Ihda Husnayani - Academia.edu (original) (raw)

Papers by Ihda Husnayani

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Indonesia possesses numerous potential sites for nuclear power plant development. A fast and comp... more Indonesia possesses numerous potential sites for nuclear power plant development. A fast and comprehensive radiological consequences analysis is required to conduct a preliminary analysis of radionuclide release into the atmosphere, including source terms estimation. One simplified method for such estimation is the use of the Relative Volatility approach by Kess and Booth, published in IAEA TECDOC 1127. The objective of this study was to evaluate the use of a simple and comprehensive tool for estimating the source terms of planned nuclear power plants to facilitate the analysis of radiological consequences during site evaluation. Input parameters for the estimation include fuel burn-up, blow-down time, specific heat transfer of fuel to cladding, and coolant debit, using 100 MWe PWR as a case study. The results indicate a slight difference in the calculated release fraction compared to previous calculations, indicating a need to modify Keywords: Source terms, Relative volatility, Rel...

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Silicon carbide (SiC) is a competitive candidate material to be used in several advanced and Gene... more Silicon carbide (SiC) is a competitive candidate material to be used in several advanced and Generation-IV nuclear reactor designs as neutron moderator, fuel coating, cladding, or core structural material. Many studies have been performed to investigate the durability of SiC in severe environment in nuclear reactor. However, the nature and behavior of defect induced by neutron irradiation are still not fully understood. This paper is aimed to study collision cascade and primary radiation damage in SiC using molecular dynamics simulation. The potential being used was a hybrid Tersoff potential modified with Ziegler-Biersack-Littmark (ZBL) screening function. The collision cascade was let evolved for 10 ps from a Si or C primary knocked atom (PKA) located initially at the top center of a system containing 960000 atoms. The simulation was carried out at room temperature as well as at several advanced fission reactor-relevant temperatures. It was obtained that the number of C point defe...

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 ...

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...

Symmetry Integrability and Geometry-methods and Applications, Jul 14, 2016

CALCULATION OF RADIONUCLIDE CONTENT OF NUCLEAR MATERIALS USING ORIGEN2.1 COMPUTER CODE. Nuclear m... more CALCULATION OF RADIONUCLIDE CONTENT OF NUCLEAR MATERIALS USING ORIGEN2.1 COMPUTER CODE. Nuclear materials contain a number of radionuclides produced from radioactive decay process. The composition of these radionuclides which are accumulated inside the nuclear materials changes over the time. The calculation of radionuclide composition inside nuclear materials is very important especially in the aspect of nuclear reactor safety evaluation, nuclear fuel behavior evaluation, and radioactive waste management. One method to calculate radionuclide content of nuclear materials is by using ORIGEN2.1 computer code. Beside radionuclide composition, this code can also calculate some characteristics related to decay process such as total radioactivity, decay heat, and neutron flux. This paper is a literature study about ORIGEN2.1 computer code. A brief description of ORIGEN2.1 and its use for calculating radionuclide content of nuclear materials are presented. Radionuclide content produced from californium-252 decay was chosen as a simple case solved by ORIGEN2.1. Californium-252 was simulated to undergo decay for 10 years. The variables which are calculated by ORIGEN2.1 in this case are radionuclide composition, total radioactivity, total alpha radioactivity, and neutron flux. From the results of this simulation, it is shown that small amount of californium-252 produces high neutron intensity so that it can be used as a reliable neutron source for many applications.

Journal of Physics: Conference Series, 2021

High Temperature Gas-cooled Reactor (HTGR) is one of the Generation-IV reactors that has many adv... more High Temperature Gas-cooled Reactor (HTGR) is one of the Generation-IV reactors that has many advantages and very promising to be utilized in the future. National Nuclear Energy Agency of Indonesia (BATAN) has a program to build a non-commercial power reactor (RDE) that will be based on the pebble-bed type of HTGR. To support this program, the research and development activities in HTGR have been conducting recently. One of the important aspects to be studied is accident analysis to assess the behavior of HTGR during the accident and to anticipate the consequences that might be caused. The purpose of this work is to study the Depressurized Loss of Force Coolant (DLOFC) accident on HTGR by using Graphite Reactor Severe Accident Code (GRSAC). The type of HTGR used as a case study is PBMR-400. This work was begun with the preparation of GRSAC input for reactor parameter including reactor design, core layout design, primary cooling loop design, graphite properties and material propertie...

FISSION PRODUCTS INVENTORY ANALYSIS OF HTGR FUEL BY USING ORIGEN2.1 COMPUTER CODE. High Temperatu... more FISSION PRODUCTS INVENTORY ANALYSIS OF HTGR FUEL BY USING ORIGEN2.1 COMPUTER CODE. High Temperature Gas-Cooled Reactor (HTGR) is one type of reactor that offers many advantages, one of which is its TRISO-coated fuel resistance to high temperature. Most of the discussions about HTGR safety are focused on the TRISO- layers reliability in containing fission products and preventing them to release out of the fuel particles. One important step before determining the number of fission products that release to the environment is the analysis of fission products inventory inside the fuel particles. This paper presents an analysis of fission products inventory inside the HTGR fuel particles by using ORIGEN2.1 computer code. The calculation of fission products inventory was conducted by using neutron cross section library data for HTGR type and then the results of this calculation were compared to the results of calculation by using neutron cross section library data for PWR type. The results...

DOSES ANALYSIS OF A HYPOTHETICAL LOCA ACCIDENT ON NUCLEAR POWER PLANT (NPPs) SITTING. Lessons lea... more DOSES ANALYSIS OF A HYPOTHETICAL LOCA ACCIDENT ON NUCLEAR POWER PLANT (NPPs) SITTING. Lessons learn from the Fukushima accident, IAEA and national regulatory agencies BAPETEN requires nuclear emergency documents with the postulation of accidents involving the release of radioactive substances into the environment. The document were for existing nuclear installations and installations are planned to be built. The study on the impact of reactor accidents for the situation in Indonesia has been done, but not comprehensively assesses the radiation dose on environment. This research will be presented in the analysis of radiation dose due to design bases accident (DBA) for PWR power reactor with a capacity of 1000 MWe are planned to be built. The research goal was to analysis of radiation dose as short-term effective dose by distance, nuclide, pathway, and by organ; long-term effective dose by distance, nuclide, pathway, and by organ; and collective dose by pathway and distance. Radiation...

Jurnal Sains dan Teknologi Nuklir Indonesia, 2020

also be considered for the simplicity it offers and thus potentially lower cost. Due to different... more also be considered for the simplicity it offers and thus potentially lower cost. Due to different neutronic and burnup profile between the two, the resulting spent fuel characteristic is also different and possibly requires different handling mechanism. This paper assesses the characteristics of OTTO-scheme RDE spent fuel using thorium fuel cycle to provide preliminary data and insight for its spent fuel management. The assessment is performed employing ORIGEN2.1 code. At day 30 of cooling after determined end-of-cycle (EOC), each spent fuel yields 234.9 Curies of radioactivity, emitting 66.26 neutrons/second, 1x10 13 photons/second, and releasing 0.7675 watts of decay heat. These numbers must be taken into consideration regarding spent fuel management and spent fuel cask design. Tl-208 isotope characteristics, whose existence is unique to thorium fuel cycle, were also determined. It is found to be yielding 3.42x10-3 Curie of radioactivity and releasing 1.2x10 8 photons/second at its peak. Understanding its high-energy gamma release, proper radiation protection mechanism must be implemented.

Malaysian Journal of Fundamental and Applied Sciences, 2019

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.

Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir, 2019

CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL. There ar... more CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL. There are several options of nuclear fuel utilisation in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilises low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilise other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fuelled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0,627 Watts of heat, 28 neutrons/s, 8.28x1012 photons/s and yiel...

Journal of Physics: Conference Series, 2018

Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a mod... more Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a moderator and helium as a coolant. In a multi-pass PBR, burnup of the fuel pebble must be measured in each cycle by online measurement in order to determine whether the fuel pebble should be reloaded into the core for another cycle or moved out of the core into spent fuel storage. One of the well-known methods for measuring burnup is based on the activity of radionuclide decay inside the fuel pebble. In this work, the activity and gamma emission of Kr-85m were studied in order to investigate the feasibility of Kr-85m as burnup measurement indicator in a PBR. The activity and gamma emission of Kr-85 were estimated using ORIGEN2.1 computer code. The parameters of HTR-10 were taken as a case study in performing ORIGEN2.1 simulation. The results show that the activity revolution of Kr-85m has a good relationship with the burnup of the pebble fuel in each cycle. The Kr-85m activity reduction in each burnup step,in the range of 12% to 4%, is considered sufficient to show the burnup level in each cycle. The gamma emission of Kr-85m is also sufficiently high which is in the order of 1010 photon/second. From these results, it can be concluded that Kr-85m is suitable to be used as burnup measurement indicator in a pebble bed reactor.

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.

AIP Conference Proceedings, 2019

The Non-Commercial Power Reactor (Reaktor Daya Non Komersial: RDNK) is a Pebble-bed type Gas-Cool... more The Non-Commercial Power Reactor (Reaktor Daya Non Komersial: RDNK) is a Pebble-bed type Gas-Cooled High Temperature Reactor with 10 MW thermal power designed by the National Nuclear Energy Agency (BATAN). RDNK reactor follows multipass fuel loading scheme with of UO 2-fueled coated with TRISO layers and duration of stay in the reactor core for 1100 days. On the working core, the scenario of fuel loading pattern is fresh fuels loading and discharge of spent fuel pebbles as many as 25 pieces every day. Thus, if the reactor operating cycle is grouped into 5 subcycles, then in 220 days/sub-cycle will be as many as 5500 spent fuels are produced. Therefore, a safe temporary spent fuels storage container must be designed in the reactor building to hold the spent fuels before sending them to further storage. This study is to design the temporary spent fuel storage to ensure radiation safety for workers. The design of spent fuels storage containers was carried out using a monte-carlo based program and the determination of the spent fuels source strength using the ORIGEN2.1 program. The results of the design of temporary storage containers to hold 5500 spent fuel pebbles using alternatives shielding materials of ordinary concrete shield as thick as 50 cm, iron-portland concrete is 20 cm thick and lead as thick as 10 cm. All these containers provide radiation safety assurance for workers after the spent fuels have decayed for 6 months. Thus, the delivery of spent fuels to further storage outside the reactor building is recommended after 6 months decayed.

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Reaktor Daya Eksperimental (RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that... more Reaktor Daya Eksperimental (RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that is planned to be constructed by National Nuclear Energy Agency of Indonesia (BATAN) in Puspiptek complex, Tangerang Selatan. RDE utilizes low enriched UO2 fuel coated by TRISO layers and loaded into the core by means of multipass loading scheme. Determination of radionuclide characteristics of RDE spent fuel; such as activity, thermal power, neutron and photon release rates; are very important because those characteristics are crucial to be used as a base for evaluating the safety of spent fuel handling system and storage tank. This study is aimed to investigate the radionuclide characteristics of RDE spent fuel at the end of cycle and during the first 5 years cooling time in spent fuel storage. The method used to investigate the radionuclide characteristics is burnup calculation using ORIGEN2.1 code. In performing the ORIGEN2.1 calculation, one pebble fuel was assumed to be irradiated ...

Jurnal Sains dan Teknologi Nuklir Indonesia, Oct 7, 2017

ANALISIS KESELAMATAN RADIASI DI TAPAK PUSAT LISTRIK TENAGA NUKLIR (PLTN) PADA KONDISI OPERASI NOR... more ANALISIS KESELAMATAN RADIASI DI TAPAK PUSAT LISTRIK TENAGA NUKLIR (PLTN) PADA KONDISI OPERASI NORMAL, STUDI TAPAK SEBAGIN. Pembangunan PLTN memerlukan analisis keselamatan radiasi untuk membuktikan bahwa PLTN dapat beroperasi secara aman dan selamat pada kondisi operasi normal dan abnormal. Analisis keselamatan radiasi PLTN diperlukan untuk melengkapi dokumen analisis tapak dan analisis keselamatan. Penelitian ini bertujuan untuk mendapatkan data dosis radiasi di lingkungan tapak PLTN pada kondisi operasi normal. Diasumsikan terdapat tiga PLTN jenis PWR daya 1000 MWe beroperasi di Tapak Sebagin, Provinsi Bangka Belitung. Data dosis dihitung dengan menggunakan perangkat lunak PC-CREAM. Metodologi yang diterapkan untuk analisis dari tiga PWR-1000 MWe tersebut yaitu mempersiapkan input PC-CREAM meliputi data sourceterm rutin, data meteorologi daerah setempat berupa frekuensi stabilitas cuaca untuk 16 sektor (arah angin) yang diambil dari data cuaca selama 1 tahun. Selain itu juga dibutuhkan data produksi pertanian dan peternakan serta data distribusi penduduk selama setahun untuk 16 sektor dan 20 arah radial. Hasil perhitungan menunjukkan bahwa dosis maksimum untuk semua jenis nuklida dan semua alur paparan yang diterima publik (dewasa) di sekitar tapak Sebagin dari lepasan tiga PWR-1000MWe untuk kondisi operasi normal adalah sekitar 0,053 mSv/tahun ke arah Utara dalam radius 1 km. Dosis ini di bawah Nilai Batas Dosis 1 mSv/tahun atau pembatas dosis 0,3 mSv/tahun (BAPETEN).

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Indonesia possesses numerous potential sites for nuclear power plant development. A fast and comp... more Indonesia possesses numerous potential sites for nuclear power plant development. A fast and comprehensive radiological consequences analysis is required to conduct a preliminary analysis of radionuclide release into the atmosphere, including source terms estimation. One simplified method for such estimation is the use of the Relative Volatility approach by Kess and Booth, published in IAEA TECDOC 1127. The objective of this study was to evaluate the use of a simple and comprehensive tool for estimating the source terms of planned nuclear power plants to facilitate the analysis of radiological consequences during site evaluation. Input parameters for the estimation include fuel burn-up, blow-down time, specific heat transfer of fuel to cladding, and coolant debit, using 100 MWe PWR as a case study. The results indicate a slight difference in the calculated release fraction compared to previous calculations, indicating a need to modify Keywords: Source terms, Relative volatility, Rel...

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Silicon carbide (SiC) is a competitive candidate material to be used in several advanced and Gene... more Silicon carbide (SiC) is a competitive candidate material to be used in several advanced and Generation-IV nuclear reactor designs as neutron moderator, fuel coating, cladding, or core structural material. Many studies have been performed to investigate the durability of SiC in severe environment in nuclear reactor. However, the nature and behavior of defect induced by neutron irradiation are still not fully understood. This paper is aimed to study collision cascade and primary radiation damage in SiC using molecular dynamics simulation. The potential being used was a hybrid Tersoff potential modified with Ziegler-Biersack-Littmark (ZBL) screening function. The collision cascade was let evolved for 10 ps from a Si or C primary knocked atom (PKA) located initially at the top center of a system containing 960000 atoms. The simulation was carried out at room temperature as well as at several advanced fission reactor-relevant temperatures. It was obtained that the number of C point defe...

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 ...

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...

Symmetry Integrability and Geometry-methods and Applications, Jul 14, 2016

CALCULATION OF RADIONUCLIDE CONTENT OF NUCLEAR MATERIALS USING ORIGEN2.1 COMPUTER CODE. Nuclear m... more CALCULATION OF RADIONUCLIDE CONTENT OF NUCLEAR MATERIALS USING ORIGEN2.1 COMPUTER CODE. Nuclear materials contain a number of radionuclides produced from radioactive decay process. The composition of these radionuclides which are accumulated inside the nuclear materials changes over the time. The calculation of radionuclide composition inside nuclear materials is very important especially in the aspect of nuclear reactor safety evaluation, nuclear fuel behavior evaluation, and radioactive waste management. One method to calculate radionuclide content of nuclear materials is by using ORIGEN2.1 computer code. Beside radionuclide composition, this code can also calculate some characteristics related to decay process such as total radioactivity, decay heat, and neutron flux. This paper is a literature study about ORIGEN2.1 computer code. A brief description of ORIGEN2.1 and its use for calculating radionuclide content of nuclear materials are presented. Radionuclide content produced from californium-252 decay was chosen as a simple case solved by ORIGEN2.1. Californium-252 was simulated to undergo decay for 10 years. The variables which are calculated by ORIGEN2.1 in this case are radionuclide composition, total radioactivity, total alpha radioactivity, and neutron flux. From the results of this simulation, it is shown that small amount of californium-252 produces high neutron intensity so that it can be used as a reliable neutron source for many applications.

Journal of Physics: Conference Series, 2021

High Temperature Gas-cooled Reactor (HTGR) is one of the Generation-IV reactors that has many adv... more High Temperature Gas-cooled Reactor (HTGR) is one of the Generation-IV reactors that has many advantages and very promising to be utilized in the future. National Nuclear Energy Agency of Indonesia (BATAN) has a program to build a non-commercial power reactor (RDE) that will be based on the pebble-bed type of HTGR. To support this program, the research and development activities in HTGR have been conducting recently. One of the important aspects to be studied is accident analysis to assess the behavior of HTGR during the accident and to anticipate the consequences that might be caused. The purpose of this work is to study the Depressurized Loss of Force Coolant (DLOFC) accident on HTGR by using Graphite Reactor Severe Accident Code (GRSAC). The type of HTGR used as a case study is PBMR-400. This work was begun with the preparation of GRSAC input for reactor parameter including reactor design, core layout design, primary cooling loop design, graphite properties and material propertie...

FISSION PRODUCTS INVENTORY ANALYSIS OF HTGR FUEL BY USING ORIGEN2.1 COMPUTER CODE. High Temperatu... more FISSION PRODUCTS INVENTORY ANALYSIS OF HTGR FUEL BY USING ORIGEN2.1 COMPUTER CODE. High Temperature Gas-Cooled Reactor (HTGR) is one type of reactor that offers many advantages, one of which is its TRISO-coated fuel resistance to high temperature. Most of the discussions about HTGR safety are focused on the TRISO- layers reliability in containing fission products and preventing them to release out of the fuel particles. One important step before determining the number of fission products that release to the environment is the analysis of fission products inventory inside the fuel particles. This paper presents an analysis of fission products inventory inside the HTGR fuel particles by using ORIGEN2.1 computer code. The calculation of fission products inventory was conducted by using neutron cross section library data for HTGR type and then the results of this calculation were compared to the results of calculation by using neutron cross section library data for PWR type. The results...

DOSES ANALYSIS OF A HYPOTHETICAL LOCA ACCIDENT ON NUCLEAR POWER PLANT (NPPs) SITTING. Lessons lea... more DOSES ANALYSIS OF A HYPOTHETICAL LOCA ACCIDENT ON NUCLEAR POWER PLANT (NPPs) SITTING. Lessons learn from the Fukushima accident, IAEA and national regulatory agencies BAPETEN requires nuclear emergency documents with the postulation of accidents involving the release of radioactive substances into the environment. The document were for existing nuclear installations and installations are planned to be built. The study on the impact of reactor accidents for the situation in Indonesia has been done, but not comprehensively assesses the radiation dose on environment. This research will be presented in the analysis of radiation dose due to design bases accident (DBA) for PWR power reactor with a capacity of 1000 MWe are planned to be built. The research goal was to analysis of radiation dose as short-term effective dose by distance, nuclide, pathway, and by organ; long-term effective dose by distance, nuclide, pathway, and by organ; and collective dose by pathway and distance. Radiation...

Jurnal Sains dan Teknologi Nuklir Indonesia, 2020

also be considered for the simplicity it offers and thus potentially lower cost. Due to different... more also be considered for the simplicity it offers and thus potentially lower cost. Due to different neutronic and burnup profile between the two, the resulting spent fuel characteristic is also different and possibly requires different handling mechanism. This paper assesses the characteristics of OTTO-scheme RDE spent fuel using thorium fuel cycle to provide preliminary data and insight for its spent fuel management. The assessment is performed employing ORIGEN2.1 code. At day 30 of cooling after determined end-of-cycle (EOC), each spent fuel yields 234.9 Curies of radioactivity, emitting 66.26 neutrons/second, 1x10 13 photons/second, and releasing 0.7675 watts of decay heat. These numbers must be taken into consideration regarding spent fuel management and spent fuel cask design. Tl-208 isotope characteristics, whose existence is unique to thorium fuel cycle, were also determined. It is found to be yielding 3.42x10-3 Curie of radioactivity and releasing 1.2x10 8 photons/second at its peak. Understanding its high-energy gamma release, proper radiation protection mechanism must be implemented.

Malaysian Journal of Fundamental and Applied Sciences, 2019

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.

Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir, 2019

CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL. There ar... more CHARACTERISTICS OF RADIONUCLIDES ON THORIUM-CYCLE EXPERIMENTAL POWER REACTOR SPENT FUEL. There are several options of nuclear fuel utilisation in the HTGR-based Experimental Power Reactor (Reaktor Daya Eksperimental/RDE). Although mainly RDE utilises low enriched uranium (LEU)-based fuel, which is the most viable option at the moment, it is possible for RDE to utilise other fuel, for example thorium-based and possibly even plutonium-based fuel. Different fuel yields different spent fuel characteristics, so it is necessary to identify the characteristics to understand and evaluate their handling and interim storage. This paper provides the study on the characteristics of thorium-fuelled RDE spent fuel, assuming typical operational cycle. ORIGEN2.1 code is employed to determine the spent fuel characteristics. The result showed that at the end of the calculation cycle, each thorium-based spent fuel pebble generates around 0,627 Watts of heat, 28 neutrons/s, 8.28x1012 photons/s and yiel...

Journal of Physics: Conference Series, 2018

Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a mod... more Pebble Bed Reactor (PBR) is a high temperature gas-cooled reactor which employs graphite as a moderator and helium as a coolant. In a multi-pass PBR, burnup of the fuel pebble must be measured in each cycle by online measurement in order to determine whether the fuel pebble should be reloaded into the core for another cycle or moved out of the core into spent fuel storage. One of the well-known methods for measuring burnup is based on the activity of radionuclide decay inside the fuel pebble. In this work, the activity and gamma emission of Kr-85m were studied in order to investigate the feasibility of Kr-85m as burnup measurement indicator in a PBR. The activity and gamma emission of Kr-85 were estimated using ORIGEN2.1 computer code. The parameters of HTR-10 were taken as a case study in performing ORIGEN2.1 simulation. The results show that the activity revolution of Kr-85m has a good relationship with the burnup of the pebble fuel in each cycle. The Kr-85m activity reduction in each burnup step,in the range of 12% to 4%, is considered sufficient to show the burnup level in each cycle. The gamma emission of Kr-85m is also sufficiently high which is in the order of 1010 photon/second. From these results, it can be concluded that Kr-85m is suitable to be used as burnup measurement indicator in a pebble bed reactor.

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.

AIP Conference Proceedings, 2019

The Non-Commercial Power Reactor (Reaktor Daya Non Komersial: RDNK) is a Pebble-bed type Gas-Cool... more The Non-Commercial Power Reactor (Reaktor Daya Non Komersial: RDNK) is a Pebble-bed type Gas-Cooled High Temperature Reactor with 10 MW thermal power designed by the National Nuclear Energy Agency (BATAN). RDNK reactor follows multipass fuel loading scheme with of UO 2-fueled coated with TRISO layers and duration of stay in the reactor core for 1100 days. On the working core, the scenario of fuel loading pattern is fresh fuels loading and discharge of spent fuel pebbles as many as 25 pieces every day. Thus, if the reactor operating cycle is grouped into 5 subcycles, then in 220 days/sub-cycle will be as many as 5500 spent fuels are produced. Therefore, a safe temporary spent fuels storage container must be designed in the reactor building to hold the spent fuels before sending them to further storage. This study is to design the temporary spent fuel storage to ensure radiation safety for workers. The design of spent fuels storage containers was carried out using a monte-carlo based program and the determination of the spent fuels source strength using the ORIGEN2.1 program. The results of the design of temporary storage containers to hold 5500 spent fuel pebbles using alternatives shielding materials of ordinary concrete shield as thick as 50 cm, iron-portland concrete is 20 cm thick and lead as thick as 10 cm. All these containers provide radiation safety assurance for workers after the spent fuels have decayed for 6 months. Thus, the delivery of spent fuels to further storage outside the reactor building is recommended after 6 months decayed.

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA

Reaktor Daya Eksperimental (RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that... more Reaktor Daya Eksperimental (RDE) is a 10 MWth pebble-bed High Temperature Gas-cooled Reactor that is planned to be constructed by National Nuclear Energy Agency of Indonesia (BATAN) in Puspiptek complex, Tangerang Selatan. RDE utilizes low enriched UO2 fuel coated by TRISO layers and loaded into the core by means of multipass loading scheme. Determination of radionuclide characteristics of RDE spent fuel; such as activity, thermal power, neutron and photon release rates; are very important because those characteristics are crucial to be used as a base for evaluating the safety of spent fuel handling system and storage tank. This study is aimed to investigate the radionuclide characteristics of RDE spent fuel at the end of cycle and during the first 5 years cooling time in spent fuel storage. The method used to investigate the radionuclide characteristics is burnup calculation using ORIGEN2.1 code. In performing the ORIGEN2.1 calculation, one pebble fuel was assumed to be irradiated ...

Jurnal Sains dan Teknologi Nuklir Indonesia, Oct 7, 2017

ANALISIS KESELAMATAN RADIASI DI TAPAK PUSAT LISTRIK TENAGA NUKLIR (PLTN) PADA KONDISI OPERASI NOR... more ANALISIS KESELAMATAN RADIASI DI TAPAK PUSAT LISTRIK TENAGA NUKLIR (PLTN) PADA KONDISI OPERASI NORMAL, STUDI TAPAK SEBAGIN. Pembangunan PLTN memerlukan analisis keselamatan radiasi untuk membuktikan bahwa PLTN dapat beroperasi secara aman dan selamat pada kondisi operasi normal dan abnormal. Analisis keselamatan radiasi PLTN diperlukan untuk melengkapi dokumen analisis tapak dan analisis keselamatan. Penelitian ini bertujuan untuk mendapatkan data dosis radiasi di lingkungan tapak PLTN pada kondisi operasi normal. Diasumsikan terdapat tiga PLTN jenis PWR daya 1000 MWe beroperasi di Tapak Sebagin, Provinsi Bangka Belitung. Data dosis dihitung dengan menggunakan perangkat lunak PC-CREAM. Metodologi yang diterapkan untuk analisis dari tiga PWR-1000 MWe tersebut yaitu mempersiapkan input PC-CREAM meliputi data sourceterm rutin, data meteorologi daerah setempat berupa frekuensi stabilitas cuaca untuk 16 sektor (arah angin) yang diambil dari data cuaca selama 1 tahun. Selain itu juga dibutuhkan data produksi pertanian dan peternakan serta data distribusi penduduk selama setahun untuk 16 sektor dan 20 arah radial. Hasil perhitungan menunjukkan bahwa dosis maksimum untuk semua jenis nuklida dan semua alur paparan yang diterima publik (dewasa) di sekitar tapak Sebagin dari lepasan tiga PWR-1000MWe untuk kondisi operasi normal adalah sekitar 0,053 mSv/tahun ke arah Utara dalam radius 1 km. Dosis ini di bawah Nilai Batas Dosis 1 mSv/tahun atau pembatas dosis 0,3 mSv/tahun (BAPETEN).