Ratna Dewi Syarifah - Academia.edu (original) (raw)

Papers by Ratna Dewi Syarifah

Research paper thumbnail of Studi Desain Boiling Water Reactor (BWR) Berdaya 100MW Berbahan Bakar Thorium Karbida

POSITRON

Reaktor air didih atau boiling water reactor (BWR) merupakan reaktor yang memanfaatkan air sebaga... more Reaktor air didih atau boiling water reactor (BWR) merupakan reaktor yang memanfaatkan air sebagai media pendingin dan juga sebagai moderator. Penelitian ini mendesain reaktor air didih berbentuk modul atau Small Modular Reactor (SMR) berdaya 100MWth berbahan bakar Thorium Karbida (Th-U233-C), selongsong Silikon Karbida (SiC) dan pendingin air (H2O). Perhitungan neutronik dilakukan dengan mengunakan program SRAC versi 2006 dengan database nuklir JENDL 4.0. Hasil yang diperoleh menunjukkan bahwa reaktor air didih berbentuk modul dengan bahan bakar Thorium Karbida yang diberi tambahan Protaktinium-231 (Pa-231) sebanyak 6% menunjukkan hasil paling kritis. Reaktor mampu beroperasi selama 20 tahun tanpa pengisian ulang bahan bakar dengan maksimum faktor multiplikasi efektif (keff) 1,00392. Nilai maximum excess reactivity yang dihasilkan sangat kecil dibawah 1% yakni 0,39% ∆k/k dan densitas daya maksimum yang dihasilkan bernilai 59 watt/cc.

Research paper thumbnail of Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Eastern-European Journal of Enterprise Technologies

Neutronic analysis on the Molten Salt Reactor FUJI-12 using the fissile material 235U in LiF-BeF2... more Neutronic analysis on the Molten Salt Reactor FUJI-12 using the fissile material 235U in LiF-BeF2-UF4 has been carried out. The problem faced in the use of thorium-based fuel is that the amount of 233U is small and not available in nature. 233U was produced through the 232Th breeding at a cost of 46million/kg.Thatisaveryhighpricewhencomparedto235Uenrichment,whichisonly46 million/kg. That is a very high price when compared to 235U enrichment, which is only 46million/kg.Thatisaveryhighpricewhencomparedto235Uenrichment,whichisonly100/kg. The MSR FUJI-12 used in this study is a generation IV reactor with a mixture of liquid salt fuel LiF-BeF2-ThF4-UF4 and thorium-based fuel (232Th+233U). In this study, neutronic analysis was carried out by replacing thorium-based fuel with uranium-based fuel (235U+238U). Neutronic analysis was performed using the OpenMC 0.13.0 code, which is a Monte Carlo simulation-based neutron analysis code. The nuclear data library used for neutronic calculations is ENDF B-VII/1. The fuel is used in a LiF-BeF2-UF4 molten salt mixture with three eutectic compositions: fuel 1, fuel 2, and fuel 3. Each fuel co...

Research paper thumbnail of Neutronics analysis of UN-PuN fuel for 300MW pressurized water reactor using SRAC-COREBN code

Eureka: Physics and Engineering, Nov 29, 2022

Nuclear Power Plants (NPP) is one of alternative energy that can be used to replace fossil energy... more Nuclear Power Plants (NPP) is one of alternative energy that can be used to replace fossil energy. NPP is a clean energy that doesn't emit CO 2 as a residue from the power plant. Pressurized Water Reactor (PWR) is the most widely used commercial reactor to date. This reactor usually uses UO 2 as fuel with enrich U-235 and produce plutonium as a waste nuclear. In this study, the fuel uses natural uranium and the addition of plutonium from the spent fuel of PWR, calling it Uranium-Plutonium Nitride (UN-PuN) fuel.

Research paper thumbnail of Preliminary study of 300 MWth pressurized water reactor with carbide fuel with addition neptunium 237 using SRAC-COREBN code

THE 3RD INTERNATIONAL CONFERENCE ON PHYSICAL INSTRUMENTATION AND ADVANCED MATERIALS (ICPIAM) 2021

Research paper thumbnail of Comparison of thorium nitride and uranium nitride fuel on small modular pressurized water reactor in neutronic analysis using SRAC code

Eastern-European Journal of Enterprise Technologies

Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutron... more Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutronic analysis has been carried out. PWR in module is one type of reactor that can be utilized because of its small size so that it can be placed on demand. Neutronic calculations were performed using SRAC version 2006, the data library using JENDL 4.0. The first calculation was fuel pin (PIJ) calculation with hexagonal fuel pin cell type. And the second calculation was reactor core (CITATION) calculation using homogeneous and heterogeneous core configurations. ThN and UN fuels use heterogeneous configurations with 3 fuel variations. The reactor geometry was used in two fuels are the same, with diameter and height active core was 300 cm and 100 cm. In this research, Np-237 was added as a minor actinide in the UN fuel to reduce the amount of Np-237 in the world and also reduce the k-eff value. For ThN fuel, Pa-231 also added in the fuel to reduce the k-eff value. The optimum configuration of...

Research paper thumbnail of The potential use of a hybrid human-detector for physical distance monitoring

THE 3RD INTERNATIONAL CONFERENCE ON PHYSICAL INSTRUMENTATION AND ADVANCED MATERIALS (ICPIAM) 2021

Research paper thumbnail of MOx Depletion Calculation Benchmark

Research paper thumbnail of Studi Karakteristik Aliran Gas Berbagai Polutan Menggunakan Simulasi CFD(Computational Fluid Dynamics)

Seminar Nasional Fisika dan Pembelajarannya, Jul 15, 2019

Research paper thumbnail of GPVDM simulation of thickness effect on power conversion efficiency of PEDOT:PSS/P3HT:PCBM solar cell performance

THE 3RD INTERNATIONAL CONFERENCE ON PHYSICAL INSTRUMENTATION AND ADVANCED MATERIALS (ICPIAM) 2021

Research paper thumbnail of Analisis Fraksi Volume Bahan Bakar Uranium Karbida Pada Reaktor Cepat Berpendingin Gas Menggunakan SRAC Code

The fuel volume fraction analysis has been carried out with uranium carbide in Gas-Cooled Fast Re... more The fuel volume fraction analysis has been carried out with uranium carbide in Gas-Cooled Fast Reactor using the SRAC Code (Standard Reactor Analysis Code). This code was developed by the Japan Atomic Energy Agency (JAEA) and uses the latest nuclear data library JENDL 4.0. There are two calculation has been used, fuel pin cell calculation (PIJ Calculation) and core calculation (CITATION Calculation). In core calculation, the leakage is calculated so the calculation more precise. The CITATION calculation use two type of core configuration, i.e. homogeneous core configuration and heterogeneous core configuration. The power density value of two type core configuration is quite difference. It is better use heterogeneous core configuration than homogeneous core configuration, because the power density of heterogeneous core configuration is flatter than the other. The optimum design based on the analysis that has been carried out is the fuel volume fraction of 49% with a heterogeneous core configuration of 3 types of fuel percentages, for Fuel type 1 = 9%, Fuel type 2 = 12% and Fuel type 3 = 15% with the reactor is cylindrical and has a core diameter of 240 cm and a core height of 100 cm.

Research paper thumbnail of Study of Neptunium, Americium and Protactinium Addition for 300MWth GFR with Uranium Carbide Fuel

Computational And Experimental Research In Materials And Renewable Energy, 2019

A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide f... more A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide fuel has been performed. The purpose of this study was to determine the characteristics of addition Neptunium, Americium, and Protactinium in a 300MWth Gas-Cooled Fast Reactor. Neutronics calculation was design by using Standard Reactor Analysis Code (SRAC) version 2006 with data nuclides from JENDL-4.0. Neutronics calculations were initiated by calculating the fuel cell calculation (PIJ calculation) and continued with the reactor core calculation (CITATION calculation). The reactor core calculation used two-reactor core configurations, namely the homogeneous core configuration and heterogeneous core configuration. The Neptunium, Americium, and Protactinium additions were performed after obtaining the optimal condition from heterogeneous core configuration. The addition of Neptunium and Americium which are Spent Nuclear Fuel (SNF) from LWR fuels, aims to reduce the amount of Neptunium and...

Research paper thumbnail of Comparative study on various thermal power for gas cooled fast reactor with uranium plutonium nitride fuel

HIGH-ENERGY PROCESSES IN CONDENSED MATTER (HEPCM 2020): Proceedings of the XXVII Conference on High-Energy Processes in Condensed Matter, dedicated to the 90th anniversary of the birth of RI Soloukhin

Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium ni... more Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium nitride fuel has been done. The purpose of this study was to compare various thermal powers on nuclear power plant (NPP) Gas Cooled Reactor type. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some parameter surveys to obtain the optimal results. The first step was calculation of fuel cell (PIJ-method) by using hexagonal cell and then followed by calculation of core reactor (CITATION-method). The power variations carried out are seven variations of power, namely from 100 MWth to 700 MWth. In this power variation all parameters are made the same. The parameters that are made are the same, namely the percentage of plutonium, the volume fraction of the fuel, the reactor geometry type, the diameter and the active core height (terrace). All power variations have an average power density value and maximum power density increases along with the increase in thermal power. When thermal power increases, the k-eff peak value will be increase too. It shows that if the thermal power increases, the burn-up fuel is also increase more than before, so that the fuel is used for the burn-up process which causes k-eff increase. The increasing graph which shows in the figure explains that the reactor breeding occurs, while the declining graph shows the reactor is burning.

Research paper thumbnail of Analysis of Tensile Strenght and Shear Modulus of GRE Pipe using Ansys

Computational And Experimental Research In Materials And Renewable Energy, 2018

Composite materials (GRE pipe) had been applied in various industries. These kind application are... more Composite materials (GRE pipe) had been applied in various industries. These kind application are based on the advantages of composite properties, that are lightweight, high corrosion resistance and low cost. In order to make a lightweight and strong materials, some materials that light and stiff had been widely used, lie fiber glass, epoxy and the other. These materials (fiber glass and epoxy) are synthetic and non-biodegradable, but give some advantages in composite to make composite more stiff, light and strong. Mechanical properties of fiber glass composites had been doing based on theory through modeling. Theoretical results obtained showed that maximum stress and shear modulus value of GRE pipe are less than each components (glass fiber and epoxy resin). Each value of maximum stress and shear modulus are 584,57 MPa and 46,15 MPa.

Research paper thumbnail of Prospect of Uranium Nitride for Pressurized Water Reactor (PWR)

Dalam studi ini telah dilakukan studi desain reaktor air bertekanan (PWR) dengan bahan bakar uran... more Dalam studi ini telah dilakukan studi desain reaktor air bertekanan (PWR) dengan bahan bakar uranium nitride (UN) dan mixed nitride (UNPuN), Pa-231 sebagai material burnable poison, dan Americium sebagai tambahan dari aktinida minor. Pa-231 mempunyai harga cross section capture yang besar dan memiliki kemampuan untuk menghasilkan material fisil sehingga material ini sangat berpotensi untuk mengurangi excess reactivity. Americium adalah salah satu minor actinide yang mempunyai waktu paruh yang lama. Tujuannya menambahkan americium adalah untuk menguragi bahan bakar bekas nuklir (spent fuel) di dunia. Analisis neutronic dari hasil perhitungan sel menunjukkan bahwa bahan bakar mixed nitride menghasilkan k-inf lebih besar daripada bahan bakar uranium nitride. Ini disebabkan karena penambahan Pu-239 di dalam bahan bakar mixed nitride. Dalam analysis fraksi bahan bakar, fraksi paling optimum adalah 50% fraksi bahan bakar, 20% cladding dan 30% moderator. Untuk penambahan Pa-231 dan Americi...

Research paper thumbnail of Neutronic Analysis of UN-PuN Fuel use FI-ITB-CHI Code for 500MWth GFR Long Life Without Refueling

Journal of Physics: Conference Series, 2018

Research paper thumbnail of Design Study of 600 MWt Long Life Modular Gas Cooled Fast Reactors

Journal of Physics: Conference Series, 2018

Research paper thumbnail of Phase Transition of Fe₃O₄ Magnetic Material Based on Observation of Curie Temperature and Hysteresis Curve: Micromagnetic Simulation Study

European Journal of Applied Physics, 2021

Phase transition yesng happens to the material magnetite (Fe3O4) is an interesting phenomenon to ... more Phase transition yesng happens to the material magnetite (Fe3O4) is an interesting phenomenon to study because it has many important applications, one of which is RAM (Radar Absorbing Material). The magnetic properties of nanomaterials are known to be influenced by their size. In this simulation research, the research objective was to analyze the temperature value of the Curie and the hysteresis curve of the Fe3O4 material with variations in the size of the material sample cube of 5 nm, 8 nm, 10 nm, 12 nm, and 15 nm. In this study, using a micromagnetic simulation method based on atomistic models with the Vampire program. The results showed that the Curie temperature value in the Fe3O4 material was influenced by variations in the size of the material. The Curie temperature values when the side sizes of the cube are 5 nm, 8 nm, 10 nm, 12 nm, and 15 nm, namely 650 K, 635 K, 650 K, 665 K and 645 K. The characteristics of the hysteresis curve for Fe3O4 material based on simulations at e...

Research paper thumbnail of Comparative study on various thermal power for gas cooled fast reactor with uranium plutonium nitride fuel

HIGH-ENERGY PROCESSES IN CONDENSED MATTER (HEPCM 2020): Proceedings of the XXVII Conference on High-Energy Processes in Condensed Matter, dedicated to the 90th anniversary of the birth of RI Soloukhin, 2020

Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium ni... more Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium nitride fuel has been done. The purpose of this study was to compare various thermal powers on nuclear power plant (NPP) Gas Cooled Reactor type. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some parameter surveys to obtain the optimal results. The first step was calculation of fuel cell (PIJ-method) by using hexagonal cell and then followed by calculation of core reactor (CITATION-method). The power variations carried out are seven variations of power, namely from 100 MWth to 700 MWth. In this power variation all parameters are made the same. The parameters that are made are the same, namely the percentage of plutonium, the volume fraction of the fuel, the reactor geometry type, the diameter and the active core height (terrace). All power variations have an average power density value and maximum power density increases along with the increase in thermal power. When thermal power increases, the k-eff peak value will be increase too. It shows that if the thermal power increases, the burn-up fuel is also increase more than before, so that the fuel is used for the burn-up process which causes k-eff increase. The increasing graph which shows in the figure explains that the reactor breeding occurs, while the declining graph shows the reactor is burning.

Research paper thumbnail of Actinide Minor Addition on Uranium Plutonium Nitride Fuel for Modular Gas Cooled Fast Reactor

Journal of Physics: Conference Series, 2020

This study has made a calculation of Actinide Minor Addition on Uranium Plutonium Nitride Fuel fo... more This study has made a calculation of Actinide Minor Addition on Uranium Plutonium Nitride Fuel for Modular Gas Cooled Fast Reactor. The purpose of this study was to determine the characteristics of minor actinides when incorporated into plutonium nitride uranium fuel. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some steps of parameter survey to obtain the ultimate optimization results. The initial calculation was calculation of fuel cell (PIJ calculation) by using hexagonal cell and then followed by calculation of core reactor (CITATION calculation) by using program code SRAC2006. The addition of minor actinide has been calculated in this research. The minor actinides which were added were americium (Am-241 and Am-243) and neptunium 237. They were put into the uranium plutonium nitri...

Research paper thumbnail of Neutronic analysis of comparation UN-PuN fuel and ThN fuel for 300MWth Gas Cooled Fast Reactor long life without refueling

Journal of Physics: Conference Series, 2020

Neutronic analysis of comparation UN-PuN fuel and ThN fuel for 300MWth Gas Cooled Fast Reactor lo... more Neutronic analysis of comparation UN-PuN fuel and ThN fuel for 300MWth Gas Cooled Fast Reactor long life without refueling has been done. Gas Cooled Fast Reactor is a Generation IV reactor with gas coolant (i.e. helium) and using fast spectrum neutron. The neutronic calculation was carried out using SRAC (Standard Reactor Analysis Code) version 2006 under the Linux Operating System with nuclear data library JENDL4.0. The first calculation is fuel pin cell calculation (PIJ-method) by using a hexagonal cell and then followed by the calculation of the core reactor (CITATION-method). The calculation of the core reactor used homogeneous and heterogeneous core configuration. The UN-PuN fuel use plutonium as a fissile material and natural uranium as a fertile material and the ThN fuel use U233 as a fissile material and natural thorium as a fertile material. The percentages of fissile material are varied in heterogeneous core configuration. It is used to decrease the peaking power in the ce...

Research paper thumbnail of Studi Desain Boiling Water Reactor (BWR) Berdaya 100MW Berbahan Bakar Thorium Karbida

POSITRON

Reaktor air didih atau boiling water reactor (BWR) merupakan reaktor yang memanfaatkan air sebaga... more Reaktor air didih atau boiling water reactor (BWR) merupakan reaktor yang memanfaatkan air sebagai media pendingin dan juga sebagai moderator. Penelitian ini mendesain reaktor air didih berbentuk modul atau Small Modular Reactor (SMR) berdaya 100MWth berbahan bakar Thorium Karbida (Th-U233-C), selongsong Silikon Karbida (SiC) dan pendingin air (H2O). Perhitungan neutronik dilakukan dengan mengunakan program SRAC versi 2006 dengan database nuklir JENDL 4.0. Hasil yang diperoleh menunjukkan bahwa reaktor air didih berbentuk modul dengan bahan bakar Thorium Karbida yang diberi tambahan Protaktinium-231 (Pa-231) sebanyak 6% menunjukkan hasil paling kritis. Reaktor mampu beroperasi selama 20 tahun tanpa pengisian ulang bahan bakar dengan maksimum faktor multiplikasi efektif (keff) 1,00392. Nilai maximum excess reactivity yang dihasilkan sangat kecil dibawah 1% yakni 0,39% ∆k/k dan densitas daya maksimum yang dihasilkan bernilai 59 watt/cc.

Research paper thumbnail of Neutronic analysis on molten salt reactor FUJI-12 using 235U as fissile material in LiF-BeF2-UF4 fuel

Eastern-European Journal of Enterprise Technologies

Neutronic analysis on the Molten Salt Reactor FUJI-12 using the fissile material 235U in LiF-BeF2... more Neutronic analysis on the Molten Salt Reactor FUJI-12 using the fissile material 235U in LiF-BeF2-UF4 has been carried out. The problem faced in the use of thorium-based fuel is that the amount of 233U is small and not available in nature. 233U was produced through the 232Th breeding at a cost of 46million/kg.Thatisaveryhighpricewhencomparedto235Uenrichment,whichisonly46 million/kg. That is a very high price when compared to 235U enrichment, which is only 46million/kg.Thatisaveryhighpricewhencomparedto235Uenrichment,whichisonly100/kg. The MSR FUJI-12 used in this study is a generation IV reactor with a mixture of liquid salt fuel LiF-BeF2-ThF4-UF4 and thorium-based fuel (232Th+233U). In this study, neutronic analysis was carried out by replacing thorium-based fuel with uranium-based fuel (235U+238U). Neutronic analysis was performed using the OpenMC 0.13.0 code, which is a Monte Carlo simulation-based neutron analysis code. The nuclear data library used for neutronic calculations is ENDF B-VII/1. The fuel is used in a LiF-BeF2-UF4 molten salt mixture with three eutectic compositions: fuel 1, fuel 2, and fuel 3. Each fuel co...

Research paper thumbnail of Neutronics analysis of UN-PuN fuel for 300MW pressurized water reactor using SRAC-COREBN code

Eureka: Physics and Engineering, Nov 29, 2022

Nuclear Power Plants (NPP) is one of alternative energy that can be used to replace fossil energy... more Nuclear Power Plants (NPP) is one of alternative energy that can be used to replace fossil energy. NPP is a clean energy that doesn't emit CO 2 as a residue from the power plant. Pressurized Water Reactor (PWR) is the most widely used commercial reactor to date. This reactor usually uses UO 2 as fuel with enrich U-235 and produce plutonium as a waste nuclear. In this study, the fuel uses natural uranium and the addition of plutonium from the spent fuel of PWR, calling it Uranium-Plutonium Nitride (UN-PuN) fuel.

Research paper thumbnail of Preliminary study of 300 MWth pressurized water reactor with carbide fuel with addition neptunium 237 using SRAC-COREBN code

THE 3RD INTERNATIONAL CONFERENCE ON PHYSICAL INSTRUMENTATION AND ADVANCED MATERIALS (ICPIAM) 2021

Research paper thumbnail of Comparison of thorium nitride and uranium nitride fuel on small modular pressurized water reactor in neutronic analysis using SRAC code

Eastern-European Journal of Enterprise Technologies

Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutron... more Comparison of thorium nitride (ThN) and uranium nitride (UN) fuel on small modular PWR in neutronic analysis has been carried out. PWR in module is one type of reactor that can be utilized because of its small size so that it can be placed on demand. Neutronic calculations were performed using SRAC version 2006, the data library using JENDL 4.0. The first calculation was fuel pin (PIJ) calculation with hexagonal fuel pin cell type. And the second calculation was reactor core (CITATION) calculation using homogeneous and heterogeneous core configurations. ThN and UN fuels use heterogeneous configurations with 3 fuel variations. The reactor geometry was used in two fuels are the same, with diameter and height active core was 300 cm and 100 cm. In this research, Np-237 was added as a minor actinide in the UN fuel to reduce the amount of Np-237 in the world and also reduce the k-eff value. For ThN fuel, Pa-231 also added in the fuel to reduce the k-eff value. The optimum configuration of...

Research paper thumbnail of The potential use of a hybrid human-detector for physical distance monitoring

THE 3RD INTERNATIONAL CONFERENCE ON PHYSICAL INSTRUMENTATION AND ADVANCED MATERIALS (ICPIAM) 2021

Research paper thumbnail of MOx Depletion Calculation Benchmark

Research paper thumbnail of Studi Karakteristik Aliran Gas Berbagai Polutan Menggunakan Simulasi CFD(Computational Fluid Dynamics)

Seminar Nasional Fisika dan Pembelajarannya, Jul 15, 2019

Research paper thumbnail of GPVDM simulation of thickness effect on power conversion efficiency of PEDOT:PSS/P3HT:PCBM solar cell performance

THE 3RD INTERNATIONAL CONFERENCE ON PHYSICAL INSTRUMENTATION AND ADVANCED MATERIALS (ICPIAM) 2021

Research paper thumbnail of Analisis Fraksi Volume Bahan Bakar Uranium Karbida Pada Reaktor Cepat Berpendingin Gas Menggunakan SRAC Code

The fuel volume fraction analysis has been carried out with uranium carbide in Gas-Cooled Fast Re... more The fuel volume fraction analysis has been carried out with uranium carbide in Gas-Cooled Fast Reactor using the SRAC Code (Standard Reactor Analysis Code). This code was developed by the Japan Atomic Energy Agency (JAEA) and uses the latest nuclear data library JENDL 4.0. There are two calculation has been used, fuel pin cell calculation (PIJ Calculation) and core calculation (CITATION Calculation). In core calculation, the leakage is calculated so the calculation more precise. The CITATION calculation use two type of core configuration, i.e. homogeneous core configuration and heterogeneous core configuration. The power density value of two type core configuration is quite difference. It is better use heterogeneous core configuration than homogeneous core configuration, because the power density of heterogeneous core configuration is flatter than the other. The optimum design based on the analysis that has been carried out is the fuel volume fraction of 49% with a heterogeneous core configuration of 3 types of fuel percentages, for Fuel type 1 = 9%, Fuel type 2 = 12% and Fuel type 3 = 15% with the reactor is cylindrical and has a core diameter of 240 cm and a core height of 100 cm.

Research paper thumbnail of Study of Neptunium, Americium and Protactinium Addition for 300MWth GFR with Uranium Carbide Fuel

Computational And Experimental Research In Materials And Renewable Energy, 2019

A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide f... more A study of Neptunium, Americium, and Protactinium addition for GFR 300MWth with Uranium Carbide fuel has been performed. The purpose of this study was to determine the characteristics of addition Neptunium, Americium, and Protactinium in a 300MWth Gas-Cooled Fast Reactor. Neutronics calculation was design by using Standard Reactor Analysis Code (SRAC) version 2006 with data nuclides from JENDL-4.0. Neutronics calculations were initiated by calculating the fuel cell calculation (PIJ calculation) and continued with the reactor core calculation (CITATION calculation). The reactor core calculation used two-reactor core configurations, namely the homogeneous core configuration and heterogeneous core configuration. The Neptunium, Americium, and Protactinium additions were performed after obtaining the optimal condition from heterogeneous core configuration. The addition of Neptunium and Americium which are Spent Nuclear Fuel (SNF) from LWR fuels, aims to reduce the amount of Neptunium and...

Research paper thumbnail of Comparative study on various thermal power for gas cooled fast reactor with uranium plutonium nitride fuel

HIGH-ENERGY PROCESSES IN CONDENSED MATTER (HEPCM 2020): Proceedings of the XXVII Conference on High-Energy Processes in Condensed Matter, dedicated to the 90th anniversary of the birth of RI Soloukhin

Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium ni... more Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium nitride fuel has been done. The purpose of this study was to compare various thermal powers on nuclear power plant (NPP) Gas Cooled Reactor type. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some parameter surveys to obtain the optimal results. The first step was calculation of fuel cell (PIJ-method) by using hexagonal cell and then followed by calculation of core reactor (CITATION-method). The power variations carried out are seven variations of power, namely from 100 MWth to 700 MWth. In this power variation all parameters are made the same. The parameters that are made are the same, namely the percentage of plutonium, the volume fraction of the fuel, the reactor geometry type, the diameter and the active core height (terrace). All power variations have an average power density value and maximum power density increases along with the increase in thermal power. When thermal power increases, the k-eff peak value will be increase too. It shows that if the thermal power increases, the burn-up fuel is also increase more than before, so that the fuel is used for the burn-up process which causes k-eff increase. The increasing graph which shows in the figure explains that the reactor breeding occurs, while the declining graph shows the reactor is burning.

Research paper thumbnail of Analysis of Tensile Strenght and Shear Modulus of GRE Pipe using Ansys

Computational And Experimental Research In Materials And Renewable Energy, 2018

Composite materials (GRE pipe) had been applied in various industries. These kind application are... more Composite materials (GRE pipe) had been applied in various industries. These kind application are based on the advantages of composite properties, that are lightweight, high corrosion resistance and low cost. In order to make a lightweight and strong materials, some materials that light and stiff had been widely used, lie fiber glass, epoxy and the other. These materials (fiber glass and epoxy) are synthetic and non-biodegradable, but give some advantages in composite to make composite more stiff, light and strong. Mechanical properties of fiber glass composites had been doing based on theory through modeling. Theoretical results obtained showed that maximum stress and shear modulus value of GRE pipe are less than each components (glass fiber and epoxy resin). Each value of maximum stress and shear modulus are 584,57 MPa and 46,15 MPa.

Research paper thumbnail of Prospect of Uranium Nitride for Pressurized Water Reactor (PWR)

Dalam studi ini telah dilakukan studi desain reaktor air bertekanan (PWR) dengan bahan bakar uran... more Dalam studi ini telah dilakukan studi desain reaktor air bertekanan (PWR) dengan bahan bakar uranium nitride (UN) dan mixed nitride (UNPuN), Pa-231 sebagai material burnable poison, dan Americium sebagai tambahan dari aktinida minor. Pa-231 mempunyai harga cross section capture yang besar dan memiliki kemampuan untuk menghasilkan material fisil sehingga material ini sangat berpotensi untuk mengurangi excess reactivity. Americium adalah salah satu minor actinide yang mempunyai waktu paruh yang lama. Tujuannya menambahkan americium adalah untuk menguragi bahan bakar bekas nuklir (spent fuel) di dunia. Analisis neutronic dari hasil perhitungan sel menunjukkan bahwa bahan bakar mixed nitride menghasilkan k-inf lebih besar daripada bahan bakar uranium nitride. Ini disebabkan karena penambahan Pu-239 di dalam bahan bakar mixed nitride. Dalam analysis fraksi bahan bakar, fraksi paling optimum adalah 50% fraksi bahan bakar, 20% cladding dan 30% moderator. Untuk penambahan Pa-231 dan Americi...

Research paper thumbnail of Neutronic Analysis of UN-PuN Fuel use FI-ITB-CHI Code for 500MWth GFR Long Life Without Refueling

Journal of Physics: Conference Series, 2018

Research paper thumbnail of Design Study of 600 MWt Long Life Modular Gas Cooled Fast Reactors

Journal of Physics: Conference Series, 2018

Research paper thumbnail of Phase Transition of Fe₃O₄ Magnetic Material Based on Observation of Curie Temperature and Hysteresis Curve: Micromagnetic Simulation Study

European Journal of Applied Physics, 2021

Phase transition yesng happens to the material magnetite (Fe3O4) is an interesting phenomenon to ... more Phase transition yesng happens to the material magnetite (Fe3O4) is an interesting phenomenon to study because it has many important applications, one of which is RAM (Radar Absorbing Material). The magnetic properties of nanomaterials are known to be influenced by their size. In this simulation research, the research objective was to analyze the temperature value of the Curie and the hysteresis curve of the Fe3O4 material with variations in the size of the material sample cube of 5 nm, 8 nm, 10 nm, 12 nm, and 15 nm. In this study, using a micromagnetic simulation method based on atomistic models with the Vampire program. The results showed that the Curie temperature value in the Fe3O4 material was influenced by variations in the size of the material. The Curie temperature values when the side sizes of the cube are 5 nm, 8 nm, 10 nm, 12 nm, and 15 nm, namely 650 K, 635 K, 650 K, 665 K and 645 K. The characteristics of the hysteresis curve for Fe3O4 material based on simulations at e...

Research paper thumbnail of Comparative study on various thermal power for gas cooled fast reactor with uranium plutonium nitride fuel

HIGH-ENERGY PROCESSES IN CONDENSED MATTER (HEPCM 2020): Proceedings of the XXVII Conference on High-Energy Processes in Condensed Matter, dedicated to the 90th anniversary of the birth of RI Soloukhin, 2020

Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium ni... more Comparative study on various thermal powers for gas cooled fast reactor with uranium plutonium nitride fuel has been done. The purpose of this study was to compare various thermal powers on nuclear power plant (NPP) Gas Cooled Reactor type. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some parameter surveys to obtain the optimal results. The first step was calculation of fuel cell (PIJ-method) by using hexagonal cell and then followed by calculation of core reactor (CITATION-method). The power variations carried out are seven variations of power, namely from 100 MWth to 700 MWth. In this power variation all parameters are made the same. The parameters that are made are the same, namely the percentage of plutonium, the volume fraction of the fuel, the reactor geometry type, the diameter and the active core height (terrace). All power variations have an average power density value and maximum power density increases along with the increase in thermal power. When thermal power increases, the k-eff peak value will be increase too. It shows that if the thermal power increases, the burn-up fuel is also increase more than before, so that the fuel is used for the burn-up process which causes k-eff increase. The increasing graph which shows in the figure explains that the reactor breeding occurs, while the declining graph shows the reactor is burning.

Research paper thumbnail of Actinide Minor Addition on Uranium Plutonium Nitride Fuel for Modular Gas Cooled Fast Reactor

Journal of Physics: Conference Series, 2020

This study has made a calculation of Actinide Minor Addition on Uranium Plutonium Nitride Fuel fo... more This study has made a calculation of Actinide Minor Addition on Uranium Plutonium Nitride Fuel for Modular Gas Cooled Fast Reactor. The purpose of this study was to determine the characteristics of minor actinides when incorporated into plutonium nitride uranium fuel. Neutronics calculation was design by using SRAC Code version 2006 (Standard Reactor Analysis Code) with the data nuclides from JENDL-4.0 under the Linux Operating System with nuclear data library JENDL4.0. Neutronic calculations were done through some steps of parameter survey to obtain the ultimate optimization results. The initial calculation was calculation of fuel cell (PIJ calculation) by using hexagonal cell and then followed by calculation of core reactor (CITATION calculation) by using program code SRAC2006. The addition of minor actinide has been calculated in this research. The minor actinides which were added were americium (Am-241 and Am-243) and neptunium 237. They were put into the uranium plutonium nitri...

Research paper thumbnail of Neutronic analysis of comparation UN-PuN fuel and ThN fuel for 300MWth Gas Cooled Fast Reactor long life without refueling

Journal of Physics: Conference Series, 2020

Neutronic analysis of comparation UN-PuN fuel and ThN fuel for 300MWth Gas Cooled Fast Reactor lo... more Neutronic analysis of comparation UN-PuN fuel and ThN fuel for 300MWth Gas Cooled Fast Reactor long life without refueling has been done. Gas Cooled Fast Reactor is a Generation IV reactor with gas coolant (i.e. helium) and using fast spectrum neutron. The neutronic calculation was carried out using SRAC (Standard Reactor Analysis Code) version 2006 under the Linux Operating System with nuclear data library JENDL4.0. The first calculation is fuel pin cell calculation (PIJ-method) by using a hexagonal cell and then followed by the calculation of the core reactor (CITATION-method). The calculation of the core reactor used homogeneous and heterogeneous core configuration. The UN-PuN fuel use plutonium as a fissile material and natural uranium as a fertile material and the ThN fuel use U233 as a fissile material and natural thorium as a fertile material. The percentages of fissile material are varied in heterogeneous core configuration. It is used to decrease the peaking power in the ce...