M. Budi Setiawan - Academia.edu (original) (raw)

Papers by M. Budi Setiawan

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Radiation protection and safety documents for routine conditions are required to support the lice... more Radiation protection and safety documents for routine conditions are required to support the licensing requirements for nuclear power plant site. This research is focused in the assessment and analysis of the results of PWR safety study related to the routine release of radioactivity from the SMR subsystems and components of the 100 MWe-type PWR along with its consequences in the site. The core inventory calculation was done using ORIGEN2 software, applying release parameters from the existing analysis and calculation results. The radiological consequences were calculated by the PC-CREAM program package. Environmental and meteorological data were obtained using Arc-GIS and spatial analysis. The Bangka Belitung (Babel) site was used as the specific footprint. Analyzing PC-CREAM output data the radiological consequences of routine operation of 3 100 MWe PWR modules on Sebagin site (South Bangka) and Muntok site (West Bangka) in 16 sectors and within a radius of 20 km were concluded. ...

Journal of Physics: Conference Series, 2019

Program of Reaktor Daya Eksperimental (RDE) construction in Indonesia is still on progress and it... more Program of Reaktor Daya Eksperimental (RDE) construction in Indonesia is still on progress and it is continuously supported by research and development activities. RDE will be built based on pebble-bed type of High Temperature Gas-cooled Reactor utilizing UO2 fuel microkernel dispersed in a 6-diameter spherical solid graphite matrix. Graphite is the major composition of the fuel pebble. In addition to the fuel pebbles, there are also dummy pebbles in the RDE core which are fully composed of graphite without fuel kernel inside. During irradiation in the core, there happen activation of graphite due to neutron captured reaction. In this study, the activation of graphite was investigated through the simulation of ORIGEN2.1 code. The graphite matrix in the pebble fuel was irradiated for five cycles and the dummy pebble was irradiated only for one cycle. From the ORIGEN2.1 simulation, the activation of graphite matrix produces many isotopes of light nuclei but the isotopes that have significant half-life and activity are only H-3, Be-10, and C-14. The activities of H-3, Be-10, and C-14 inside the graphite matrix of one fuel pebble are 3.98E-08 Ci, 2.45E-08 Ci, and 8.99E-07 Ci, respectively. The results for the activation of graphite of one dummy pebble for the same isotopes are 3.48E-10 Ci, 5.13E-09 Ci, and 1.89E-07 Ci, respectively. These isotopes deposit in the graphite matrix and might be released into the primary coolant through some mechanisms such as pebble crack, graphite corrosion, and graphite abrasion due to the friction during the pebble shuffling in the core. However, since the activity of isotopes is small, it can be stated that the fuel pebble of RDE is safe.

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

One of the barriers on the implementation of nuclear energy in Indonesia is public perception tow... more One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available...

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

High Temperature Gas-cooled Reactor (HTGR) design has an improved safety which depends on its TRI... more High Temperature Gas-cooled Reactor (HTGR) design has an improved safety which depends on its TRISO coated fuel particles that are considered not to be damaged even in accident condition. However, the radiological impacts from accident condition in HTGR is still important to be assessed. This research is aimed to perform radiological impacts assessment of two postulated accidents of HTGR, which are depressurization and water ingress accident. As a case study, a 10-MWTh pebble-bed HTGR design named Reaktor Daya Eksperimental with the planned site located in Serpong Nuclear Area was chosen. The source terms from the accident conditions were estimated using mechanistic source term model and the dose consequences were calculated using PC-COSYMA. The input data for PC COSYMA, which are meteorological, population distribution, agricultural and local farm data, were compiled based on the site data of Serpong Nuclear Area. The radiological impacts were assessed based on individual and colle...

THE 4TH INTERNATIONAL CONFERENCE ON NUCLEAR ENERGY TECHNOLOGIES AND SCIENCES (ICoNETS) 2021

Preliminary Study on Monitoring Radiation Release from Multipurpose Research Reactor GA Siwabessy... more Preliminary Study on Monitoring Radiation Release from Multipurpose Research Reactor GA Siwabessy at Serpong Nuclear Facility Area is performed in order to evaluate the use of Atmospheric Dispersion formula for predicting gamma radiation dose released from a certain nuclear facility to environment during normal operation. The release from reactor were recorded by four radiation monitoring stations available at the distance of 110 m, 135 m, 150 m and 544 m from MPS-GAS Serpong. The purpose of this study is to evaluate the how Gaussian Dispersion Model can describe the recorded gamma radiation coming from Multipurpose Reactor GA Siwabessy which are close to radiation monitoring stations for modelling purpose. The study is started by identifying the strength of source of Argon-41 and Iodine-131, identifying and differentiating measured gross gamma radiation dose rate during operation and shutdown recorded by each monitoring stations using student distribution test or t-test, regression model and compare the result with the Gaussian theoretical model. The study showed that Gaussian Dispersion Model must be modified for short distance prediction by considering building effect surround the nuclear facility.

THE 4TH INTERNATIONAL CONFERENCE ON NUCLEAR ENERGY TECHNOLOGIES AND SCIENCES (ICoNETS) 2021

THE 4TH INTERNATIONAL CONFERENCE ON MATERIALS AND METALLURGICAL ENGINEERING AND TECHNOLOGY (ICOMMET) 2020, 2021

The High Temperature Gas-Cooled Reactor (HTGR) requires the use of the material as a core structu... more The High Temperature Gas-Cooled Reactor (HTGR) requires the use of the material as a core structure which is resistant to operating temperature reaching more than 500°C. The fine-grained isotropic graphites of IG-110 is a graphite material that has been applied to existing HTGR. The reactor core structure has a function to ensure reactor coolant flow, control rod movement, and reactor fuel conditions. In the implementation of reactor component inspections, various nondestructive test methods have been developed which aim to find cracks that are the initial failure of components. Besides, a non-destructive test method was also developed to measure the stress acting on the reactor components in order to determine the aging process. Currently, stress measurement is generally done indirectly using strain gage. The purpose of this study is to determine the effect of the load on ultrasonic wave velocity. Ultrasonic testing is carried out by using a portable ultrasonic flaw detector, using two types of waves, longitudinal wave and transverse wave, with a frequency of 5MHz. The stress acting on graphite material is compressive. The load was varied at several levels. The test results show that compressive stress in graphite material reduces ultrasonic wave propagation, for both types of waves used. The higher the compressive stress in the material, the more significant the decrease in ultrasonic wave propagation. The testing system used in this study has a level of measurement accuracy when the load value is higher than 20MPa.

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

The design process of Experimental Power Reactor (Reaktor Daya Eksperimental/RDE) has been carrie... more The design process of Experimental Power Reactor (Reaktor Daya Eksperimental/RDE) has been carried out by BATAN for the last five years, adopting HTGR-type reactor with thermal power of 10 MW. RDE is designed with the reference of similar reactor, namely HTR-10. During this process, source term estimation is required to prove the safety of RDE design, as well as to fulfill the concept of As Low As Reasonably Achievable (ALARA) in radiation protection. The source term is affected by the magnitude of the radioactive substances released from the reactor core due to an accident. Conservative accident postulations on the RDE are water ingress and depressurization accidents. Based on these postulations, source term estimation was performed. It follows the mechanistic source term flow, with conservative assumptions for the radioactive release of fuel into the coolant, reactor building, and finally discharged into the environment. Assumptions for the calculation are taken from conservative ...

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Re... more One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach a...

Sensors

An oscillating piezoresistive microcantilever (MC) coated with an aluminum (Al)-doped zinc oxide ... more An oscillating piezoresistive microcantilever (MC) coated with an aluminum (Al)-doped zinc oxide (ZnO) nanorods was used to detect carbon monoxide (CO) in air at room temperature. Al-doped ZnO nanorods were grown on the MC surface using the hydrothermal method, and a response to CO gas was observed by measuring a resonant frequency shift of vibrated MC. CO gas response showed a significant increase in resonant frequency, where sensitivity in the order of picogram amounts was obtained. An increase in resonant frequency was also observed with increasing gas flow rate, which was simultaneously followed by a decrease in relative humidity, indicating that the molecular interface between ZnO and H2O plays a key role in CO absorption. The detection of other gases of carbon compounds such as CO2 and CH4 was also performed; the sensitivity of CO was found to be higher than those gases. The results demonstrate the reversibility and reproducibility of the proposed technique, opening up future ...

Jurnal Sains Materi Indonesia

MECHANICAL PROPERTIES PREDICTION OF IG-110 GRAPHITE BY NON DESTRUCTIVE INSPECTION USING ULTRASONI... more MECHANICAL PROPERTIES PREDICTION OF IG-110 GRAPHITE BY NON DESTRUCTIVE INSPECTION USING ULTRASONIC METHOD. The core structure of high temperature gas cooled reactor is the most important part of the reactor which its integrity must be ensured during operation stage. The structure of reactor core must ensure the position of fuel to be kept in its position, to ensure the control rods can get into the guiding canal, and to ensure the flow of the gas coolant. One of the stressor of the graphite material degradation is neutron exposure. The impact of neutron exposure is the change in mechanical properties such as modulus of elasticity. In order to ensure the integrity of the materials, an in-service non-destructive inspection is implemented. The aim of this study is to develop non-destructive inspection method in order to predict mechanical properties of graphite materials. Inspections were done using Ultrasonic Flaw Detector with 35×35×55 [mm] block-shaped specimens made of graphite IG-110. Two types of transducer were used to generate longitudinal and transversal waves with the same frequency of 5 MHz. Two mechanical properties were predicted, that are isotropic characteristic and the modulus of elasticity. The predicted value of the modulus of elasticity was verified by conducting compressive tests using 10×10×10 [mm] cube specimens. According to the ultrasonic propagation velocities resulted from ultrasonic inspection results showed that the graphite IG-110 is an isotropic material. From the calculation of the modulus of elasticity based on measurement results of transversal and longitudinal waves propagation, IG-110 graphite has a value of modulus of elasticity of 9.1 GPa. Compared to the modulus of elasticity measured from compressive test, this value was 10% lower. It can be concluded that the ultrasonic non-destructive inspection can be used to predict mechanical properties of the IG-110 graphite.

Distribusi - Journal of Management and Business

Tujuan dari penelitian ini adalah untuk mengetahui tingkat pengembalian dana Simpan Pinjam Kelomp... more Tujuan dari penelitian ini adalah untuk mengetahui tingkat pengembalian dana Simpan Pinjam Kelompok Perempuan (SPP) di Kabupaten Sumbawa dan menganalisis strategi Inovasi pengelolaan dana Simpan Pinjam Kelompok Perempuan (SPP) di Kabupaten Sumbawa.Metode yang digunakan dalam penelitian ini adalah metode kuantitatif yang menggunakan pendekatan Ex post Facto di lima Kecamatan yaitu : Kecamatan Alas Barat, Kecamatan Utan, Kecamatan Labuhan Badas, Kecamatan Lopok, dan Kecamatan Lape. Pengumpulan data dengan metode observasi, kuisioner, dokumentasi dan wawancara.Hasil penelitian ini menunjukkan bahwa dengan melakukan Inovasi Pengelolaan Dana Simpan Pinjam Kelompok Perempuan yang semula tingkat kolektibilitas diatas 90% menjadi 43% ini berarti turun sekitar 47% sehingga terjadi peningkatan dalam pengembalian pinjaman. Sehingga mempunyai hasil yang signifikan terhadap peningkatan kolektibilitas.Kesimpulan dari penelitian ini bahwa ada beberapa penyebab terjadinya kredit macet diantaranya: ...

Malaysian Journal of Fundamental and Applied Sciences

The evaluation of RSG-GAS research reactor for transmutation reactor was proposed to study its ef... more The evaluation of RSG-GAS research reactor for transmutation reactor was proposed to study its effectiveness to transmute minor actinides (MA), specifically Am-241, to support geologic storage/disposal. The Am-241 radionuclide was assumed to be discharged from 1000MWe PWR’s spent fuel. The mass of Am-241 discharged from within a year operation of 1000MWe PWR was 1.65E+03 gram, while the optimum Am-241 mass which can be transmuted in RSG-GAS - and still meet the safety requirements of reactivity - was 8.0E+03 gram. This was equivalent to about cumulative Am-241 discharged from 5 units of 1000MWe PWR. In 10 cycles of RSG-GAS operation (about 2 years), the remaining of Am-241 is only about 100 grams. The ratio of Am-241 transmuted (8.0E+03 gram) and Am-241 produced in the RSG-GAS core (1.98E-02 gram) within 1-year operation shows the effectiveness of RSG-GAS as a transmutation reactor.

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

FLEXBLUE reactor is France's small modular power reactor type of pressurized water reactor with 1... more FLEXBLUE reactor is France's small modular power reactor type of pressurized water reactor with 160 MWe. This reactor is based on ocean site which is suitable to be built in archipelagoes like Indonesia. The purpose of the present work is to evaluate radioactivity inventory in FLEXBLUE reactor core. The analysis was carried out by calculations using the ORIGEN-2 program. Calculations were carried out based on pin cell calculations, where in the pin cell there were 4 regions namely UO2 fuel region with enrichment of 4.95%, region 2 vacuum area containing He, region 3 in the form of cladding Zr-4 and Region 4 are extra regions consisting of homogenization of Gd2O3 and H2O. The results obtained were radionuclide inventory in a reactor consisting of 8 radionuclide groups namely Tritium, Noble gas, Halogen, Alkali metal, Tellurium, Strontium and Barium, Nobel metal, Lanthanide and Cerium. The biggest radionuclide activity was in the Halogen group namely nuclei I-134 of 1.12E+18 Bq. In addition, it also obtained activity from large and dangerous gamma transmitter nuclides for the body, namely I-131 and Cs-137, each of which had activities of 4.92E+17 Bq and 2.72E+16 Bq.

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with n... more Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with nominal power of 30 MW and currently utilized for material irradiation and other research purposes. When a sample material is put in the core of RSG-GAS, there will be some amount of nuclear heating generated in the sample material induced by interaction of gamma rays with the sample material. Evaluating the nuclear heating is one of the important aspects regarding the safety of reactor operation and the safety of the sample material itself. In this work, the nuclear heating of several sample materials commonly irradiated in the RSG-GAS core were evaluated using GAMSET code. The sample materials taken as the case study is sample for radioisotope production (TeO2, MoO3, UO2, Sm2O3, Yb2O3, Zn, S), sample for research purpose (C, AlMg3, Hg), topaz, and sample for cladding material (Al, Zr, Fe, SS304L). The sample materials were irradiated in 3 positions in the core, i.e. E6, D9, and B1, for 5 days with thermal power of 15 MW. From the results of nuclear heating calculation, it was found that the nuclear heating generated in sample material in certain position is greatly determined by the type of core structure that surrounding the material position. The difference of nuclear heating generated in the position of D9 has a higher amount of 5% compared to the nuclear heating generated in the position of E6, while for the position of B1 the amount of nuclear heating generated is much lower. Among all the material samples, UO2 has the highest nuclear heating since it contains fissile material, white for the other sample material the amount of nuclear heating varied between between 3 ~ 11 watts/gram in the E6 and D9 position, and between 0.4 ~ 1.4 watts/gram in the B1 position. The results of nuclear heating obtained in this work can be used as a database for the purpose of evaluating the safety of reactor operation and sample material irradiated in RSG-GAS. The data of the nuclear heating in this work can also be used to complement the RSG-GAS safety analysis report.

Journal of Physics: Conference Series

A new fuel management pattern has been applied to the RSG-GAS Research reactor operation. The new... more A new fuel management pattern has been applied to the RSG-GAS Research reactor operation. The new fuel management pattern is done so that the replacement of fuel in every operating cycle is more efficient. The new fuel replacement pattern required various safety analyzes, among others related to the fuel fraction and radioactivity inventory of the RSG-GAS. Both of those are the main elements in dose acceptance analysis for workers and communities around the reactor when the reactor operates both in normal or in abnormal conditions. Fuel burn-up and inventory analysis was performed using ORIGEN2 computer program. Inputs for ORIGEN2 are the fuel mass, the time required for one operating cycle as well as the peak power factor in each fuel in the specified fuel management pattern. The result for the 97 th Core configuration (T97) is that the average burn-up fraction of each cycle is 6.79%, and the maximus fuel fraction is 52.36% for Fuel Elemet and 56.52% for Contro Elemet. It is also obtained that the largest and dangerous human inventory activity at the end of cycle (EOC) for Iodine radionuclide group is I-131 for 5.18E+04 Ci, and for the alkali metal radionuclide group Cs-137 of 7.65E+03 Ci.

Journal of Physics: Conference Series

BATAN priority activity supported by the Center for Nuclear Reactor Technology and Safety is the ... more BATAN priority activity supported by the Center for Nuclear Reactor Technology and Safety is the conceptual design documents and evaluation of experimental power reactor (RDE). Based on the design established, the radiation safety involving the dispersion of radioactive into the site and environment shall be calculated. The objective of this study is to obtain the radioactivity impact of RDE due to hypothetical accident in Serpong II Nuclear Area (KNS-II) in Puspiptek Area. Postulated hypothetical accidents are water ingress and depressurization accident. The sourceterm input data taken are based on HTR-10 hypothetical accident. Meteorological and environmental data are taken from available data of KNS Serpong. The calculation is carried out using PC-Cosyma code. The highest radioactivity of air dispersion for Kr-87due to depressurization accident is 1.96E+04 Bq/m 3 and due to water ingress accident is 1.96E+04 Bq/m 3. The highest radioactivity of surface deposition due to depressurization accident for Cs-137 is 1.51E+03 Bq/m 2 and due to water ingress accident for I-131 is 1.5E+01 Bq/m 2. The impact of RDE radioactivity for hypothetical accidents on the KNS-II site shows lower than BAPETEN regulatory requirements.

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Radiation protection and safety documents for routine conditions are required to support the lice... more Radiation protection and safety documents for routine conditions are required to support the licensing requirements for nuclear power plant site. This research is focused in the assessment and analysis of the results of PWR safety study related to the routine release of radioactivity from the SMR subsystems and components of the 100 MWe-type PWR along with its consequences in the site. The core inventory calculation was done using ORIGEN2 software, applying release parameters from the existing analysis and calculation results. The radiological consequences were calculated by the PC-CREAM program package. Environmental and meteorological data were obtained using Arc-GIS and spatial analysis. The Bangka Belitung (Babel) site was used as the specific footprint. Analyzing PC-CREAM output data the radiological consequences of routine operation of 3 100 MWe PWR modules on Sebagin site (South Bangka) and Muntok site (West Bangka) in 16 sectors and within a radius of 20 km were concluded. ...

Journal of Physics: Conference Series, 2019

Program of Reaktor Daya Eksperimental (RDE) construction in Indonesia is still on progress and it... more Program of Reaktor Daya Eksperimental (RDE) construction in Indonesia is still on progress and it is continuously supported by research and development activities. RDE will be built based on pebble-bed type of High Temperature Gas-cooled Reactor utilizing UO2 fuel microkernel dispersed in a 6-diameter spherical solid graphite matrix. Graphite is the major composition of the fuel pebble. In addition to the fuel pebbles, there are also dummy pebbles in the RDE core which are fully composed of graphite without fuel kernel inside. During irradiation in the core, there happen activation of graphite due to neutron captured reaction. In this study, the activation of graphite was investigated through the simulation of ORIGEN2.1 code. The graphite matrix in the pebble fuel was irradiated for five cycles and the dummy pebble was irradiated only for one cycle. From the ORIGEN2.1 simulation, the activation of graphite matrix produces many isotopes of light nuclei but the isotopes that have significant half-life and activity are only H-3, Be-10, and C-14. The activities of H-3, Be-10, and C-14 inside the graphite matrix of one fuel pebble are 3.98E-08 Ci, 2.45E-08 Ci, and 8.99E-07 Ci, respectively. The results for the activation of graphite of one dummy pebble for the same isotopes are 3.48E-10 Ci, 5.13E-09 Ci, and 1.89E-07 Ci, respectively. These isotopes deposit in the graphite matrix and might be released into the primary coolant through some mechanisms such as pebble crack, graphite corrosion, and graphite abrasion due to the friction during the pebble shuffling in the core. However, since the activity of isotopes is small, it can be stated that the fuel pebble of RDE is safe.

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

One of the barriers on the implementation of nuclear energy in Indonesia is public perception tow... more One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA. Due to the limitations of available...

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

High Temperature Gas-cooled Reactor (HTGR) design has an improved safety which depends on its TRI... more High Temperature Gas-cooled Reactor (HTGR) design has an improved safety which depends on its TRISO coated fuel particles that are considered not to be damaged even in accident condition. However, the radiological impacts from accident condition in HTGR is still important to be assessed. This research is aimed to perform radiological impacts assessment of two postulated accidents of HTGR, which are depressurization and water ingress accident. As a case study, a 10-MWTh pebble-bed HTGR design named Reaktor Daya Eksperimental with the planned site located in Serpong Nuclear Area was chosen. The source terms from the accident conditions were estimated using mechanistic source term model and the dose consequences were calculated using PC-COSYMA. The input data for PC COSYMA, which are meteorological, population distribution, agricultural and local farm data, were compiled based on the site data of Serpong Nuclear Area. The radiological impacts were assessed based on individual and colle...

THE 4TH INTERNATIONAL CONFERENCE ON NUCLEAR ENERGY TECHNOLOGIES AND SCIENCES (ICoNETS) 2021

Preliminary Study on Monitoring Radiation Release from Multipurpose Research Reactor GA Siwabessy... more Preliminary Study on Monitoring Radiation Release from Multipurpose Research Reactor GA Siwabessy at Serpong Nuclear Facility Area is performed in order to evaluate the use of Atmospheric Dispersion formula for predicting gamma radiation dose released from a certain nuclear facility to environment during normal operation. The release from reactor were recorded by four radiation monitoring stations available at the distance of 110 m, 135 m, 150 m and 544 m from MPS-GAS Serpong. The purpose of this study is to evaluate the how Gaussian Dispersion Model can describe the recorded gamma radiation coming from Multipurpose Reactor GA Siwabessy which are close to radiation monitoring stations for modelling purpose. The study is started by identifying the strength of source of Argon-41 and Iodine-131, identifying and differentiating measured gross gamma radiation dose rate during operation and shutdown recorded by each monitoring stations using student distribution test or t-test, regression model and compare the result with the Gaussian theoretical model. The study showed that Gaussian Dispersion Model must be modified for short distance prediction by considering building effect surround the nuclear facility.

THE 4TH INTERNATIONAL CONFERENCE ON NUCLEAR ENERGY TECHNOLOGIES AND SCIENCES (ICoNETS) 2021

THE 4TH INTERNATIONAL CONFERENCE ON MATERIALS AND METALLURGICAL ENGINEERING AND TECHNOLOGY (ICOMMET) 2020, 2021

The High Temperature Gas-Cooled Reactor (HTGR) requires the use of the material as a core structu... more The High Temperature Gas-Cooled Reactor (HTGR) requires the use of the material as a core structure which is resistant to operating temperature reaching more than 500°C. The fine-grained isotropic graphites of IG-110 is a graphite material that has been applied to existing HTGR. The reactor core structure has a function to ensure reactor coolant flow, control rod movement, and reactor fuel conditions. In the implementation of reactor component inspections, various nondestructive test methods have been developed which aim to find cracks that are the initial failure of components. Besides, a non-destructive test method was also developed to measure the stress acting on the reactor components in order to determine the aging process. Currently, stress measurement is generally done indirectly using strain gage. The purpose of this study is to determine the effect of the load on ultrasonic wave velocity. Ultrasonic testing is carried out by using a portable ultrasonic flaw detector, using two types of waves, longitudinal wave and transverse wave, with a frequency of 5MHz. The stress acting on graphite material is compressive. The load was varied at several levels. The test results show that compressive stress in graphite material reduces ultrasonic wave propagation, for both types of waves used. The higher the compressive stress in the material, the more significant the decrease in ultrasonic wave propagation. The testing system used in this study has a level of measurement accuracy when the load value is higher than 20MPa.

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

The design process of Experimental Power Reactor (Reaktor Daya Eksperimental/RDE) has been carrie... more The design process of Experimental Power Reactor (Reaktor Daya Eksperimental/RDE) has been carried out by BATAN for the last five years, adopting HTGR-type reactor with thermal power of 10 MW. RDE is designed with the reference of similar reactor, namely HTR-10. During this process, source term estimation is required to prove the safety of RDE design, as well as to fulfill the concept of As Low As Reasonably Achievable (ALARA) in radiation protection. The source term is affected by the magnitude of the radioactive substances released from the reactor core due to an accident. Conservative accident postulations on the RDE are water ingress and depressurization accidents. Based on these postulations, source term estimation was performed. It follows the mechanistic source term flow, with conservative assumptions for the radioactive release of fuel into the coolant, reactor building, and finally discharged into the environment. Assumptions for the calculation are taken from conservative ...

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Re... more One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach a...

Sensors

An oscillating piezoresistive microcantilever (MC) coated with an aluminum (Al)-doped zinc oxide ... more An oscillating piezoresistive microcantilever (MC) coated with an aluminum (Al)-doped zinc oxide (ZnO) nanorods was used to detect carbon monoxide (CO) in air at room temperature. Al-doped ZnO nanorods were grown on the MC surface using the hydrothermal method, and a response to CO gas was observed by measuring a resonant frequency shift of vibrated MC. CO gas response showed a significant increase in resonant frequency, where sensitivity in the order of picogram amounts was obtained. An increase in resonant frequency was also observed with increasing gas flow rate, which was simultaneously followed by a decrease in relative humidity, indicating that the molecular interface between ZnO and H2O plays a key role in CO absorption. The detection of other gases of carbon compounds such as CO2 and CH4 was also performed; the sensitivity of CO was found to be higher than those gases. The results demonstrate the reversibility and reproducibility of the proposed technique, opening up future ...

Jurnal Sains Materi Indonesia

MECHANICAL PROPERTIES PREDICTION OF IG-110 GRAPHITE BY NON DESTRUCTIVE INSPECTION USING ULTRASONI... more MECHANICAL PROPERTIES PREDICTION OF IG-110 GRAPHITE BY NON DESTRUCTIVE INSPECTION USING ULTRASONIC METHOD. The core structure of high temperature gas cooled reactor is the most important part of the reactor which its integrity must be ensured during operation stage. The structure of reactor core must ensure the position of fuel to be kept in its position, to ensure the control rods can get into the guiding canal, and to ensure the flow of the gas coolant. One of the stressor of the graphite material degradation is neutron exposure. The impact of neutron exposure is the change in mechanical properties such as modulus of elasticity. In order to ensure the integrity of the materials, an in-service non-destructive inspection is implemented. The aim of this study is to develop non-destructive inspection method in order to predict mechanical properties of graphite materials. Inspections were done using Ultrasonic Flaw Detector with 35×35×55 [mm] block-shaped specimens made of graphite IG-110. Two types of transducer were used to generate longitudinal and transversal waves with the same frequency of 5 MHz. Two mechanical properties were predicted, that are isotropic characteristic and the modulus of elasticity. The predicted value of the modulus of elasticity was verified by conducting compressive tests using 10×10×10 [mm] cube specimens. According to the ultrasonic propagation velocities resulted from ultrasonic inspection results showed that the graphite IG-110 is an isotropic material. From the calculation of the modulus of elasticity based on measurement results of transversal and longitudinal waves propagation, IG-110 graphite has a value of modulus of elasticity of 9.1 GPa. Compared to the modulus of elasticity measured from compressive test, this value was 10% lower. It can be concluded that the ultrasonic non-destructive inspection can be used to predict mechanical properties of the IG-110 graphite.

Distribusi - Journal of Management and Business

Tujuan dari penelitian ini adalah untuk mengetahui tingkat pengembalian dana Simpan Pinjam Kelomp... more Tujuan dari penelitian ini adalah untuk mengetahui tingkat pengembalian dana Simpan Pinjam Kelompok Perempuan (SPP) di Kabupaten Sumbawa dan menganalisis strategi Inovasi pengelolaan dana Simpan Pinjam Kelompok Perempuan (SPP) di Kabupaten Sumbawa.Metode yang digunakan dalam penelitian ini adalah metode kuantitatif yang menggunakan pendekatan Ex post Facto di lima Kecamatan yaitu : Kecamatan Alas Barat, Kecamatan Utan, Kecamatan Labuhan Badas, Kecamatan Lopok, dan Kecamatan Lape. Pengumpulan data dengan metode observasi, kuisioner, dokumentasi dan wawancara.Hasil penelitian ini menunjukkan bahwa dengan melakukan Inovasi Pengelolaan Dana Simpan Pinjam Kelompok Perempuan yang semula tingkat kolektibilitas diatas 90% menjadi 43% ini berarti turun sekitar 47% sehingga terjadi peningkatan dalam pengembalian pinjaman. Sehingga mempunyai hasil yang signifikan terhadap peningkatan kolektibilitas.Kesimpulan dari penelitian ini bahwa ada beberapa penyebab terjadinya kredit macet diantaranya: ...

Malaysian Journal of Fundamental and Applied Sciences

The evaluation of RSG-GAS research reactor for transmutation reactor was proposed to study its ef... more The evaluation of RSG-GAS research reactor for transmutation reactor was proposed to study its effectiveness to transmute minor actinides (MA), specifically Am-241, to support geologic storage/disposal. The Am-241 radionuclide was assumed to be discharged from 1000MWe PWR’s spent fuel. The mass of Am-241 discharged from within a year operation of 1000MWe PWR was 1.65E+03 gram, while the optimum Am-241 mass which can be transmuted in RSG-GAS - and still meet the safety requirements of reactivity - was 8.0E+03 gram. This was equivalent to about cumulative Am-241 discharged from 5 units of 1000MWe PWR. In 10 cycles of RSG-GAS operation (about 2 years), the remaining of Am-241 is only about 100 grams. The ratio of Am-241 transmuted (8.0E+03 gram) and Am-241 produced in the RSG-GAS core (1.98E-02 gram) within 1-year operation shows the effectiveness of RSG-GAS as a transmutation reactor.

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

FLEXBLUE reactor is France's small modular power reactor type of pressurized water reactor with 1... more FLEXBLUE reactor is France's small modular power reactor type of pressurized water reactor with 160 MWe. This reactor is based on ocean site which is suitable to be built in archipelagoes like Indonesia. The purpose of the present work is to evaluate radioactivity inventory in FLEXBLUE reactor core. The analysis was carried out by calculations using the ORIGEN-2 program. Calculations were carried out based on pin cell calculations, where in the pin cell there were 4 regions namely UO2 fuel region with enrichment of 4.95%, region 2 vacuum area containing He, region 3 in the form of cladding Zr-4 and Region 4 are extra regions consisting of homogenization of Gd2O3 and H2O. The results obtained were radionuclide inventory in a reactor consisting of 8 radionuclide groups namely Tritium, Noble gas, Halogen, Alkali metal, Tellurium, Strontium and Barium, Nobel metal, Lanthanide and Cerium. The biggest radionuclide activity was in the Halogen group namely nuclei I-134 of 1.12E+18 Bq. In addition, it also obtained activity from large and dangerous gamma transmitter nuclides for the body, namely I-131 and Cs-137, each of which had activities of 4.92E+17 Bq and 2.72E+16 Bq.

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

THE 4TH BIOMEDICAL ENGINEERING’S RECENT PROGRESS IN BIOMATERIALS, DRUGS DEVELOPMENT, HEALTH, AND MEDICAL DEVICES: Proceedings of the International Symposium of Biomedical Engineering (ISBE) 2019

Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with n... more Reaktor Serba Guna GA Siwabessy (RSG-GAS) is a multipurpose Material Testing Reactor (MTR) with nominal power of 30 MW and currently utilized for material irradiation and other research purposes. When a sample material is put in the core of RSG-GAS, there will be some amount of nuclear heating generated in the sample material induced by interaction of gamma rays with the sample material. Evaluating the nuclear heating is one of the important aspects regarding the safety of reactor operation and the safety of the sample material itself. In this work, the nuclear heating of several sample materials commonly irradiated in the RSG-GAS core were evaluated using GAMSET code. The sample materials taken as the case study is sample for radioisotope production (TeO2, MoO3, UO2, Sm2O3, Yb2O3, Zn, S), sample for research purpose (C, AlMg3, Hg), topaz, and sample for cladding material (Al, Zr, Fe, SS304L). The sample materials were irradiated in 3 positions in the core, i.e. E6, D9, and B1, for 5 days with thermal power of 15 MW. From the results of nuclear heating calculation, it was found that the nuclear heating generated in sample material in certain position is greatly determined by the type of core structure that surrounding the material position. The difference of nuclear heating generated in the position of D9 has a higher amount of 5% compared to the nuclear heating generated in the position of E6, while for the position of B1 the amount of nuclear heating generated is much lower. Among all the material samples, UO2 has the highest nuclear heating since it contains fissile material, white for the other sample material the amount of nuclear heating varied between between 3 ~ 11 watts/gram in the E6 and D9 position, and between 0.4 ~ 1.4 watts/gram in the B1 position. The results of nuclear heating obtained in this work can be used as a database for the purpose of evaluating the safety of reactor operation and sample material irradiated in RSG-GAS. The data of the nuclear heating in this work can also be used to complement the RSG-GAS safety analysis report.

Journal of Physics: Conference Series

A new fuel management pattern has been applied to the RSG-GAS Research reactor operation. The new... more A new fuel management pattern has been applied to the RSG-GAS Research reactor operation. The new fuel management pattern is done so that the replacement of fuel in every operating cycle is more efficient. The new fuel replacement pattern required various safety analyzes, among others related to the fuel fraction and radioactivity inventory of the RSG-GAS. Both of those are the main elements in dose acceptance analysis for workers and communities around the reactor when the reactor operates both in normal or in abnormal conditions. Fuel burn-up and inventory analysis was performed using ORIGEN2 computer program. Inputs for ORIGEN2 are the fuel mass, the time required for one operating cycle as well as the peak power factor in each fuel in the specified fuel management pattern. The result for the 97 th Core configuration (T97) is that the average burn-up fraction of each cycle is 6.79%, and the maximus fuel fraction is 52.36% for Fuel Elemet and 56.52% for Contro Elemet. It is also obtained that the largest and dangerous human inventory activity at the end of cycle (EOC) for Iodine radionuclide group is I-131 for 5.18E+04 Ci, and for the alkali metal radionuclide group Cs-137 of 7.65E+03 Ci.

Journal of Physics: Conference Series

BATAN priority activity supported by the Center for Nuclear Reactor Technology and Safety is the ... more BATAN priority activity supported by the Center for Nuclear Reactor Technology and Safety is the conceptual design documents and evaluation of experimental power reactor (RDE). Based on the design established, the radiation safety involving the dispersion of radioactive into the site and environment shall be calculated. The objective of this study is to obtain the radioactivity impact of RDE due to hypothetical accident in Serpong II Nuclear Area (KNS-II) in Puspiptek Area. Postulated hypothetical accidents are water ingress and depressurization accident. The sourceterm input data taken are based on HTR-10 hypothetical accident. Meteorological and environmental data are taken from available data of KNS Serpong. The calculation is carried out using PC-Cosyma code. The highest radioactivity of air dispersion for Kr-87due to depressurization accident is 1.96E+04 Bq/m 3 and due to water ingress accident is 1.96E+04 Bq/m 3. The highest radioactivity of surface deposition due to depressurization accident for Cs-137 is 1.51E+03 Bq/m 2 and due to water ingress accident for I-131 is 1.5E+01 Bq/m 2. The impact of RDE radioactivity for hypothetical accidents on the KNS-II site shows lower than BAPETEN regulatory requirements.