.36 E-111.35 E-1 8.13 E-12 3.50 E-1 1.41 E-111.45 E-1 8.50 E-12 3.70 E-1 1.46 E-11

25

Neutron Energy (Mev)

Kerma Coefficient (Gy.cm2)

Neutron Energy (Mev)

Kerma Coefficient (Gy.cm2)

3.90 E-1 1.52 E-11 3.50 E+0 4.29 E-114.20 E-1 1.66 E-11 3.70 E+0 4.40 E-114.60 E-1 1.64 E-11 3.90 E+0 4.33 E-115.00 E-1 1.65 E-11 4.20 E+0 4.43 E-115.40 E-1 1.71 E-11 4.60 E+0 4.43 E-115.80 E-1 1.77 E-11 5.00 E+0 4.68 E-116.20 E-1 1.83 E-11 5.40 E+0 4.57 E-116.60 E-1 1.89 E-11 5.80 E+0 4.77 E-117.00 E-1 1.95 E-11 6.20 E+0 4.92 E-117.40 E-1 2.00 E-11 6.60 E+0 5.07 E-117.80 E-1 2.06 E-11 7.00 E+0 5.19 E-118.20 E-1 2.11 E-11 7.40 E+0 5.42 E-118.60 E-1 2.16 E-11 7.80 E+0 5.47 E-119.00 E-1 2.23 E-11 8.20 E+0 5.41 E-119.40 E-1 2.33 E-11 8.60 E+0 5.56 E-119.80 E-1 2.50 E-11 9.00 E+0 5.66 E-111.05 E+0 2.52 E-11 9.40 E+0 5.83 E-111.15 E+0 2.52 E-11 9.80 E+0 5.96 E-111.25 E+0 2.63 E-11 1.05 E+1 6.01 E-111.35 E+0 2.71 E-11 1.15 E+1 6.38 E-111.45 E+0 2.76 E-11 1.25 E+1 6.38 E-111.55 E+0 2.83 E-11 1.35 E+1 6.54 E-111.65 E+0 2.94 E-11 1.45 E+1 6.61 E-111.75 E+0 2.99 E-11 1.60 E+1 6.77 E-111.85 E+0 3.12 E-11 1.80 E+1 6.95 E-111.95 E+0 3.13 E-11 2.00 E+1 7.04 E-112.10 E+0 3.24 E-11

Reference: [17]2.30 E+0 3.29 E-112.50 E+0 3.44 E-112.70 E+0 3.59 E-112.90 E+0 3.75 E-113.10 E+0 3.85 E-113.30 E+0 4.19 E-11

26

Table 4.4. Kerma coefficients for photons.

Photon Energy (Mev)

Kerma Coefficient (Gy.cm2)

Photon Energy (Mev)

Kerma Coefficient (Gy.cm2)

1.00 E-3 5.63 E-10 2.00 E-1 9.43 E-131.50 E-3 2.83 E-10 3.00 E-1 1.52 E-122.00 E-3 1.68 E-10 4.00 E-1 2.09 E-123.00 E-3 8.07 E-11 5.00 E-1 2.62 E-124.00 E-3 4.70 E-11 6.00 E-1 3.13 E-125.00 E-3 3.02 E-11 8.00 E-1 4.08 E-126.00 E-3 2.09 E-11 1.00 E+0 4.93 E-128.00 E-3 1.16 E-11 1.25 E+0 5.89 E-121.00 E-2 7.24 E-12 1.50 E+0 6.76 E-121.50 E-2 4.04 E-12 2.00 E+0 8.29 E-122.00 E-2 2.64 E-12 3.00 E+0 1.09 E-113.00 E-2 7.02 E-13 4.00 E+0 1.31 E-114.00 E-2 4.23 E-13 5.00 E+0 1.52 E-115.00 E-2 3.25 E-13 6.00 E+0 1.71 E-116.00 E-2 2.98 E-13 8.00 E+0 2.09 E-118.00 E-2 3.27 E-13 1.00 E+1 2.47 E-111.00 E-1 4.03 E-13 1.50 E+1 3.39 E-111.50 E-1 6.61 E-13 2.00 E+1 4.33 E-11

Reference: [17]

Flux-to-kerma conversion was done by using DEn/DFn cards.

IV.2.4. Beam Criteria

It was said in the IAEAs technical document that most practitioners would

rather have better quality of the neutron beam than more intensity. It was also

emphasised that the beam quality was determined by four parameters, in order of

importance: fast neutron component, -ray component, thermal neutron component, and directionality. Thus the designing process was done according to

this rule. Table 4.5 shows the desired BNCT-purpose beam in this study.

27

Table 4.5. Beam criteria.

Nomenclature Valueepi (n.cm-2.s-1) > 1.0 x 109Df / epi (Gy.cm2.n-1) < 2.0 x 10-13D / epi (Gy.cm2.n-1) < 2.0 x 10-13

th / epi < 0.05 J / epi > 0.7

Reference: [6]

IV.2.5. Collimator Conceptual Designing

Here discussed the consideration of materials chosen and thickness

variations made. Determination of the size variation was based on the mean free

path of neutrons within the materials. Mean free path was calculated by using

several formulas as follows. First, for getting the atomic density, [18]

= ,where wfi is the weight fraction, Ni(atoms.cm

-3) is the atom density, and Mi

(g.mole-1) is the atomic weight of ith element. (g.cm-3) is the density of the material (mixture), NA is the Avogadros number, 6.02 x 10

23 atoms.mole-1. Then,

the macroscopic cross section of phenomenon of interest was calculated,

= s

,

where in cm-1 is the macroscopic cross section of the material. Ni is the atom density and i (cm2) is the microscopic cross section of ith element. Then the mean free path is

= 1 .The mean free path used depended on the role of each material. Scattering mean

free path should be used for moderator and collimator wall materials. For filters,

absorption mean free path should be used. For beam delimiter which would both

(4.3)

(4.2)

(4.1)

28

moderate and absorb neutrons in the same time, the total mean free path was used.

Total cross section in Equation 4.3 was replaced by attenuation coefficient in -ray shielding variation calculation. The data of cross sections and attenuation coefficient were gained respectively from Reference 19 and 20.

1. Beam delimiter, 6Li2CO3-polyethylene

As discussed earlier, 6Li was the best material to be located near the patient.

The combination between C, H, and O resulted in a good moderation effects for

the neutrons meanwhile 6Li would absorb the neutrons. The minimum thickness

the beam delimiter should be equivalent to the total cross section of the

compound.3 cm thick of 6Li2CO3-polyethylene compound was used.

2. Collimator wall, Ni

Among all collimator wall materials suggested, Ni was found outperformed

other materials. The minimum thickness of collimator wall should be, at least,

equivalent to the scattering mean free path of high energy neutrons, 3 cm. Since

the thickness variation for every 3 cm was considered too large for collimator

wall, it was varied for every 1 cm rather than 3 cm.

3. Moderator (Al/AlF3/Al2O3)

Materials for moderating fast neutrons were compared. The thickness

variation made depended on its fast neutron scattering cross section. Since Al

were being the main component, the main free path of Al considerably used for

this purpose. Thus, moderator thickness was varied for every 5 cm. After the best

moderator was chosen among 3 candidates, within the same principal of formula,

the mean free path of moderator material used was calculated to be used as

variation difference.

4. Filter,60Ni

60Ni was said to be the best material for absorbing fast neutrons. More over

in fact, it also reduced the thermal neutrons intensity dramatically. Thus, no

thermal neutron filter needed in this study. The variation depended on the fast

29

neutron absorption cross section of 60Ni. In fact, the calculation resulted in 953 cm

of mean free path. The variation of about 950 cm was unacceptable since the size

of collimator itself had been limited as short as 100cm. Hence, the variation was

changed to be equivalent to the fast neutron total cross section. Variation of

absorber thickness of 3 cm was considerably much more sensible than 950 cm.

5. Gamma-ray shielding, Bi

Bi was more preferable rather than Pb because of its lower cross section in

epithermal energy range compared to Pb. This was an advantage of using Bi as

material for -ray shielding in the collimator since lower cross section would cause lower decrease of neutrons. With attenuation coefficient of 0.614 cm-1, the

mean free path of a high energy (20 MeV) -ray was found 2 cm. Thus, the variation used was 2 cm.

As the first step, a rough collimator design was made by using MCNP5

codes, with 100 cm length of collimator, since it is the shortest length known for a

design of collimator, and considering the low number of neutrons produced from

a reactor with thermal power of 100 kW. Based on the mean free path calculation,

3 cm thick of beam delimiter was used, made of 6Li2CO3-polyethylene.Maximum

collimator diameter used was 54 cm. For the outlet, 3 cm of aperture diameter was

used. In designing collimator, one should start with the varied size of collimator

wall. The best thickness would be that the thickness which provided the highest

epithermal neutron flux. Then, moderator material was varied. Due the tendency

of the usage of Al and its composites such as AlF3 and Al2O3, these materials

were compared. The best material was that with highest epithermal neutron flux

for comparable value of fast neutron components. Best material gained from this

step then used and the thickness was varied until the fast neutron component

decrease became no longer significant. The increment of collimator wall thickness

would decrease the collimator inner diameter. After that, 60Ni, was started to be

used and varied until the fast and thermal neutron components desired reached.

The next step was to employ -ray shielding into the collimator and alter itsthickness until the desired -ray component gained. The last parameter of beam

30

quality, the directionality, was checked right after. If it is still below the desired

value, then the thickness of beam delimiter would be increased for higher value of

directionality. The last step conducted was varying the aperture to find out the

performance of the collimator design in different aperture size. Aperture size was

altered as needed; 1, 2, 3, 4, and 5 cm. 1 and 2 cm diameter are for irradiating the

tumour cell samples, meanwhile 3, 4, and 5 cm are for irradiating the tumour cells

within the animals.

IV.3. Results Analysis

In this study, data analysis was done during the simulation, since one part of

collimator depended on or affect to the other parts. It would be very convenient

to make graphs from the data resulted from the simulations, so that the tendency

of the phenomena could be visually and, thus, easily examined. For the collimator

wall, the graph (wall thickness versus epithermal neutron flux) had a peak which

depicts the highest flux in a certain wall thickness. The best thickness was that

provided the highest epithermal neutron flux. Different to the wall collimator, the

variation of moderator, filter, and -ray shielding resulted in graphs (material thickness versus parameter of interest) in exponential trend. The thickness used

was that which provided the desired value for each parameter of interest. [9]

31

CHAPTER V

RESULTS AND ANALYSIS

V.1. Reactor Criticality

The criticality calculation by using MCNP5 gave result 1.007 + 0.000,

which was a good approach to the criticality value of 1.000 +0.010. The thermal

neutron flux in Ring B of the reactor core was (14.30 +0.00) x 1011 n.cm-2.s-1,

mean while the real value, which was detected by a study, was approximately

(12.45 +0.23) x 1011n.cm-2.s-1 [16]. This difference might be caused by the

multiplication factor inputted into the MCNP codes that did not quite depict the

real number of neutrons. With these results, collimator designing was then

conducted.

V.2. Collimator Conceptual Design

Neutron beam which comes into the collimator must be dominated by

middle- to high-energy neutrons since the low energy neutrons must be reflected

back into the reactor core by radial reflector. Sufficient moderation and absorption

effects by the materials consisted in the collimator results in a middle-energy

neutrons dominated flux within good quality. This section explains further about

the results of the simulations and the final conceptual design.

V.2.1. Collimator Wall

Natural nickel is a very good material to be employed as a neutron

collimator wall. Its atomic mass which is not too small, that would make too much

energy decrement of neutrons, and yet not too high, that only would slightly shift

the energy spectrum of neutrons. Hence without moderator, the natural nickel

itself already produce epithermal neutron-dominated beam, but still needs more

32

materials to raise its quality. The results of simulation for wall thickness variation

are depicted in Figure 5.1.

Figure 5.1. Epithermal neutron flux for various thickness of wall (Ni).

As shown in Figure 5.1, the flux increases when 3 to 5 cm of wall thickness is

used. The thicker the collimator wall, the more neutrons would be reflected. The

flux reaches its highest value (2.67 n.cm-2.s-1) in thickness of 5 cm. At this point,

the energy spectrum shifts of fast neutrons to become epithermal neutrons is

optimum. In 6, 7, 8 cm of wall thickness and so on, epithermal neutron flux

decreases monotonically. In fact, as the thickness of collimator wall increases, the

inner diameter of collimator decreases, causing more collisions occurred between

the neutrons and the wall. Thus the energy spectrum shift becomes further, and

the epithermal neutrons more reduced, instead. Figure 5.2 shows the scattering

cross section of 58Ni. Since the natural nickel consists of about 80% 58Ni and 20% 60Ni, it is considerably assumed that the 58Ni cross section does depict the natural

nickel cross section. From Figure 5.2 it can be seen that 58Ni has scattering cross

section about 20 to 30 barns for epithermal neutrons. Just for comparison, Pb and

Bi which are recommended by the IAEA, have about 9 to 13 barns [18]. This is a

very good argument why natural nickel reflects more neutrons than Pb or Bi does.

0.0

0.5

1.0

1.5

2.0

2.5

3.0

3.5

0 2 4 6 8 10 12

ep

i(

x 10

9n

.cm

-2.s

-1)

Wall thickness (cm)

33

Figure 5.2. Scattering cross sections of 58Ni.

Reference: [19]

V.2.2. Moderator

The simulations proved that Al outperform the other materials, as depicted

by the data written in Table 5.1. For a comparison, with fast neutron component

of about 25 x 10-13 Gy.cm2.n-1, Al, AlF3, and Al2O3, produced epithermal neutron

flux of, respectively, 1.67 n.cm-2.s-1, 1.04 n.cm-2.s-1, and 0.92 n.cm-2.s-1. Thus Al

was chosen as material for moderator.

34

Table 5.1. Comparison of moderator materials.

Illuminator Thickness

(cm)

Al AlF3 Al2O3epi

(x 109 n.cm-2.s-1)Df / epi

(x 10-13 Gy.cm2.n-1)epi

(x 109 n.cm-2.s-1)Df / epi

(x 10-13 Gy.cm2.n-1)epi

(x 109 n.cm-2.s-1)Df / epi

(x 10-13 Gy.cm2.n-1)

5 2.23 77.13 1.98 76.85 1.60 62.4210 2.04 60.65 1.49 49.54 1.24 41.4115 1.91 45.41 1.24 35.11 0.92 25.5320 1.79 33.38 1.04 24.87 0.71 14.7525 1.67 26.58 0.81 18.78 0.56 11.00

35

The results of simulations for varied moderator thickness are depicted in Figure

5.3. It shows nicely how the ratio between fast neutron dose rate per epithermal

neutron flux decreases exponentially. With no moderator, the fast neutron

component is 1.08 x 10-11 Gy.cm2.n-1 or, approximately, 50 times higher than the

desired value, 2.0 x 10-13 Gy.cm2n-1.

Figure 5.3. Fast neutron component for various thickness of moderator (Al).

Al performs very well moderation effect that it reduces the fast neutron dose more

rapidly without much decrease of epithermal neutron flux up to 60 cm thickness.

After that, the addition of moderator is no longer effective since the fast neutron

component only slightly decreases, as shown in Table 5.2.

Table 5.2. Results of moderator (Al) thickness variations.

Moderator Thickness (cm)

epi(x 109 n.cm-2.s-1)

Df / epi(x 10-13 Gy.cm2.n-1)

55 1.33 5.7960 1.27 4.0765 1.21 3.5870 1.11 3.0475 1.06 2.6380 0.98 2.33

0

20

40

60

80

100

120

140

0 20 40 60 80 100

f/

epi(x

10-

13G

y.cm

2 .n

-1)

Moderator thickness (cm)

36

60 cm thick Al is used as moderator, with fast neutron component of 4.07 x 10-13

Gy.cm2.n-1 and epithermal neutron flux of 1.27 x 109n.cm-2.s-1.

V.2.3. Filter

Usage of 60Ni as filter gave results as shown in Figure 5.4 and 5.5.

Figure 5.4. Fast neutron component for various thickness of filter (60Ni).

Figure 5.5. Thermal neutron component for various thickness of filter ( 60Ni).

0

1

2

3

4

5

6

0 5 10 15 20

f/

epi(x

10-

13G

y.cm

2 .n

-1)

Filter thickness (cm)

0

0.01

0.02

0.03

0.04

0.05

0.06

0.07

0.08

0 5 10 15 20

th

/ep

i

Filter thickness (cm)

37

Figure 5.4 depicts that the fast neutron component, once again, decreases

exponentially. 12 cm thick of filter is actually enough to decrease the fast neutron

component to 1.84 x 10-13 Gy.cm2.n-1, below the upper limit recommended, but

according to the simulations done it eventually increased exceeding 2.0 x 10-13

Gy.cm2.n-1 when Bi as -ray shielding is added. Thus 15 cm thick of filter is preferred, with 1.70 x 10-13 Gy.cm2.n-1fast neutron component and 9.99 x 108

n.cm-2.s-1 epithermal neutron flux.

Thermal neutron component also decreases exponentially as more 60Ni

added into the collimator, as shown in Figure 5.5. With 15 cm thick of 60Ni, it is

reduced from 0.061 to 0.008, which is far below the recommended maximum

value, 0.05.

Figure 5.6. Absorption cross sections of 60Ni.

Reference: [19]

The reason why these phenomena happen is because of the absorption cross

section of 60Ni. As shown in Figure 5.6, 60Ni has minimum absorption cross

38

section for epithermal neutrons. Hence, 60Ni undergoes minimum interactions

with epithermal neutrons, and it increases the beam quality of the existence of

neutrons in energy beyond the epithermal spectrum range.

V.2.4. Gamma-ray Shielding

The effects of Bi addition in the collimator are shown in Figure 5.7.

Figure 5.7. Gamma-ray component for various thickness of shielding (Bi).

The -ray component is reduced exponentially by using Bi. With thickness of 2 cm, the -ray component remains 1.44 x 10-13 Gy.cm2.n-1. The addition for more thickness will, of course, decrease the -ray component. 4 and 6 cm thick of Bi results in 0.79 x 10-13 and 0.40 x 10-13 Gy.cm2.n-1 -ray components, respectively. Unfortunately, as Bi made thicker, the fast neutron component increases, as

shown in Table 5.3.

Table 5.3. Results of -ray shielding (Bi) thickness variations.

Shielding Thickness (cm)

epi(x 108 n.cm-2.s-1)

Df / epi(x 10-13 Gy.cm2.n-1)

D / epi(x 10-13 Gy.cm2.n-1)

0 9.99 1.70 2.97

2 7.48 1.80 1.44

4 5.95 1.90 0.79

6 4.88 2.02 0.40

0

0.5

1

1.5

2

2.5

3

3.5

4

0 1 2 3 4 5 6 7

/

epi (

x 10

-13

Gy.

cm2 .

n-1

)

Shielding thickness (cm)

39

Figure 5.8 shows that its total cross section declines for energy above 1

MeV. Hence, Bi undergoes more interactions with neutrons in 1 MeV and lower.

It leads to the increment of higher-energy neutrons components which is highly

avoided. Thus 2 cm thick of Bi is used rather than 4 or 6 cm. With 2 cm thick of

Bi, the epithermal neutron flux decrease to 7.48 x 108 n.cm-2.s-1.

Figure 5.8. Total cross sections of Bi.

Reference: [19]

V.2.5. Aperture

Diameter of aperture was altered in 1, 2, 3, 4, and 5 cm. The results are

collected in Table 5.4. Data in Table 5.4 show that, generally, the aperture size

apparently does not cause any certain effect to the beam. Almost all parameters

show fluctuating results.

40

Table 5.4. Results of beam characteristics for different aperture diameter.

Aperture diameter (cm) 1 2 3 4 5

epi (n.cm-2.s-1) 7.55x108 7.61x108 7.48x108 7.65x108 7.57x108Df / epi (Gy.cm2.n-1) 1.80x10-13 1.85x10-13 1.80x10-13 1.76x10-13 1.81x10-13D / epi (Gy.cm2.n-1) 1.47x10-13 1.45x10-13 1.44x10-13 1.34x10-13 1.32x10-13th / epi 0.010 0.010 0.009 0.009 0.008J / epi 0.72 0.72 0.73 0.72 0.73

This collimator design does not fully pass the IAEAs criteria, since the

epithermal neutron flux is always below the recommended value of 1.0 x 109

n.cm-2.s-1. Nonetheless, the beam is still usable with epithermal neutron flux

exceeding 5.0 x 108 n.cm-2.s-1.

V.2.6. Environment Surrounding the Collimator

During the designing process, the environment surrounding the collimator

was neglected, so that the results depicted the collimators single performance.

This is a kind of prevention to the dependency of the collimator design to the

environment.

When it was assumed that the graphite inside the thermal column was not

discharged but only the part which is going to be replaced by the collimator, the

performance of the collimator became better, as depicted in Table 5.5.

Table 5.5. Results of beam characteristics for different aperture diameter of graphite-surrounded collimator.

Aperture diameter (cm) 1 2 3 4 5

epi (n.cm-2.s-1) 1.60x109 1.63x109 1.64x109 1.68x109 1.65x109Df / epi (Gy.cm2.n-1) 1.56x10-13 1.69x10-13 1.61x10-13 1.61x10-13 1.59x10-13D / epi (Gy.cm2.n-1) 1.25x10-13 1.18x10-13 1.24x10-13 1.26x10-13 1.16x10-13th / epi 0.006 0.007 0.007 0.007 0.007J / epi 0.73 0.73 0.72 0.72 0.72

With the graphite thickness of about 8 cm, the epithermal neutron flux increases

dramatically up to 1.68 x 109 n.cm-2.s-1 which is exceeding the recommended

41

value of 1.0 x 109 n.cm-2.s-1, accompanied by relatively better beam quality. The

graphite is, in fact, also reflects more neutrons into the collimator; the same role

as collimator wall. IAEA does not recommend graphite to be used as a material

for collimator wall since it has low atomic weight that will cause energy drop to

the neutrons. This is unacceptable since the desired distance from the reactor core

to the treated patient is as far as possible, and graphite usage would make the

neutrons lose most of its energy as they undergo some collisions until then reach

the outlet of collimator. As a rough estimation, with the same collimator length,

thermal neutrons might dominate the other energy spectrums of neutrons. In this

case, graphite was considered only as the environment (hence it was not included

in the designing process) which gave positive effect to the collimator whenever it

did really exist and was not being neglected during the simulations. Table 5.5

shows that graphite contributes to reflect more neutrons. Some neutrons leak from

the collimator would then interact with the graphite which located exactly outside

the collimator, and reflected back. Those neutrons reflected by the graphite mostly

are high energy neutrons that they do not interact with the collimator wall, Ni,

hence they have longer free path than the others. Graphite, since it has low atomic

mass, will decrease the neutrons energy more than the natural nickel did.

Neutrons which just scattered with the graphite could be back into the collimator,

meanwhile the rest leak to the outer side of the thermal column. This results in the

enhanced epithermal neutron beam intensity and thus its quality, generally, and

also passes all the IAEAs criteria. Moreover, it is also possible to prolong the

collimator length to minimize the unwanted radiation from the core which may

still be able to penetrate through the wall.

CHAPTER VI

CONCLUSION AND RECOMMENDATION

VI.1. Conclusion

A conceptual design of collimator which is proper to be implanted in the

thermal column of Kartini Research Reactor has been made. It consists of:

1. 5 cm thick of Ni, as collimator wal