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In research reactor, gamma heating deposited in the samples is an important issue because it is related to the samples and the reactor operational safety. Multi Purpose Reactor (MPR) 30 MWth or its called Reaktor Serba Guna G.A. Siwabessy (RSG-GAS) is a research reactor that play a role as a place to irradiate various target material type. RSG-GAS has been operating since 1987 and the last measurement of gamma heating in the core was conducted around twenty years ago in oxide core. In 1996, RSG-GAS core was converted from oxide fuel type to silicide fuel type. The purpose of the conversion are to improve the performance and efficiency of RSG-GAS. As part of RSG-GAS irradiation facilities safety analysis, the gamma heating measurement was re-conducted in order to obtain latest data as benchmark data in silicide core. This paper presents the results of gamma heating measurement of LEU silicide RSG-GAS core in Central Irradiation Position (CIP) at 15 MW and 30 MW power level using gamma calorimeter. There were four types of calorimeter used, which were calorimeter with graphite (C) sample, iron (Fe) sample, aluminum (Al) sample and zirconium (Zr) sample. Gamma heating calculations using GAMSET code were performed to verify the measurement results. The measurement results are lower than the GAMSET results and the gamma heating value increases in proportion to the increase of calorimeter sample atomic number. This results are corresponds to gamma heating benchmarking results of RSG-GAS oxide core. Several optimization efforts both measuring and modeling with GAMSET were conducted as an evaluation and justification of the results. The best optimization results are achieved using the maximum value of the measurement and adjusting the power peaking factor (PPF) distribution. The calculated gamma heating value optimization results at 15 MW power are 2.78 W/g (C sample), 2.74 W/g (Al sample), 3.36 W/g (Fe sample) and 4.60 W/g (Zr sample) while at 30 MW power level are 5.57 W/g (C sample), 5.49 W/g (Al sample), 6.75 W/g (Fe sample) and 9.23 W/g (Zr sample). The best optimization results serve as a benchmark data for developing new gamma heating calculation programs based on 18 gamma energy groups.

A pulse height analyzer (PHA) is used to measure the frequency distribution of the signal output height of an event detector. However, the typical PHA algorithm has limitations in measurement; it cannot distinguish multiplet due to coincidence and singlet pulse. Coincidences will pile up the signals that make the pulse height value higher than it should be. Therefore, this study has developed a simple algorithm to calculate pulse height distribution with coincidence correction (PHA-CC). The developed algorithm firstly identified the coincidence pulse by considering its pulse width. The result shows that the developed algorithm could successfully distinguish multiplet and singlet pulses. The significance of pulse height distribution measured by the PHA-CC compared to regular PHA was also discussed in this paper.

Molten Salt Reactor (MSR) used a freeze valve to handle accident conditions when the reactor overheated and the active safety control systems failed. The freeze valve will melt if the reactor core temperature increases and exceeds the freeze valve’s melting temperature. In this research, a numerical investigation of freeze valve melting behavior was conducted. Freeze valve geometry used geometry proposed by Oak Ridge National Laboratory (ORNL) in Molten Salt Reactor Experiment (MSRE) and also used in test loop facility in Shanghai Institute of Applied Physics (SINAP) to develop Thorium Molten Salt Reactor (TMSR). The material used FLiNaK salt (LiF-NaF-KF, 46.5-11.5-42.0 mol %). Numerical calculation performs using OpenFOAM. BuoyantPimpleFoam was used to analyze the heat transfer mechanism. Isothermal phase-change simulated by adding solidificationMeltingSource terms into fvOptions as a flexible framework in OpenFOAM.

Molten salt has attracted the attention of researchers, especially as a coolant because of its characteristics. This study aimed to analyse the transient and steady-state of the natural circulation system in the molten salt loop with vertical heater and horizontal cooler using computational fluid dynamics. The computational fluid dynamics module in the COMSOL Multiphysics software was used to solve governing equations. Experimental data from several references were used as input data in modelling. The loop was modelled as a rectangular loop with Inconel 625 as the pipe material and molten salt (NaNO3+KNO3) as the working fluid. The variation of heater power was also carried out in modelling. The modelling results showed a good agreement with the experimental results. The results provided that the molten salt natural circulation loop with vertical heater and horizontal cooler showed good flow stability without reversal of flow direction.

The comparison of four blanket modules in DEMO made up the optimization material having a reasonable requirement as blanket material in this study. Either neutron flux distribution in blanket material or material endurance under neutron irradiation, from four modules, the WCLL has a high tolerance neutron distribution and the best neutron irradiation endurance. Furthermore, many suggestions closed to the statement to use the benefits of water coolant and lithium lead (compose Li-6) as a material component in the blanket.

Series of MPS simulations have been conducted using two-dimensional geometry. The simulation was based on Sudha’s experiment (2018) about initial velocity variations on molten Wood’s Metal (WM). The molten WM would be flowed through nozzle with the diameter was 6 mm. It would impinge to the Woods Metal Plate (WMP) which 270 mm below the nozzle. The WMP diameter was 470 mm. The temperature of molten WM and WMP were set at 573 K and 300 K, respectively. The initial velocity of molten WM was varied at 0.327 m/s, 0.397 m/s, 0.498 m/s in the y-negative direction. The simulation was calculated by using 2D MPS with additional procedures such as heat transfer calculation and defining a new type of wall particle. The results showed some different spread patterns, leading edge and phase fraction change for each initial velocity. It can be concluded that with varying the initial velocity will affect on the radial spread pattern but not so much effect occurs on the phase volume fraction change.

Variation of solid ceramic breeding might be one of the excellent candidates in a fusion reactor. The LiAlO2, Li4SiO4, Li2O, and Li2ZrO3 show pretty good requirements in tritium breeding capability and thermodynamic behavior. Especially for LiAlO2 and Li2ZrO3, in which they could be possible to breed without neutron multiplying needed as blanket used generally in order to reach the self-sufficiency reactor. So that, it makes up the material could be possible as high-estimation breeder material.

An advanced nuclear reactor Generation IV, called Molten Salt Reactor (MSR), has been developed with Thorium utilization for a sustainable energy system. In this paper, the study is focused in the neutronic calculation of 250 MWt/100 MWe MSR with Thorium-Plutonium (Th-Pu) fuel salt for 5 years time operation. Fuel salt is composed of a eutectic FLiBe, Th, and Pu, as a coolant, fertile, and fissile nuclide, respectively. Pu loaded is a weapon-grade which consist of 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am. The reactor is operated in the thermal energy range and moderated by the graphite. The reactor design is calculated in neutronic terms with program code CITATION in SRAC 2006 with JENDL 4.0 as a nuclear data library. The result shows some neutronic parameter changes with increasing Plutonium loaded and the reactor criticality is obtained for 5 years by minimal loaded of PuF4 of 2.41%mol. The utilization of Plutonium is described as a capability of MSR in burning a high-level waste of nuclear and radioactive isotopes. This system can be dedicated to future cleaning energy production in a nuclear reactor.

The problem of food waste is increasing as food needs increase as a source of human protein. Therefore, biogas conversion can also answer the problem of household food waste treatment. The purpose of this study is to find out the producing of biogas from a mixture of food waste and liquid waste tofu in anaerobic biodegester at thermophylic temperature. The manufacture of biogas in this study using batch feeding, which is a type of digester that replenishes organic matter is done once full and then waited until biogas is produced. The manufacture of biogas is carried out for 5 weeks. The observed parameters are COD, BOD, TS, VS, and the resulting volume of biogas. From the observations obtained that, by varying substrate can decrease COD, BOD, TS, while substrate temperature can increase VS. The volume of biogas produced in the mixture of liquid waste tofu 60% and food waste 30% is 22.600 ml. The results showed that the percentage of substrate from the mixture of food waste with liquid waste tofu had an effect on increasing biogas production.

Finite element method with linear triangular shape function was used to numerically investigate steady state heat conduction in composite layer. This method divides each layer into a number of small elements. By defining elements, local node, global node, and temperature boundary conditions, an element coefficient matrix can be formed. Element coefficient matrix needs to be assembled to form a global coefficient matrix. The solution of the global coefficient matrix will provide the temperature value at each node. In this research, variations of the thermal conductivity value and temperature boundary conditions are applied. The results of the study will show the effect of different thermal conductivity value against temperature gradient.

The process of fluid descend vertically on static fluid has been studied through experiment and simulation by using Moving Particle Semi-Implicit (MPS) method. MPS has been developed by Koshizuka and Oka (1996). MPS method is a particle method based on the Lagrangian calculation for incompressible medium and does not rely on grid system. This study has been done by doing simulation first then validated by experiment. Water and cooking oil were used in the experiment. This study was carried out to analyze distance and density influence in the height of static fluid bursting process. The experiments were conducted in an acrylic box with the dimension 150 mm x 40 mm x 30 mm and a bottle with height of 100 mm and the nozzle of bottle was 26 mm. The distance between nozzle and surface of fluid inside box is varied for 100 mm and 200 mm. The results show that in the same height, water will be more difficult to be moved than cooking oil because water has bigger density. The distance between nozzle and fluid surface inside the box will affect at the pressure which is received by the fluid. Higher distance will create higher pressure so that fluid inside the box will be moved easier and more fluid will split out to the box. It also gives some influence in the final condition of remaining fluid in the acrylic box.

Molten Salt Reactor (MSR) is one of the next-generation nuclear power systems known as Generation IV. In this paper, the FUJI-U3 MSR type is designed for neutronic analysis that natural uranium is used as a fuel. The standard reactor core model is adopted due to the flux flattening effect in the active core, as reported from previous research. FLiBe eutectic is used both as the coolant salt and moderator in the reactor core. The neutronic calculation is performed using the SRAC 2006 program: PIJ for pin cell calculation and Citation for core calculation. JENDL 4.0 is used as a nuclear data library that provides the cross-section data of the nuclides. The neutronic parameter results, such as the effective multiplication factor, conversion ratio, neutron spectrum, and power density, are obtained from the calculation. The reactor can achieve the criticality condition with a minimum loading Uranium Tetrafluoride (UF4) of 3.6%mol for 2000 effective full power days (EFPD). The results calculation provide some parameter survey for the research and the future development of MSR.

Indonesia has a resource of nuclear fuel which spread around the country, especially in Bangka Belitung, West Kalimantan, West Sulawesi, and Papua. Therefore, Indonesia has a potential in the nuclear energy development. The Generation IV nuclear power system has been developed by the Generation IV Forum (GIF) with its benefits such as highly economical, enhanced safety, low waste production, and proliferation resistant. MSR is one of Generation IV nuclear power system which uses molten salt as a fuel and a coolant. Based on RUPTL (Electricity Supply Business Plan) PLN 2018-2027, Indonesia needs 100 MWe power harness from nuclear energy. This study focuses on a FUJI-U3 reactor, a type of MSR which developed in Japan. Power operation varied from 120-150 MWth 60-100 MWe and using FLiBe as molten salt. The calculation was performed with PIJ and CITATION modules in SRAC 2006 and JENDL 4.0 as nuclear data library.

The design study of GFR concepts comprises neutronic analysis of fuel pin and fuel assembly. The Monte Carlo method has advantages in three-dimensional (3D) geometry modeling but requires a high computation time. In this research, the comparative study of Gas-cooled Fast Reactor (GFR) using the Monte Carlo code. The GFR feasibility design study will be carried out with natural uranium with plutonium as fuel cycle inputs. The Monte Carlo method simulates GFR model at full-scale and heterogeneous three-dimensional (3D) using Evaluated Nuclear Data File (ENDF/B-VIII.b5) nuclear data. The code of Monte Carlo methods will be used in this research are the Monte Carlo N - Particle (MCNP) and OpenMC. The comparison of the GFR fuel assembly calculation simulation results is made between the MCNP and OpenMC code. The equilibrium cycle configuration is used as the basis model for the comparisons. The comparison of MCNP and OpenMC code gives a good agreement in criticality calculation of GFR that achieves delta kinf less than 1%. The (U-Pu)N fuel is a good candidate to be chosen in GFR research that gives kinf more than 1.1 in fissile contain 10%Pu and has the highest thermal conductivity. The Zircaloy-4 is the best candidate for material cladding in GFR design that provides the highest kinf.

In this article, we employ a numerical calculation to determine the Pm2O3-Si p-n junction betavoltaic electrical performance using Finite Difference Minority Carrier Diffusion Equation (FD-MCDE) method. In order to verify the method, the comparation result is presented according to the basic planar design in the experimental study as well as the analytical calculation. At 0.8 Ci/cm2 and 1 Ci/cm2 activity content, we obtained 8.04% and 7.03% error in short-circuit current calculation, respectively. Furthermore, the variation in the number of n-type and p-type arrays was done to detect the shifting effect in the simulation result. The V oc, Vmp , and FF for both activity contents have achieved the optimum values using 500 x 500 grids, despite of having more iterations to reach the desired convergency level.

The limiting efficiency of betavoltaic battery can be affected by the self-absorption of radioisotope source. Moreover, this effect can be more detrimental for a low energy beta emitter such as Ni-63. The cylindrical design is known to have a benefit over rectangular one in response to the self-absorption effect. In this study, the model of betavoltaic was built under MCNPX which also capable to do many particle simulation based on Monte Carlo method. The interaction of beta particle and its average energy deposition in 4H-SiC semiconductor as the radiation-electricity converter was calculated, along with photonic emission due to the bremsstrahlung event. Basically, the average energy deposition shows a deterioration along with the addition of source mass thickness. However, an optimum peak was observed at about 0.1424 mg/cm2 source mass thickness in the N+ substrate average beta energy deposition. The cylindrical surface magnification also give a positive influence to the energy deposition. In addition, the absorbed dose in each layers of semiconductor was also investigated, and the results were consistent to the semi-empirical equation of self-absorption theory.

Augmented reality (AR) technology is a visual technology that combines virtual world objects into real world views in real time. By utilizing augmented reality technology on an android smartphone, a geometry object of space can be visualized concretely through three-dimension virtual modeling that is similar to the original object. This research incorporates AR technology into 3dimension space geometry objects and then it is implemented in three dimension objects lectures for cadets of Ship Machinary Study Program. The purpose of this research is to find out: how the cadets response in using the AR application. This application is an application that runs on the android mobile platform, where the AR application requires video streaming taken from a smartphone camera as an input source, then this application will detect markers using a tracking system, after the markers have been detected, the object model in the geometry of the space will appear above the marker as if the object model is real. The methods or stages in designing this application use the Prototype Model, whereas the implementation of the application uses a one-group pretest-posttest design pattern. The results showed that the use of the AR application on the Android platform was able to get a positive response from cadets in lecturing the Concept of Geometry Course, especially three-dimension objects.

Two-dimensional simulation of Wood’s Metal Plate (WMP) Melting Process has been conducted by varying the molten temperature. This simulation was performed by using Moving Particle Semi-Implicit (MPS) method which can calculate the Navier-Stokes and heat transfer equation without divining mesh system. As a substitute, particles are selected to represent every part of simulation such as fluid, wall and dummy type. The experiment of WMP melting process has been conducted by Sudha (2018). The molten Wood’s Metal was originally set at 573 K. In this simulation, the holes formation in WMP will be investigated as a consequence of molten temperature variation. Furthermore, temperature distribution at top side of WMP and solid-liquid phase change will be analyzed for initial temperature of molten are 523 K, 473 K and 423 K. The results show that the phases change significantly in a short time when the molten temperature is increased. However, until the end of simulation which on 1.5 s just for molten temperature 573 K shows the breaching process to the WMP. This study can be implemented when the reactor nuclear severe accident happened.

Modular Gas-cooled Fast Reactor (GFR) is one of six advanced reactor concepts set by the generation IV international forum. Modular GFR has the potential for use actinide recycling and closed fuel cycle as well as applying fast reactor, using helium gas as the main coolant, high working temperature and low void reactivity effect. The neutronic analysis of nuclear reactor means behavior study of subatomic particles that interact with matter. In this paper, the feasibility of plutonium fuel in modular Gas-cooled Fast Reactor (GFR) was investigated. The Monte Carlo method has advantages in full-scale and heterogeneous three-dimensional (3D) geometry modeling using Evaluated Nuclear Data File (ENDF/B-VIII.b5) nuclear data but requires a highly computation time. Since the progress of high performance computing, the reactor physicist community began proposing to use Monte Carlo method for nuclear reactor simulation through the parallelization of calculations. The GFR feasibility design study will carried out with plutonium fuel as fuel cycle inputs with 5-25% of fissile contain. The most important neutronic parameters characterizing of GFR core are determined for beginning of life (BOL) and during burnup calculation conditions. The results of calculation series in parallel computing give the good agreement will be faster of calculation time when more threads. Materials of (U-Pu)C and (U-Pu)N fuel are the good candidates to be chosen in GFR research that give keff more than 1.2 in fissile contain 20%Pu. The variation fissile contain gives the linearity with keff. In depletion simulation, the core reactor still in critical during 20 years operation, burn up values linear with operation time and mass evolution of plutonium and uranium from start-up core to equilibrium core.

As a free CO2 emission energy source, nuclear energy can supply the world electricity for very long time, even for thousand years. However, to increase the public acceptance, nuclear spent fuel (high level nuclear waste) should be manage to minimize its radio-toxicity. There are several methods of processing nuclear spent fuel including geological repositories, reprocessing it to be used as fuel again or doing transmutation. In this preliminary study, the nuclear high level waste is transmuted by using an Accelerator Driven Subcritical System (ADS). ADS is a system consisting of a subcritical reactor core and a particle accelerator. Nuclear waste which will be transmuted is plutonium (Pu) and minor actinides (MA). Pu and MA are mixed in the ADS fuel that consists of U-ThO2(NO3)2, PuMA(NO3)4 dan HNO3. Two type of plutonium vectors have been evaluated, those are reactor grade Pu (RGPu) and weapon grade Pu (WGPu). The ADS can maintain it sub-criticality condition for RGPuMA and WGPuMA with the concentration of PuMA-Nitrate are above 62.6 kmol/m3 and 65.6 kmol/m3, respectively.

Reaktor Serba Guna G.A. Siwabessy (RSG-GAS, previous name MPR-30) is the largest research reactor in Southeast Asia that acts as a national facility to irradiate material. Gamma heat is a very important factor for the safety analysis in every material irradiation activity at the irradiation facility. Gamma heat is the main research topic of several forms of safety investigations at world research reactors. Gamma heat information is useful to predict the temperature of the material to be irradiated and negligence in predicting gamma heat can cause overheating. Gamma heat values are very dependent on the characteristics of the reactor core. Changes in reactor power can affect the core characteristics. RSG-GAS is designed to have a 30 MWth of nominal power but it is currently operated at 15 MWth power level. In this study observed changes in gamma heat as a function of reactor power and material target on the Central Irradiation Position (CIP) silicide core of RSG-GAS by using a modified GAMSET program. Modifications are made by adjusting the material and power configuration in the core. The results of the analysis show that gamma heat will be increase in accordance with the increase in power level in various material targets such as graphite (C), aluminium (Al), iron (Fe) and zirconium (Zr). Gamma heat also tends to increase according to the increase of atomic number target. Verification has been carried out with the results of calculations in the 35 MWth CEA Grenoble reactor with the smallest yield difference of 1.23% in the graphite target and 2.71% in the iron target.

The research on the developments of thermoluminescence (TL) material in nano- and microstructure are increasing in recent years. There are many new techniques developed for preparing TL material that aim to improve and investigate the structure, morphology, and TL characteristics. In this study, the TL characteristics of CaSO4 doped with Dysprosium (Dy) were being observed. CaSO4:Dy was performed by co-precipitation method with 0.1 mol% Dy concentration. The obtained sample was irradiated with Cs-137 for gamma source and Sr-90 for beta source with doses of 5, 10, 15, 20 and 25 mGy. After irradiation, the structure, morphology, and TL response were observed by using XRD, SEM, and TLD Reader Harshaw 3500. XRD results showed that the average crystallite size for CaSO4:Dy were 58 nm. But, the presence of dysprosium could not be clearly observed from XRD results. The characterization results from SEM for CaSO4:Dy had shape resembling needle-like crystal. The TL response of CaSO4:Dy was compared from beta and gamma radiation source. TL responses for Cs-137 source were 26.56, 34.09, 51.85, 75.97, and 86.29 nC for 5, 10, 15, 20, and 25 mGy, respectively. Sr-90 source with the same dose obtained TL responses as follows: 35.67, 68.60, 100.87, 121.69, and 179.85 nC. The shapes of the glow curves were almost identical, but these curves had different intensities. The TL response from β-ray source obtained a higher intensity than from γ-ray source because the beta source is a particle with mass so that it will have more interaction than gamma source in TLD material.

(Th,U)O2 fuel has been utilized for neutronic analysis of HTTR 30 MWt which is a high-temperature test reactor with helium gas as a coolant that has been built in Japan. HTTR has a Triple-isotropic Coated Fuel Particle (TRISO CFP) type of fuel. Homogenous fuel of (Th,U)O2 in the active core reactor give some different results for neutronic analysis such as effective multiplication factor (k-eff) each block fuel and whole of active core, conversion ratio, and spectrum of neutrons. The calculation was calculated by using neutronic SRAC2006 code and based on JENDL4.0 nuclear data which has been developed by JAERI. The results show that the optimum core may be obtained when (Th,U)O2 has 60% of ThO2 and 16.8% enrichment of UO2. The spectrum neutron shows that (Th,U)O2 fuel results in a softer spertrum.