186gRe (T1/2 = 3.7183 d, E(β-)mean = 346.7 keV, I(β-)mean = 92.59%), a mixed beta and γ-emitter shows great potential for use in theranostic applications. The dominant 185Re(n,γ) route, via use of a nuclear reactor, provides 186gRe in carrier added form with low specific activity, while cyclotrons offer no carrier-added (NCA) high specific activity production of 186gRe. However, to be able to select the best possible nuclear reaction and to optimize the production route via the use of a cyclotron, information on the excitation function for the reaction of interest as well as for the competing reactions is necessary. Accordingly, we have conducted a detailed study of the excitation functions for natW(d, x) reactions in seeking optimized parameters for the NCA production of 186gRe. Noting a discrepancy among the experimental data, we made an evaluation of the available literature, finally selecting optimum parameters for the production of 186gRe via the 186W(d,2n)186Re reaction. These beam parameters were then used for batch production of 186gRe by irradiating an enriched 186W metallic powder target, followed by a subsequent automated chemical separation process. The preliminary results show 98.1% radionuclidic purity of 186gRe at 8 h subsequent to the End of Bombardment (EOB), offering the potential for use in clinical applications.
The excitation functions were measured for the (nat)Cu(α,x)(66,67)Ga,(65)Zn,(57,58,60)Co reactions in the energy range of 16.5 -50MeV. A conventional stacked-foil activation technique combined with HPGe γ-ray spectrometry was employed to determine cross-sections. The measured cross-sections were critically compared with relevant previous experimental data and also with the evaluated data in the TENDL-2014 library. Present results confirmed some of the previous experimental data, whereas only a partial agreement was found with the evaluated data. The measured data are useful for reducing the existing discrepancies in the literature, to improve the nuclear reaction model codes, and to enrich the experimental database towards various applications.
One of the most prevailing issues in the operation of Nuclear Reactor 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. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades ID system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented.
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.
A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (k eff ) for TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of k eff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of k eff from MCNP5 were obtained. When the control-fuel rods were fully inserted, the k eff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370 ± 0.00054. When the control-fuel rods were fully withdrawn the value of k eff value indicates the RTP reactor is in the supercritical condition, that is 1.10773 ± 0.00083.
PUSPATI TRIGA Reactor (RTP) is the only nuclear research reactor in Malaysia. It has been safely operated and maintained since 28 June 1982. Over 28 years of operation, some of the reactor systems have been upgraded or replaced to ensure the functionality and safety of the reactor. One of the major reactor systems which is primary cooling system is used to remove heat generated in the reactor core. The former primary cooling system consisting of single unit of shell-and-tube heat exchanger, three centrifugal pumps and piping system was replaced with a new system due to decreasing of the cooling performance. The new primary cooling system, consisting of two units of the 1.5-MW plate-type heat exchangers, new three primary pumps and new piping system was installed in accordance to the specified AELB requirements and guidelines of Nuclear Malaysia Safety, Health and Environment Committee (JKSHE). This paper summarises the replacement process of the former RTP primary cooling system. The activities involved preparation before and during construction and installation phases as well as safety consideration based on International Atomic Energy Agency (IAEA), Atomic Energy Licensing Board (AELB) requirements and Occupational Health and Safety Act (Act 514) were discussed and evaluated.
The miniaturization boiling (micro-bubble emission boiling [MEB]) phenomenon, with a high heat removal capacity that contributes considerably to the cooling of the divertor of the nuclear fusion reactor, was discovered in the early 1980s. Extensive research on MEB has been performed since its discovery. However, the progress of the application has been delayed because the generation mechanism of MEB remains unclear. Reasons for this lack of clarity include the complexity of the phenomenon itself and the high-speed phase change phenomenon in which boiling and condensation are rapidly generated. In addition, a more advanced thermal technique is required to realize the MEB phenomenon at the laboratory scale. To the authors' knowledge, few studies have discussed the rush mechanism of subcooled liquid to the heating surface, which is critical to elucidating the mechanism behind MEB. This study used photographic images to verify that the cavitation phenomenon spreads to the inside of the superheated liquid on the heating surface and thus clarify the mechanism of MEB.
Present work shows the development of nuclear technology in Malaysia and highlights its
applications that have been developed by using the instrumental neutron activation analysis
(INAA) method. In addition, present study exhibits a comprehensive review of INAA for
calculation of neutron flux parameters and concentration of elements. The INAA is a
powerful method to analyse the sample which identifies qualitative and quantitative of
elements present in a sample. The INAA is a working instrument with advantages of
experimental simplicity, high accuracy, excellent flexibility with respect to irradiation and
counting conditions, and suitability for computerization. In INAA, sample is irradiated and
measured directly. In practical. INAA is based on an absolute, relative and single-comparator
standardisation method. The INAA has been developed since 1982 when the
TRIGA Mark II reactor of Malaysia has commissioned. The absolute method was less
utilised, the relative method has been used since 1982, and the ko-INAA method is derived
from single-comparator standardization method has been developed since 1996 in Malaysia.
The relative method, because of its advantages, such as high accuracy, easy for using, has
many applications in Malaysia. Currently, local universities and Malaysian Nuclear Agency
(MNA) research reactor use INAA method in Malaysia.
Mechanical properties of blended polyethylene (PE) containing the antioxidant Irganox 1010 and the UV-absorber Tinuvin 326 were studied for future use as radiation capsule material for the TRIGA Mark II research reactor. High density and low density polyethylene were blended with the additives and tested for elongation at break, impact strength and gel content, before and after irradiation inside the nuclear reactor. Characterization via FTIR as well as determination of crystallization and melt transition temperatures through DSC were also conducted. It was found that the addition of the antioxidant at different amounts (from 0 to 4 phr) had various effects on the properties of the blended PE, with 0 phr being the amount at which there was the biggest increase in elongation at break and impact strength, post-irradiation.
Thorium is a fertile material that can undergo transmutation for it to become a fissile material,
uranium-233. The fissile material can go through a fission process in order to generate heat energy
and eventually electricity. Most nuclear reactors use uranium as their fission source. The use of
thorium as nuclear fuel has been only investigated for few types of reactors such as a high, temperature
gas reactor (HTGR), fast breeder reactor, light water reactor (LWR) and heavy water reactor
(HWR). For research reactors specifically, there are limited academic publications related to the
la,test u.se of thorium. Hence, the main, interest, of this work is to compile and review the latest
academic publications related to the active use of thorium, for research reactors in particular. The
reviewed studies have been, divided into two categories which are experimented and simulation projects.
The experimental projects are a,bold the ongoing thorium fuel tests that have been carried out. in an
actual, research reactor. On the hand, the simulation work: is related to the computational analysis
performed in predicting the neutronic behaviour of thorium based fuel in research reactors. The
experimented study of thorium is currently active for the KAMINI research reactor. Additionally, most,
simulation works focus on finding criticality and neutron spectra.
The influence of water-to-cement ratio (w/c) on the compressive strength of cement-biochar-spent resins matrix was
investigated. Spent resins waste from nuclear reactor operation was solidified using cement with w/c ranging from 0.35
to 0.90 by weight. In this study, biochar was used as a cement admixture. Some properties of spent resins and biochar
were determined prior to the formulation study. Compressive strength of harden cement-biochar-spent resins matrix
was determined at 28 days. The compressive strength of cement-biochar-spent resins matrix was found to depend on the
w/c and the amount of spent resins added to the formulation. The immersion test of cement-biochar-spent resins matrix
showed no significant effects of cracking and swelling. The compressive strength of the cement-biochar-spent resins
matrix increased after two weeks in water immersion test.
On March 11, 2011, a serious accident occurred in Daiichi nuclear reactor plant, Fukushima,
Japan which caused radioactive materials been released into the atmosphere in the form of
aerosols and dust particles. Sea water around the plant was also found contaminated with high
radioactivity readings. These radioactive materials could be transported by the winds and ocean
current across international borders and cannot be controlled by human. Thus, a continuous
monitoring activity of radionuclide content in the air and sea water needs to be conducted by the
authorities. In addition to radioactivity monitoring, Malaysia should also control the entry of
contaminated food in order to prevent radionuclide ingestion by human. The radionuclide 131I,
134Cs and 137Cs were used as a measure of pollution levels and counted with gamma spectrometry
using standard analysis method suggested by AOAC International. In this paper, details description
of the role of Radiochemical and Environment Group, Nuclear Malaysia who’s responsible in
analyzing the radioactivity in the food samples due to Fukushima Daiichi, Japan accident was
included. The radioactivity limit adopted and analysis results from this monitoring were discussed
Since the world’s first nuclear reactor major breakthrough in December 02, 1942, the nuclear power industry has undergone tremendous development and evolution for more than half a century. After surpassing moratorium of nuclear power plant construction caused by catastrophic accidents at Three-Mile Island (1979) and Chernobyl (1986), today, nuclear energy is back on the policy agendas of many states, both developed and developing nations, signaling nuclear revival or nuclear renaissance. Selection of suitable nuclear power technology has thus been subjected to primary attention. This short paper attempts to draw preliminary technology assessment for the first nuclear power reactor technology for Malaysia. Methodology employed is qualitative analysis collating recent finding of TNB-KEPCO Preliminary Feasibility Study for Nuclear Power Program in Peninsular Malaysia and other published presentations and/or papers by multiple experts.
The results suggested that the pressurized water reactor (PWR) is the prevailing technology in terms of numbers and plant performances, and while the commercialization of Gen IV reactors is remote (e.g. not until 2030), Generation III/III+ NPP models are commercially available on the market today. Five (5) major steps involved in reactor technology selection were introduced with a focus on introducing important aspects of selection criteria. Three (3) categories for the of reactor technology selection were used for the cursory evaluation. The outcome of these analyses shall constitute deeper and full review analyses of the recommended reactor technologies for the intended full feasibility study in the near future. Recommendations for reactor technology option were also provided for both strategic and technical recommendations. The paper shall also postulate or rather implore what could be the best way for Malaysian and also other aspiring new entrant nations to select systematically their first civilian nuclear power reactor.
An instrumental neutron activation analytical (INAA) technique is used for the determination of thirty elements in five coal samples collected from Kapar power station, imported from Indonesia and Australia. Analyses of the samples are being associated with standards. All irradiations were performed in the nuclear reactor of Malaysia Nuclear Agency (MNA). Samples were counted by Hyper Pure Germanium (HPGe) detector for short period irradiations at MNA, while for the long period irradiations the samples were counted at Universiti Kebangsaan Malaysia (UKM). The concentrations of thirty elements have been determined: The major components are Cl, Ca, Mg, K, Fe, Ti and Na with the mean concentrations in the range between 70±69 ppm- 6100±1639 ppm; and the trace elements are Zr, V, Mn , Sc, Cr, Co, As ,Br ,Rb ,Sb ,Ba , La, Ce, Nd, Sm, Eu, Tb, Yb, Lu, Hf, Th, U and Ta with the mean concentrations in the range between 0.1381±0.0202 - 69.0±2.8 ppm. The results have been compared to the reported data of eight coal samples from the United States and the reported data of Australian bituminous coal.
The flow rate or fluid velocity measurement is important to maintain fluid flow quality performance in the systems. This study focuses on determination of volumetric flow rate measurement and to calibrate the conventional flowmeter using industrial radiotracer approach in quadrilateral gas-liquid bubble column reactor. In this work, two different radioisotopes which emit γ-ray have been chosen as radioactive tracer which is 99mTc produced from 99Mo/99mTc radioisotope generator and 198Au nanoparticle form neutron activation at research nuclear reactor TRIGA Mark II. Both radioisotopes representing liquid and solid tracer purposely designed for tracing liquid flow. The peak to peak radiotracer method known as pulse velocity method was applied to determine the volumetric flow rate. The radiation signals were monitored using 4 unit NaI scintillation detectors located at 4 different points nearby the inlet and outlet of the quadrilateral bubble column reactor process stream. The water volume inside the bubble column reactor was fixed at 0.04 m3 and liquid flow rates in this reactor were specified on installed flowmeter at different reference value which is 4 lpm, 8 lpm, and 12 lpm, respectively. The experimental result shows very good linearity and repeatability by following the theoretical equations with less uncertainty in volumetric flow rate measurement. The obtained results also validated the effectiveness of the proposed method for the installed flowmeter calibration efficiency.