Indonesia Conference Directory

Corresponding Author
Nining Yuningsih

Institutions
aPhysics Study Program
bNuclear Physics and biophysics Research Division, Laboratory,Institut Teknologi Bandung,
Jl. Ganesha no. 10 Bandung 40132, Gedung Fisika FMIPA ITB Indonesia


1)ninng.yuningsih[at]students.itb.ac.id
2)dirwanto[at]fi.itb.ac.id

Abstract
Sabu Raijua is one of the small areas in the East Nusa Tenggara (NTT) region with an electrification ratio that has not reached 100%. Additionally, the East Nusa Tenggara (NTT) region also faces a clean water crisis. High-Temperature Gas Reactor (HTGR) is a type of reactor that produces not only electricity but also could be used for cogeneration applications, such as desalination of seawater. In this research, the evaluation of the energy and seawater desalination process using HTGR was analyzed using the Desalination Economic Evaluation Program (DEEP) code. Meanwhile, calculation of reactor design performed by Standard Thermal Reactor Analysis Code (SRAC) code and using Japanese Evaluated Nuclear Data Library (JENDL) 4.0 as nuclear data library. The reactor was designed to produced 150 MWt power while seawater desalination used Multi-Effect Desalination (MED) method. As a result, this reactor design can meet electricity demand in the Sabu Raijua region. Also, seawater desalination yields 110000 cubic meters per day which are meet the needs of clean water.

Keywords
High Temperature Gas Reactor (HTGR), seawater desalination, Multi Effect Desalination (MED)

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/NPxG8Cdr6QwD

Corresponding Author
fauzan ghilman anshari

Institutions
a) Bandung Institute of Technology, Department of Physics, Jl. Ganesa 10, 40132, Bandung, Indonesia
fauzan.ghilman[at]gmail.com
b) Bandung Institute of Technology, Nuclear Physics and Biophysics Research Division, Department of Physics, Jl. Ganesa 10, 40132, Bandung, Indonesia

Abstract
High Temperature Test Reactor (HTTR) is one type of IV generation reactor with high temperature gas-cooled designed with low-enriched uranium. The purpose of this study is to compare various coolants in the reactor to obtain optimal coolant. Coolant is a material used to cool the reactor. In this study compared three types of material, namely Helium, Pb-Bi and CO2. In this study also calculated with three types of reactor fuel, namely UO2, (Th,U-233)O2, and (U,Pu)O2. The parametric survey conducted is to do enrichment variations between 1 - 10%. The parametric survey results that meet the operating targets for UO2 and (Th, U-233)O2 fuels are 5% for Helium coolers, 6% for Pb-Bi coolers and 4% for CO2 coolers. Whereas fuel (U,Pu)O2 only reached the operating target for 9% enrichment on CO2 coolers. It was found that CO2 coolers were the best coolers among the three coolers compared. The reactor cell calculations are performed using the SRAC 2006 program, and utilize JENDL4.0 nuclide data.

Keywords
coolant, enrichment, HTTR, neutronic aspects

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/xYZtnuDyqVW4

Corresponding Author
Melati Ifthacharo

Institutions
(a)Master Program of Physics Department,
Faculty of Mathematics and Natural Science, Bandung Institute of Technology
Jalan Ganesha 10, Bandung 40132, Indonesia
*Melati.ifthacharo[at]student.itb.ac.id

(b)Nuclear Physics and Biophysics Research Division, Physics Department,
Faculty of Mathematics and Natural Science, Bandung Institute of Technology
Jalan Ganesha 10, Bandung 40132, Indonesia

(c)Doctoral Program of Physics Department,
Faculty of Mathematics and Natural Science, Bandung Institute of Technology
Jalan Ganesha 10, Bandung 40132, Indonesia

Abstract
There are many systems in a reactor shutdown function in MSR in addition to inherent self-stabilization. One of those systems is the fuel-salt drain system. The present study focused on the melting and solidification phenomenon that occurs in the freeze valve. An experiment was performed to investigate the erosion behavior of a solid plate by an impinging liquid with respect to time. In addition, a numerical modelling based on MPS method to visualize the heat distribution in the plate will also be carried out. The experiment will be conducted by varying the parameters such as the liquids, temperature, and diameter. Hot water (70 and 90oC), molten paraffin, and cooking oil, will be used while both molded candle and paraffin wax will serve as the target plates. The dimension of the target plate is a cylindrical with 44 mm in thickness and 140 mm in length for both paraffin and candle wax. Heat distribution images will be acquired using FLIR thermal video camera. The data then will be compared to the MPS simulation. Recent results showed that cooking oil outperformed hot water and liquid paraffin by having the shortest penetration time but not necesarrily the fastest erotion rate.

Keywords
Erosion behavior; Heat transfer ; MPS

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/9ZarTymB3cQU

Corresponding Author
Duwi Hariyanto

Institutions
(1) Nuclear Physics and Biophysics Research Division, Physics Department, Institut Teknologi Bandung
(2) Nuclear Science and Engineering Department, Institut Teknologi Bandung
Jl. Ganesha 10 Bandung, 40132, Indonesia
Email : (a) duwi_hariyanto[at]students.itb.ac.id, (b) psidik[at]fi.itb.ac.id

Abstract
The natural circulation loop is one of the design concepts of a cooling system in new advanced reactors that has attracted many researchers to develop it. This study aimed to perceive the effect of horizontal width variation on the thermal behavior of a single-phase natural circulation loop (NCL). NCL apparatus with a vertical heater and a vertical cooler was designed for experimental study. The height of the loop was 1.000 mm while the width of the loop was varied at 500 mm and 1.000 mm. The heater was designed using nichrome wire on the outside of the stainless pipe while the cooler was designed using pipe-in-pipe with water flowing through the annulus. Arduino microcontroller and K-type thermocouple sensors were used in temperature data acquisition. XAMPP software was used in data recording. The direction of the fluid flow was in a clockwise direction. The result in this study was the fluid temperature in a 1.000 mm width lower than the fluid temperature in a 500 mm width. This study is supposed to be one of the references for a single-phase natural circulation loop.

Keywords
natural circulation, single-phase, temperature, microcontroller, K-type thermocouple

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/nPFwA83VRmKL

Corresponding Author
Bilal El Bari

Institutions
a)Nuclear Physics and Biophysics Research Group, Department of Physics,
Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology
Jalan Ganesha 10, Bandung 40132, Indonesia
*bilalelbari[at]students.itb.ac.id

Abstract
Thermal-hydraulics aspect is one of the crucial aspects that must be considered when reactor design and operation analysis were performed, because this aspect involves security, safety, and efficiency factor that must have to be examined. In this study, thermal-hydraulics aspect of the HTR-10MW (which classified into Pebble-Bed Reactor or PBR) were analyzed by reviewing the fast neutron irradiation influence using the modified PEBBLE program (or mPEBBLE), this program using finite-difference numerical method for solving the differential equation of the system. In reviewing fast neutron irradiation, the thermal conductivity value of fuel is induced by fast neutron irradiation dose. The study concludes that fast neutron can influence the fuel thermal conductivity. The value of fuel thermal conductivity can induce the thermal-hydraulics aspect of PBR core which is so important to be understanding more the safety ability of PBR core.

Keywords
Fast Neutron Irradiation, Finite-Difference, Thermal-Hydraulics Aspect, Pebble-Bed Reactor, Wall Effect.

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/tK84FnVUdR9q

Corresponding Author
Andrey Kosasih

Institutions
(a) Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Jl. Ganesa 10, Bandung 40132, Indonesia
(b) Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of
Mathematics and Natural Sciences, Bandung Institute of Technology, Jl. Ganesa 10, Bandung 40132, Indonesia
*awaris[at]fi.itb.ac.id

Abstract
The high temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR) developed by Japan. This reactor can operate with outlet temperature of 950°C and the heat can be used for the cogeneration of electricity and hydrogen production. In this study, the neutronic analysis is carried out for the helium cooled HTTR reactor with (Th, U-233)O2 fuel. The fuel blocks used in the core vary from 3,3% to 7,5% of U-233 content. The calculations are performed by PIJ and CITATION modules on SRAC2006 code system with JENDL-4.0 as the nuclear data library. The cell-burnup calculations are conducted with two models, with and without microscopic cell definition in the fuel compact. The core calculations are conducted with triangular-z and hexagonal-z core geometry. Several neutronic parameters are analyzed, such as effective multiplication factor (k-eff), conversion ratio, changes in atomic density for fissile and fertile materials, neutron spectrum, power density distribution, and power peaking factor. The results show similar neutronic parameters with model 1 and 2. The k-eff is greater in model 1. The neutron spectrum is dominant in the thermal energy. Both core geometries show similar results with greater k-eff in the triangular-z geometry. The maximum power density is located at the fuel block with 5,5% of U-233 content.

Keywords
HTTR, JENDL-4.0, k-eff, neutronic, SRAC2006

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/BYbgkPpJxrnV

Corresponding Author
Yanti Yulianti

Institutions
Department of Physics University of Lampung
Jl. Sumantri Brojonegoro No.1 Bandar Lampung Indonesia

Abstract
This paper will present solving coolant dynamics using Lattice Boltzmann Method (LBM) for single heated channel. Advanced Computational Fluid Dynamics (CFD) using finite difference, finite element or finite volume has been widely used for solving coolant dynamics. However, two-phase coolant flow such as Boiling Water Reactor (BWR) case has accurate model challenge. LBM is an alternative way to solve transient coolant flow. Simulation was done for D1Q2, D2Q3, D2Q4 and D2Q5.

Keywords
Lattice Boltzman; Fluid Dynamics; Heated Channel; Transient Analysis

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/XTJxLMZ4zyqf

Corresponding Author
Imam Ghazali Yasmint

Institutions
1Nuclear Physics Laboratory,
Nuclear Physics and Biophysics, Department of Physics
Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology,
Ganesha No. 10 Bandung, Indonesia, 40132

2Environmental 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
Apart from nuclear reactors, natural radiation sources contribute to the radiation received by humans. One of them is internal radiation due to the process of entry of food into the human body. There are some food ingredients that naturally emit natural radiation such as meat, milk, etc. West Java is one of the biggest cows milk producers in Indonesia. Therefore testing of cow’s milk produced is needed so that it can be seen how much natural radiation enters the human body from milk. In this study several milk samples were taken from a large farm in Lembang. The sample came from three different cowsheds. Then, it measured using an ORTEC gamma spectrometer with HPGe detector. Radionuclides observed in this study were K-40 and Ra-226. Based on measurement results, the natural radioactivity contained in cows milk is below the permissible standard and safe for consumption.

Keywords
Cow’s Milk Samples, Gamma Spectrometer, HPGe Detector, Natural Radiation

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/gcLB6XxyKmUF

Corresponding Author
Cici Wulandari

Institutions
a) Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132, Indonesia
b) Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Jl. Ganesa 10 Bandung 40132, Indonesia

*E-mail: awaris[at]fi.itb.ac.id

Abstract
An advanced nuclear reactor Generation IV, called MSR, has been developed with Thorium utilization for a sustainable energy system. In this paper, the reactor is designed with power operation of 250 MWt/100 MWe in five years without refueling. Fuel salt in the reactor is composed of a eutectic FLiBe, Thorium, and Plutonium, as a coolant, fertile, and fissile nuclide, respectively. Plutonium loaded is a weapon-grade which consist of 238Pu, 239Pu, 240Pu, 241Pu, 242Pu, and 241Am. Graphite is used as moderator; therefore, the reactor is operated in thermal energy range. The reactor design is calculated in neutronic terms with program code CITATION in SRAC 2006 with JENDL 4.0 as nuclear data library. The result shows some neutronic parameter changes with increasing Plutonium loaded. The utilization of Plutonium, in this case, is described as a capability of MSR in burning a high-level waste of nuclear and radioactive isotopes. This system can be dedicated to future cleaning energy production in a nuclear reactor.

Keywords
MSR; Neutronic design; SRAC;Thorium; Weapon Grade Plutonium

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/frXBULzjt7RN

Corresponding Author
Andrey Kosasih

Institutions
(a) Department of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Jl. Ganesa 10, Bandung 40132, Indonesia
(b) Nuclear Physics & Biophysics Research Division, Department of Physics, Faculty of
Mathematics and Natural Sciences, Bandung Institute of Technology, Jl. Ganesa 10, Bandung 40132, Indonesia
*awaris[at]fi.itb.ac.id

Abstract
The high temperature engineering test reactor (HTTR) is a block-type high-temperature gas-cooled reactor (HTGR) developed by Japan. The HTTR is a graphite-moderated and helium-cooled HTGR with a thermal power of 30 MW and a maximum outlet temperature of 950°C. HTTR uses UO2 fuel with enrichment vary from 3,4% to 9,9%. Neutronic calculations are performed by PIJ and CITATION modules on SRAC2006 code system with JENDL-4.0 as the nuclear data library. In this study, two models are used for the cell-burnup calculations. Microscopic cell in the fuel compact is defined in model 1, whereas it is not defined in model 2. The core geometry used in these calculations are triangular-z and hexagonal-z. The neutronic analysis includes several parameters such as effective multiplication factor (k-eff), conversion ratio, changes in atomic density for fissile and fertile materials, neutron spectrum, power density distribution, and power peaking factor. The results show a quite different neutronic parameters with model 1 and 2. Model 1 achieved the first criticality at the enrichment of 6,3%, whereas model 2 at 6,7%. The conversion ratio tends to increase during burnup and greater in model 2. Both core geometries show similar results with greater k-eff in the triangular-z geometry. The maximum power density is located at the fuel block with an enrichment of 6,7%.

Keywords
HTTR, JENDL-4.0, k-eff, neutronic, SRAC2006

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/tMpkbA9BqDv6

Corresponding Author
Imam Ghazali Yasmint

Institutions
1)Nuclear Physics Laboratory,
Nuclear Physics and Biophysics, Department of Physics
Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology,
Ganesha No. 10 Bandung, Indonesia, 40132

2)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
Today, nuclear technology has become one of the most interesting research objects in the world. With many nuclear reactors operating in the world, both power reactors and research reactors, and the application of nuclear technology in all fields can have a detrimental effect in the form of releasing radioactive material into the environment. Other than that, there are also natural sources of radiation that have emerged naturally due to natural symptoms that occur on earth and radiation that is in the human body (internal radiation). Natural radiation is the biggest contributor to the radiation source that humans received in one year. So it is important to monitor radiation in an area, especially radiation from natural sources. Natural radiation monitoring is carried out by mapping natural radiation of certain areas, so that spots of the area are found which have anomalies in environmental radioactivity. In addition, it is necessary to study radionuclide transfer factors for environmental elements such as soil, plants, etc. This study used an experimental method by measuring radionuclides in several samples. Sampling is done in an area in Lembang. The samples taken were soil and grass, then measured using an ORTEC gamma spectrometer with HPGe detector. Based on the measurement of samples, it can be concluded that in the area there is no anomaly of radioactivity in the environment and still at a safe level of radiation.

Keywords
Gamma Spectrometer, HPGe Detector, Natural Radiation, Soil and Grass Samples, Transfer Factors

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/uXaprPh9ATBz

Corresponding Author
Rindi Wulandari

Institutions
(1)(2)Nuclear Physics and Biophysics Research Division, Institut Teknologi Bandung
Jl. Ganesha 10, Bandung 40132, Gedung Fisika FMIPA ITB Indonesia

(3) Theoretical High Energy Physics and Instrumentation Division, Institut Teknologi Bandung
Jl. Ganesha 10, Bandung 40132, Gedung Fisika FMIPA ITB Indonesia

1) wulandarindi[at]gmail.com (corresponding author)
2) psidik[at]fi.itb.ac.id
3) supri.haryono[at]gmail.com

Abstract
One of the problems in fullfing energy needs in Indonesia is marked by the low electrification ratio, which is 60%. Many researchs dan various studies of alternative energy has been conducting to solve these problems. One of them is nuclear energy. The development of nuclear power plant (NPP) is very rapid. Nowdays, many studies of 4th Generation nuclear reactor which focus on improving safety is conducted. The characteristic of some IV generation nuclear reactors is the use of molten salt as a coolant. The purpose of this study is to determine the heat transfer of molten salt in the natural circulation system for steady state analysis and transient characteristic with COMSOL Multiphysics method. The selected module is the Non-Isothermal FLow (NITF) module. This module is a combination of three basic equations, namely the continuity equation, the Navier-Stokes equation, and the dynamic equation of heat transfer in fluid. The simulation model measures 1.5 x 2 (m) with sodium (Na) as a fluid. The simulation demonstrates 4 conditions: 1) Steady state; 2) Transient I; 3) Transient II; 4) Heater Trip. The conditions of transient I, and transient II, show the system is still in a safe condition because the temperature value is still below the value of liquid sodium boiling and SS316 pipe melting point. In the heater trip condition, liquid sodium has a temperature drop to near freezing.

Keywords
heat transfer, natural circulation, COMSOL Multiphysics method

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/6rk3aR4nEGqt

Corresponding Author
Robi Dany Riupassa

Institutions
a) Department of Physics, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia
b) Department of Physics and Department of Nuclear Science & Engineering, Faculty of Mathematics and Natural Sciences, Institut Teknologi Bandung, Indonesia
*robiriu[at]students.itb.ac.id

Abstract
The phenomenon of natural circulation is used as a mechanism of passive cooling in nuclear reactors. This mechanism will help to dispose of residual heat in nuclear reactors when emergency conditions for example due to pump failure. Testing natural circulation systems can be done with experiments or closed-loop system simulations. Previous studies have conducted experiments with closed loop systems with variations in temperature differences between heaters and coolers. Water is used as a fluid in this experiment. The experimental results show that the temperature difference between the heater and the cooler influences to the velocity of fluid flow. For a maximum temperature difference of 80 oC, the flow velocity reaches 3 cm/s. For single phase closed loop systems with water as the fluid used, this value is no longer possible to be enlarged. In this study, a single phase closed loop system will be further investigated using computational fluid dynamics (CFD). Tests are carried out for several types of fluids to see the effect of fluid density on temperature differences in heating and cooling. Simulation results will show the temperature distribution and velocity of fluid flow in a closed loop system.

Keywords
natural circulation; CFD

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/74Q93N8hXbEK

Corresponding Author
Sabiq Fatoni

Institutions
Institut Teknologi Bandung

Abstract
Nuclear energy is one of the alternative energy sources to resolve the increasing needs of energy sources. Small Modular reactor (SMR) is one type of small nuclear reactor that can be resolve energy needs for remote areas. In this study, variations of fuel were carried out on the NuScale SMR reactor were Thorium with Plutonium (reactor grade plutonium), Thorium with enrichment Uranium, and Thorium with U-233. This study aims to determine the composition of the fuel that makes the reactor capable in a critical state for 10 years with a power of 150MWth. Variations in the composition concentration of Plutonium, concentration and enrichment of natural uranium, and concentration U233 were carried out on the NuScale reactor core. Neutronic analysis was carried out using the SRAC2006 program with JENDL4.0 on the LINUX operating system, which resulted in the value of the multiplication factor, conversion ratio, and reactor reactivity values. Based on calculations and simulations critical reactor (Keff ≥ 1) was produced for 10 years with a composition of 4.47% Plutonium in Th-Pu, 14%, 13%, and 12% Uranium with 17%, 18%, and 19% enrichment in Th-U, and 1,8% Uranium-233 in Th-U233. The reactor reactivity shows that the value is always positive and tends to decrease every year so the reactor is safe. Value of conversion ratio (CR) < 1 so that it can be used to reduce plutonium waste from other reactors

Keywords
Conversion Ratio , Multiplication Factor, NuScale, SRAC2006, Thorium

Topic
Nuclear Science and Engineering

Link: https://ifory.id/abstract/QUP9EFh7ZXyJ