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1	Biophysics and Medical Nuclear Physics	ABS-40
Measurement and Identification of Human Brain Wave Signals while Listening to Al-Quran Recitation using Electroencephalogram (EEG) and Fourier Analysis Elin Yusibani, Ayu Mahara, Fahyumi Senye, Fashbir, Adi Rahwanto, Evi Yufita Physics Department Faculty of Mathematics and Natural Sciences Universitas Syiah Kuala Banda Aceh Indonesia, 23111 Abstract Brain signals are not entirely random- they exhibit periodic properties in their components. Various studies have been conducted to determine the clinical, physiological, and psychological effects of these signals. However, the underlying mechanisms behind these signals have not been widely explored. The recorded Electroencephalogram (EEG) signal patterns are associated with mental activity, levels of consciousness, physiological conditions, and brain pathology. Generally, the human brain produces five basic signals: delta, theta, alpha, beta, and gamma. These signals function optimally in different situations, and disruptions in their production can lead to various issues. An increase in the amplitude of the theta and alpha waves can indicate a person^s comfort level. Individuals with higher alpha levels tend to be less easily agitated, which indirectly supports the immune system. Fourier Analysis was applied to identify EEG signal patterns while listening to Quran recitation. Electrodes were placed using the 10 - 20 system, and brain wave signals were recorded for approximately 3 minutes. The signals were then analyzed to assess the impact of listening to Quran recitation. Keywords: EEG, Fourier analysis, brain wave signal, al quran recital Share Link | Plain Format | Corresponding Author (Elin Yusibani)

2	Biophysics and Medical Nuclear Physics	ABS-51
Design and testing of the oblique and V-shaped Cu-ITO microelectrode arrays to generate dielectrophoretic force on red blood cells Edwar Iswardy, Khazanna, Elin Yusibani, Mursal, Kurnia Lahna, Sri Fitriyani 1 Department of Physics Faculty of Mathematics and Natural Sciences, Banda Aceh, Indonesia, 23111 2 Master Program in Physics, Department of Dental Materials, Faculty of Dentistry, Banda Aceh, Indonesia, 23111 3 Department of Dental Materials, Faculty of Dentistry, Banda Aceh, Indonesia, 23111 Abstract Dielectrophoresis (DEP) is a phenomenon in which a force is applied to a particle induced by a gradient of an electric field. The dielectrophoretic technique is popular for manipulating bioparticles, because it requires only a small sample, is label-free, rapid, and inexpensive. Manipulation of biosample can be in the form of monitoring, separation, sorting, capturing, etc., so that the DEP method can be applied as a biosample analysis tool. However, research on the application of the DEP method is still developing on various electrode arrays and bioparticles. In this work, a lab-on-chip device with an oblique and V-shaped 3D microelectrode array has been developed to manipulate red blood cells using the dielectrophoresis (DEP) method. The microelectrodes were fabricated with copper and indium tin oxide films on a glass substrate, while the microchannel was constructed using double-sided tape insulators. Red blood cell samples were prepared in deionized water and EDTA medium with an electrical conductivity of 1.5 S/m. The test of dielectrophoretic force characteristics on red blood cells was carried out by applying an AC signal to the microelectrode, and the phenomenon was observed using a microscope with a CCD camera. The results showed that negative DEP forces were observed at frequencies of 5-7 MHz, 3.5-5 MHz, and 2-4 MHz in the oblique electrode spacing area and in the middle area of the V-shaped electrode curve. While positive DEP forces were observed at frequencies of 8-14 MHz, 6-13 MHz, and 5-11 MHz in the edge area of the oblique electrode and in the inner tip area of the V-shaped electrode curve, respectively at voltages of 5 Vpp, 10 Vpp, and 15 Vpp. The results of this work show the promising potential of lab-on-chip devices with oblique and V-shaped microelectrode arrangements to manipulate bioparticles so that biosamples can be further analyzed. Keywords: Oblique and v-shaped microelectrodes, Dielectrophoresis, Red blood cells, Non-uniform electric field Share Link | Plain Format | Corresponding Author (Edwar Iswardy)

3	Biophysics and Medical Nuclear Physics	ABS-52
Comparison of Dose Distribution Between Proton and Photon Radiotherapy on Organ at Risk (OAR) Based On The Number of Beams Jesse Owen, Rena Widita Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Jalan Ganesha 10, Bandung 40132, Indonesia Abstract Proton and photon radiotherapy have become essential treatment methods in treating cancer. The number of radiation angles in therapy planning significantly influences the success of the therapy. Therefore, the aim of this study is to investigate the dose distribution of proton and photon radiotherapy on Organs at Risk (OAR) based on the number of radiation beam angles. This study involves dosimetry planning using matRad to create a Treatment Planning System (TPS) to observe and compare the dose distribution on the tumor target and OAR. By varying the number of radiation angles, this study will evaluate the effectiveness of both radiotherapy methods in delivering an optimal dose to the target while minimizing the dose to OAR. The results of the study are expected to provide further insights into the impact of the number of angles on dose distribution, providing a basis for selecting a more accurate radiotherapy method and improving the quality of cancer care. Keywords: Dose Distribution, Photon, Proton, Radiation Angle Share Link | Plain Format | Corresponding Author (Jesse Owen)

4	Biophysics and Medical Nuclear Physics	ABS-53
Addition of Repopulation on Tumor Control Probability (TCP) Model in Prostate Cancer Salma Aqilah Zahroh (a*), Rena Widita (a) a) Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jalan Ganesha 10, Bandung, 40132, Indonesia salmaqilahz[at]gmail.com Abstract When designing fractionation schedules for radiotherapy, one critical factor to account for is overall treatment time. Accelerated repopulation of tumor cells can significantly impact biological effectiveness of treatment. To address this, incorporating a repopulation factor into tumor control probability (TCP) model becomes essential. In this study, we analyzed data from 9,690 patients treated with external beam radiotherapy. Patients were categorized into three risk groups: 28% low risk, 53% intermediate risk, and 19% high risk. TCP model was fitted to clinical outcomes, specifically 5-year biochemical relapse-free survival (5y-bRFS). We employed maximum likelihood estimation (MLE) with Nelder-Mead simplex algorithm to maximize likelihood function produced by TCP model. Our results show that dose required to counteract daily repopulation is 0.54 Gy/day, with a kick-off time of 23 days. Our analysis reveals that there was no significant difference in kick-off time and repopulation rate across different risk groups. Incorporating repopulation factor into the TCP model yielded a good fit to data, as indicated by Akaike Information Criterion (AIC). Strategies to mitigate tumor cell repopulation during radiotherapy include employing accelerated fractionation, which reduces overall treatment time and minimizes opportunities for accelerated tumor repopulation. Keywords: Maximum likelihood estimation, Nelder-Mead simplex algorithm, Prostate cancer, Repopulation, Tumor control probability model Share Link | Plain Format | Corresponding Author (Salma Aqilah Zahroh)

5	Energy Management, Regulation, and Policy	ABS-22
Periodic Safety Review of RSG-GAS: Strategy and Evaluation to Manage Research Reactor Aging Endiah Puji Hastuti(a*,b), Topan Setiadipura(b), Dwi Irwanto(a), Abdul Waris(a) a) Bandung Institute of Technology, Master of Computational Science, Master of Nuclear Science and Engineering, and Doctor of Nuclear Engineering Division, Faculty of Mathematics and Natural Sciences Dept., Bandung-Indonesia-401323 * endi001[at]brin.go.id b) 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, Indonesia-15314 Abstract IAEA^s research reactor data base indicates that ~50% of 225 units of research reactor in the world have been in operation for more than 50 years and about 20% of them are more than 60 years, including research reactors in Indonesia. The latest research reactor in Indonesia is multi-purpose reactor RSG-GAS 30 MW that has been operating for 37 years (Its first criticality was in 1987). In accordance with the regulation of BAPETEN, the regulatory body, to ensure sustainable safe reactor operation, a periodic evaluation on reactor safety should be performed. This periodic evaluation is one technique to evaluate nuclear reactor safety whether the reactor design, the status of the structure system and component (SSC), equipment qualification, component aging, safety performance and operation feedback, safety management and nuclear preparedness program, and radiological impact on environment still fulfil the existing safety criteria. This paper is aimed to describe the strategy and evaluation on the periodic safety review (PSR) of RSG-GAS using the reactor operation report in year 2005 - 2015, which is 10 years as regulated. The results of the reactor operation evaluation based on these data show that RSG-GAS still meets the reactor operation safety criteria and the reactor is allowed to operate until 2030. Henceforth, evaluation on the reactor operation safety for 2016 - 2025 period is necessary. In addition, safety analysis report, aging management program, and the latest operation report are also required. Keywords: RSG-GAS, research reactor, PSR, aging management Share Link | Plain Format | Corresponding Author (Endiah Puji Hastuti)

6	Energy Management, Regulation, and Policy	ABS-26
Integrating New and Renewable Energy in Indonesia^s Post-Pandemic Energy Landscape: The Role of Nuclear Power Sunarko Polytechnic Institute of Nuclear Technology Babarsari, Yogyakarta Abstract The past COVID-19 pandemic has profoundly impacted the global economy and altered electricity use patterns, including in Indonesia. Government policies enforcing social restrictions and remote work led to increased household energy consumption while reducing commercial energy use. The pandemic caused a sharp economic downturn, with growth plummeting to -5.32%. Although the economy is recovering, some changes in energy use patterns persist. Utilities and energy companies have adjusted their infrastructure and investment plans in response to these shifts. The pandemic underscored the need for resilient and sustainable energy systems. The recent initiative to phase out coal in the energy sector paves the way for New and Renewable Energy (NRE) to provide more environmentally friendly electricity, helping to reduce greenhouse gas (GHG) emissions. This paper examines how NRE, particularly nuclear energy, can be integrated into the national energy mix, considering the recovering demand and the shift to new electricity consumption patterns post-pandemic. Keywords: Sustainable energy- Coal phase-out- NRE- Nuclear energy Share Link | Plain Format | Corresponding Author (Sunarko Sunarko)

7	Energy Management, Regulation, and Policy	ABS-34
Preparing Decommissioning Regulatory Infrastructure for Embarking Countries with Ageing Nuclear Facilities: A Case Study of Indonesia Reno Alamsyah, Sidik Permana dan AbdulWaris 1. BAPETEN. Nuclear Energy Regulatory Agency 2. and 3. Bandung Institute of Technology Abstract Indonesia is considered as among nuclear embarking country. Besides that, in Indonesia currently a government research institute operates three research reactors, and six fuel cycle facilities in addition to one facility operated by a state-own company. All of these installations are more than 36 years of operating ages. Decommissioning issue for the two areas, i.e., embarking and ageing facility, is a very important challenge for the regulator to be prepared. What national policy available and what to be prepared? Has pertinent international obligation fulfilled? What regulation and law existed and what to be prepared? How regulator build their capacity and what HRD Plan should be prepared? This study answered all of the above typical questions with a comparative, gap analysis and descriptive method. Data and information from reliable publications, such as from the IAEA and NEA or from scientific papers of recognized publication, were used in this study. Interview to senior re-searchers and managers of the engaged organizations was also carried out to identify potential impediments and solution with sufficient risk assessment discussion. The description of the existing installations, current status of legislation, international instruments, and regulation and guides were provided together with the regulatory frame-work and updates of their capacity, training and education arrangements, and human re-sources. Regulatory approach chosen by the regulator and the application graded approach were evaluated, including with the availability of the TSO and its competency on decommissioning. This paper reviewed as well the regulator^s integrated management system, regulatory policies, infrastructure for consultation and communication to interested parties in developing regulation and decision making or licensing, and the infrastructure for oversight and enforcement. Finally, the paper appraises the core challenges in developing a better capacity building and human resources plan. Acceptance criteria in evaluating of these regulatory topics were developed from national common legislation and law, international agreements, international standards, especially from the IAEA, and good international practices. This study concluded that Indonesia has been the parties of the required international agreement related to decommissioning. Indonesia could solve its challenges at the same time with the establishment of national policy in waste management, decommissioning and environmental remediation, in addition to the stablished nuclear safety policy. Furthermore, there are a plenty room for improving the regulation, regulator capacity and it^s HRD plan in some specific areas. This study also suggest that the regulator body may use many international and bilateral cooperation to help improve their regulatory infrastructure. Keywords: Decommissioning, Regulation, Embarking Country, Ageing Facility Share Link | Plain Format | Corresponding Author (Reno Alamsyah)

8	Energy Management, Regulation, and Policy	ABS-66
Comparison of the Use of Fissile Material U-233 and Plutonium in Thorium Nitride (ThN) Fuel in Small Long-Life Modular PWR Cores at 300 MWth Power Level Boni Pahlanop Lapanporo (a,b*), Zaki Su^ud (a), Asril Pramutadi Andi Mustari (a) (a) Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung Jl. Ganesa 10, Bandung 40132, Indonesia (b) Department of Physics, Faculty of Mathematics and Natural Sciences, Universitas Tanjungpura, Jl. Prof. Dr. H. Hadari Nawawi, Pontianak 78124, Indonesia * boni8poro[at]physics.untan.ac.id Abstract In this study, neutronic calculations were conducted on thorium nitride (ThN)-based fuel with different fissile materials (U-233, reactor-grade plutonium (RG-Pu), weapon-grade plutonium (WG-Pu), and a mixture of RG-Pu and WG-Pu) in a small long-life modular PWR core with 300 MWth power. We calculated using the Standard Reactor Analysis Code (SRAC), employing the PIJ module for fuel cells and the CITATION module for the reactor core. The 3D XYZ reactor core model consisted of 89 fuel assemblies, each with 17x17 pins (264 fuel pins and 25 guide tubes). We composed the fuel pins of Th-232, fissile material, and Pa-231 as a burnable poison. We divided the core into three regions (F1, F2, F3) with increasing fissile material percentages from the center outward, differing by 1-2%. Results showed that U-233 and the mixture of RG-Pu and WG-Pu performed best, maintaining excess reactivity below 1.00% dk/k for 20 years. Power density distribution at a 65% fuel volume fraction was similar for all fissile materials, with plutonium being more uniform than U-233. All fissile materials had low and safe PPF values for PWRs. U-233 had the highest Doppler coefficient, RG-Pu the lowest, and WG-Pu increased at high volume fractions. The mixture of RG-Pu and WG-Pu had a moderate Doppler coefficient. Burnup levels were slightly lower for RG-Pu at 60% fuel volume and lower for the mixture of RG-Pu and WG-Pu at 65% fuel volume fraction. Keywords: Long-life PWR, Neutronics, SRAC, Thorium, Uranium, Plutonium, BP Share Link | Plain Format | Corresponding Author (Boni Pahlanop Lapanporo)

9	Energy Management, Regulation, and Policy	ABS-69
Analysis of weather field data at the TRIGA 2000 reactor site to enhance public safety measures Haryo Seno (a*), Muhamad Hilmi Haidar (b), Nabila Putri Rihan (b), Djoko Prakoso Dwi Atmojo (a), Jibran Alfandi Rachman (c), Prasetyo Basuki (d) a) Nuclear Energy Research Organization, National Research and Innovation Agency (BRIN) *haryo.seno[at]brin.go.id b) Telkom University c) Gadjah Mada University d) National Research and Innovation Agency (BRIN) Abstract Planning for a radiological safety and emergency preparedness program in the nuclear reactor facility requires a comprehensive understanding of environmental impact assessment. The radioactive plume dispersion emitted from nuclear reactors is becoming the most significant source for environmental risk in the atmospheric pathway. Many factors affect the radioactive plume dispersion into the environment. However, weather data, in particular wind speed and direction, significantly influence the dispersion of radioactive plumes into the environment. Since it is difficult to determine the wind speed and direction immediately after an emergency situation, a thorough grasp of previous wind data with a machine learning approach would be helpful to forecast subsequent and forthcoming wind data. This could predict the movement of the radioactive plume as precisely as possible, thus being able to facilitate proper countermeasure processes and protect the nearby population. This study examines a dataset of weather sensors from the last three years of the TRIGA 2000 nuclear reactor site. The characteristics of the wind field blowing from the reactor site to the surrounding heavily inhabited places are particularly investigated. In conclusion, suggestions for developing countermeasure strategies are offered. Keywords: radiological safety, wind data, machine learning, emergency countermeasure Share Link | Plain Format | Corresponding Author (Haryo Seno)

10	Energy Management, Regulation, and Policy	ABS-79
Study of the Impact of Nuclear Power Plants (NPP) and NuScale Reactors on Economic and Environmental Aspects Kevin Wijaya(a) and Sidik Permana(a,b) a) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia. b) Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia. Abstract Nuclear Power Plants (NPPs) are one type of power plant that can produce large amounts of electricity and do not produce carbon. Although when operating, NPPs can compete with power plants from other energy sources, they require quite large capital costs. In addition, there have been cases of nuclear accidents that have an impact on the environment. Therefore, this article discusses the economic and environmental analysis of the existence of NPPs. In addition, the NuScale Reactor is also analyzed from the design and development, seawater desalination, and the impact of NuScale on the economy and the environment. From the results of the study, it was found that NPPs require large capital costs but when operating they tend to be smaller, NPPs can also open up jobs from various sectors, have a significant impact on environmental objects such as water quality, soil, air, ecosystems, and habitats. Also obtained are design features that are only owned by NuScale and can answer the Fukushima disaster problem. Then, information was obtained that there are three desalination and reverse osmosis (RO) processes that are more efficient in terms of economy and produce clean water, while if water quality is desired, the multi-effect distillation (MED) process can be used. Keywords: Desalination, economy, environment, nuclear power plant, NuScale Share Link | Plain Format | Corresponding Author (Kevin Wijaya)

11	Innovative Nuclear Energy Systems	ABS-7
Preliminary Numerical Investigation of Cooling Fins Feature in the Frozen Salt Melting Performance Anni Nuril Hidayati (a*), Asril Pramutadi Andi Mustari (b), Yulia Mifftah Huljanah (c), Nina Widiawati(a) a) Research Center for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency, Building No. 80, BJ Habibie Integrated Science Area, South Tangerang 15310, Indonesia *anni030[at]brin.go.id b) Nuclear Physics and Biophysics Research Division, Physics Department Faculty of Mathematics and Natural Science, Institut Teknologi Bandung, Jalan Ganesha 10, Bandung, 40132, Indonesia c) Bachelor Program of Physics Department, Faculty of Mathematics and Natural Science, Institut Teknologi Bandung, Jalan Ganesha 10, Bandung, 40132, Indonesia Abstract The opening time of the freeze plug safety feature in a molten salt reactor is a crucial aspect that must be carefully considered. This relates to the reactor core^s drainage during a blackout or abnormal temperature rise due to a LOCA, Loss of Coolant Accident, while in operation. This study analyzes the impact of varying the geometry of the distance between the freeze plug and cooling fins on the opening time. Calculations were performed using a particle-based Lagrangian numerical simulation called MPS, Moving Particle Semi-implicit. The calculations were conducted using two-dimensional geometry, with the initial design being an in-line arrangement. The approach involves forced convective heat transfer between the liquid fuel, the freeze plug, and the cooling fins, while conductive calculations were performed among the freeze plug particles during melting. This study is essential to determine the variations that yield optimal values for supporting the safety system of molten salt reactors. Keywords: Forced Convection, In-line arrangement, Molten Salt, Particle, Plug Share Link | Plain Format | Corresponding Author (Anni Nuril Hidayati)

12	Innovative Nuclear Energy Systems	ABS-8
Comparative Analysis of Molten Salt Reactor (MSR) Using FLiBe, FLiNaK, and FNaBe with Power Output of 100 MWe Cici Wulandari(a,b*), Sidik Permana(a,b), Dwi Irwanto(a,b), Abdul Waris(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 *cici.wulandari01[at]itb.ac.id Abstract This study presents the neutronic design and analysis of a Molten Salt Reactor (MSR) with an electrical output of 100 MWe, specifically tailored for deployment in a low populated regions such as West Kalimantan or small islands where electricity grid connectivity is unreliable. The selection of a 100 MWe power level aims to meet the energy needs of these areas while minimizing the infrastructural and environmental impacts. The comparative analysis was conducted on various molten salt types, including FLiBe (LiF-BeF2), FLiNaK (LiF-NaF-KF), and FNaBe (NaF-BeF2), to determine the most efficient salt for reactor operations. Neutronic calculations were carried out using the SRAC (Standard Reactor Analysis Code) program, employing the PIJ (cell calculation module) and CITATION (core calculation module) to evaluate reactor performance. The results demonstrate that FLiBe molten salt offers significant advantages in terms of neutron economy, contributing to higher reactor efficiency and improved fuel utilization. These findings suggest that FLiBe is a superior choice for MSRs in the regions, providing a reliable and sustainable energy. Keywords: FLiBe, FLiNaK, FNaBe, MSR, Neutronic Share Link | Plain Format | Corresponding Author (Cici Wulandari)

13	Innovative Nuclear Energy Systems	ABS-10
DESIGN AND ANALYSIS OF VODO-VODYANOI ENYERGETICHESKIY REAKTOR-1000 (VVER-1000) WITH MONTE CARLO N-PARTICLE (MCNP) Kevin Wijaya(a*), Dwi Irwanto(a,b), and Sidik Permana(a,b) a) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia. *kevinwijaya3001[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 Currently, electrical energy is still one of the sources of human needs in everyday life. With the depletion of fossil energy, nuclear power plants can become one of the suppliers of electricity that can be used. In Indonesia, there is still no nuclear reactor for nuclear power plants, only for research purposes. One type of nuclear reactor is the Vodo-Vodyanoi Enyergeticheskiy Reaktor-1000 (VVER-1000).VVER-1000 is a generation III fission reactor with a thermal power of 3000 MWth. In this final project, we will discuss the design and analysis of the VVER-1000 using a program based on Monte Carlo N-Particle (MCNP) which is a license code from the Los Alamos National Laboratory (LANL). The purpose of this study was to determine the fuel assembly model, determine the multiplication factor, and determine the multiplication factor as a function of burnup and burnup rate with variations of uranium enrichment and gadolinium variations using MCNP. Broadly speaking, this research is divided into three stages, namely the literature study stage, the MCNP simulation and modeling stage, and the processing and analysis stage. Based on the experiments carried out, obtained a model of the fuel assembly along with the fuel cell, guide tube cell, and central tube cell. Then, it was obtained information that variations in uranium and gadolinium enrichment affect the value of the multiplication factor which affects the amount of neutron production. Then the results of the multiplication are obtained which changes the value of each factor and the variations used do not affect the burnup rate value. Keywords: Burnup, fuel assembly, MCNP, multiplication factor, VVER-1000 Share Link | Plain Format | Corresponding Author (Kevin Wijaya)

14	Innovative Nuclear Energy Systems	ABS-11
STUDY OF THE EFFECT OF THORIUM AS A FUEL AND PROTACTINIUM-231 AS A BURNABLE POISON ON VODO-VODYANOI ENYERGETICHESKIY REACTOR-1000 FUEL ASSEMBLY USING MONTE CARLO N-PARTICLE (MCNP) SIMULATION Kevin Wijaya(a), Dwi Irwanto(a,b), and Sidik Permana(a,b) a) Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia. *kevinwijaya3001[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 One type of nuclear reactor is the Vodo-Vodyanoi Enyergeticheskiy Reactor-1000 (VVER-1000). VVER-1000 is a generation III fission reactor with a thermal power of 3000 MWth. In this final project, we will discuss the effect of giving thorium as a fuel and protactinium as a combustible poison in the VVER-1000 fuel assembly using a simulation program based on Monte Carlo N-Particle (MCNP) which is a license code from the Los Alamos National Laboratory (LANL). The purpose of this study was to determine the multiplication factor of the fuel assembly with variations in ThO2 and variations in the fractions of gadolinium and protactinium-231, as well as the burnup rate for 5 years. Thorium-232 is predicted to be a substitute for UO2 fuel in reactors because of its abundant material and can produce uranium-233 which is a better fuel than uranium-235. While protactinium-231 has a function other than absorbing neutrons, generating it can further produce fissile material that aids in chain reactions. Simulation of using MCNP to determine the effective multiplication factor with variations of ThO2 fraction for 3 cases of administration of combustible poison, as well as variations of the fraction of combustible poison itself. Results Based on the simulation, the fuel assembly with Pa-231 as burnable poison has a higher value than gadolinium. In addition, the larger the burnable poison fraction, the smaller the multiplication factor value is. In addition, the multiplication factor and burnup rate were also obtained for 5 years. Keywords: Burnable poison, burnup, MCNP, thorium, uranium Share Link | Plain Format | Corresponding Author (Kevin Wijaya)

15	Innovative Nuclear Energy Systems	ABS-17
ANALYSIS OF BURNABLE POISON DOPING ON FUEL MOLTEN MIXTURE FOR MOLTEN SALT REACTOR (MSR) FUJI-12 Hilarius Adiwarna(b), Cici Wulandari(a,b), Syeilendra Pramuditya(a,b), Sidik Permana(a,b*), Syaiful Bakhri (c) a)Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi, Bandung, Indonesia b)Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi, Bandung, Indonesia c)Research Center for Nuclear Material and Radioactive Waste Technology (PRTBNLR), Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), South Tangerang 15314, Indonesia *psidik[at]itb.ac.id Abstract Nuclear Power Plants (NPPs) are a type of power generator that supports Sustainable Development Goals (SDGs) by producing clean, affordable, and climate-friendly energy. This study simulated the MSR FUJI-12 to determine suitable SRAC 2006 parameters with a PIJ module for cell calculation and CITATION module for core calculation. The simulation results showed an initial keff of 1.013046 for a U-233 F4 concentration of 0.22%, with a deviation of 0.0354% from the target of 1.0134. However, the final keff remained below 1, at 0.9455, prompting an increase in the U-233 F4 concentration to 0.28%. One of the challenges identified was the high Power Peaking Factor (PPF) of 3.49, that may damage the inside coating of reactor chamber. Using Protactinium as burnable poison and actinide producer to stabilize the keff, it was concluded that the doping on molten mixture is 0.003% to avoid uncriticality on MSR FUJI-12. However, the impact of the doping is unsignificant to the PPF. It is then proposed using other type of burnable poison, such as Americium (Am) and Gadolinium (Gd). However, as it was known that only Protactinium that produces burnable actinide to fuel the core, the impact is negligeble except for the Gd gas, as its doping is enough to decrease the initial keff and to increase the conversion ratio (CR) in MSR from 0.835 to 0.933. Keywords: Burnable poison, MSR FUJI-12, Protactinium, U-233 F4 Share Link | Plain Format | Corresponding Author (Hilarius Adiwarna)

16	Innovative Nuclear Energy Systems	ABS-27
Analysis of Uranium-Plutonium Based Fuel Performance in Gas-Cooled Fast Reactors (GFR) With Various Core Geometries Using OpenMC Ratna Dewi Syarifah, Ahmad Haris Rasidi, Nurul Kholifah, Fajri Prasetya, Ahmad Muzaki Mabruri, Mohammad Ali Shafii, Dian Fitriyani 1Department of Physics, Universitas Jember 2Department of Nuclear Science and Engineering, Institut Teknologi Bandung 3Department of Physics, Universitas Andalas *Corresponding Author e-mail: mashafii[at]sci.unand.ac.id Abstract Analysis of thorium-based fuel performance in Gas-cooled Fast Reactors (GFR) with various core geometries has been conducted. GFR is a type of Generation-4 nuclear reactor with a fast neutron spectrum and helium gas coolant, enabling it to operate at high temperatures and effectively facilitate the breeding of U-233 from Th-232. This research aims to analyze the performance of thorium-based fuel in GFR core geometries: balanced, tall, and pancake shapes. The performance analysis includes neutronics aspects such as criticality, neutron behavior, and material transmutation resulting from burn-up. The analysis was conducted using the Monte Carlo approach with the OpenMC code, which is an open-source tool for neutron and photon transport analysis. The research findings indicate that the balanced geometry variation yields the best performance, requiring less fuel to achieve criticality over the same period, providing a more uniform neutron flux distribution, and achieving the most optimal breeding ratio. These results highlight the potential of the balanced design as the preferred choice to enhance efficiency and reliability in GFR reactors. Keywords: GFR, Thorium fuel, Variation Core Geometries Share Link | Plain Format | Corresponding Author (Ratna Dewi Syarifah)

17	Innovative Nuclear Energy Systems	ABS-30
ANALYSIS OF POWER PEAKING FACTOR REACTOR CORE TRIGA 2000 BANDUNG USING OPENMC PROGRAM Fajri Prasetya (a*), Nuri Trianti (b), Ratna Dewi Syarifah (c), Zaki Su^ud (a), Efrizon Umar (b) (a) Department of Nuclear Science and Engineering, Institut Teknologi Bandung, Ganesha Street No. 10, Siliwangi, Coblong Subdistrict, Bandung City, West Java, Indonesia, 40132 *prasfajri[at]gmail.com (b) Nuclear Reactor Technology, National Research and Innovation Agency (BRIN), Tamansari Street No. 71, Siliwangi, Coblong Subdistrict, Bandung City, West Java, Indonesia, 40132 (c) Department of Physics, Universitas Jember, Kalimantan Street, No 37, Krajan Timur, Sumbersari Subdistrict, Jember, East Java, Indonesia, 68121 Abstract Currently, the TRIGA reactor has been reshuffling several fuel elements in order to maintain its criticality during operation. The reshuffling includes the filling of heterogeneous fuel elements with an enrichment of 8.5%wt- 12%wt- 20%wt for the isotope U-235. The heterogeneity in the preparation of fuel elements has the opportunity to produce a fairly high PPF (Power Peaking Factor) value. This is the main focus that needs to be discussed in this study. The purpose of the study is to conduct a more comprehensive neutronic analysis of fuel elements by calculating the criticality value (k-eff) and PPF which are then compared with the results of the reference calculation (MCNP). The results of the validation of OpenMC and MCNP k-eff values showed accurate values with an error percentage (%&#8710-k/k) of < 1%, namely 0.1159% (60% of control rod withdrawal) and 0.1988% (100% control rod withdrawal). The results of the PPF calculation for the axial direction (APF) show a similar trend pattern for each variation of power used. In the middle position, the active fuel element has the highest PPF value than other regions. This shows that the fission rate in the central area of the fuel element is very dominant. The power used in this calculation is varied in the range of 223-640 kW because it still meets the safety limits of the current TRIGA 2000 Bandung reactor operation. Keywords: K-eff- OpenMC- PPF Share Link | Plain Format | Corresponding Author (Fajri Prasetya)

18	Innovative Nuclear Energy Systems	ABS-31
Analysis of Criticality and Fissile Material Production in Pebble Fuel of HTR-10 Using OpenMC Code Ahmad Muzaki Mabruri (a*), Nuri Trianti (b), Zaki Suud (a,c), Efrizon Umar (b), Ratna Dewi Syarifah (d) a) Department of Nuclear Science and Engineering. Bandung Institute of Technology, Bandung, Indonesia *amuzaki073[at]gmail.com b) Nuclear Reactor Technology, National Research and Innovation Agency (BRIN), Bandung, Indonesia c) Department of Physics, Bandung Institute of Technology, Bandung, Indonesia d) Department of Physics, University of Jember, Jember, Indonesia Abstract Analysis of Criticality and Fissile Material Production in Pebble Fuel of HTR-10 Using OpenMC Code has been conducted. Pebble fuel is a type of nuclear reactor fuel consisting of TRISO particles randomly dispersed within a kernel and surrounded by several layers, forming a unified pebble. The analysis was performed using Monte Carlo calculations with OpenMC. The pebble fuel model is assumed to operate at a constant power of 0.54 kW and 4.00 kW for 12 months, representing the average operating power of the HTR-10 design and the maximum operating power for a single pebble. Calculation results with various nuclear data libraries indicate that at maximum power, the pebble can only last for 7 months. Meanwhile, at 0.54 kW power, the pebble can maintain criticality until the end of the burn-up period, with the production of fissile plutonium material amounting to 0.8% of the total U-238 present at the beginning of the burn-up. This shows that the pebble fuel design has good proliferation resistance and is able to maintain criticality longer under actual operating conditions Keywords: HTR-10- Pebble fuel- Proliferation- Triso. Share Link | Plain Format | Corresponding Author (Ahmad Muzaki Mabruri)

19	Innovative Nuclear Energy Systems	ABS-32
OPTIMIZATION OF ACTIVE CORE DESIGN ON NEUTRONIC ASPECTS FOR MOLTEN SALT REACTOR (MSR) FUJI-12 Hilarius Adiwarna(b), Cici Wulandari(a,b), Marisa Variastuti(a,b), Syeilendra Pramuditya(a,b), Sidik Permana(a,b*) a)Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi, Bandung, Indonesia b)Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi, Bandung, Indonesia *psidik[at]itb.ac.id Abstract The Sustainable Development Goals (SDGs) are aided by nuclear power plants (NPPs), a form of power generator that generates economical, clean, and climate-friendly electricity. To find appropriate parameters for the SRAC-2006 and JENDL-4.0 modules with PIJ module for cell calculation and CITATION module for core calculation. this study is simulated for MSR FUJI-12 as one of the fourth generation of reactor. This study continued to optimize the core reactor to achieve a Conversion Ratio (CR) approaching 1 and a Power Peaking Factor (PPF) approaching 2. Variations in the number of core regions-two and three regions-with different fuel percentages were explored. By variating the composition of fuel and moderator or Moderator to Fuel Ratio (MFR) for each region, it was found that the MSR FUJI-12 with three regions and fuel percentages of 50%-30%-40% achieved a CR of 0.962 and a PPF of 2.579. However, this variation impacted the life cycle of the reactor from its adjustment on fuel and moderator to only 1400 days.This limited lifespan could be addressed with the concept of online refueling, applicable to MSR technology. Keywords: MFR, MSR FUJI-12, Protactinium, U-233 F4 Share Link | Plain Format | Corresponding Author (Hilarius Adiwarna)

20	Innovative Nuclear Energy Systems	ABS-36
Computational Fluid Dynamics for Irregular Pentagon Natural Circulation Loop Rahmad Dahana Saputra (a), Sidik Permana (a,b*), Duwi Hariyanto (c), Syeilendra Pramuditya (a,b), Syaiful Bakhri (d) (a)Nuclear Physics and Biophysics Research Division, Physics Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia (b)Nuclear Science and Engineering Department, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Jalan Ganesha 10 Bandung 40132, Indonesia (c) Industrial Technology Faculty, Institut Teknologi Sumatera, Jl. Terusan Ryacudu, Lampung, Indonesia (d) Research Center for Nuclear Material and Radioactive Waste Technology (PRTBNLR), Research Organization for Nuclear Energy, National Research and Innovation Agency (BRIN), South Tangerang 15314, Indonesia *psidik[at]itb.ac.id Abstract Electricity is an important needs for everyday life. Nuclear Reactor is a power plant that have low CO2 emission and the energy is available at any time. Generation IV nuclear reactor have a passive safety system that allow it to run without external power source or any human intervention. One of the design of generation IV nuclear reactor is Molten Salt Reactor (MSR) that have natural circulation as its passive safety system. A study of irregular pentagon natural circulation loop using water as its coolant has been done previously. In this study the natural circulation loop is simulated using ANSYS Fluent. A mesh independece study is done for obtaining mesh with accurate result but at effective computational cost. After that some flow parameters is obtained, namely cold leg temperature, hot leg temperature, and the flow Reynolds number with a value of 34.29 C, 39.93 C, and 86.8-132.6 respectively . Those parameters is compared to a reference study and an error less than 3.5% are obtained. Those result mean that this simulation model can be used for further study. Keywords: Natural Circulation, Ansys Fluent, Pentagon Loop Share Link | Plain Format | Corresponding Author (Rahmad Dahana Saputra)

21	Innovative Nuclear Energy Systems	ABS-37
Study of Prospecting NPP Cogeneration Potential as District Heating in Indonesia Muksin Aji Setiawan(a,b*)- Prof. Sidik Permana (a)- Dr. Topan Setiadipura (b) (a) Nuclear Science and Engineering Faculty of Mathematics and Natural Sciences Bandung Institute of Technology Basic Science Center A Building Jalan Ganesa 10, Bandung, 40132, Indonesia- (b) Center for Nuclear Reactor Technology Research Organization of Nuclear Energy National Research and Innovation Agency Jl. Kawasan PUSPIPTEK Serpong building 80 Tangerang Selatan, Banten, 15310, Indonesia Abstract Metropolitan City, such Jakarta as always linked to large population and huge energy consumption challenges. During working hours, up to 4 million people move around this city. These phenomena make energy demand pattern quiet slightly different compare to the satellite. The demand for energy spikes, especially in the summer months when air conditioning use surges, necessitating a reliable electricity supply. Currently reliant on coal as its primary energy source, Jakarta faces issues when coal availability is disrupted or economically unfeasible. This dependence becomes problematic when coal supply is disrupted or when coal is no longer economically viable. The use of coal also has negative long-term environmental impacts. To support the government^s goal of achieving net-zero carbon emissions, there is a need to increase the use of renewable energy sources. Nuclear energy emerges as a potential option for adoption in Indonesia due to its substantial production capacity. Utilizing Nuclear Power Plants (NPPs) cogeneration aspect for district heating and cooling can reduce dependence on fossil fuels, particularly for cooling needs. This will lead to a reduction in air pollution resulting from fossil fuel combustion, resulting in improved air quality in Jakarta. This study evaluates the potential utilization of NPPs in the context of large cities in Indonesia, taking into consideration factors such as geography, population, economy, and environmental impact. Consequently, Jakarta is expected to move toward cleaner and more sustainable air quality. Keywords: Energy Consumption- Energy Demand- Cogeneration- District Heating Share Link | Plain Format | Corresponding Author (Muksin Aji Setiawan)

22	Innovative Nuclear Energy Systems	ABS-42
INFLUENCE OF PIPE DIAMETER IN ADVANCED REACTORS NATURAL CIRCULATION SYSTEM Amna Yasya Mubarok, (a*), Sidik Permana (a), Dwi Irwanto (a) a) Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology Jalan Ganesha 10, Bandung 40132, Indonesia *amnayasya[at]gmail.com Abstract 4th gen reactors are focused on developing 4 aspects- highly economical, enhanced safety, minimal waste, and proliferation resistance. 4th gen reactor using natural circulation on coolant as an enhanced safety system. Natural circulation makes sure coolant is still able to release heat since the coolant is still flowing. This research is done to study the natural circulation using COMSOL Multiphysics and experiment. Model used in this research is a reactor coolant which is 100 cm in height and 50 cm in width. In this research the diameter of the pipe varied at 1/2, 1/4, and 1 inch. Simulation and experiment able to perform natural circulation. The result shows changes in heat dissipation on each configuration. Bigger pipes have more water volume, meanwhile the heater and cooler give and take the same amount of heat, the result is bigger pipes have higher difference on water lowest and highest temperature. Keywords: Natural circulation, Pipe diameter, Reactors Share Link | Plain Format | Corresponding Author (Amna Yasya Mubarok)

23	Innovative Nuclear Energy Systems	ABS-44
Akademik Lomonosov FPU: operating experience Ivan Renev (a*), Alexander Renev (a), Alexander Tsibulya (a) Rosatom Energy Projects JSC, Leninskaya Sloboda 26 bld. 5, Moscow, Russian Federation Abstract Over four years of operation, the world^s only floating nuclear power plant Akademik Lomonosov generated more than 860 million kWh of electricity. Moreover, every year the share of carbon-neutral electricity from floating nuclear power plants in the overall energy balance of the region is growing steadily. For the entire 2023, the operation of the floating nuclear power plant made it possible to generate 28.5% of all electricity produced in Chukotka. In the future, after 2025, when the Bilibino NPP is shut down and the restoration of the floating nuclear power plant is completed, the station load will reach peak values and approach the maximum power of 70 MW, which will have a positive impact on reducing power tariffs. The purpose of the report is to familiarize students with the experience of operating the world^s only floating nuclear power plant Akademik Lomonosov. Keywords: Akademik Lomonosov, SMR, FNPP, FPU Share Link | Plain Format | Corresponding Author (Ivan Renev)

24	Innovative Nuclear Energy Systems	ABS-45
Civil marine reactor evolution. From OK-150 to RITM-200 Ivan Renev (a*), Alexander Renev (a), Alexander Tsibulya (a) Rosatom Energy Projects JSC, Leninskaya Sloboda 26 bld. 5, Moscow, Russian Federation Abstract The RITM-200 reactor unit (RP) with a capacity of 175 MW is a fundamentally new step in the development of the icebreaker fleet. The reactor plant has a unique energy-efficient integral layout, which ensures that the main equipment is placed directly inside the steam generating unit housing. This allows to classify the RITM-200 reactor as the 3rd generation of civil ship-class reactor unit. In contrast to the 2nd generation of reactor plants (OK-900 and KLT-40), in the 3rd generation a transition was made from a block layout to an integral one. New engineering solutions made it possible to implement the strict restrictions on weight and size characteristics laid down in the technical specifications. Thanks to the integrated layout, the RITM-200 reactor turned out to be twice lighter, one and a half times more compact and 25 MW more powerful than the KLT-40 reactor units. The RITM-200 nuclear power plant is capable of ensuring more economical operation of the new nuclear icebreaker compared to existing ones with increased reliability and safety. The improvement of the reactor plant proceeded in the following directions: - Reducing the composition of equipment and its weight and size characteristics - Increased maneuverability - Increasing equipment life - Reduction of own energy consumption Keywords: SMR, FPU, RITM, KLT-40 Share Link | Plain Format | Corresponding Author (Ivan Renev)

25	Innovative Nuclear Energy Systems	ABS-47
Enhancing Safety in Molten Salt Reactors: Understanding How Design Choices Affect Freeze Plug Performance Muhammad Ilham Sample Abstract Molten salt reactors (MSRs) represent a promising advancement in nuclear technology, offering potential benefits such as improved safety and efficiency. A critical component of MSRs is the freeze plug, which acts as a safety mechanism installed between the reactor vessel and the drain tank. During an emergency, external cooling stops, and the salt inside the freeze plug melts, allowing the fuel salt to be discharged. The melting time (opening time) and the drain time, which are key factors in ensuring MSR passive safety, depend on the drain tube design and the initial shape of the frozen salt formed in the drain tube. Given this, a simulation of the solidification process was first performed, followed by melting and drain time approximation. Systematic numerical simulations were conducted to explore the effects of wall thickness, inner diameter, and tube inclination on the freeze plug used in MSRs. Furthermore, a novel jacket design for melting acceleration and a conical tube for plug structural integrity were proposed. It was found that in the solidification process, the tube design significantly impacted the equilibrium shape of the frozen salt. Melting simulations showed that a small tube shortened the opening time. However, it took a long time to drain the liquid salt from the reactor core into the drain tank after opening. The results indicated that all aspects of the solidification, melting, and drainage processes should be sufficiently understood to utilize the freeze plug as an effective passive safety system in MSRs. Accordingly, a simplified analytical model was developed for a rough estimation of the opening time that reasonably agreed with the full simulation results. Keywords: Molten salt reactor, Freeze plug, Solidification, Melting, Opening time, Drain time, Inclination, Numerical simulation Share Link | Plain Format | Corresponding Author (Muhammad Ilham)

26	Innovative Nuclear Energy Systems	ABS-55
Neutronic & Dynamic Analysis of TMSR-500 Reactor Core Alessandro Widjati (a,1), Sidik Permana (b,1,2,3), Syeilendra Pramuditya (1,2,3), Cici Wulandari (1,2,3) 1) Department of Physics, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Bandung 40132 Indonesia 2) Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Bandung 40132 Indonesia 3) Nuclear Physics and Biophysics Research Group, Faculty of Mathematics and Natural Science, Bandung Institute of Technology, Bandung 40132 Indonesia a) alswidjati[at]gmail.com b) psidik[at]fi.itb.ac.id Abstract TMSR-500 designed by ThorCon is a generation IV nuclear reactor. This reactor has potential to provide energy that are sustainable, safe, and low in CO2 emissions. Therefore, study about TMSR-500 will have a significant impact on the development of nuclear energy. Neutronic and dynamic analysis will be conducted in this study. For neutronic analysis, influence of reactor design parameters against neutronic aspects will be evaluated. The design parameters evaluated are fuel composition, fuel volume fraction, coolant quantity, and material temperature, while the neutronic aspects are effective multiplication factor (keff). For dynamic analysis, response of thermal power, fuel salt temperature, and graphite temperature to positive reactivity insertion will be evaluated. The method used to obtain the neutronic aspects of the reactor is neutronic calculation by computer simulation using the SRAC2006 code system and the JENDL-4.0 nuclear library. While to obtain the dynamic response, dynamic calculation is performed by solving the reactor point kinetics equation and the reactor heat transfer equation numerically. The study shows that the concentration of fissile material is directly proportional to keff. The fuel volume fraction will be inversely proportional to the keff at the beginning of operation. TMSR-500 requires a minimum U-235 concentration of 1.28% with an optimal fuel volume fraction of 27%. TMSR-500 has a negative coolant nuclide density and temperature reactivity coefficient, while also has a good safety response to positive step reactivity insertion in the 50 pcm to 500 pcm range because the temperature of the materials in those dynamic conditions are still within a safe range. Keywords: coolant, dynamic, fuel composition, fuel volume fraction, material temperature, neutronic, reactivity insertion, TMSR Share Link | Plain Format | Corresponding Author (Alessandro Widjati)

27	Innovative Nuclear Energy Systems	ABS-57
Parameter Justification to Validate Neutronic Calculation Using MCNP6 for CANDLE Reactor Case Maryam Afifah(a)*, Zaki Su^ud(a), Nuri Trianti(b) , Dwi Irwanto(a) a) Department of Nuclear Science and Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia 40132 * 34922002[at]mahasiswa.itb.ac.id b) Research Centre for Nuclear Reactor Technology, Research Organization for Nuclear Energy, National Research and Innovation Agency, PUSPIPTEK, Indonesia 15314 Abstract This paper presents the survey parameters to justify several factors with the minimum parameters that should be applied to validate neutronic calculation for CANDLE reactor case. In particular, we investigate the effect of neutron number history, inactive-active cycle, the burnup time step length and the recommendation of those factors as a way to improve the computing efficiency using MCNP6 code. This study presented the minimum number of neutron history and the total cycle to minimize the oscillation. However, the keff value with larger length step has a difference of 1% compared to smaller length step application. In this case we explore the methodology that used in MCNP6 to calculate burnup to make sure the numerically stable method was applied. Lastly, this study concluded some recommendations that should be applied for complex core geometry to achieve accuracy and efficiency for neutronic calculation using Monte Carlo method especially using MCNP6 for CANDLE reactor case. Keywords: CANDLE reactor, Monte Carlo, Depletion, time step length, MCNP6 Share Link | Plain Format | Corresponding Author (Maryam Afifah)

28	Innovative Nuclear Energy Systems	ABS-58
Effect of the Number of Freeze Valve Channels on Core Discharge Rate and Pressure in Core and Drain Tank Ahmad Muzaki Mabruri (a), Amna Yasya Mubarok (a), Bagja M B Kertasafari (a), Asril Pramutadi Andi Mustari (a), Sidik Permana (a) a) Department of Nuclear Science and Engineering. Bandung Institute of Technology, Bandung, Indonesia *amuzaki073[at]gmail.com Abstract Analysis of the Influence of the Number of Freeze Valve Channels on Core Discharge Rate and Pressure in Core and Drain Tank has been conducted. This research employed a 2D freeze valve design representation using the Moving Particle Semi-implicit Method. This method utilizes the behavior and movement of particles within a predefined geometry, influenced by various physical parameters within the system. The aim of this study is to observe velocity profiles and pressure profiles during the core discharge process with variations in freeze valve geometries V1, V2, and V3. The results indicate that the core discharge times for V1, V2, and V3 are 4.75 s, 6.40 s, and 6.15 s respectively. V3 exhibits a lower average system pressure compared to V1 and V3, suggesting that V3 is more effective in managing pressure during the core discharge process. Keywords: Core Discharge- Drain Tank- Freeze Valve- MPS- MSR Share Link | Plain Format | Corresponding Author (Bagja Muhammad Busyro Kertasafari)

29	Innovative Nuclear Energy Systems	ABS-63
Nuclear Energy: A Strategic Path Towards Indonesia^s Sustainable Energy Future Nining Yuningsih1, Indah Rosidah Maemunah2, Fisca Dian Utami3 1 SMAN 5 Sukabumi, Indonesia 2 National Research and Innovation Agency, Indonesia 3 Politeknik Transportasi Sungai, Danau, dan Penyeberangan Palembang, Palembang, Indonesia Abstract Indonesia^s evolving energy landscape highlights nuclear energy as a viable solution for future energy demands. Nuclear power offers a reliable and sustainable alternative, crucial for balancing increasing urban energy consumption and variable rural needs. This study underscores nuclear energy^s potential role in meeting Indonesia^s future energy requirements, stressing the necessity of strategic energy policy planning to align with these projections. Comparing energy sources like oil, gas, and nuclear power plants reveals nuclear energy^s robustness as an alternative. A 1000 MWth nuclear power plant can fulfill approximately 46-53% of the energy demands in Nusa Tenggara, Maluku, and Papua. Moreover, nuclear power boasts minimal carbon emissions, approximately 0.029 kg/kWh, positioning it favorably in achieving Indonesia^s 2060 energy targets and fostering a sustainable energy future. Keywords: Energy, Emission, Nuclear, Sustainable Share Link | Plain Format | Corresponding Author (Indah Rosidah Maemunah)

30	Innovative Nuclear Energy Systems	ABS-73
Dr. Indarta Kuncoro Aji PT Kakiatna Enjiniring, Molten Salt Lab. Inc. Abstract Abstract is submitted as file Keywords: molten salt, solid-liquid interface, natural circulation Share Link | Plain Format | Corresponding Author (Indarta Kuncoro Aji)

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