A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results.
The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II REACTOR were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06E8 n/cm2/s. According to IAEA (2001) flux of 1.00E9 n/cm2/s requires three hours of treatment. Few modifications were needed to get higher flux.
An electronic database has been developed and implemented for ko-INAA method in Malaysia. Databases are often developed according to national requirements. This database contains constant nuclear data for ko-INAA method; Hogdahl-convention and Westcott-formalism as 3 separate command user interfaces. It has been created using Microsoft Access 2007 under a Windows operating system. This database saves time and the quality of results can be assured when the calculation of neutron flux parameters and concentration of elements by ko-INAA method are utilised. An evaluation of the database was conducted by IAEA Soil7 where the results published which showed a high level of consistency.
Determination of thermal to fast neutron flux ratio (f(fast)) and fast neutron flux (ϕ(fast)) is required for fast neutron reactions, fast neutron activation analysis, and for correcting interference reactions. The f(fast) and subsequently ϕ(fast) were determined using the absolute method. The f(fast) ranged from 48 to 155, and the ϕ(fast) was found in the range 1.03×10(10)-4.89×10(10) n cm(-2) s(-1). These values indicate an acceptable conformity and applicable for installation of the fast neutron facility at the MNA research reactor.
A bulk of used paper supplied to recycling industry may contain water in their internal voids. This is because the price of the used paper is currently based on their weight and it has a huge potential of suppliers to add with water in order to increase the price. Currently used methods for detecting moisture content in a paper are restricted to a sheet of paper only. This paper presents a non-intrusive method for quick and in-situ measurement of water content in a bulk of used paper. The proposed method extends the capability of common paper moisture gauge, by using a neutron device. A fast neutron source (Am-Be 241) and a portable backscattering neutron detector are used for water measurement. It theoretically indicates that the slow neutron counts can be correlated to the hydrogen or water level in a paper. The method has the potential of being used by the paper-recycling industry for rapid and non-destructive measurement of water in a bulk of used paper.
Paper recycling plants usually buy their raw material from suppliers. More than often, bulk used paper supplied to the plant contains some significant quantity of water in its internal voids. It may be included intentionally or unintentionally. The price of used paper depends on its weight, thus adding water will help to increase weight and consequently increase the price. In this way, plant owner who purchase the used paper suffers a significant of financial lost. The objectives of our experiment are to establish a calibration curve that correlate between the amount of neutron backscattered and water content, and finally to develop a correction factor that need to be introduced to the measured values of water content. A fast neutron source (Am-Be 241) and a portable backscattering neutron detector were used for water measurement. The experiments were carried out by measuring neutron backscattering from used paper that has been added with different amount of water. As a result, a neutron calibration curve that provides a correlation between neutron backscattering and water content was established.
This work main aim is to study the analysis of slow neutrons which include thermal and
epithermal neutrons and also analysis on fast neutrons. The outcome from this work showed that
the comparison result between fast and slow neutrons. The safety assessment at reactor TRIGA
FUSFATI (RTF) is one of the main objectives of the work and there is a detailed discussion on it
which helped in accomplishing the task. Gamma Rays produced in this experiment was high and in
the experiment and it is realized that the shielding plays a vital role in the success of this
experiment which prevents all the radiations. From the results of the experiment it is realized that
these gamma rays are not suitable for the application of Boron Neutron Capture Therapy
(BNCT). However, these radiations are suitable for the application of Neutron Radiography (NR).
The study on this work will help in study of nuclear applications such as BNCT, NR, SANS etc.
These applications are using in medical and nuclear fields. The electronic device used in the
experiment to detect neutron is Neutron Spectrometer. The results from Neutron Spectrometer
and TLDs are very similar which showed that the experiment is a success. Numerical results were
compared with those available in literature for validation.
In industrial plants such as electricity generating, petroleum, chemical and petrochemical plants, pipelines are used extensively to transport liquid from one location to another. In radiation vulcanization of natural rubber latex (RVNRL) plants, pipelines are also used to transport latex to storage tank. During one of its maintenance activities, a pipeline intelligent gauge (PIG) that was used to monitor pipe integrity jammed inside the pipe causing interruption to its operation or loading activities. Sealed source technology was utilized to determine the location of jammed PIG in the pipeline. Fast neutrons from a 50 mCi Americium Beryllium (AmBe241), with energy range between 0.5 to11 MeV, were used for the study. Helium 3 (He3) detector was used to detect slow neutrons having a range of energy of 30 eV- 0.5 MeV. The investigation was carried out using neutron backscatter technique scanner. By adopting back-scattered technique, the location of jammed PIG in the pipeline has been successfully determined.
Boron niride microflakes of 2-5 μm in diameter and greater than 40 μm in length with multilayer structure and highly crystalline nature are synthesized in two states of catalysts and dual role of nitrogen at 1100 °C. Most of the microflakes are flat, smooth and vertically aligned with a wall-like view from the top. Transmission electron microscopy shows overlapped layers of microflakes with an interlayer spacing of 0.34 nm. The h-BN components of the synthesized microflakes are verified from B 1s and N1 s peaks at 190. 7 and 397.9 eV. Raman shift at 1370 (cm(-1)) and sharp peaks in the XRD pattern further confirm the h-BN phase and crystalline nature of the synthesized microflakes. Microflakes of h-BN with the above characteristics are highly desirable for the development of a solid state neutron detector with higher detection efficiency.
Non-destructive and real time method becomes a well-liked method to researchers in the oil palm
industry since 2000. This method has the ability to detect oil content in order to increase the
production of oil palm for better profit. Hence, this research investigates the potential of neutron
source to estimate oil content in palm oil fruit since oil palm contains hydrogen with chemical
formula C55H96O6. For this paper, oil palm loose fruit was being used and divided into three
groups. These three groups are ripe, under-ripe and bruised fruit. A total of 21 loose fruit for each
group were collected from a private plantation in Malaysia. Each sample was scanned using
neutron backscattered technique. The higher neutron count, the more hydrogen content, and the
more oil content in palm oil fruit. The best correlation result came from the ripe fruits with r2=0.98.
This research proves that neutron backscattered technique can be used as a non-destructive and
real time grading system for palm oil.
The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial
criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of
basic nuclear research, manpower training, and production of radioisotopes. This
paperdescribes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP);
focusing on the application of the developed reactor 3D model for criticality calculation,
analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D
continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full
model of the TRIGA reactor. The consistency and accuracy of the developed RTP MCNP model
was established by comparing calculations to the experimental results and TRIGLAV
code.MCNP and TRIGLAV criticality prediction of the critical core loading are in a very good
agreement with the experimental results.Power peaking factor calculated with TRIGLAV are
systematically higher than the MCNP but the trends are the same.Depletion calculation by both
codes show differences especially at high burnup.The results are conservative and can be
applied to show the reliability of MCNP code and the model both for design and verification of
the reactor core, and future calculation of its neutronic parameters.
Boron carbide (B4C) is a ceramic material which is effective to absorb thermal neutron due to wide neutron absorption cross section. In this work, B4C is added into concrete as fine aggregates to test the attenuation properties by getting the attenuation coefficient of the concrete/B4C. The samples of concrete/B4C were exposing to the thermal neutron radiation source (241-Americium-Berylium) at the dos rate of 29.08 mR/h. The result show that the attenuation coefficient of the sample with 20wt% B4C is 0.299cm -1 and the sample without B4C is 0.238cm -1 and hence, concrete/B4C is suitable as a shield for thermal neutron radiation.
The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the calculation of neutron flux and power distribution in PUSPATI TRIGA REACTOR (RTP) 14th core configuration. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core and fuels. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data as well as S (α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed.
Muhammad Akmal Asraf Mohamad Sharom, Zainol Abidin Ibrahim, Wan Ahmad Tajuddin Wan Abdullah, Megat Harun Al Rashid Megat Ahmad, Faridah Mohamad Idris, Abdul Aziz Mohamed
Small angle neutron scattering (SANS) is used for probing the microstructure of materials in the range between 1-100 nm in dimension. The scattered neutrons from the target material were detected by a 128 x 128 array area sensitive, Helium gas-filled proportional counter, which is known as Position Sensitive Detector (PSD). The small angle neutron scattering (SANS) facility in Malaysian Nuclear Agency has been developed since 1995. The data acquisition system of this prototype facility consists of the two-dimensional Position Sensitive Detector (2D-PSD) and neutron monitor as a data grabber, TDC Histogram as a memory processing processor, two units of ORTEC 994 as a counter and timer and a computer as a data acquisition controller via GPIB interfacing protocol. This paper will describe on the development of GPIB interface for data acquisition of the SANS instrument on Windows based platform. The GPIB device interface and graphical user interface (GUI) for this data acquisition is developed using WaveMetrics Igor software.
L18 orthogonal array in mix level of Taguchi robust design method was carried out to optimize experimental conditions for the preparation of polymer blend composite. Tensile strength and neutron absorption of the composite were the properties of interest. Filler size, filler loading, ball mixing time and dispersion agent concentration were selected as parameters or factors which are expected to affect the composite properties. As a result of Taguchi analysis, filler loading was the most influencing parameter on the tensile strength and neutron absorption. The least influencing was ball-mixing time. The optimal conditions were determined by using mix-level Taguchi robust design method and a polymer composite with tensile strength of 6.33 MPa was successfully prepared. The composite was found to fully absorb thermal neutron flux of 1.04 x 105n/cm2/s with only 2 mm in thickness. In addition, the filler was also characterized by scanning electron microscopy (SEM) and elemental analysis (EDX).
Cement and concrete has been widely used as shielding material in reactor nuclear in order to minimize exposure to individuals. In this paper we present boron based concrete as neutron shielding for nuclear reactor applications. Concrete specimens with dimension of 10x10x10 cm were used and irradiated with neutron radiation of 252-californium. Characterization of physical, mechanical and radiation attenuation properties of concrete were carried out. The results show that the shielding performance is better than ordinary concrete. From the result, we confirmed that the performance of the concrete/boron carbide is suitable for practical use.
Molybdenum-99 is one of the most important radionuclides for medical diagnostics. In 2015, the International Atomic Energy Agency organized a round-robin exercise where the participants measured and calculated specific saturation activities achievable for the (98)Mo(n,γ)(99)Mo reaction. This reaction is of interest as a means to locally, and on a small scale, produce (99)Mo from natural molybdenum. The current paper summarises a set of experimental results and reviews the methodology for calculating the corresponding saturation activities. Activation by epithermal neutrons and also epithermal neutron self-shielding are found to be of high importance in this case.
Neutrons can be generated in medical linear accelerators (Linac) due to the interaction of high-energy photons (> 10 MeV) with the components of the accelerator head. The generated photoneutrons may penetrate the treatment room if a suitable neutron shield is not used. This causes a biological risk to the patient and occupational workers. The use of appropriate materials in the barriers surrounding the bunker may be effective in preventing the transmission of neutrons from the treatment room to the outside. In addition, neutrons are present in the treatment room due to leakage in the Linac's head. This study aims to reduce the transmission of neutrons from the treatment room by using graphene/hexagonal boron nitride (h-BN) metamaterial as a neutron shielding material. MCNPX code was used to model three layers of graphene/h-BN metamaterial around the target and other components of the linac, and to investigate its effect on the photon spectrum and photoneutrons. Results indicate that the first layer of a graphene/h-BN metamaterial shield around the target improves photon spectrum quality at low energies, whereas the second and third layers have no significant effect. Regarding neutrons, three layers of the metamaterial results in a 50% reduction in the number of neutrons in the air within the treatment room.
Theranostics in nuclear medicine refers to personalized patient management that involves targeted therapy and diagnostic imaging using a single or combination of radionuclide (s). The radionuclides emit both alpha (α) or beta (β-) particles and gamma (γ) rays which possess therapeutic and diagnostic capabilities, respectively. However, the production of these radionuclides often faces difficulties due to high cost, complexity of preparation methods and that the products are often sourced far from the healthcare facilities, hence losing activity due to radioactive decay during transportation. Subject to the availability of a nuclear reactor within an accessible distance from healthcare facilities, neutron activation is the most practical and cost-effective route to produce radionuclides suitable for theranostic purposes. Holmium-166 (166Ho), Lutetium-177 (177Lu), Rhenium-186 (186Re), Rhenium-188 (188Re) and Samarium-153 (153Sm) are some of the most promising neutron-activated radionuclides that are currently in clinical practice and undergoing clinical research for theranostic applications. The aim of this paper is to review the physical characteristics, current clinical applications and future prospects of these neutron activated radionuclides in theranostics. The production, physical properties, validated clinical applications and clinical studies for each neutron-activated radionuclide suitable for theranostic use in nuclear medicine are reviewed in this paper.
Nuclear radiation shielding capabilities for a glass series 20Bi2O3 - xPbO - (80 - 2x)B2O3 - xGeO2 (where x = 5, 10, 20, and 30 mol%) have been investigated using the Phy-X/PSD software and Monte Carlo N-Particle transport code. The mass attenuation coefficients (μm) of selected samples have been estimated through XCOM dependent Phy-X/PSD program and MCNP-5 code in the photon-energy range 0.015-15 MeV. So obtained μm values are used to calculate other γ-ray shielding parameters such as half-value layer (HVL), mean-free-path (MFP), etc. The calculated μm values were found to be 71.20 cm2/g, 76.03 cm2/g, 84.24 cm2/g, and 90.94 cm2/g for four glasses S1 to S4, respectively. The effective atomic number (Zeff)values vary between 69.87 and 17.11 for S1 or 75.66 and 29.11 for S4 over 0.05-15 MeV of photon-energy. Sample S4, which has a larger PbO/GeO2 of 30 mol% in the bismuth-borate glass, possesses the lowest MFP and HVL, providing higher radiation protection efficiency compared to all other combinations. It shows outperformance while compared the calculated parameters (HVL and MFP) with the commercial shielding glasses, different alloys, polymers, standard shielding concretes, and ceramics. Geometric Progression (G-P) was applied for evaluating the energy absorption and exposure buildup factors at energies 0.015-15 MeV with penetration depths up to 40 mfp. The buildup factors showed dependence on the MFP and photon-energy as well. The studied samples' neutron shielding behavior was also evaluated by calculating the fast neutron removal cross-section (ΣR), i.e. found to be 0.139 cm-1 for S1, 0.133 cm-1 for S2, 0.128 cm-1 for S3, and 0.12 cm-1 for S4. The results reveal a great potential for using a glass composite sample S4 in radiation protection applications.