[en] A toroidal discharge apparatus has been made as an initial research in magnetic confinement system. This system consists of a capacitor, a RF source, an igniter system, a primary coil, a torus, and completed by Rogowski probe as a current detector. In this system, the discharge occurs when the minimum voltage is operated at 5 kV. The experiment result shows that the coupling factor is 0.35, it is proved that there is an equality between estimated and measurement results of the primary inductance i.e 8.5 μH

[en] In order to prepare the reactor core conversion from oxide to silicide, analysis of the gamma heat generation in the fuel plate and its influence on the gamma density in the reactor core using the GAMSET computer code have been done. The heat generation was evaluated for oxide (U₃O₈-Al) and silicide (U₃Si₂-Al) plate for different uranium loading. The calculation result shows that the heat generation in the silicide fuel plate contains 400 gram of U-235 per fuel element increase of 10.64% related to the normal oxide plate. This means that the gamma density in the reactor core will consequently decrease. Regarding this result, it can be concluded that the core conversion from oxide to silicide fuel with higher uranium loading will be followed by the heat generation increases in the fuel plate and the gamma density decreases in the reactor core

[en] Determination of thermal, epithermal and total self shielding factor and cadmium ratio of cylindrical probe has been done by measurement and calculation. Self shielding factor can be determined by dividing probe activity to Al-alloy probe activity. Due to the lack of cylindrical probe made of Al-alloy, self shielding factor can be determined by parabolic extrapolation of measured activities to 0 cm radius to divide those activities. Theoretically, self shielding factor can be determined by making numerical solution of two dimensional integral equations using Romberg method. To simplify, the calculation is based on single collision theory with the assumption of monoenergetic neutron and isotropic distribution. For gold cylindrical probe, the calculation results are quite close to the measurement one with the relative discrepancy for activities, cadmium ratio and self shielding factor of bare probe are less then 11.5%, 3,5% and 1.5% respectively. The program can be used for the calculation of other kinds of cylindrical probes. Due to dependency to radius, cylindrical probe made of copper has the best characteristic of self shielding factor and cadmium ratio

[en] The results of the influence of nitridation temperature and time on the surface hardness of AISI 1010 low carbon steels nitrided by means of plasma glow discharge technique are presented in this paper. The results are the changing of surface hardiness, the changing of surface microstructure and the penetration profile depth. The experiment has been carried out at the temperature 400^oC, 450^oC, 500^oC, 550^oC, 570^oC and 600^oC, whereas the time is 5 minutes, 15 minutes, 40 minutes, 90 minutes and 180 minutes. All the experiments have been carried out at the optimum plasma density condition. The optimum plasma density condition is achieved at the pressure of p = 0.2 torr, when thr gas flow of nitrogen is 0.6 liter/minute and the distance of electrode plate is 4.5 cm. It was found that the optimum hardness of the surface was achieved at the temperature of 570^oC and the time of nitridation was 90 minutes, i.e. 190 KHN

[en] Measurement of the Argon ion distribution in plasma focus by using Faraday cup has been done. The intensity of ion beam followed the I Rn rule, n=1,02. In the operation condition of 0,8 mbar and 12,5 kV the current sheath spen 2.2 to 2.4 μsecond in the rundown phase. Cu ion was also been observed in the Faraday cup

[en] Automation of nuclear reactors, especially research reactors, needs to be done in order to avoid human error that raise from human. One of the important operation parameters that should be automatically controlled that of a reactor power. Until now, the reactor power control still need an operator intervention due to lack of recently used conventional control method. This paper describes the automation of a reactor power controlled by reactivity constraint. Reactivity constraint is needed to solve the problem that is present in the classic control method, e.g, overshoot and/or undershoot at transient condition. Reactivity constraint method is developed from the dynamic periods equation, which gives the instantaneous reactivity period as a function of the reactivity and the rate of change of reactivity. By experimental testing in the MITR-II research reactor, automatic controlling of the reactor power by reactivity constraint method proved to be good

[en] Effects of fast neutron 14 MeV irradiation on electronic components especially semiconductor diode commercially available in the market have been investigated. The investigation was done by comparing volt-ampere of diode characteristic before and after irradiation for various neutron fluent. The effect occurred is in the form of the change of the diode characteristic which is the increased current that flows at forward bias. Minimum neutron fluent required to induce some effects on diode was found to be 2.8 x 1012 n/cm² for BZX29C6V8 and IN935 silicon diode and 1.6 x 1012 n/cm² for OA germanium diode

[en] Three dimensional residual stress distribution in the heat affected zone of 10 mm thick welded steel by means of neutron diffraction technique has been measured. The results showed that the residual stress was distributed near the welded metal, namely within about 46,25 mm. The major tensile stresses occurred in the X-direction, and they attained a level greater than 2000 MPa through the position far away fram the weld. The tensile stresses in the Y and Z- directions lied between 500 and 1500 MPa, The results also suggest that the stress in the surface was greater than that in the middle of the sample

[en] Analysis of the efficiency improvement of the shell and tube type of the Kartini reactor's Heat Exchanger (HE) have been carried out after the flow direction system was modified from the parallel flow to the counter flow system. The HE was tested by operating the reactor at the power level of 100 k W, until the temperature of the water coolant reached the steady state condition. The efficiency and other HE's parameters was investigated by using the ^SIMULTAN^method. From the experiment it is known that the inlet and outlet primary and secondary water coolants are T_i = 38^oC, T_o = 35^oC, t_i 32^oC and t_o = 33^oC respectively. The investigation and analysis show that that HE's efficiency is η= 45,5 % due to U a= 674,79 W/m K, LMT = 3,27 and NTU 0,835. From the analysis can be concluded that the increase of the HE's efficiency is 2.5 % compared to parallel flow and the decrease is 6.7% compared to the HE's efficiency as soon as after having been cleaned in 1994

[en] Accident transient due to the loss of flow in the Process Inherent Ultimate Safety (PIUS) reactor has been simulated. This simulation constitutes a part of the safety analysis in the advanced nuclear reactor design which is inherently safe such as PIUS reactor having the thermal power of 2000 M Wth (PIUS2000). The analysis was focused on the accidental transient of loss of flow due to the loss of power supply to the primary pumps that can result from a loss of off-site power. Simulation was done in Nagoya University of Japan by using computer code RELAP5/MOD2. The effect of heat structures on the coolant flow rate through the core during the phase of natural circulation through density lock and borated water pool has been analyzed. The simulation results showed that the cyclic behavior of natural circulation through the pool, density lock and riser (core) occurred due to the failure of the primary pump