JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
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THE PRELIMINARY STUDY ON IMPLEMENTING A SIMPLIFIED SOURCE TERMS ESTIMATION PROGRAM FOR EARLY RADIOLOGICAL CONSEQUENCES ANALYSIS
No full textIndonesia possesses numerous potential sites for nuclear power plant development. A fast and comprehensive radiological consequences analysis is required to conduct a preliminary analysis of radionuclide release into the atmosphere, including source terms estimation. One simplified method for such estimation is the use of the Relative Volatility approach by Kess and Booth, published in IAEA TECDOC 1127. The objective of this study was to evaluate the use of a simple and comprehensive tool for estimating the source terms of planned nuclear power plants to facilitate the analysis of radiological consequences during site evaluation. Input parameters for the estimation include fuel burn-up, blow-down time, specific heat transfer of fuel to cladding, and coolant debit, using 100 MWe PWR as a case study. The results indicate a slight difference in the calculated release fraction compared to previous calculations, indicating a need to modify Keywords: Source terms, Relative volatility, Release fraction, PWR, SMAR
A SIMULATION OF IRRADIATION CALCULATIONS ON LUTETIUM-177 PRODUCTION IN RSG-GAS USING U3SI2-AL AND U9MO-AL FUELS
Full text linkThis research is a simulation of irradiation calculations on the production of the radioisotope Lutetium-177 (177Lu) in the G.A Siwabessy Reactor (RSG-GAS). This study aims to analyze the comparative calculation of 177Lu activity and its purity. One of the production methods of 177Lu in RSG-GAS is carried out by irradiating Lu2O3 targets. This Lu2O3 target irradiation produced the radioisotope 177Lu along with 177mLu as an impurity. For Medical treatment using radioisotopes, the minimum activity for 177Lu is 20 GBq/mg, and the impurity should not exceed 0.1%. Calculations were carried out with thermal neutron flux input at 15 MWt operational power for the RSG-GAS core with U3Si2-Al fuel (density 2.96 gU/cc and 3.55 gU/cc) and U9Mo-Al fuel (density 3.55 gU/cc). Calculations were carried out by simulating 8 days of irradiation using ORIGEN2.1. The results showed that the 177Lu activity resulting from irradiation of Lu2O3 targets at various CIP positions in the U9Mo-Al reactor core was larger than that of the U3Si2-Al core. Until the 30th day, the 177Lu product resulting from irradiation on the U3Si2-Al and U9Mo-Al cores still meets the minimum value of 20 GBq/mg for treatment needs in nuclear medicine, with the activity value of 177Lu resulting from irradiation on the U3Si2-Al core ranging from 241-403 GBq/mg, while the activity of irradiated 177Lu in the U9Mo-Al core ranges from 335-561 GBq/mg. In addition, until the 30th day of decay, 177Lu has a percentage value of 177mLu irradiated in the U9Mo-Al and U3Si2-Al cores of 0.0346% and 0.0344%, respectively. The results are still below the maximum impurity value of 0.1% and thus safe to use as a therapeutic agent. Keywords: 177Lu, Activity, RSG-GAS, ORIGEN2, Irradiatio
Neutronic Analysis of the RSG-GAS Fuel Using Burnable Poison
Full text linkControl and safety of nuclear reactors are significantly influenced by the determination of safety parameters. The three most crucial safety factors for assessing reactor status are the infinite multiplication factor, reactivity coefficients, and power peaking factor. The objective of the present study is to examine how the RSG-GAS fuel safety parameters behave in a typical reactor operation state. A lattice cell fuel model of the fuel lattice of the RSG-GAS reactor core was modeled using WIMSD-5Bwith cross-section library data based on ENDF/B-VIII.0. The value of the infinite multiplication factor with various burnable poison concentrations, as well as the moderator and fuel temperatures, were the variables that were examined. The reactivity coefficient parameters were similarly analyzed. By comparing the WIMSD-5B code results with information from the SAR document, the WIMS model for RSG-GAS fuel was verified, and it was inferred that the parameters are in good agreement. Safe behavior uses the predicted reactivity coefficient values as an example
DETERMINING GAMMA SOURCE IN URANIUM MOLYBDENUM OF FUEL IN G.A SIWABESSY MULTI PURPOSE REACTOR
Full text link Nuclear fission reactions produce a lot of radionuclides that release energy, one of which is in the form of gamma radiation. Gamma radiation is produced by various types of radionuclides, and nuclear reactor fuel will produce different values of gamma intensity. Uranium Molybdenum (U7Mo-Al) is the type of nuclear fuel for future research reactors that possesses many advantages. For the application of molybdenum-based fuel, it is necessary to determine the resulting gamma radiation. The purpose is to determine the gamma radiation produced from molybdenum-based fuel with various densities. This study begins with the determination of the mass composition of the reactor component, calculations with ORIGEN2.1, and data output analysis. The U7Mo-Al density was varied, namely 2.96 gU/cm3, 3.85 gU/cm3, 4.44 gU/cm3, 5.43 gU/cm3, 6.91 gU/cm3, and 8.29 gU/cm3. The gamma radiation yield of U7Mo-Al is lower than that of uranium silicide (U3Si2) with the same density of 2.96 gU/cm3. The result will add to the justification for the superiority of U7Mo-Al compared to U3Si2/Al. For U7Mo-Al with densities of 3.85 gU/cm3, 4.44 gU/cm3, 5.43 gU/cm3, 6.91 gU/cm3, and 8.29 gU/cm3, the one that produced the lowest gamma radiation intensity is 3.85 gU/cm3 while the highest is 8.29 gU/cm3. This explains that the intensity of the gamma radiation produced is directly proportional to the fuel density. The low intensity of gamma radiation in molybdenum-based fuel can be used as a suggestion in shielding design to ensure the operational safety of reactors.
An Approach for Integration of User Requirement and Anthropometry Data in The Process Design of Reactor Main Control Room
No full textThe construction of a Nuclear Power Plant (NPP) using Small Modular Reactor (SMR) technology is an interesting scheme to support Net-Zero Carbon Emission. The SMR design is an advanced generation reactor with high safety and utilization features, especially the electricity needed and industry. Its modular size can also be applied to remote areas with lower construction costs compared to other types of power plants. Considering the geographical location and territory of Indonesia which is an archipelagic country, this type of reactor is suitable for application in Indonesia. To ensure safety and increase mastery of technology, it is necessary to create a simulator to support this program. Nonetheless, specific regulations govern human-machine interactions (HMI) which is covering the nuclear reactor simulators in Indonesia is not yet available. The research carried out is a review of the regulations that have been implemented in other countries, then provides a choice of operator condition designs, which are adjusted to the average size of Indonesian by considering anthropometric aspects and ergonomic aspects
DESIGN OF HELICAL TYPE STEAM GENERATOR FOR EXPERIMENTAL POWER REACTOR
Full text linkReaktor Daya Eksperimental (RDE) is a high-temperature gas-cooled reactor (HTGR) for electricity generation, heat generation, and hydrogen production by Batan. Empirical and numerical calculations are needed to strengthen the existing design. The numerical method by computational fluid dynamic (CFD) analyzes temperature distribution and pressure drop along the pipe. The Batan RDE steam generator design has a seven-layer helical pipe model, while this research uses a one-layer helix pipe. In empirical calculations, the heat transfer region has three sections; single-phase liquid, two-phase, and single-phase vapor heat transfer. In numerical calculations, apply the assumption of constant heat flux and constant working fluid properties. The results of empiric calculations data showed that the helical pipe height was 3.98 m, shorter than the Batan design, which is 4.97 m. This considerable difference due to empirical calculations did not cover the safety factor. The results of numerical calculations show that in the single-phase, empiric calculation data were acceptable since the different values of numerical calculations for empiric calculations data were below 10%. Meanwhile, the case of the two-phase numerical calculations is not satisfactory and needs further research to obtain optimal results
GAMMA RADIATION EFFECTS ON THE PERFORMANCE OF MONO-CRYSTALLINE SOLAR CELLS
Full text linkIn this study, we present examples of solar cells that were subjected to various levels of 60Co gamma radiation. The solar cells we use are mono-crystalline, which has a stable crystal structure and high efficiency compared to polycrystalline. Prior to and during gamma irradiation, the current-voltage characteristics of monocrystalline silicon solar cells under AM1.5 light conditions and their photon spectral currents were examined. The results of the experiment demonstrate that as the dose of gamma radiation increases, solar cell metrics including open circuit voltage (Voc), short circuit current (Isc), and efficiency (η) drop. The photon spectral current demonstrates that as dose gamma is increased, the current decreases at shorter wavelengths and the defects are primarily produced near the solar cell's surface. Our findings demonstrate the gamma irradiation-induced breakdown of silicon solar cells and the minority carrier lifetime which demonstrates that the minority carrier lifetimes sharply decline with increasing radiation dose
SYNTHESIS AND CHARACTERIZATION OF CESIUM SILICATE TO DETERMINE ITS DETAILED PROPERTIES AS CHEMISORBED ONTO STRUCTURAL MATERIALS OF LIGHT WATER REACTOR DURING SEVERE ACCIDENT CONDITIONS
Full text linkCesium chemisorption phenomenon strongly contributes to the source terms transport during light water nuclear reactor accidents. Large amounts of cesium silicates are identified to be chemisorbed onto structure material, reduce cesium volatility, and affect the late release and re-vaporization phenomena. Although it has been studied for a long time, several characteristics of these compounds are still under discussion. In this study, Cs2SiO3, Cs2Si2O5, and Cs2Si4O9 were synthesized through the solid-state method and the results have been confirmed using X-Ray Diffraction (XRD) measurement. Furthermore, their crystal structures have been refined based on the XRD analysis. The crystal structure refinement of these compounds proves the previous studies, but with minor distinctions in the lattice parameters. XRD patterns changing over time when measured in the open-air environment also show that Cs2Si4O9 is the most stable species among other cesium silicate species. This indicates that the chemisorbed Cs-Si-O compound onto the structural material as identified by previous studies is most likely Cs2Si4O9 rather than Cs2SiO3 or Cs2Si2O5. Therefore, detailed Cs2Si4O9 identification including its thermodynamic properties characterization could be very useful to enhance the database that is being built to improve current source terms transport codes