Megat Harun Al Rashid Megat Ahmad

Characterization of alumina-based LTCC composite materials: Thermal and electrical properties

The study on Low temperature co-fired ceramic (LTCC) has been widely investigated due to its potential as an advanced approach for packaging of electronics devices. This paper evaluates and discusses about the firing temperatures effect on the densification, thermal and electrical properties of the prepared LTCC material. The LTCC tape, which consists mostly of Al2O3 was fabricated using tape casting technique. Five layers of tapes were then stacked and air-fired at different temperatures of 840, 880, 920 and 960°C. All samples were characterized for thermal and electrical properties. The crystallinity and microstructure of all samples were studied using XRD and SEM respectively. Current investigations show that the changing of density or volume fraction of pores is the main reason for the dielectric properties shift. Meanwhile, the changing in lattice parameters is observed as the main reason in thermal diffusivity shift. The sample fired at 880°C has the best thermal and electrical properties with thermal diffusivity of about 0.62 mm2/s, dielectric constant (εr) of about 5.65 and dielectric loss, tan (δ) of about 0.003.

Effects of Neutron and Electron Irradiation on 4H-SiC Diodes

4H-SiC Schottky barrier diodes (SBDs) were irradiated to neutron fluence of 3.55 x1016 cm-2 and 6.6 x 1015 cm-2 (15,000 kGy) electrons respectively. In general, characterization of the irradiated samples show that the current characteristics of the diodes decreased. The performance of Schottky gate contact is less for electron irradiated sample compared to neutron irradiated sample. The d-spacing, crystallite sizes and lattice strains were calculated from X-ray diffraction (XRD) measurements. SiC Schottky interface damage and radiation defects, as observed in atomic force microscopy (AFM) topography and scanning electron microscope (SEM) morphology images is possibly the main reason for this reduction in performance.

Study of pipe thickness loss using a neutron radiography method

The purpose of this preliminary work is to study for thickness changes in objects using neutron radiography. In doing the project, the technique for the radiography was studied. The experiment was done at NUR-2 facility at TRIGA research reactor in Malaysian Nuclear Agency, Malaysia. Test samples of varying materials were used in this project. The samples were radiographed using direct technique. Radiographic images were recorded using Nitrocellulose film. The films obtained were digitized to processed and analyzed. Digital processing is done on the images using software Isee!. The images were processed to produce better image for analysis. The thickness changes in the image were measured to be compared with real thickness of the objects. From the data collected, percentages difference between measured and real thickness are below than 2%. This is considerably very low variation from original values. Therefore, verifying the neutron radiography technique used in this project.

Preliminary study of neutron absorption by concrete with boron carbide addition

Concrete has become a conventional material in construction of nuclear reactor due to its properties like safety and low cost. Boron carbide was added as additives in the concrete construction as it has a good neutron absorption property. The sample preparation for concrete was produced with different weight percent of boron carbide powder content. The neutron absorption rate of these samples was determined by using a fast neutron source of Americium-241/Be (Am-Be 241) and detection with a portable backscattering neutron detector. Concrete with 20 wt % of boron carbide shows the lowest count of neutron transmitted and this indicates the most neutrons have been absorbed by the concrete. Higher boron carbide content may affect the concrete strength and other properties.

Effect of Boron Carbide Content on Properties of Thermoplastic Natural Rubber Composite as Thermal Neutron Shielding

In this work, thermoplastic natural rubber (TPNR) composites were produced through melt blending method. Boron carbide (B4C) as filler was added into the polymer blend (TPNR) with different weight percent from 0% to 30% and the effect of different B4C contents on mechanical and thermal neutron attenuation properties of TPNR composites has been studied. The phase formation in composites was analyzed using XRD technique. From the results, it showed that the incorporation of B4C fillers into TPNR matrix has enhanced the macroscopic cross section of the composites, however it lessens the tensile strength. Macroscopic cross section of the composites were increased from 3.34 cm-1 to 14.8 cm-1, while the tensile strength of the composites decreased from 3.79 MPa to 1.06 MPa with increasing B4C from 0 wt% to 30 wt%. B4C diffraction peaks were also increased in intensity with increasing B4C content.

Characterization of ZnBr2 solution as a liquid radiation shield for mobile hot cell window

The Mobile Hot Cell (MHC) has a viewing window which is usually made of almost transparent radiation shield material for the safety of MHC operators. Mobility is the main criterion for MHC; therefore liquid solution that can act as a radiation shield is usually selected as the window for MHC due to ease of transportation instead of a solid glass. As reported, Zinc Bromide (ZnBr2) solution was successfully used in viewing window for MHCs in South Africa and China. It was chosen due to its transparent solution, excellent performance as radiation shielding for gamma radiation, ease in preparation, handling, storage and treatment. Nevertheless, data and baseline studies on ZnBr2 as radiation shield are quite few. Therefore, a study on this matter was carried out. The preparation of ZnBr2 solution was processed at laboratory scale and the radiation shielding experiments were carried out using Cs-137 as radiation source. ZnBr2 solution was prepared by mixing ZnBr2 powder with distilled water. The mixing percentage of ZnBr2 powder, (%wt.) was varied to study the effect of density on the attenuation coefficient. The findings from this study will be used as a guideline in the production and management of ZnBr2 solution for MHC applications.The Mobile Hot Cell (MHC) has a viewing window which is usually made of almost transparent radiation shield material for the safety of MHC operators. Mobility is the main criterion for MHC; therefore liquid solution that can act as a radiation shield is usually selected as the window for MHC due to ease of transportation instead of a solid glass. As reported, Zinc Bromide (ZnBr2) solution was successfully used in viewing window for MHCs in South Africa and China. It was chosen due to its transparent solution, excellent performance as radiation shielding for gamma radiation, ease in preparation, handling, storage and treatment. Nevertheless, data and baseline studies on ZnBr2 as radiation shield are quite few. Therefore, a study on this matter was carried out. The preparation of ZnBr2 solution was processed at laboratory scale and the radiation shielding experiments were carried out using Cs-137 as radiation source. ZnBr2 solution was prepared by mixing ZnBr2 powder with distilled water. The mixing percent...

Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried out using a neutron spectrometer.

Proposal for grid computing for nuclear applications

The use of computer clusters for computational sciences including computational physics is vital as it provides computing power to crunch big numbers at a faster rate. In compute intensive applications that requires high resolution such as Monte Carlo simulation, the use of computer clusters in a grid form that supplies computational power to any nodes within the grid that needs computing power, has now become a necessity. In this paper, we described how the clusters running on a specific application could use resources within the grid, to run the applications to speed up the computing process.

Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

Microstructural investigations of materials for low temperature co-fired ceramic (LTCC) based fuel cell using small angle neutron scattering

The concept and the realization fuel cell based on LTCC technology require the investigations of fired LTCC microstructures. The majority of the works involved using small angle neutron scattering studies on the microstructural of LTCC ceramic tape and development of neutron tomography for future tool to visualize channels inside the fired tape. Most SANS characterization were carried out at Smarter SANS instrument at BATAN, Indonesia. Standard sample for resolving tens of micron of object size were measured using simple neutron tomography setup utilizing monochromatic SANS beam at Malaysian Nuclear Agency. The initial microstructural findings indicates that organic additives shape the final microstructural of LTCC after firing with the glassy material possibly fill the space left by the burned organic additives. The tomography results showed that 40 micron size object can be differentiated. The conductor deposited on LTCC is preliminary investigated which will later be used as support for catalyst.