[en] In order to convert the RSG-GAS core fuel element from oxide to silicide, accordingly the reactivity change and neutron flux depression is needed to measure. Reactivity change measurement using control rod calibration method was done last year. Using foil activation method did continuing those measurements, therefore in this year neutron flux distribution flux were done. The positions of measurement are at the fuel elements of A-6, C-6, F-7, H-7, E-3, E-5, D-8 and D-10 and in irradiation position of B-6, G-7, E-4, E-6, D-7 and D-9 at 100 kW reactor power. The averages of measurement result are between 2,47E12 and 3,55E12 n/cm exp.2/s at the fuel elements positions, and for the irradiation positions are between 6,67E12 and 8,18E12 n/cm exp.3/s at that power

[en] Foil activation method is commonly used in flux and neutron spectrum measurement in nuclear reactor and other research. The effect of the thickness, type of foil material and neutron spectrum shape on the self shielding correction and activities correction on the edges of the foil have been analyzed. Also the effect of neglecting those correction factors on neutron flux and spectrum measurement were analyzed. The calculation of the correction factor has been done by using the program which had been verified for several foils. The foils used are Au, In. Cu, Co and Dy of 0.00254 cm -0.127 cm thickness and 1.27 cm diameter. The result showed that the correction factor foils were not similar due to the variation of activation cross section and neutron spectrum shape. For the neutron spectrum in RS-2 multi purpose reactor GAS using foils of 0.00254 cm thick. The effect of neglecting correction factor on thermal flux measurement for Au, In, Co and Cu were less than -6%, for Dy was about -25%. On epithermal flux measurement for Au and In were about -60%, Co and Dy was -12% and -6%, for Cu less than -2%. The effect of neglecting correction factor on spectrum measurement was the change on the neutron flux density values along neutron energy region

[en] In order to 4.8 and 5.2 gr U/cm exp 3 loading of U₃Si₂--Al fuel plates characterization, he core reactivity change and neutron flux depression had been done. Control rod calibration method was used to reactivity change measurement and neutron flux distribution was measured using foil activation method. Measurement of insertion of A-type of testing fuel element with U-loading above cannot be done due to technical reason, so the measurement using full type silicide fuel element of 2.96 gr U/cm exp 3 loading. The reactivity change measurement result of insertion in A-9 and C-3 is + 2.67 cent. The flux depression at silicide fuel in A-9 is 1.69 times bigger than oxide and in C-3 is 0.68 times lower than oxide

[en] Determination of self-shielding factor and cadmium ration of foil and cylindrical probe have been done by measurement and calculation. Self-shielding factor was determined by dividing the activity of detector with its Al-alloy. Theoretically, self-shielding factor can be determined by numerical solution of two-dimensional integral equations in FORTRAN. For gold foil and wire, the calculation result are quite close to the measurement. The relative difference between calculation and measurement of activity, self-shielding factor and cadmium ratio are respectively less than 11%, 9% and 4%. It is therefore, the calculation program can be used for calculation of other kinds of activation detectors. The application in neutron flux measurement gives a better result especially for epithermal flux. For neutron spectrum measurement, self-shielding correction can avoid resonance peaks in epithermal region due to absorption by activation detectors. (author)

[en] Determination of neutron flux thermal to epithermal ratio f and parameter a and k_o factor has been done in irradiation facility of RSG G.A. Siwabessy reactor. Those parameters are needed to determine the concentration of an element in a sample using k_o NAA method. While parameters f were measured using foil activation method. α parameter was obtained from power function fitting at epithermal neutron spectrum. Based on the fitting method, the α parameter (non 1/E epithermal spectrum) was determined of 0.0267, 0.0255 and -0.0346 at system rabbit, IP2 and CIP irradiation position, respectively. The k_o factor is dependent on absolute gamma fraction, thermal cross section and resonance integral. The neutron flux thermal to epithermal ratio f at all rabbit system is about 40, which is met for neutron activation analysis

[en] The research to develop detection radiation system using coincidence method has been done to determine thermal neutron flux in RS1 and RS2 irradiation facilities RSG-GAS. At this research has arranged beta-gamma coincidence equipment system and parameter of measurement according to Au-198 beta-gamma spectrum. Gold foils that have irradiated for period of time, counted, and the activities of radiation is analyzed to get neutron flux. Result of research indicate that systems measurement of absolute activity with gamma beta coincidence method functioning well and can be applied at activity measurement of gold foil for irradiation facility characterization. The results show that thermal neutron flux in RS1 and RS2, respectively is 2.007E+12 n/cm²s and 2.147E+12 n/cm²s. To examine the system performance, the result was compared to measure activity using high resolution of Hp Ge detector and achieved discrepancy is about 1.26% and 6.70%. (author)

[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] In accordance with the PTRKN strategic plan, it was obtained the conceptual design of innovative research reactor. In terms of the reactor core, the conceptual design of the core design, thermohydraulic design and radiation shield design. The main objective of this research is the design of the shield and analysis of radiation dose rate to obtain boundary conditions that meet radiation safety in the work area of innovative research reactor. Analysis of the radiation dose rate is calculated using MCNPX program package by varying the wall thickness and type of radiation shielding. From the calculation of gamma dose can be selected type and thickness of radiation shielding required for a research reactor in the form of ordinary concrete with a thickness of 200 cm. The analysis showed that the shield design produces maximum dose rate 15.34 μSv/hour in near the shield wall in the experiment hall. The dose rate in compliance with radiation safety limits set by the Nuclear Energy Regulatory Agency (BAPETEN). (author)

[en] Highlights: • A systematic theoretical design via transfer matrix method of the metal-insulator-metal biosensing platform was proposed. • Insulator layer with suitable refrextive index, eg. SiO₂ (refrextive index = 1.50) was employed in the nanostructure. • The adhesive Cr thin layer between SiO₂ and Au was found to deteriorate the plasmon effect by the factor of |ε₁/ε₂| ratio. • (3-mercaptopropyl)trimethoxysilane was used, instead, as an excellent replacement to the Cr layer. • The nanostructure with (3-mercaptopropyl)trimethoxysilane as adhesion layer showed a 106% higher signal intensity. Advanced techniques in nanofabrication are essential in order to manifest theoretical works in establishing nanobio-sensor devices. Metal-insulator-metal (MIM) geometry has been effectively used for tuning plasmonic effect that dramatically altered biosensor sensitivity. Normally, adhesive metal at metal-insulator interface; sub-nm chromium is required to stabilize the metallic nano-layer structure. This deteriorates the performances of the MIM sensors, by severely damping the plasmon modes. We replaced the commonly used sub-nm Cr with an organic molecular layer of (3‑Mercaptopropyl) trimethoxysilane; MPTMS, as an adhesive molecule to strengthen the structure of the MIM-based biosensor. This successfully demonstrated a non-deterioration plasmon effect due to no imaginary part exist in the refractive index of the MPTMS. The new structure of metal/(MPTMS)/insulator/(MPTMS)/metal, instead of metal/(Cr)/insulator/(Cr)/metal or metal/poly(methyl methacrylate) or PMMA/metal was used as biosensing platform to detect molecular interactions. We simulated the biosensing concept using transfer matrix methods and construct the theoretically optimized MIM structure. The metal (Au) thin-layer was prepared by means of thermal evaporation method on a pre-cleaned glass slide. MPTMS is then deposited using solution deposition technique. This enables larger usable area on the MIM platform that can be utilized as the sensing area compared to an MIM using PMMA as insulator layer.

[en] The reaktor daya eksperimental (RDE) is 10 MWth experimental power reactor HTR Pebble-bed type. Comparing with water pool research reactors, the distance between the core and the reactor vessel is very close while the medium mainly is graphite. Thus, the main role of the radiation shield relies on biological shields to obstruct radiation exposure coming from the reactor core. Meanwhile, the RDE reactor should be met with the operation standard safety limits of a nuclear reactor. Therefore, the purpose of this study is to estimate the radiation dose rates in the working area and outside the RDE reactor building. By analyzing the distribution of radiation dose rates can be estimated radiation safety levels for workers and society and the surrounding environment. The radiation dose rates calculation was done using MCNP. By determining the tally position, it can be determined the radiation dose rates inside and outside the RDE reactor building. The results show that the maximum dose rates on the outer surface of the biological shielding (working area) under normal operating conditions are 8 μSv/h. That dose rates are below the limit value determined by BAPETEN showing the radiation shielding design make the RDE safe for workers and the community and the surrounding environment from radiation hazards. (paper)