Table of contents

PREFACE

All praises to God the Al-Mighty for giving us the strengths and the His blessing in making Nuclear Science, Technology and Engineering Conference 2018 (iNuSTEC2018) a success.

The history in organizing Nuclear Science, Technology and Engineering Conference started back in 2010. During that year Then, it was called NuSTEC2010 and was organized only at national and regional levels. From that year onwards, we succeeded in organizing our second (iNuSTEC2011), third (iNuSTEC2012), fourth (international iNUSTEC2013), fifth (NuSTEC2014), sixth (international iNuSTEC2015), seventh (international iNuSTEC2016) and eighth (international iNuSTEC2017) conferences. Based on these achievements, we have decided to reconvene the conference sustainability by organizing regional/international iNuSTEC on annual basis.

The iNuSTEC2018 was held from 23-25 November 2018 at Universiti Teknologi Malaysia (UTM) in Skudai, Johor, Malaysia. The conference continued the tradition of continuous effort by Malaysian Nuclear society to bring together the nuclear communities in an annual event to discuss and to discourse on progress in nuclear researches and related fields. The conference particularly encouraged the interaction of research students, research institutions and academics with the more established community. iNuSTEC2018 also played an essential role in introducing and fostering new research talents and in advancing the future profile of nuclear science, technology and engineering regionally and globally.

List of Message from: President, Malaysian Nuclear Society (MNS), Contents, Proceeding Editorial Committee, iNuSTEC2018 Organising Committee, Malaysia Nuclear Society, International Advisory Committee, National Advisory Committee, Scientific and Publication, Nuclear Engineering Student Society (NESS) Secretariat, Acknowledgement, MNS Council Members 2017/2019 and MNS Chapters are available in this pdf.

Cover page is available in this pdf.

All papers published in this volume of IOP Conference Series: Materials Science and Engineering have been peer reviewed through processes administered by the proceedings Editors. Reviews were conducted by expert referees to the professional and scientific standards expected of a proceedings journal published by IOP Publishing.

Spent fuel pool uses water to temporarily cool spent nuclear fuels before final disposal. The Malaysian Nuclear Agency is recently working on its spent fuel pool facility to store irradiated TRIGA fuel rods. These irradiated fuels still contain significant amount of fissile material which are capable to induce criticality in the fuel storage facility. This study is therefore performed to investigate the safety criticality analysis for the proposed fuel storage rack design of 5x5 arrays. MCNP5 code was used to model the 5x5 array of fuel storage rack in order to determine the optimal fuel pitch distance. The analysis was based on the standard UZrH1.6 TRIGA fuel cladded with stainless steel used in PUSPATI TRIGA Reactor, which is 19.9% enriched 235U and 20% weight of Uranium. By varying the pitch distance between 4 cm to 14 cm with no neutron absorber included, the optimum fuel rack design was determined where sub-criticality condition can be maintained.

Fuel Temperature reactivity coefficient is one of the important operational and safety parameters in nuclear reactors to monitor for normal operation and transient safety analysis in research as well as in power reactors. The aim of this work is to estimate the temperature reactivity coefficient by using adaptive neuro-fuzzy inference system (ANFIS) as a preliminary assessment for developing an online monitoring system in research reactors such as the Reactor TRIGA PUSPATI (RTP). The input for the learning mechanism is the fuel temperature recorded at Ch A of the instrumented fuel element (IFE) and control rod positions with the fuel temperature reactivity coefficient as the targeted output. The ANFIS model was validated using testing data that consisted of fuel temperature measured at Ch B and results on -0.00571 $°C⁻¹ (-3.997 pcm°C⁻¹) of actual output and -0.00592 $°C⁻¹ (4.144 pcm°C⁻¹) of predicted output. Through the developed model, it showed that the ANFIS also can be used as a tool to estimate the fuel temperature reactivity coefficient with good prediction and accuracy shown from the calculation of mean absolute error (MAE) of 0.00349.

This paper presents the performance analysis of two random sampling algorithms, the inverse-transform method and the Vose aliasing method, on GNU Octave. The Monte Carlo code MCS developed by UNIST uses random sampling methods to simulate the physics of neutron and photon transport [1]. The goal is to optimize the sampling time of MCS for cases when the probability density function is a constant function throughout the simulation. For this purpose, the runtime of the inverse-transform method and Vose aliasing method are compared for increasing input size with scripts developed on GNU Octave. To compare the execution time, the initialization and generation time of both methods are determined and discussed.

An assumption system with a one-body model of spent nuclear fuel (SNF) assemblies was developed to predict the thermal distribution in the spent fuel pool (SFP) in the absence of external cooling system. It was based on the three-dimensional (3D) thermal hydraulic behaviour computed using the computational fluid dynamic (CFD) software, Fluent 18.0. This study aims to determine the applicability of the one-body model (average decay heat) by comparing it with the individuals-body model (variation of decay heat) computation results. The thermal distribution for both models were compared and evaluated, neglecting other effects such as position of the SNF assemblies, decay heat, and axial heat flux profile. It was found that the one-body model was applicable to predict the thermal distribution of the SFP during loss of active cooling system with error less than 1°C.

Conventional nuclear power reactors convert between 30 – 35 % only of the total energy input into electricity while the remaining was wasted. The waste heat is sometimes used for desalination processes while the remaining heat is transferred to a cooling media or lost to the surrounding. Therefore, heat from nuclear reactor can be used to produce heat and power, as well as for cooling in a trigeneration system. This paper presents the Trigeneration System Cascade Analysis (TriGenSCA) for an optimal Pressurised Water Reactor (PWR) design. The TriGenSCA framework allows engineers to determine an optimum utility generation system size and estimate the required amount of external utilities. The analysis includes data extraction, cascade analysis for size estimation and calculation of the new trigeneration system size. The technique also enables users to determine accurate results for energy minimisation based on demand fluctuations. Application of the framework on a case study presented on this paper demonstrates the trigeneration PWR system successfully saved energy of 328GWh/y (97 %).

In the beginning of the 21^st century, several research institutions and also private companies had proposed various design of thorium-based nuclear reactor, ranging from solid fuel to molten fuel, fast and thermal neutron spectrum and also various path of waste management. This paper studies 10 of the proposed reactor designs by 10 different organizations, three key aspects analysed quantitatively namely price per kilowatt, safety features and spent fuel managements. Corresponding factors contributing to the key aspects mentioned above were gathered, weighted based on evidence available and analysed using decision matrix. Based on the information collected, preliminary ranking were constructed based on trends between various factors.

This paper discusses the overview of the development of TRIMON, a next-generation core management code for TRIGA reactors. The newly developed code employs Monte Carlo method which uses diffusion theory type homogenized neutron cross section data, thus, speeding up the stochastic simulation via simplification of a complex core geometry. It has the capability of running three-dimensional core simulation up to 16 neutron energy groups. In general, the group neutron cross sections are pre-processed and homogenized via the Effective Diffusion Homogenization (EDH) method. The important features of TRIMON include core eigenvalue calculation, fuel burnup management, three-dimensional power and neutron flux distribution prediction. In this paper, the numerical results obtained using TRIMON were compared with the actual measured operational data. The performance comparison was also made between TRIMON and MCNP.

Radiation shielding is a body of material that is placed between a radiation source and an object to be protected with the aim of reducing the intensity of radiation at the object's location. It can be made from various materials. These materials can be stacked into a multilayer shield or they can be mixed into a composite shield. The main objective of the present study is to review the list of multilayer shield combinations that have been studied and to highlight the findings on material arrangement and consequent buildup factor. The scope of the study is limited to the results of the performed studies. It was observed that there was no clear method on arranging the layer. Buildup factor was also found to be complicated in multilayer shields. Future studies may focus on new multilayer shielding design with unlisted materials, complementary buildup calculations, and applications of metaheuristics in shielding optimization.

A couple of elevated concentrations of noble gas Xenon-133 were detected at the Takasaki Radionuclide Monitoring Station (JPX38) on 2 October 2017. Xenon-133 is among four relevant radioxenon isotopes that are used for monitoring and verification of nuclear explosions including Xenon-13, Xenon-133m and Xenon-131m. These radioxenon gases could be emitted either from nuclear explosion or civil nuclear facilities including medical isotopes production facilities, nuclear power reactors and nuclear research reactors. This paper presents the Malaysian CTBT National Data Centre (MY-NDC) findings in identifying the possible source regions for elevated concentrations of Xenon-133 detected at JPX38 on 2 October 2017. The findings will be useful in understanding the possible source of release for such Xenon-133 detection, whether coming from a late release of the 3 September 2017 North Korea nuclear test site or from civil nuclear facilities.

The Comprehensive Nuclear-Test-Ban Treaty (CTBT) is briefly characterized as a global arms control and disarmament initiative, which complements the goal of the Non-Proliferation Treaty (NPT). The verification regime of the CTBT is designed to monitor countries' compliance with the CTBT by detecting any nuclear explosion conducted on Earth - underground, underwater or in the atmosphere. This paper briefly goes through the International Monitoring System (IMS) of CTBT's verification regime that uses four different technologies - seismic, hydroacoustic, infrasound and radionuclide - to monitor the planet for nuclear explosions. It pays particular attention to the seismology monitoring techniques of CTBT IMS, including the practical steps in discriminating underground explosions and earthquakes using data mainly from the IMS seismic network. It also highlights the seismic analyses performed by the Malaysian CTBT National Data Centre (MY-NDC) that in principle, applies the practical steps of CTBT nuclear explosion seismology monitoring. In summary underground explosions produced seismic waves with unique characteristics which allowed the discrimination between explosions and earthquakes.

Biochar (a by-product of biomass pyrolysis) is developed as a renewable, low-cost, and promising carbon-based electrode for electrochemical remediation of Cr, Mn and Fe in polluted river water. The biochar - electrode were prepared with two types of tailored sieved size medium (2.0 mm) and fine (150 µm). The effect of the different sieved size on the electrochemical reactor was set at 60V of voltage supply, 6 cm inter-electrode distance within 105 minutes electrolysis operational time. The results indicated that the highest removal efficiency for Cr was achieved at 95% after 105 minutes followed by Mn at 99.9% in 65 minutes and Fe at 92.76% in 105 minutes. Interestingly, the electrode with larger sieved size showed the highest removal efficiency for all three elements. These results emphasize the importance of tailoring the porous structure of the biochar electrode material as a function of the specific size of adsorbate ions to improve the electrochemical water treatment performance for polluted river water.

Ground Penetrating Radar (GPR) is a high resolution electromagnetic techniques that is designed primarily to investigate the shallow subsurface of the earth, building material, roads and bridges. GPR was also able to detect water leaks in the underground distribution system. A series of laboratory experiments were conducted to determine the validity and effectives of GPR technology in detecting water leakage using two different types of pipes which are metal and polyvinyl chloride (PVC) pipes. Experiment was conducted using GPR 800 MHz antenna using 4 inch pipe with diameter of hole is ¼ inch to stimulate the leakage. GPR identify leaks in buried water pipes either by detecting underground voids created by the leaking water as it erodes the material around the pipe or by detecting anomalous change in the properties of the materials around pipes due to the water saturation.

Fly ash is a solid waste from a coal combustion product and about 43% is recycled. Fly ash is often used as a replacement of partial replacement for Portland cement in concrete production. Concrete has been widely used as radiation materials due to its lower cost. Five samples of fly ash concrete mixture were prepared and investigated by using small angle scattering (SAN) for neutron irradiation and gamma radiation test for gamma irradiation. It was found that 20% of fly ash concrete mixture absorbed more neutron radiation whereas 40% of flyash concrete mixture absorbed more gamma radiation.

One of the primary concerns in the operation of nuclear related facilities is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Thus, investigation and analysis on the safety system shall be continuously conducted to determine the safety condition of the facilities. In the early days, analytical and experimental methods were employed. However, this approach is limited to low risk experimentation. With the advancement of computer technology, specific computer codes were used to predict and analyse high risk safety issues such as the fall-out from a nuclear explosion or analysis of melt down occurrence in a reactor core. This paper discusses a prediction model on a gamma irradiation facility source pit which having failure on its air cooling system. The CFD model predicts the temperature of the gamma sources under such situation to determine its integrity.

This paper contains the research done concerning natural rubber. This included its origin, economic importance, and how the functional groups in the natural rubber reacted with selected actinides. The selected actinides discussed were Thorium, Uranium and Plutonium. The information was retrieved from different sources including information from laboratory experiments. The results obtained were gathered when conducted under different conditions. The conditions varied based on the temperatures and the pH. The functional groups exhibited different outcomes when reacted with different actinides under different conditions. The functional groups namely amine carboxyl and hydroxyl groups have different efficiency and capacity of adsorption. Besides economic importance, the environmental significance of rubber also has been discussed.

Neutron diffraction (ND) is an application of non-destructive test which involves the process of interference between neutron and atoms within materials. This technique has drawn many attentions especially in industrial, metallurgical and nuclear sectors. Neutron diffractometer system (NDS) is a system which includes all the instrumentation to perform ND technique. One of the most important instruments in NDS set-up is neutron collimator which is used to reduce the size and steer the direction of the beam. Generally, neutron collimator used in NDS is based on step convergent collimator type. Currently, the NDS facility is planned to be built at one of the radial beam port of TRIGA MARK II PUSPATI research reactor (RTP) in Malaysia. The aim of this research is to characterize the materials suitable for neutron collimator and to obtain the ideal geometry for neutron collimator design. In order to achieve these aims, several neutron collimators with different geometries have been designed using Monte Carlo Simulation Codes MCNPX. The results are then compared to obtain the best design with high thermal neutron flux and low gamma contamination at the object plane. This research may be useful in the selection of neutron collimator design for NDS facility at TRIGA research reactor.

Boron Neutron Capture Therapy (BNCT) is a promising method to cure cancer by using interaction between boron compound and the slow neutron beam obtained from nuclear power reactor, nuclear research reactor and neutron generator. Based on the previous feasibility study shows that the thermal column can generate higher thermal neutrons for BNCT application at the TRIGA MARK II reactor. Currently, the facility for BNCT are planned to be developed at thermal column. Thus, the main objective was focused on the optimization of thermal neutron and epithermal neutron flux at the thermal column by designing collimator. This paper briefly goes through the characterization of collimator for the ideal neutron collimator design. Several collimators with variable geometries have been designed using Monte Carlo Simulation Codes of MCNPX. The results are then compared in term of thermal neutron, epithermal neutron and gamma fluxes yielded and the ideal collimator then optimize based on aperture size, shielding thickness and collimator condition This research is useful in the selection of neutron collimator design BNCT facility at the TRIGA MARK II reactor.

Characterization of neutron beam compositions is important in order to capture great radiographic images. The quality of the neutron beam has significant impacts to the radiographic images produced using that beam. Characterizing of the beam, including determination of the effective value for the thermal neutron content, scattered neutron content, gamma content, and pair production content as well as beam sensitivity, is vital for producing quality radiographic images. This paper provides a characterization of the neutron radiography's beam port (beam port 3) at TRIGA Mark II PUSPATI research reactor (RTP). The experiments to determine the effective beam contents and neutron radiograph sensitivity at the beam are based on American Society for Testing and Materials (ASTM) standards document E545. The Beam Purity Indicator (BPI) and Sensitivity Indicator (SI) are the standard used in these experiments. Neutron radiographic imaging of this standard was performed using direct radiographic method with 0.1 mm thickness Gadolinium (Gd) metal converter and film. Characterization of the beam was carried out for 10, 15, 20, 25 and 30 minutes exposure time at 750 kW reactor operating power. The standard was placed at 110 cm from the beam opening. The current neutron beam performances are compared with previous results.

The Prototype and Plant Development Center (PDC) of Malaysian Nuclear Agency is developing a hot cell facility which is meant for research activities such as spent fuel inspection, fuel fabrication and post irradiation material examination/inspection. The hot cell is of alpha-gamma shielding hot cell type. The size of the hot cell is 4.4 x 3.9 x 3.825 meter. The 1-meter thickness hot cell wall consists of 89 pieces of interlocking curve design concrete. The 1-meter hot cell roof consists of 41 pieces of square shape concrete. The density of all concrete blocks is 2350kg/m³. The hot cell is designed to handle up to 227.5 Ci ⁶⁰Co activity. This paper will discuss the detailed design of the shielding wall, as well as the shielding evaluation/analysis against gamma and fission products of 50% burnup RTP 20wt% fuel type using MCNP.

Particle and Heavy Ion Transport Code System (PHITS) is a Monte Carlo simulation program used for transport and collision of nearly all particles over wide energy range. One of its useful features for accelerator facility design is the calculation of dose distribution. This paper describes how the geometry of the bunker for low electron beam accelerator in Nuklear Malaysia (NM) was modelled and used to calculate the equivalent dose caused by the radiation for specific region in the bunker by using PHITS. The dose distribution and energy deposition were obtained. Finally, the simulation results are compared to the measured value from the dose measurement testing inside the bunker facility and found both results are in a good agreement.

This paper discusses the design and development of a portable PC-based Automated Ultrasonic Testing (AUT) system, which is capable of classifying defect types on a welded metal components in real-time. Three types of welding defects such as a crack, slag inclusion and cluster porosity plus the normal condition (no defect) were used as the test states for the classification process. The system utilizes an ultrasonic pulse-receiver card model attached to computer notebook for signal display. A new specific software package (RUSS) was developed to control the operation of the scanner, displaying the data, extract the signal defect and classify type of defect. The proposed algorithm and complete system were implemented in a computer software developed using Microsoft Visual BASIC 6.0. The classification process NNTool in MATLAB component was manipulated by using the ActiveX. Some result shows the successful of the calibration module and real – time extraction for classification module type of defect

This paper describes the study and development of the filament power supply. This locally designed low energy electron accelerator with the present energy of 140 keV will be upgraded to 300 keV. The filaments are used to heating-up and produce the electron. The produced electron will be accelerated through a potential different by a DC High Voltage Power Supply (DCHVPS). Whereby this filament power supply is connected with the DCHVPS, therefore the isolation of high voltage is required. For such purpose, the specifications and the isolation mechanisms of the filament power supply have been identified. As the result, the filament power supply by using an air core transformer has been selected and developed with the required heating power.

In this paper, the effect of contact and non-contact antenna system of ground penetrating radar (GPR) for buried pipes are presented. The frequency for GPR antenna used in this project is 400MHz which is suitable for GPR system to detect buried pipes for the tested surfaces, since the soil involved is sand which has low electrical conductivity. The results of the radargram obtained from this experiment and the limitations occurred are discussed.

Gamma and neutron irradiation effect on sand and activated carbon were studied in this paper intensively and, were found from various studies to have strong correlation between neutron fluence or gamma energy on physical and mechanical damage of materials. Sand and activated carbon were irradiated with neutron at Reactor TRIGA PUSPATI (RTP) and gamma at Sinagamma Malaysia Nuclear Agency (Nuklear Malaysia). Morphological analysis was carried out on the irradiated elements using the SEM, XRD and SAXS which showed changes on the microstructure of irradiated samples. Irradiation of activated carbon showed increase in the numbers of pores while change of textural profile of the surface take place at gamma irradiated sand. XRD pattern graph did not indicate any changes, however the specific surface area for both irradiation materials decreased.

In several studies on the impact of radiation towards superconductor, there were evidences that changes may occur in superconducting properties due to alteration in lattice structure of HTS materials. In this work, the solid-state reaction method was employed in preparing the bulk samples of Bi_1.6 Pb_0.4Sr₂Ca₂Cu₃O₁₀ (hereafter Bi-2223) superconductor. The samples were then irradiated with gamma and electron sources with dosage of 100 kGray each. X-ray powder diffraction (XRD) patterns were used to determine the phases that present in the samples. Results of XRD patterns for both the non-irradiated and irradiated samples showed well-defined peaks all of which could be indexed based on the Bi-2223 phase structure. Besides, the XRD patterns indicate that gamma and electron irradiations did not affect the Bi-2223 superconducting phase structure. Measurements of the critical temperature, T_C and the critical current density, J_C of all the samples were conducted before and after irradiation using AC susceptibility technique. T_C for non-irradiated, electron irradiated and gamma irradiated sample is 97 K, 95 K and 103 K, respectively. On the other hand the J_C for non-irradiated, electron irradiated and gamma irradiated sample is 1618 A/cm² 1729 A/cm² and 1549 A/cm² respectively. Results of scanning electron microscopy (SEM) micrographs show improvement in the grain growth and disorientation in the texture of all the gamma and electron irradiated samples.

To enhance the crosslinking of prevulcanized natural rubber latex, combination of Irradiation and peroxide vulcanizations were used. Through this method, hexanediol diacrylate (HDDA) from irradiation vulcanization acted as the main sensitizer, while cumene hydroperoxide (CHPO) from peroxide vulcanization would act as the co-sensitizer. The effects of irradiation doses on the mechanical properties of latex film were investigated. 16 kGy irradiation dose, 2.5 parts perhundred rubber (pphr) of HDDA, 0.1 pphr of CHPO and 2.5 phr of Aquanox LP antioxidant were found to be the optimum conditions for compounding formulation. The rubber film obtained had tensile strength, modulus at 500% and modulus at 700% of 27.7, 3.5 and 12.4 MPa respectively, which is more than 21% increment compared to control film. Besides, the crosslink percentage of the rubber film showed 7 % increment from 90.7% to 97.7%.

Adhesive property could be described as an interchangeably with some ink and substrate which was applied to one surface of two separate items that bonded together. Lanthanum oxide (La₂O₃) had been used as a rare earth metal candidate as printing ink. This metal deposit was printed on carbon-silicon substrate and carbon-glass substrate using Magnetron Sputtering technique. The choose of Lanthanum Oxide as a target is due to its wide application in producing micro-electronic devices such as thin film battery, electronic circuit and printed circuit board. The La₂O₃ was deposited on the surface of carbon-silicon and carbon-glass substrate were then analyzed using Angle Resolve X-Ray Photoelectron Spectroscopy (ARXPS). The position for each synthetic component in the narrow scan of Lanthanum (La) 3d and O 1s were referred to the electron binding energy (eV). This research was focused on the narrow scan regions which were O 1s and La 3d. Further discussion of the spectrum evaluation was discussed in detail. Here, it was proposed that from the adhesive and surface chemical properties of La was suitable as an alternative medium for micro-flexographic printing technique in printing multiple fine solid lines image at micro to nano scale feature. Micro-flexographic printing was developed in patterning technique from micron to nano scale range to be used for graphic, electronic and bio-medical device on variable substrates. Hence, this paper will describe the capability of this particular metal as rare earth metal in a practice of micro-flexographic printing process.

The measurements of mass attenuation coefficients by using Compton scattering technique was carried out by using the Ludlum configuration. A ¹³⁷Cs sealed sourced was used and attenuated at angles between 30 and 75° to provide scattered gamma energies between 337.72 and 564.09 keV. The mass attenuation coefficients of solid water and perspex were measured and compared to the calculated value of water by using XCOM. The results showed that the measured mass attenuation coefficients of solid water and Perspex® phantoms were in agreement to the values of water within 6.84 and 7.20% average percentage of discrepancies. The results indicated the suitability of the Ludlum configuration for the measurement of mass attenuation coefficients by using the Compton scattering method.

The mass attenuation coefficients of the fabricated Rhizophora spp. particleboards were measured by using the Compton scattering technique and Ludlum configurations. The mass attenuation coefficents of the particleboards were measured at scattered gamma energies of ¹³⁷Cs between 337.72 and 564.09 keV and compared to the its theoretical values and the values of water by using XCOM calculation. The results showed that the measured mass attenuation coefficients of the fabricated Rhizophora spp. particleboards were in good agreement to its theoretical values and water (XCOM) within 3.8 and 5.9% discrepancies. The results showed the near attenuation properties of the fabricated Rhizophora spp. particleboards to water at scattered gamma energies and the suitability of the Ludlum configuration for the measurement of mass attenuation coefficients of materials.

An X-ray fluorescent (XRF) configuration was used to measure the mass attenuation coefficient of materials at low energy photons. An Americium-241 (²⁴¹Am) annular source was used in conjunction with niobium, palladium, molybdenum and tin plates that provided K_α1 photons with energies between 16.59 and 25.26 keV. The transmitted photons were measured by using a low energy Germanium (LEGe) detector and analyzed by using MAESTRO software. A calibration with ²⁴¹Am showed good linearity between photon energies and peak channels with R² value near to 1. The measurement of mass attenuation coefficient was carried out on Al plates and Perspex based on the transmitted photons on the samples. The measured mass attenuation coefficients were compared to the XCOM calculations. The results showed that the measured mass attenuation coefficients of aluminum and Perspex® were in good agreement to their XCOM values within 10% percentage of discrepancies at all experimented photon energies. The overall results indicated the suitability of the XRF configuration for the measurement of mass attenuation coefficients at low energy photons.

The aim of this study was to evaluate the performance of corn starch-bonded Rhizophora spp. particleboards as phantom for SPECT/CT including the image contrast, source dimension ratio and recovery coefficient values. Phantom set made of the Rhizophora spp. particleboards was constructed according to the Jaszczak phantom commonly used in SPECT/CT with external dimension of 22 and 18 cm diameter and length respectively. Cylindrical vials with different diameter sizes filled with ^99mTc unsealed source were inserted in drilled holes of the constructed particleboards phantom. The particleboards phantom was scanned by using SPECT/CT scanning mode and the SPECT images were reconstructed using clinical protocol. The results showed that the contrast value, source dimension ratio and recovery coefficient of Rhizophora spp. particleboards in good agreement with Jaszczak phantom. The overall results showed an excellent agreement of performance between corn starch-bonded Rhizophora spp. particleboards to the Jaszczak (water) as standard phantom material. The overall results indicated that the corn starch-bonded Rhizophora spp. can be highly recommended to be used as tissue-equivalent phantom material for Quality Assurance and dosimetry works in SPECT/CT imaging.

The development of nuclear medicine has led to an entire subset of extremely effective strategies in diagnosis and disease therapy. The emerging advancements in nuclear medicine are vast and diverse. Nuclear medicine tests and treatments involve the use of small amounts of radioactive material in order to diagnose and determine the severity of a variety of conditions and diseases. Basically, nuclear medicine involves the use of the latest imaging facilities such as magnetic resonance imaging, computed tomography scans and positron emission tomography which will find in many modern medical institutions today. One significant advantage of using nuclear medicine is the ability to make complex medical procedures simpler and safer for patients. With this technology, physicians and medical professionals are able to examine in great detail the most sensitive areas. All examinations can happen without subjecting patients to possibly dangerous and invasive procedures and surgeries. This medical technology has enabled the development of treatment options for patients suffering from serious illnesses, such as cancer either through radiation or chemotherapy. Health care professionals now face many challenges in their profession such as new rules and the latest practices that enable them to learn new skills and new ways of practicing their daily tasks in nuclear medicine areas. As a result, they are confronted with new skills and new ways of practicing in nuclear medicine area every day.