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61	Radiation Physics	ABS-68
Data Acquisition of Gamma Radiation Measurement Using Long Range (LoRa) Radio Communication Santiko Tri Sulaksono (a,b*), Zalfa Raihan Salsabila (c), Danang Supriyanto (d), Haryo Seno (e), Dani Muliawan (d), Gallant Tsany Abdillah (d), Satrio Aris Setiawan (d) a) Department of Doctoral Nuclear Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Ganesha 10 Street, Bandung, 40132, Indonesia b) Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Tamansari 71 Street, Bandung, 40132, Indonesia *sant012[at]brin.go.id c) Department of Electrical Engineering, Faculty of Science and Technology, Universitas Muhammadiyah Bandung, Soekarno-Hatta 752 Street, Bandung, 40614, Indonesia d) National Research and Innovation Agency (BRIN), Tamansari 71 Street, Bandung, 40132, Indonesia e) Research Center for Nuclear Beam Analysis Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Tamansari 71 Street, Bandung, 40132, Indonesia Abstract Measurement and monitoring of gamma radiation in nuclear installation areas or other areas are commonly carried out for research or radiation protection programs. Measurements in large areas require real-time data acquisition and long-distance data transmission using Long Range (LoRa) radio communication. This study aims to obtain the effectiveness of gamma radiation measurement data transmission using LoRa. Gamma radiation is measured using a Geiger Muller (GM) counter based on the ATMega 328 microcontroller and data transmission using the LoRa RF96 915MHz module. The distances, obstacles, and the length of data delivery time are the parameters observed in this study. The comparison between the GM counter measurement data, as a transmission module, and the data on the receiver module shows no significant differences. Therefore, it can be concluded that the gamma radiation measurement data acquisition system using LoRa functions effectively and can be implemented for radiation monitoring and long-distance measurements. Keywords: gamma radiation monitoring, radiation protection, Long Range radio communication, LoRa Share Link | Plain Format | Corresponding Author (Santiko Tri Sulaksono)

62	Radiation Physics	ABS-70
A Comparative Analysis of Organic and Inorganic Scintillation Detectors for Monitoring Environmental Radiation Using Geant4 Simulation Imam Ghazali Yasmint1,2, Sidik Permana2,3, Zulfahmi3 1 Joint Doctoral Program for Sustainability Research, The University of Electro-communications, Tokyo, Japan 2 Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia 3 Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia Abstract The number of nuclear physics technologies worldwide is steadily growing. Not just in the domain of nuclear reactors, but also in various other domains such as nuclear physics for medical, agricultural, industrial use, and more. As nuclear utilizes continue to grow, the flow of radioactive material will become more extensive. This provides additional difficulties in terms of monitoring. The purpose of this monitoring is to guarantee that no highly radioactive materials are being freely distributed throughout the community. Radiation detectors are essential for carrying out monitoring. A scintillation detector is currently being developed as a kind of radiation detector due to its superior efficiency compared to other detector designs. Scintillation detectors utilizing organic materials, such as plastic, are specifically designed to monitor radiation levels in the environment. This study aims to conduct a comparative analysis of organic and inorganic scintillation detectors using the Geant4 simulation code. The outcomes of this comparison will serve as a benchmark for evaluating the sensitivity of the organic scintillation detector. Keywords: Efficiency, Geant4, Nuclear Radiation, Radiation Monitoring, Scintillation Detectors Share Link | Plain Format | Corresponding Author (Imam Ghazali Yasmint)

63	Radiation Physics	ABS-71
Assessment of Transfer Factors for Natural Radionuclides and Radiocesium from Soil-to-Plant and Plant-to-Cow^s Milk on a Cattle Farm in Lembang Imam Ghazali Yasmint1,2, Abdul Waris2,3,4, Sidik Permana2,3,4, Eko Pudjadi5, Ismail Humolungo3, Deni Karsa Sondana4 1 Joint Doctoral Program for Sustainability Research, The University of Electro-communications, Tokyo, Japan 2 Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia, 40132 3 Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Ganesha No. 10 Bandung, Indonesia, 40132 4 Department of Physics Teaching, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia, 40132 5 Environmental Laboratory, Subs-section of Environmental Safety, Radioecology Section, Center for Technology of Safety and Radiation Metrology, National Nuclear Energy Agency of Indonesia, Lebak Bulus Raya No.49 Jakarta, Indonesia, 12440 Abstract Besides nuclear reactors, humans are exposed to radiation from natural sources. Natural radioactivity in the environment originates from Uranium and Thorium Radionuclides exposure, typically found in the soil. The detection of radiocesium in the soil signifies the existence of residual byproducts resulting from nuclear reactor disasters and nuclear weapons experiments. Radionuclides present in the soil may enter a process known as soil-to-plant transfer, where they are transmitted to plants such as grass and vegetables. Cows and other animals commonly consume grass as their primary food source. The existence of radionuclides in the grass might indirectly impact human health when people consume meat and milk from these animals (via the transfer of radionuclides from plants to cow^s milk). Hence, it is necessary to conduct studies on soil, grass, and cow^s milk to determine the amount of natural radiation that enters the human body and to understand the transfer of radionuclides. Lembang, a sub-district in West Java Province, Indonesia, has emerged as the focal point for agricultural and veterinary education and research in the region. This study involved the collection of soil, grass, and cow^s milk samples from a cattle farm located in Lembang. Subsequently, the radioactive radiation was quantified with an ORTEC gamma spectrometer equipped with an HPGe detector. The radionuclides detected in this investigation were 226Ra, 232Th, 40K, and 137Cs. The activity concentration in soil is lower than the global average. The transfer factor obtained for soil-to-plant and plant-to-cow^s milk is consistent with the findings of prior research, which have demonstrated that 40K exhibits the highest transfer factor compared to other radionuclides. The absence of 137Cs in milk samples enabled the determination of its transmission mechanism solely from soil to plant. Keywords: Activity Concentration, Gamma Spectroscopy, Natural Radionuclides, Plant-to-Cow^s Milk, Radiocesium, Soil-to-Plant, Transfer Factor Share Link | Plain Format | Corresponding Author (Imam Ghazali Yasmint)

64	Radiation Physics	ABS-72
Evaluation of a PVT-Based Scintillation Detector as a Cost-Effective Early Detection Solution in Nuclear Reactor Safety Systems Imam Ghazali Yasmint1,2, Sidik Permana2,3, Rasito Tursinah3,4, Amila Amatullah3, Ismail Humolungo3 and Marisa Variastuti3 1 Joint Doctoral Program for Sustainability Research, The University of Electro-communications, Tokyo, Japan 2 Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia 3 Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Ganesha No. 10 Bandung, Indonesia 4 Nuclear Energy Research Organization, National Research and Innovation Agency (BRIN), Indonesia Abstract The safety measures for nuclear reactors are constantly advancing in line with the progress in utilizing nuclear reactors as power sources. Furthermore, the capabilities of early detection systems in nuclear reactors are always advancing to provide the most efficient detectors. Gamma radiation rise serves as an indicator in the safety system of nuclear reactors, specifically in relation to the gamma radiation detector. An outcome of the advancement of this detector is the creation of detectors that possess exceptional capabilities but come with an expensive cost. Poly Vinyl Toluene (PVT) plastic can function as a gamma detector and is cost-effective. This study aims to investigate the possibility of utilizing PVT-based scintillation detectors as a cost-effective solution for incorporating early detection capabilities into nuclear reactor safety systems. The work involved the characterization of PVT-based scintillation detectors, followed by a comparison with commercially available detectors. The primary focus of this research is the efficiency of the detector parameter. Keywords: Efficiency, Gamma Radiation, PVT, Scintillation Detector Share Link | Plain Format | Corresponding Author (Imam Ghazali Yasmint)

65	Radioactive Wastes	ABS-24
Study Of Transfer Factors For Radionuclide And Heavy Metal In Bananas (Musa Paradisiaca) At Tarahan-Lampung As Supporting Data For National Food Security Rusbani Kurniawan, Radhia Pradana, Agus Nur Rachman, Ilma Dwi Winarni, Riyaz Yusuf, Nicha Miranda Sari, Ahmad Anggraria Jaya Agung, Wahyudi, Oumar Bobbo Modibo, Ilsa Rosianna, Eka Nugraha Djatnika National Research And Innovation Agency - BRIN Abstract The Tarahan-Lampung Coal Steam Power Plant (CFPPs) is one of the largest CFPPs on the island of Sumatra. CFPPs may produce hazardous trace elements (HTEs) and contribute to air pollution. Since bananas are a major commodity in Lampung, it^s crucial to assess the presence of radionuclides and heavy metals in bananas intended for consumption. To accomplish this, a study was conducted to analyze the activity of radionuclides (226Ra, 228Th, 40K, 137Cs) and the heavy metal content (Fe, As, Pb, Hg, Zn) in soil and banana samples from the industrial area around CFPPs-Tarahan. The analysis involved using a High Purity Germanium Detector (HPGe) gamma spectrometer to determine radionuclide activity concentration and X-Ray Fluorescence (XRF) to measure heavy metal concentrations. The highest activity concentration for 226Ra, 228Th, 40K, and 137Cs in soil were 54, 64, 714, and 0.4 Bq/kg, respectively. The concentrations of heavy metals Fe, As, Pb, Hg, and Zn were also determined, producing the following values: 39100, 14, 27, 3, and 221 &#956-g/g. For bananas, the highest activity values of 226Ra and 40K were 15 and 414 Bq/kg, respectively, with no activity of 228Th and 137Cs found. The highest concentrations of heavy metals Fe, Pb, and Zn in dried and wet banana samples were 68, 1.9, 11, and 46.9, 1.4, 7.3 &#956-g/g, respectively. The average value of the radionuclide transfer factor from soil to plants for each radionuclide 226Ra and 40K was 0.3 and 1.2, respectively, while transfer factor values for 228Th and 137Cs were not detected. Keywords: CFPPs, Gamma spectrometer, X-ray fluorescence, Erica Tools, Transfer factor, radionuclides Share Link | Plain Format | Corresponding Author (Rusbani Kurniawan)

66	Radioactive Wastes	ABS-35
Distribution of Radon-222 Radioisotope in Soil at Siwabessy Science and Technology Area Lebak Bulus Jakarta Satrio Gilang Ismaya (a*), Rasi Prasetio (b), Agus Budhie Wijatna (a) (a) Department of Nuclear Engineering and Engineering Physics, Faculty of Engineering, Universitas Gadjah Mada, Indonesia 55281 *satrio.gilang.i[at]mail.ugm.ac.id (b) Research Center for Radiation Process Technology, National Research and Innovation Center, Indonesia 15324 Abstract Two temporary storage sites for nuclear mineral waste at KST Siwabessy Lebak Bulus, Jakarta have the potential to increase radon gas, a natural radioactive gas resulting from uranium decay. Radon-222 (\(^{222}Rn\)) is an alpha emitter contributing 1.15 mSv/year per capita to natural radioactivity in the world. Research at KST Siwabessy found that the temporary storage of nuclear mineral waste influences the distribution of \(^{222}Rn\). The soil radon concentration ranges from 680 to 18,410 Bq/\({m^3}\), with anomalies at points 14 and 19 (volleyball court and waste pool). These two points have a much higher value compared to the rest of the measurement points at KST Siwabessy, but are still within typical levels of 2,000 to 50,000 Bq/\({m^3}\). The temporary nuclear waste storage site can cause high radon concentrations, but the direction is random and limited to a distance of 75 meters. Keywords: radon- distribution- concentration- soil- gas Share Link | Plain Format | Corresponding Author (Satrio Gilang Ismaya)

67	Theoretical and Computational Nuclear Physics	ABS-18
ASSESSMENT OF GAMMA HEATING IN THE GRAPHITE-MODERATED MOLTEN SALT REACTOR TMSR-500 USING OPENMC SIMULATIONS Nurul K Fitriyani*, Rida S.N. Mahmudah, Azizul Khakim *Physics departemant, faculty of mathematics and natural science (FMIPA) Yogyakarta State University, Sleman 55581, Indonesia Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Building No. 80 BJ Habibie Science and Technology Area, Setu, South Tangerang 15314, Indonesia. Abstract Ensuring the safety and efficiency of the TMSR-500 reactor, a Generation IV Molten Salt Reactor (MSR), is crucial for its planned development in Indonesia. This study evaluates gamma heating exposed to the graphite moderator, contributing to the main heat source in the graphite. The TMSR-500, an advancement of the Molten Salt Reactor Experiment (MSRE), utilizes molten salt fuel, is graphite-moderated, and operates at high temperatures. The effective moderation by graphite reduces neutron energy from fast to thermal levels, ensuring optimal fission sustainability and impacting gamma heating. OpenMC simulations were employed using the Python programming language and Monte Carlo methods to assess gamma heating with ENDF/B-VII.1 nuclear data and a particle count of one million. The simulations were conducted on The National Research and Innovation Agency^s computer facilities. The maximum gamma flux was observed at energies greater than 1 eV, with a measured value of 1.8096 x 10^(14) photons/cm^2 sec. The gamma heating in graphite was found to be 3.4917 W and the corresponding volumetric gamma heating was 9.1 x 10^(-3) W/cm^3. The results indicate that gamma heating in the TMSR-500 has a percentage of 2.0138 of the reactor^s total power, within the level of common knowledge. Keywords: TMSR-500, gamma flux, gamma heating, graphite moderator, OpenMC Share Link | Plain Format | Corresponding Author (Nurul Kumala Fitriyani)

68	Theoretical and Computational Nuclear Physics	ABS-20
Validation of TRIAC Code for Tangential Stress of TRISO Coated-Particle Based on The Experimental Result at Tecdoc. IAEA 1674 Chapter 9t Muhammad Ilham Bayquni (a), Fajar Al Afghani (a,b*), Rohmat Sigit Eko (a), Fitri Miftasani (c), Nini Widiawati (c), Topan Setiadipura (c), Arya adhyaksa waskita (d), Anik Purwaningsih (c), Helmi Fauzi (a), Imam Abdurrosyid (a), Putra Octavianto (a) a) Research Center for Nuclear Material and Radioactive Waste, National Research and Innovation Agency, Tangerang Selatan, 15314, Indonesia *faja020[at]brin.go.id b) Department of Physics, Faculty of Mathematics and Natural Science, Universitas Indonesia, Depok, 16424, Indonesia * c) Nuclear Reactor Technology Research Center, National Research and Innovation Agency, Tangerang Selatan, 15314, Indonesia d) Data and Information Science Research Center, National Research and Innovation Agency, Tangerang Selatan, 15314, Indonesia Abstract TRISO-coated particle fuel is extensively utilized in high-temperature gas-cooled reactors and other advanced reactors. The performance of these coated fuel particles is crucial for reactor safety. It^s essential to assess and determine the failure probability of coated fuel particles using appropriate fuel performance models and methods under both normal and accident conditions. To enhance the design process of coated particle fuel, a new TRISO fuel performance code-named TRIAC-BATAN was developed. This code is designed to calculate internal gas pressure, mechanical stress, and failure probability of coated fuel particles. This paper introduces TRIAC and benchmarks it against IAEA CRP-6 (tecdoc IAEA 1674 chapter 9) benchmark cases for analyzing coated particle failure, especially for tangential stress maximal. TRIAC-BATAN^s results align well with benchmark values, demonstrating its accuracy and applicability, particularly for tangential stress maximal. This work establishes a reliable foundation for the application of TRIAC-BATAN. Keywords: validation, TRIAC-BATAN, tangential stress, computational Share Link | Plain Format | Corresponding Author (Fajar Al Afghani)

69	Theoretical and Computational Nuclear Physics	ABS-25
Thermalhydrolic Analysis on Hot Channels of TRIGA Kartini Reactor with Computational Fluid Dynamics (CFD) Using OpenFOAM Aisah Rizqi Amaliyah (a), Ratna Dewi Syarifah (a*), Nuri Trianti (b), Tri Nugroho Hadi Susanto (c) a) Department of Physics, Faculty of Mathematics and Natural Sciences, University of Jember, Jember, Indonesia *rdsyarifah.fmipa[at]unej.ac.id b) Research Centre for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency, Bandung, Indonesia c) Kartini Reactor Installation, Directorate of Nuclear Facility Management, National Research and Innovation Agency, Yogyakarta, Indonesia Abstract To address the growing energy demands in Indonesia, nuclear power plants, including the development of nuclear reactors, are being explored. One such reactor is the TRIGA Kartini reactor in Yogyakarta, primarily utilized for research and educational purposes. This study analyzes the thermal-hydraulic behavior of hot channels within the TRIGA Kartini reactor, operating at 100 kW, with a focus on temperature distribution and fluid flow-critical factors for ensuring reactor safety and efficiency. The research employs a geometric model, power-to-temperature conversion calculations, and numerical simulations using Computational Fluid Dynamics (CFD) in OpenFOAM. Results from these simulations are compared with experimental data from IFE, which recorded a temperature of 305K in the upper section of the reactor. The simulations are executed using two solvers, Fluid and XiFluid, under four flow conditions: laminar, k-epsilon, k-omega, and k-omegaSST. The findings reveal that the laminar flow condition produces the largest error, exceeding 1% for both solvers, followed by k-omegaSST with errors above 0.6%. For the Fluid solver, the k-omega model shows a 0.30% error, while the k-epsilon model yields the lowest error at 0.23%. Conversely, for the XiFluid solver, the k-epsilon model results in a 0.62% error, with k-omega closely behind at 0.29%. The k-epsilon model using the Fluid solver provides the closest alignment with the experimental data, making it the most accurate among the tested conditions. Keywords: Thermal-hydraulic- TRIGA Kartini reactor- OpenFOAM- CFD Share Link | Plain Format | Corresponding Author (Aisah Rizqi Amaliyah)

70	Theoretical and Computational Nuclear Physics	ABS-28
Neutronic Analysis of Small Long-Life Modular Boiling Water Reactor (BWR) with Thorium Nitride Fuel Using OpenMC M. Khanifuddin Zuhri (a), Ratna Dewi Syarifah (a*), Nuri Trianti (b) a) Department of Physics, Faculty of Mathematics and Natural Science, Universitas Jember, Jember, Indonesia b) Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency, Bandung, Indonesia *rdsyarifah.fmipa[at]unej.ac.id Abstract Nuclear power plants (NPPs) are an alternative to fossil fuel-based power plants, such as coal, which harm the environment and human health. NPPs are built to meet electricity needs and have advantages like being environmentally friendly and economical. One reactor type in NPPs is the boiling water reactor (BWR), currently being developed into a small modular reactor (SMR). The operational duration of an SMR-type BWR can be determined by analyzing the effective multiplication factor (keff) values generated during the combustion process. This research used a homogeneous fuel design with thorium nitride, 8% enriched U-233, and 6% Pa-231. This research aims to determine the most optimal operational outcome for a small long-life BWR without refueling for 30 years. The research was conducted through simulations using the OpenMC program to obtain the most optimal keff value. The most optimal keff value at each parameter optimization stage was then used in the subsequent stages. The final results show that the BWR achieved the most optimal keff value using parameters axial steam percentage, fuel compound density, ENDF-8 library, fuel pin arrangement without assembly, nitrogen-14 nuclide, thermal scattering data, an input power of 100 MWth, and a 66% fuel fraction. These parameters allow the BWR to operate for 30 years with a maximum excess reactivity of 1.96% &#8710-k/k. Keywords: Neutronics, Boiling water reactor, Thorium nitride, OpenMC Share Link | Plain Format | Corresponding Author (M. Khanifuddin Zuhri)

71	Theoretical and Computational Nuclear Physics	ABS-29
Neutronics Analysis of Heterogeneous Core of Small Modular Reactor Type GFR Thorium Nitride Fueled using OpenMC Arroofi Candra Kartiko1, Ratna Dewi Syarifah1*, Nina Widiawati2, Artoto Arkundato1, and Fajri Prasetya3 1 Department of Physics, Faculty of Mathematics and Natural Sciences, Jember University, Jember, Indonesia. 2 Research Center for Nuclear Reactor Technology, BRIN, Indonesia 3 Department of Nuclear Science and Engineering, Bandung Institute of Technology, Bandung, Indonesia Abstract The use of mixed fuel in the form of thorium-uranium nitride (ThN-UN) offers the potential to improve thermal efficiency and optimize criticality, especially in fast reactors like Gas-cooled Fast Reactors (GFRs). To maintain its criticality, Th-232 absorbs neutrons and transforms into a new fissile material U-233. The inclusion of fertile material Th-232 necessitates a more precise reactor geometry design to ensure the reactor remains critical state until the end of the burn-up period. This study aims to compare variations in the percentage of U-233 enrichment in heterogeneous core geometries with five fuel variations (F1:F2:F3:F4:F5) using the OpenMC program at a power of 100 MWth. Benchmarking is performed by measuring the effective multiplication factor (k-eff) over 5 years of burn-up using reference data from previous studies. If the error value is less than 2%, further calculations will be conducted for both homogeneous and heterogeneous core configurations. The homogeneous calculations indicate that the U-233 enrichment percentage of 8.5% yields the most optimal results. This homogeneous data then is used for calculations in the heterogeneous core with 5 fuel variations. The heterogeneous core configuration is designed using five types different fuel percentages cases, each with variations in ring geometry. The comparison results show that the case 5 heterogeneous core geometry design, with a percentage distribution of 7%:7.5%:8.5%:9%:10,5%, achieves good optimization in terms of k-eff value, excess reactivity, neutron flux, and extended burn-up over a 10-years period. Keywords: GFR- SMR- OpenMC- Thorium Nitride Share Link | Plain Format | Corresponding Author (Arroofi Candra Kartiko)

72	Theoretical and Computational Nuclear Physics	ABS-41
Phenomenon Study of Heat Transfer from Clad Surface to Coolant in Pressurized Water Reactor (PWR) Duwi Hariyanto(a), Sidik Permana(b)(c), Alan Maulana(d) (a) Industrial Technology Faculty, Institut Teknologi Sumatera, Jl. Terusan Ryacudu, Lampung, Indonesia (b)Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesha 10, Bandung, Indonesia (c)Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesha 10, Bandung, Indonesia (d)Research Center for Nuclear Reactor Technology, National Research and Innovation Agency, Indonesia Abstract The heat transfer study from the clad surface to the coolant has become a concern in the thermal analysis of nuclear reactors. The study was proposed to model the phenomenon of heat transfer from the clad surface to the coolant in a pressurized water reactor (PWR). Finite element-based modelling was carried out using the computational fluid dynamics (CFD) module in the Fluent software. In the modelling, heat generation was considered only in the fuel pellets, and the coolant flow rate was varied from bottom to top with the help of a pump. The modelling results showed that the lower the coolant flow rate, the higher the temperature on the clad surface. The coolant flow rate of 5,0 m/s can keep the coolant in the liquid phase. Keywords: pressurized water reactor, thermal analysis, clad, heat transfer, computational fluid dynamics, Fluent Share Link | Plain Format | Corresponding Author (Duwi Hariyanto)

73	Theoretical and Computational Nuclear Physics	ABS-50
Investigative Study of Fuel Salt Drainage from Molten Salt Reactor (MSR) to Drain Tank Using Moving Particle Semi-Implicit (MPS) Method Dhiya Salma Salsabila (a), Fajri Prasetya (a*), Fabian Yoga Prastha (a), Tsania Eksa Angelina (a), Asril Pramutadi (b), Sidik Permana (b) a) Master Program in Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10 Bandung 40132, Indonesia *prasfajri[at]gmail.com b) Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology Jalan Ganesha 10 Bandung 40132, Indonesia Abstract This study investigates the dynamic behavior of fuel salt drainage within a molten salt reactor system, focusing on pressure distribution, velocity profiles, and volume clearance rates over time. Detailed analyses at discrete time points reveal distinct pressure patterns influenced by the density and viscosity of three different fuel salt compositions (A, B, C). Initial high pressures at 0.2 seconds are attributed to collisions as the fuel salt first interacts with pipe corners. Subsequent time intervals show varying pressure ranges, with fuel salt A consistently exhibiting elevated pressures in regions 2 and 3, indicative of its high viscosity and density characteristics. Velocity profiles highlight that fuel salt C has higher flow velocities, attributed to its lower viscosity, which significantly influences fluid dynamics. Additionally, volume clearance rates indicate faster reactor clearance for fuel salt C at 800K, owing to its lower density and consequent lower kinematic viscosity. These findings underscore the complex relationship between fuel salt composition, viscosity, and reactor system dynamics, which is crucial for optimizing reactor design and operational efficiency in molten salt reactor technology. Keywords: MSR, MPS, Fuel Salt, Pressure, Velocity Share Link | Plain Format | Corresponding Author (Fabian Yoga Prastha)

74	Theoretical and Computational Nuclear Physics	ABS-56
Simulation on Neutron Imaging Facility Based on 30 MeV Cyclotron Dian Adi Prastowo, Fahrurrozi Akbar, Andeka Tris Susanto, Rissa Damayanti, Anas Fahmi Imron, Haryo Seno, Rasito, Ridwan, Mujamilah, Abu Khalid Rivai Research Center for Nuclear Beam Analysis Technology National Research and Innovation Agency of Indonesia Abstract A new research facility based on 30-MeV cyclotron is now being developed at the National Research and Innovation Agency (BRIN) of Indonesia. One of the instruments at this facility will be neutron imaging (NI). This study aims to design an optimal neutron delivery system for NI using the Particle and Heavy Ion Transport code System (PHITS). The cyclotron accelerates 30 uA protons up to 30 MeV and then subsequently bombarded these protons onto a beryllium target. The neutrons then are moderated by high-density polyethylene. The beamtube is 1-meter long and is constructed based on the refined version of the existing neutron imaging facility at BRIN^s G.A. Siwabessy Research Reactor. Through simulations, we investigated the influence of different reflector and filter materials on the neutron flux. At the start of the beam tube, the overall neutron flux is 3.74 x 10^6 neutrons/(cm^2.s), of which 42.7% are thermal. Our findings suggest that an aluminium reflector coupled with a 5-cm thick graphite filter is the most effective configuration for minimizing neutron flux loss. This configuration delivers 1.5 x 10^4 thermal neutrons/(cm^2.s), which is 69.02% of overall neutrons exiting the beam tube. This number is comparable to other thermal neutron radiography facilities worldwide. Keywords: Neutron Radiography, Cyclotron, Beamline, PHITS Share Link | Plain Format | Corresponding Author (Dian Adi Prastowo)

75	Theoretical and Computational Nuclear Physics	ABS-64
NEUTRONICS ANALYSIS OF THE FUJI-U3 REACTOR WITH VARIOUS MIXTURES OF THORIUM, URANIUM AND PLUTONIUM FUEL Faris Rasyad Mutashim (a,b), Cici Wulandari (a,b), Sidik Permana (a,b*), Dwi Irwanto (a,b) a. Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi, Bandung, Indonesia b. Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi, Bandung, Indonesia *psidik[at]itb.ac.id Abstract Indonesia experiences limited fossil energy sources which creates a scarcity of electrical energy. The use of new renewable energy has been considered to overcome this problem. Nuclear power plants are one of the appropriate solutions for utilizing new, renewable energy using nuclear power. National Energy Council and the Ministry of Energy and Mineral Resources are targeting that by 2032 nuclear power plants will be commercialized in Indonesia. Of the various types of nuclear reactors, the Molten Salt Reactor is the most suitable type to be developed in Indonesia, which is rich in thorium and uranium. In the development of MSR, the Fuji-U3 reactor scheme was created which has 3 types of reactor cores with varying amounts of different fuel. In this research, MSR Fuji-U3 will be simulated using the SRAC program utilizing the JENDL-4.0 nuclear data library. The results obtained will produce various types of reactor neutronic data. This data is then analyzed in such a way as to obtain a final conclusion regarding the Fuji-U3 reactor. The Keff of the reactor has a value that decreases during operation and greatly influences the amount of fissile substances in the fuel. The CR value is the opposite of Keff. The highest burn-up was obtained by a mixture of thorium-uranium 233 fuel. The smallest peak power ratio was obtained by a mixture of natural uranium fuel and weapons grade plutonium. Keywords: nuclear power plant, FUJI-U3, neutronic analysist, K-eff Share Link | Plain Format | Corresponding Author (Faris Rasyad Mutashim)

76	Theoretical and Computational Nuclear Physics	ABS-65
Study of The Effect of Gap Width on APWR Passive Containment Cooling System Santiko Tri Sulaksono (a,b*), Zaki Suud (a), Efrizon Umar (b) a. Department of Doctoral Nuclear Engineering, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Ganesha 10 Street, Bandung, 40132, Indonesia *sant012[at]brin.go.id b. Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Tamansari 71 Street, Bandung, 40132, Indonesia Abstract Advance Pressurized Water Reactor is equipped with a safety feature called Passive Containment Cooling System (PCCS). This system uses natural air circulation that cools the outer surface of the container to remove decay heat released inside the containment vessel. The air in the gap between the outer wall of the containment and the baffle absorbs heat and moves up towards the chimney. Cold air will enter through the inlet so that a natural circulation cycle occurs. To obtain the optimal gap width, it is necessary to analyze the heat transfer on the surface of the containment. In this study, a CFD analysis has been carried out with a 1:40 model dimension of the original containment with a gap width of 1 cm, 2 cm and 3 cm. Based on the results of the analysis, the optimum steady-state temperature of the outer surface of the containment occurs at a gap width of 2 cm. Keywords: passive containment cooling system, natural circulation, CFD Share Link | Plain Format | Corresponding Author (Santiko Tri Sulaksono)

77	Theoretical and Computational Nuclear Physics	ABS-81
Distribution of Volumetric Gamma Heating in the Graphite Moderator of TMSR-500 using OpenMC Codes Nurul K Fitriyani1* , Rida S.N. Mahmudah1, Azizul Khakim2 1 Physics department, faculty of mathematics and natural science (FMIPA) Yogyakarta State University, Sleman 55581, Indonesia 2Research Centre for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), Building No. 80 BJ Habibie Science and Technology Area, Setu, South Tangerang 15314, Indonesia. *E-mail : aziz003[at]brin.go.id Abstract Ensuring the safety and efficiency of the TMSR-500 reactor, a Generation IV Molten Salt Reactor (MSR), is crucial for its planned development in Indonesia. This study evaluates gamma heating exposed to the graphite moderator, contributing to the main heat source in the graphite. The TMSR-500, an advancement of the Molten Salt Reactor Experiment (MSRE), utilizes molten salt fuel, is graphite-moderated, and operates at high temperatures. The effective moderation by graphite reduces neutron energy from fast to thermal levels, ensuring optimal fission sustainability and impacting gamma heating. OpenMC simulations were employed using the Python programming language and Monte Carlo methods to assess gamma heating with ENDF/B-VII.1 nuclear data and a particle count of one million. The simulations were conducted on The National Research and Innovation Agency^s computer facilities. The maximum gamma flux was observed at energies greater than 1 eV, with a measured value of 1.8096 x 10^(14) photons/cm^2 sec. The gamma heating in graphite was found to be 3.4917 W and the corresponding volumetric gamma heating was 9.1 x 10^(-3) W/cm^3. The results indicate that gamma heating in the TMSR-500 has a percentage of 2.0138 of the reactor^s total power, within the level of common knowledge. Keywords: Gamma heating, TMSR, OpenMC, Graphite Share Link | Plain Format | Corresponding Author (Azizul Khakim)

78	Theoretical and Computational Nuclear Physics	ABS-82
Neutronics Designs of the Indonesian Experimental Power Reactor/RDE (Comprehensive Review and Future Challenges) tract Liem Peng Hong 1,2 * and Donny Hartanto 3 1 Nippon Advanced Information Service (NAIS Co., Inc.), 416 Muramatsu, Tokaimura, Naka-gun, Ibaraki, Japan 2 Tokyo City University, Cooperative Major in Nuclear Energy, Graduate School of Engineering, 1-28-1 Tamazutsumi, Setagaya, Tokyo, Japan 3 Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, Tennessee, USA Abstract In this paper, several design results of the 10 MWth Reaktor Daya Eksperimental (RDE) (Experimental Power Reactor), so far conducted, are reviewed and compared from the neutronics, reactor types, refuelling schemes, and fuel cycle points of view. The review covers the multipass and once-through then-out (OTTO) pebble-bed cores, as well as block/prismatic type cores with several fuel shuffling options. As for the fuel cycle, the uranium and thorium fuels are considered. The fuel burnup performance and power distribution are evaluated and compared among other important design parameters. Reactor physics codes, nuclear data libraries, and calculation models & procedures used for the design and analysis are reviewed and challenges for future improvements are discussed. Keywords: RDE, OTTO Share Link | Plain Format | Corresponding Author (Liem Peng Hong)

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