Nuclear Science and Technology

After reconstruction, the Dalat Nuclear Research Reactor (DNRR) was inaugurated on March 20th, 1984 with the nominal power of 500 kW. Since then the production of radioisotopes and labelled compounds for medical use was started. Up to now, DNRR is still the unique one in Vietnam. The reactor has been operated safely and effectively with the total of about 37,800 hrs (approximately 1,300 hours per year). More than 90% of its operation time and over 80% of its irradiation capacity have been exploited for research and production of radioisotopes. This paper gives an outline of the radioisotope production programme using the DNRR. The production laboratory and facilities including the nuclear reactor with its irradiation positions and characteristics, hot cells, production lines and equipment for the production of Kits for labelling with 99mTc and for quality control, as well as the production rate are mentioned. The methods used for production of 131I, 99mTc, 51Cr, 32P, etc. and the procedures for preparation of radiopharmaceuticals are described briefly. Status of utilization of domestic radioisotopes and radiopharmaceuticals in Vietnam is also reported.

In order to clarify the effect of SFE on SCC resistance of austenitic stainless steels and to develop the alternative material of Type 316LN stainless steel for BWR application, the effect of chemical composition and heat treatment on SFE value and SCCGR in oxygenated high temperature water  were studied. The correlation factors between SFE values for 54 heats of materials and their chemical compositions for nickel, molybdenum, chromium, manganese, nitrogen, silicon and carbon were obtained. From these correlation factors, original formulae for SFE values calculation of austenitic stainless steels in the SHTWC, SHTFC and AGG conditions were established. The maximum crack length, average crack length and cracked area of the IGSCC for 33 heats were evaluated as IGSCC resistance in oxygenated high temperature water. The IGSCC resistance of strain hardened non-sensitized austenitic stainless steels in oxygenated high temperature water increases with increasing of nickel contents and SFE values.  From this study, it is suggested that the SFE value is a key parameter for the IGSCC resistance of non-sensitized strain hardened austenitic stainless steels. As an alternative material of Type 316LN stainless steel, increased SFE value material, which is high nickel, high chromium, low silicon and low nitrogen material, is recommendable.

Handling and improving the quality of medical images with the help of computer software is one of the important stages in the diagnosis and treatment. In this article, we focus on describing the new morphological algorithms by ITK (Insight Segmentation and Registration Toolkit). These morphological operators eliminate noise, detect good edges, and overcome the drawback of traditional edge detection methods.

A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA).The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducted to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed

Zinc scrap is a source of raw material for zinc oxide production. However, to qualify the requirement of raw material for zinc oxide (99.5%) production, refining this source is needed. Many methods are considered such as rectification, chemical method, etc., but difficult to apply on an industrial scale. This workfocused on the investigation of the influence of temperature and time factors for asessing the possibility of applying liquation method for the purification of impurities from scrap zinc.The experiment results show that the optimum temperature of liquation to remove Pb,Fe from zinc scrap is in the range of 440-450°C, the optimal time of the process is 8h for the pot with 8cm in height and 6cm in diameter (the quantity of raw zinc sample is about 2kg / batch), then we can obtain about 80% of zinc metal with an average Zn content of about 97, 0%, both Pb and Fe content decreased to a range from 0.35 to 0.4%, and 1.0 to 1.1%, respectively, which meet the requirement of raw materials for the production of high quanlity ZnO ( 99,5 %). Based on the parameters obtained on lab-scale, a trial on pilot scale of 250 kg / batch was conducted, The result confirms that the quality of the products meets the requirement of raw materials for production of high quality ZnO (99.5%) and a technology process for refining zinc scrap by the liquation was proposed.

87Sr/86Sr isotopic ratio is one of the useful tools that can authenticate the original source of the natural products from the earth-created and/or geological processes. However, the effect of interferences in petroleum drill-holes water sample such as thickness of sample matrix or isotopic signal of 87Rb might cause the low precision of 87Sr/86Sr determination using quadrupole inductively coupled plasma spectrometry (ICP-MS). The elimination of these mentioned effects was thus studied by using the ion - exchange chromatography. Calcium in sample matrix was separated on anionite column (Bio-Rad AG1-X8 resin) in methanol medium with the high efficiency while rubidium was removed from strontium on cation exchange resin (Bio-Rad AG50-X8) with strontium recovery over 99%. The isotope dilution technique with 86Sr - enriched isotopic standard solution was used for the control of separation process. The 87Sr/86Sr isotopic ratio was thus determined using ICP-MS with the signal correction by a strontium isotopic ratio standard reference material (NIST SRM 987).

The isomeric ratio (IR) of isomeric pair 109m,gPd, produced in 110Pd(γ, n)109m,gCd reaction and 108Pd(n, γ)109m,gPd neutron capture reactions, induced by thermal, epithermal and mixed thermal-epithermal neutrons have been determined. The off-line activation technique using a spectroscopic system consisting of a HPGe semiconductor detector with high energy resolution and a PC based 8192 channel analyzer (CANBERRA) was applied. The investigated samples were prepared from the 99.99 % purity PdO and irradiated at the electron accelerator Microtron MT-25 of the Joint Institute for Nuclear Research Dubna, Russia. The data analysis and necessary corrections were made to upgrade the precision of the experimental method. The obtained results were discussed, compared and combined with those from other authors to point out the role of the reaction channels in nuclear reactions.

Sodium bentonite (NaBent) was modified by poly(acrylic acid) (PAAc) to prepare the carriers for immobilization of Bacillus subtilis. Different mixtures of NaBent/AAc were regularly dispersed in distilled water and irradiated under gamma rays at an absorbed dose of 6.5 kGy with dose rate of 0.85 kGy/hr in air for polymerization of acrylic acid and formation of poly(acrylic acid)/sodium bentonite (PAAc–NaBent). The reaction yield was determined with the initial concentration of acrylic acid (AAc). The functional group properties of the resulting PAAc–NaBent were analyzed by Fourier Transform Infrared spectra (FTIR). Bacillus subtilis cells were immobilized on both NaBent and PAAc–NaBent as carriers by adsorption method for treating the sludge contaminated by fish feces and residual feed from the Pangasius farming ponds. The results showed that immobilization capacity of Bacillus subtilis on the PAAc–NaBent was better than that on non-modified NaBent. Analysis of BOD for the farming pond water containing Bacillus subtilis and the bacteria immobilized carriers with time revealed the lower BOD values obtained with the samples containing PAAc–NaBent, suggested that degradation of organic pollutants by Bacillus subtilis immobilized on the PAAc–NaBent was faster than that by free bacteria.

This paper introduces a new controller module based on a high-speed field-programmable gate array (FPGA) and digital signal processing (DSP) using moving average (MA) filters for calculation of the reactor power and period at the start range of the Dalat nuclear research reactor (DNRR). The reactor power is proportional to the neutron flux in the reactor core, and the reactor period is the time that the reactor power changes by a factor of 2.718. In the control and protection system (CPS) of the DNRR, the reactor power and period have been monitored by the 8-bit microprocessor controller named BPM-107R. There are two main functions of the BPM-107R controller including 1) measurement and determination of reactor power and period and 2) generation of warning and emergency protection signals by reactor power or/and by reactor period. Those discrete signals will access to the logical processing unit of the CPS to prohibit the upward movement of control rods or to shut down the reactor. The CPS has three BPM-107R controllers corresponding to three independent neutron flux measurement equipment (NFME) channels working by logic voting “2 out-of 3”. Each NFME channel was designed for detection of neutron flux density in the full range from 1×100 to 1.2×1010 n/cm2 ×s, which is divided into two sub-ranges named start range (SR) and working range (WR). The designed FPGA-based controller module was tested using simulated signals as well as signals from the CPS in comparison with the original controller BPM-107R. The experimental results show that the characteristics and functions of the two controllers are equivalent.

The article presents the multipole expansion for the electron-nucleus scattering cross-section at high energies within framework of the unified electroweak theory. The electroweak currents of  the nucleus is expanded into simple components with definite angular momentum, which are called the multipole form factors. The multipole expansion of the cross-section is a consequence of the above expansion. Besides the familiar electromagnetic form factors FxL , there are new form factors VLX and ALX related to weak interactions, corresponding to the vector and axial weak currents. The obtained general expressions are applied to the nucleus 36Li , where the partial form factors are computed in the multiparticle shell model.