Sri Kuntjoro - Academia.edu (original) (raw)

Papers by Sri Kuntjoro

Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian ... more Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA) Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahw...

Antisipasi kecelakaan parah suatu rancangan PLTN menjadi sangat penting dalam penerimaan suatu te... more Antisipasi kecelakaan parah suatu rancangan PLTN menjadi sangat penting dalam penerimaan suatu teknologi PLTN setelah terjadinya kecelakaan parah reaktor nuklir Fukushima di Jepang. Antisipasi tersebut tertuang dalam rencana strategi manajemen kecelakaan reaktor. Berdasarkan permasalahan tersebut maka dilakukan penelitian tentang analisis konsekuensi kecelakaan parah PWR ( Pressurized Water Reactor ) belajar dari Fukushima untuk manajemen kecelakaan reaktor dengan metode baru yang dikenal dengan nama backwards method (metode hitung mundur). Backwards method adalah menghitung berdasarkan hasil paparan radiasi terukur yang diterima publik untuk mengestimasi besarnya kerusakan pada teras reaktor sebagai sumber utama radiasi di reaktor PLTN. Analisis konsekuensi kecelakaan parah untuk reaktor daya PWR dengan bakwards method telah dilakukan untuk calon tapak potensial di Indonesia seperti Semenajung Muria, Pesisir Serang, tapak dengan Stabilitas C, Stabilitas D, Stabilitas E, dan Stabili...

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

The study of atmospheric dispersion requires meteorological data from the nuclear reactor site. P... more The study of atmospheric dispersion requires meteorological data from the nuclear reactor site. Parameter data that is very important to solve the atmospheric dispersion equation is weather stability. Weather stability, among others, requires solar radiation data. One way to measure solar radiation is to use an actinograph equipment like in BMKG Mempawah. The actinograph measurement results are curve, the edge of the curve is calculated using a planimeter to get solar radiation value. In this study, a code was developed to calculate the curve edge detection with phyton. The data used is Mempawah BMKG Station for one month periodical hourly on March 2021. The stages of processing the image to get the solar radiation value are pre-processing, processing and post processing. In the pre processing stage, the canny edge detection method is used to get the edge of the curve, as the basis to get solar radiation value. The regression model between the results of the planimeter calculation and the code has a correlation coefficient of 88.6 % with R-square of 78.5 %, which means that the result of the code calculations can explain the results of planimeter measurements of 78.5 %. Based on the result shows that canny edge method can be used to simplify the calculation of solar radiation. Further research with other method needs to be done to get better results. The Solar radiation value obtained from calculations using planimeter or code, based on table 1 is in the range between 175-675. The solar radiation value in meteorological data is used to calculate atmospheric stability

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 ...

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...

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.

Journal of Physics: Conference Series, 2019

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, 2019

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.

Journal of Physics: Conference Series, 2019

Currently plate-type fuels are planned to use in the research reactor TRIGA-2000 Bandung, Indones... more Currently plate-type fuels are planned to use in the research reactor TRIGA-2000 Bandung, Indonesia (further called as TRIGA-Plate reactor). Plate type fuel consists of enriched uranium sandwiched between metal cladding. Plate-type fuel is used in some research reactors to obtain high neutron flux and is used to study material irradiation or isotope production. For the purpose of Safety Analysis Report (SAR) on the fuel-type modification plan of TRIGA-2000 Bandung, it is necessary to calculate radioactivity inventory from the reactor core under operating conditions. The radioactivity inventory of materials irradiated in a reactor depends on the initial material composition at the BOC, burn-up history from BOC to EOC and the power peaking factor. The purpose of the present work is to evaluate the fuel burn-up and radioactivity inventory in fuel materials of TRIGA-Plate research reactor. The mass of U-235 consumed and the mass of Pu-239 and Th-232 produced are also evaluated. The calculation results obtained using ORIGEN2 code are: the mass of U-235 consumed is 1.40E+02 gram; while the mass of Pu-239 produced is 1.36E+00 gram and Th-232 produced is 1.03E-06 gram. The largest radioactivity inventories of the reactor at the EOC sequentially are: Kr group is about 2.29E+02 Ci; for group I is 7.77E+03 Ci and for groups Cs is 1.92E+02Ci.

Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir, 2016

ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI. Berkaitan deng... more ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI. Berkaitan dengan rencana Badan Tenaga Nuklir Nasional (BATAN) untuk mengoperasikan reaktor eksperimental jenis Reaktor Gas Temperatur Tinggi (RGTT), maka diperlukan analisis keselamatan terhadap reaktor terutama yang berkaitan dengan issue lingkungan. Analisis sebaran radionuklida dari reaktor ke lingkungan pada kondisi operasi normal atau abnormal diawali dengan estimasi sumber radionuklida di teras reaktor (inventori teras) berdasarkan pada tipe, daya, dan operasi reaktor. Tujuan penelitian adalah melakukan analisis inventori teras untuk disain Reaktor Daya Eksperimental (RDE) jenis reaktor gas temperature tinggi berdaya 10 MWt, 20 MWt dan 30 MWt. Analisis dilakukan menggunakan program ORIGEN2 berbasis pustaka penampang lintang pada temperatur tinggi. Perhitungan diawali dengan membuat modifikasi beberapa parameter pustaka tampang lintang berdasarkan temperatur rata-rata teras sebesar 5700 °C dan di...

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA, 2015

ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAA... more ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe. Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR) diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor. Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Da...

Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir, 2017

Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Te... more Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source term released into the environment based on postulation of normal or routine operations of RDE. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitation during the fabrication. Mechanism of fission products release from the fuel to the environment was cr...

Atom Indonesia, 2016

The Fukushima accident resulted in the melting of the reactor core due to loss of supply of coola... more The Fukushima accident resulted in the melting of the reactor core due to loss of supply of coolant when the reactor stopped from operating conditions. The earthquake and tsunami caused loss of electricity due to the flooding that occurred in the reactor. The absence of the coolant supply after reactor shutdown resulted in heat accumulation, causing the temperature of the fuel to rise beyond its melting point. In the early stages of the accident, operator could not determine the severity of the accident and the percentage of the reactor core damaged. The available data was based on the radiation exposure in the environment that was reported by the authorities. The aim of this paper is to determine the severity of the conditions in the reactor core based on the radiation doses measured in the environment. The method is performed by backward counting based on the measuring radiation exposure and radionuclides releases source term. The calculation was performed by using the PC-COSYMA code. The results showed that the core damage fraction at Dai-ichi Unit 1 was 70%, and the resulting individual effective dose in the exclusion area is 401 mSv, while the core damage fraction at Unit 2 was 30%, and the resulting individual effective dose was 99.1 mSv, while for Unit 3, the core damage fraction was 25% for an individual effective dose of 92.2 mSv. The differences between the results of the calculation for estimation of core damage proposed in this paper with the previously reported results is probably caused by the applied model for assessment, differences in postulations and assumptions, and the incompleteness of the input data. This difference could be reduced by performing calculations and simulations for more varied assumptions and postulations.

Jurnal Teknologi Pengelolaan Limbah, Mar 24, 2015

PENENTUAN ZONA KEDARURATAN NUKLIR OFF-SITE (LUAR TAPAK) DI INDONESIA. Belajar dari kejadian Fukus... more PENENTUAN ZONA KEDARURATAN NUKLIR OFF-SITE (LUAR TAPAK) DI INDONESIA. Belajar dari kejadian Fukushima terutama dalam manajemen penanganan kedaruratan nuklir setelah kejadian perlu dibahas kembali untuk meningkatkan persepsi masyarakat Indonesia tentang keamanan dan keselamatan nuklir Pembangkit Listrik Tenaga Nuklir (PLTN). Zonasi kedaruratan sangat penting dalam manajemen kecelakaan nuklir, karena mempercepat dan lebih tepat untuk pengambilan tindakan protektif terhadap masyarakat dan lingkungan. Untuk itu, makalah ini bertujuan untuk mengetahui zona perencanaan kedaruratan nuklir ( Emergency Planning Zones , EPZ) untuk luar tapak ( off-site ) di Indonesia. Metodologi penelitian adalah melakukan simulasi perhitungan untuk reaktor PLTN tipe PWR-1000 MWe (± 3300 MWth) pada kondisi abnormal yang dipostulasikan sebagai kecelakaan dasar disain ( Design Basis Accident , DBA) dan kecelakaan luar dasar disain ( Beyond Design Basis Accident BDBA) di tapak Ujung Lemah Abang, Bojanegara, Bangka Barat dan Bangka Selatan. Perhitungan dan simulasi menggunakan modul countermeasure dari program PC-Cosyma. Diperoleh bahwa Tapak Ujung Lemah Abang mempunyai zona perencanaan kedaruratan nuklir EPZ paling sederhana. Pengaruh kondisi tapak lebih berpengaruh dibandingkan dengan besarnya aktivitas lepasan radioaktif, terutama kondisi meteorologi dan kondisi lingkungan. Kata kunci: Indonesia , luar tapak, zona-kedaruratan, PLTN . A B S T R A CT THE DETERMINATION OF NUCLEAR EMERGENCY ZONE FOR OFF SITE IN INDONESIA. Learning from the Fukushima accident, especially in the nuclear emergencies management after the accident needs to be reviewed which improve Indonesian perceptions of nuclear power plant (NPP) safety. Zoning is very important for the nuclear emergency management, as it accelerates and more precise in taking protective actions on society and the environment. This paper aims to determine the nuclear emergency planning zone EPZ for off-site in Indonesia. The research methodology is to calculations for PWR- 1000 MWe (±3300 MWth) under abnormal conditions postulated as a design basis accident, DBA and beyond design basis accidents BDBA on the site of Ujung Lemah Abang, Bojanegara, West and South Bangka. Calculations and simulations using countermeasures module of PC-Cosyma programme. The result that Ujung Lemah Abang site has the simplest nuclear emergency planning zone EPZ. Site conditions is more influential than the magnitude of the activity radioactive releases, especially the meteorological and environmental conditions. Keywords: Indonesia , off-site , emergency zone , NPP.

Jurnal Pengembangan Energi Nuklir, 2015

ABSTRAKAKTIVITAS DAN KONSEKUENSI DISPERSI RADIOAKTIF UNTUK DAERAH KOTA DAN PEDESAAN. Konsekuensi ... more ABSTRAKAKTIVITAS DAN KONSEKUENSI DISPERSI RADIOAKTIF UNTUK DAERAH KOTA DAN PEDESAAN. Konsekuensi karena lepasan kontaminan radioaktif oleh manusia dipengaruhi oleh banyak faktor seperti besarnya aktivitas kontaminan yang tersebar dan kondisi lingkungan. Kondisi lingkungan meliputi kondisi meteorologi, kontur tapak, dan pathway kontaminan ke manusia. Tujuan penelitian ini adalah analisis aktivitas dan konsekuensi radionuklida waktu paruh panjang akibat kecelakaan di daerah perkotaan dan pedesaan. Tujuan khusus adalah menghitung aktivitas dispersi udara dan deposisi permukaan, prediksi laju dosis dan risiko yang ditimbulkan untuk daerah perkotaan dan pedesaan sebagai fungsi lokasi. Metode yang digunakan adalah simulasi estimasi konsekuensi dari dispersi produk fisi di atmosfer akibat kecelakaan terpostulasi Beyond Design Basis Accident, BDBA. Perhitungan dilakukan untuk lepasan radioaktif akibat kecelakaan PWR 1000 MWe yang disimulasikan untuk area pedesaan dan perkotaan Tapak Bojaneg...

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

Estimation of Core inventory simulation for PWR 160MWt has been done. Core inventory is one of th... more Estimation of Core inventory simulation for PWR 160MWt has been done. Core inventory is one of the parameters related to the release of radioactive materials be used as data for further calculation or analysis related to nuclear safety. PWR 160 MWt has a core configuration consist of 37 fuel assemblies with three types of Uranium enrichment of the fuel. Core fission product inventory calculation was performed using ORIGEN2.1 by assuming the reactor irradiated for 2 years (fuel cycle) without cooling time. Simulation results in many radionuclides, some nuclides have long decay times (e.g. Cs-135; Cs-137), some have short-lived (e.g. I-136, Xe-140), some have high radioactivity (e.g. Xe-133; I-134) and some nuclides indicate a significant contribution to determining source term (e.g. I-131; Cs-137). Simulation results obtained can be used for further calculations for safety analysis such as determination of source term, dose calculation, and precautions to minimize the occurrence of radioactive to the environment.

Symmetry Integrability and Geometry-methods and Applications, 2017

Reaktor riset RSG - GAS merupakan reactor jenis MTR dengan bahan bakar plat U 3 Si 2 - Al dan ber... more Reaktor riset RSG - GAS merupakan reactor jenis MTR dengan bahan bakar plat U 3 Si 2 - Al dan beroperasi dengan daya nomi- nal 30 MWt. Berbagai aktifitas dilakukan di reaktor antara lain penelitian bahan, penelitian reaktor serta produksi radioisotop. Isotop Mo - 99 merupakan salah satu isotop yang diproduksi di reaktor RSG - GAS dan merupakan isotope yang dibutuhkan dalam bidang kesehatan dalam jumlah besar. Produksi isotop Mo - 99 dicapai dengan cara melakukan iradiasi pada LEU ( Low Enriched Uranium ) berbentuk plat di teras reaktor. Tujuan dari penelitian ini adalah untuk menganalisis aktivitas isotop Mo - 99 sebesar 300 Ci hasil dari iradisai target plat LEU yang diiradiasi di teras reaktor RSG - GAS dengan program ORIGEN2. Sebagai masukan untuk program tersebut adalah fluks neutron di posisi LEU yang diiradiasi, lama iradiasi serta massa U - 235 dan U - 238 yang diiradiasi. Selain itu analisis dilakukan berdasarkan hasil pengolahan beberapa target LEU yang telah diiradiasi seb...

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...

Journal of Physics: Conference Series, 2018

Abstract A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-G... more Abstract A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

AIP Conference Proceedings, 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.

Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian ... more Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA) Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahw...

Antisipasi kecelakaan parah suatu rancangan PLTN menjadi sangat penting dalam penerimaan suatu te... more Antisipasi kecelakaan parah suatu rancangan PLTN menjadi sangat penting dalam penerimaan suatu teknologi PLTN setelah terjadinya kecelakaan parah reaktor nuklir Fukushima di Jepang. Antisipasi tersebut tertuang dalam rencana strategi manajemen kecelakaan reaktor. Berdasarkan permasalahan tersebut maka dilakukan penelitian tentang analisis konsekuensi kecelakaan parah PWR ( Pressurized Water Reactor ) belajar dari Fukushima untuk manajemen kecelakaan reaktor dengan metode baru yang dikenal dengan nama backwards method (metode hitung mundur). Backwards method adalah menghitung berdasarkan hasil paparan radiasi terukur yang diterima publik untuk mengestimasi besarnya kerusakan pada teras reaktor sebagai sumber utama radiasi di reaktor PLTN. Analisis konsekuensi kecelakaan parah untuk reaktor daya PWR dengan bakwards method telah dilakukan untuk calon tapak potensial di Indonesia seperti Semenajung Muria, Pesisir Serang, tapak dengan Stabilitas C, Stabilitas D, Stabilitas E, dan Stabili...

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

The study of atmospheric dispersion requires meteorological data from the nuclear reactor site. P... more The study of atmospheric dispersion requires meteorological data from the nuclear reactor site. Parameter data that is very important to solve the atmospheric dispersion equation is weather stability. Weather stability, among others, requires solar radiation data. One way to measure solar radiation is to use an actinograph equipment like in BMKG Mempawah. The actinograph measurement results are curve, the edge of the curve is calculated using a planimeter to get solar radiation value. In this study, a code was developed to calculate the curve edge detection with phyton. The data used is Mempawah BMKG Station for one month periodical hourly on March 2021. The stages of processing the image to get the solar radiation value are pre-processing, processing and post processing. In the pre processing stage, the canny edge detection method is used to get the edge of the curve, as the basis to get solar radiation value. The regression model between the results of the planimeter calculation and the code has a correlation coefficient of 88.6 % with R-square of 78.5 %, which means that the result of the code calculations can explain the results of planimeter measurements of 78.5 %. Based on the result shows that canny edge method can be used to simplify the calculation of solar radiation. Further research with other method needs to be done to get better results. The Solar radiation value obtained from calculations using planimeter or code, based on table 1 is in the range between 175-675. The solar radiation value in meteorological data is used to calculate atmospheric stability

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 ...

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...

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.

Journal of Physics: Conference Series, 2019

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, 2019

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.

Journal of Physics: Conference Series, 2019

Currently plate-type fuels are planned to use in the research reactor TRIGA-2000 Bandung, Indones... more Currently plate-type fuels are planned to use in the research reactor TRIGA-2000 Bandung, Indonesia (further called as TRIGA-Plate reactor). Plate type fuel consists of enriched uranium sandwiched between metal cladding. Plate-type fuel is used in some research reactors to obtain high neutron flux and is used to study material irradiation or isotope production. For the purpose of Safety Analysis Report (SAR) on the fuel-type modification plan of TRIGA-2000 Bandung, it is necessary to calculate radioactivity inventory from the reactor core under operating conditions. The radioactivity inventory of materials irradiated in a reactor depends on the initial material composition at the BOC, burn-up history from BOC to EOC and the power peaking factor. The purpose of the present work is to evaluate the fuel burn-up and radioactivity inventory in fuel materials of TRIGA-Plate research reactor. The mass of U-235 consumed and the mass of Pu-239 and Th-232 produced are also evaluated. The calculation results obtained using ORIGEN2 code are: the mass of U-235 consumed is 1.40E+02 gram; while the mass of Pu-239 produced is 1.36E+00 gram and Th-232 produced is 1.03E-06 gram. The largest radioactivity inventories of the reactor at the EOC sequentially are: Kr group is about 2.29E+02 Ci; for group I is 7.77E+03 Ci and for groups Cs is 1.92E+02Ci.

Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir, 2016

ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI. Berkaitan deng... more ANALISIS INVENTORI REAKTOR DAYA EKSPERIMENTAL JENIS REAKTOR GAS TEMPERATUR TINGGI. Berkaitan dengan rencana Badan Tenaga Nuklir Nasional (BATAN) untuk mengoperasikan reaktor eksperimental jenis Reaktor Gas Temperatur Tinggi (RGTT), maka diperlukan analisis keselamatan terhadap reaktor terutama yang berkaitan dengan issue lingkungan. Analisis sebaran radionuklida dari reaktor ke lingkungan pada kondisi operasi normal atau abnormal diawali dengan estimasi sumber radionuklida di teras reaktor (inventori teras) berdasarkan pada tipe, daya, dan operasi reaktor. Tujuan penelitian adalah melakukan analisis inventori teras untuk disain Reaktor Daya Eksperimental (RDE) jenis reaktor gas temperature tinggi berdaya 10 MWt, 20 MWt dan 30 MWt. Analisis dilakukan menggunakan program ORIGEN2 berbasis pustaka penampang lintang pada temperatur tinggi. Perhitungan diawali dengan membuat modifikasi beberapa parameter pustaka tampang lintang berdasarkan temperatur rata-rata teras sebesar 5700 °C dan di...

JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA, 2015

ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAA... more ABSTRAK PENGARUH KONDISI ATMOSFERIK TERHADAP PERHITUNGAN PROBABILISTIK DAMPAK RADIOLOGI KECELAKAAN PWR 1000-MWe. Perhitungan dampak kecelakaan radiologi terhadap lepasan produk fisi akibat kecelakaan potensial yang mungkin terjadi di Pressurized Water Reactor (PWR) diperlukan secara probabilistik. Mengingat kondisi atmosfer sangat berperan terhadap dispersi radionuklida di lingkungan, dalam penelitian ini akan dianalisis pengaruh kondisi atmosferik terhadap perhitungan probabilistik dari konsekuensi kecelakaan reaktor. Tujuan penelitian adalah melakukan analisis terhadap pengaruh kondisi atmosfer berdasarkan model data input meteorologi terhadap dampak radiologi kecelakaan PWR 1000-MWe yang disimulasikan pada tapak yang mempunyai kondisi meteorologi yang berbeda. Simulasi menggunakan program PC-Cosyma dengan moda perhitungan probabilistik, dengan data input meteorologi yang dieksekusi secara cyclic dan stratified, dan disimulasikan di Tapak Semenanjung Muria dan Pesisir Serang. Da...

Urania Jurnal Ilmiah Daur Bahan Bakar Nuklir, 2017

Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Te... more Experimental power reactor (RDE) which is planned to be constructed by BATAN is a kind of High Temperature Gas Cooled Reactor (HTGR) with 10 MWth power. HTGR is a helium gas-cooled reactor with TRISO-coated fuel that is able to confine fission products remained in the core. Although the fission products released into the environment are very small, in order to comply the regulations the study about environmental radiation on normal or routine operation condition need to be performed. Estimation of radiology in the environment involves the source term released into the environment under routine operation condition. The purpose of this study is to estimate the source term released into the environment based on postulation of normal or routine operations of RDE. The research approach starts with an assumption that there are defects and impurities in the TRISO fuel because of limitation during the fabrication. Mechanism of fission products release from the fuel to the environment was cr...

Atom Indonesia, 2016

The Fukushima accident resulted in the melting of the reactor core due to loss of supply of coola... more The Fukushima accident resulted in the melting of the reactor core due to loss of supply of coolant when the reactor stopped from operating conditions. The earthquake and tsunami caused loss of electricity due to the flooding that occurred in the reactor. The absence of the coolant supply after reactor shutdown resulted in heat accumulation, causing the temperature of the fuel to rise beyond its melting point. In the early stages of the accident, operator could not determine the severity of the accident and the percentage of the reactor core damaged. The available data was based on the radiation exposure in the environment that was reported by the authorities. The aim of this paper is to determine the severity of the conditions in the reactor core based on the radiation doses measured in the environment. The method is performed by backward counting based on the measuring radiation exposure and radionuclides releases source term. The calculation was performed by using the PC-COSYMA code. The results showed that the core damage fraction at Dai-ichi Unit 1 was 70%, and the resulting individual effective dose in the exclusion area is 401 mSv, while the core damage fraction at Unit 2 was 30%, and the resulting individual effective dose was 99.1 mSv, while for Unit 3, the core damage fraction was 25% for an individual effective dose of 92.2 mSv. The differences between the results of the calculation for estimation of core damage proposed in this paper with the previously reported results is probably caused by the applied model for assessment, differences in postulations and assumptions, and the incompleteness of the input data. This difference could be reduced by performing calculations and simulations for more varied assumptions and postulations.

Jurnal Teknologi Pengelolaan Limbah, Mar 24, 2015

PENENTUAN ZONA KEDARURATAN NUKLIR OFF-SITE (LUAR TAPAK) DI INDONESIA. Belajar dari kejadian Fukus... more PENENTUAN ZONA KEDARURATAN NUKLIR OFF-SITE (LUAR TAPAK) DI INDONESIA. Belajar dari kejadian Fukushima terutama dalam manajemen penanganan kedaruratan nuklir setelah kejadian perlu dibahas kembali untuk meningkatkan persepsi masyarakat Indonesia tentang keamanan dan keselamatan nuklir Pembangkit Listrik Tenaga Nuklir (PLTN). Zonasi kedaruratan sangat penting dalam manajemen kecelakaan nuklir, karena mempercepat dan lebih tepat untuk pengambilan tindakan protektif terhadap masyarakat dan lingkungan. Untuk itu, makalah ini bertujuan untuk mengetahui zona perencanaan kedaruratan nuklir ( Emergency Planning Zones , EPZ) untuk luar tapak ( off-site ) di Indonesia. Metodologi penelitian adalah melakukan simulasi perhitungan untuk reaktor PLTN tipe PWR-1000 MWe (± 3300 MWth) pada kondisi abnormal yang dipostulasikan sebagai kecelakaan dasar disain ( Design Basis Accident , DBA) dan kecelakaan luar dasar disain ( Beyond Design Basis Accident BDBA) di tapak Ujung Lemah Abang, Bojanegara, Bangka Barat dan Bangka Selatan. Perhitungan dan simulasi menggunakan modul countermeasure dari program PC-Cosyma. Diperoleh bahwa Tapak Ujung Lemah Abang mempunyai zona perencanaan kedaruratan nuklir EPZ paling sederhana. Pengaruh kondisi tapak lebih berpengaruh dibandingkan dengan besarnya aktivitas lepasan radioaktif, terutama kondisi meteorologi dan kondisi lingkungan. Kata kunci: Indonesia , luar tapak, zona-kedaruratan, PLTN . A B S T R A CT THE DETERMINATION OF NUCLEAR EMERGENCY ZONE FOR OFF SITE IN INDONESIA. Learning from the Fukushima accident, especially in the nuclear emergencies management after the accident needs to be reviewed which improve Indonesian perceptions of nuclear power plant (NPP) safety. Zoning is very important for the nuclear emergency management, as it accelerates and more precise in taking protective actions on society and the environment. This paper aims to determine the nuclear emergency planning zone EPZ for off-site in Indonesia. The research methodology is to calculations for PWR- 1000 MWe (±3300 MWth) under abnormal conditions postulated as a design basis accident, DBA and beyond design basis accidents BDBA on the site of Ujung Lemah Abang, Bojanegara, West and South Bangka. Calculations and simulations using countermeasures module of PC-Cosyma programme. The result that Ujung Lemah Abang site has the simplest nuclear emergency planning zone EPZ. Site conditions is more influential than the magnitude of the activity radioactive releases, especially the meteorological and environmental conditions. Keywords: Indonesia , off-site , emergency zone , NPP.

Jurnal Pengembangan Energi Nuklir, 2015

ABSTRAKAKTIVITAS DAN KONSEKUENSI DISPERSI RADIOAKTIF UNTUK DAERAH KOTA DAN PEDESAAN. Konsekuensi ... more ABSTRAKAKTIVITAS DAN KONSEKUENSI DISPERSI RADIOAKTIF UNTUK DAERAH KOTA DAN PEDESAAN. Konsekuensi karena lepasan kontaminan radioaktif oleh manusia dipengaruhi oleh banyak faktor seperti besarnya aktivitas kontaminan yang tersebar dan kondisi lingkungan. Kondisi lingkungan meliputi kondisi meteorologi, kontur tapak, dan pathway kontaminan ke manusia. Tujuan penelitian ini adalah analisis aktivitas dan konsekuensi radionuklida waktu paruh panjang akibat kecelakaan di daerah perkotaan dan pedesaan. Tujuan khusus adalah menghitung aktivitas dispersi udara dan deposisi permukaan, prediksi laju dosis dan risiko yang ditimbulkan untuk daerah perkotaan dan pedesaan sebagai fungsi lokasi. Metode yang digunakan adalah simulasi estimasi konsekuensi dari dispersi produk fisi di atmosfer akibat kecelakaan terpostulasi Beyond Design Basis Accident, BDBA. Perhitungan dilakukan untuk lepasan radioaktif akibat kecelakaan PWR 1000 MWe yang disimulasikan untuk area pedesaan dan perkotaan Tapak Bojaneg...

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

Estimation of Core inventory simulation for PWR 160MWt has been done. Core inventory is one of th... more Estimation of Core inventory simulation for PWR 160MWt has been done. Core inventory is one of the parameters related to the release of radioactive materials be used as data for further calculation or analysis related to nuclear safety. PWR 160 MWt has a core configuration consist of 37 fuel assemblies with three types of Uranium enrichment of the fuel. Core fission product inventory calculation was performed using ORIGEN2.1 by assuming the reactor irradiated for 2 years (fuel cycle) without cooling time. Simulation results in many radionuclides, some nuclides have long decay times (e.g. Cs-135; Cs-137), some have short-lived (e.g. I-136, Xe-140), some have high radioactivity (e.g. Xe-133; I-134) and some nuclides indicate a significant contribution to determining source term (e.g. I-131; Cs-137). Simulation results obtained can be used for further calculations for safety analysis such as determination of source term, dose calculation, and precautions to minimize the occurrence of radioactive to the environment.

Symmetry Integrability and Geometry-methods and Applications, 2017

Reaktor riset RSG - GAS merupakan reactor jenis MTR dengan bahan bakar plat U 3 Si 2 - Al dan ber... more Reaktor riset RSG - GAS merupakan reactor jenis MTR dengan bahan bakar plat U 3 Si 2 - Al dan beroperasi dengan daya nomi- nal 30 MWt. Berbagai aktifitas dilakukan di reaktor antara lain penelitian bahan, penelitian reaktor serta produksi radioisotop. Isotop Mo - 99 merupakan salah satu isotop yang diproduksi di reaktor RSG - GAS dan merupakan isotope yang dibutuhkan dalam bidang kesehatan dalam jumlah besar. Produksi isotop Mo - 99 dicapai dengan cara melakukan iradiasi pada LEU ( Low Enriched Uranium ) berbentuk plat di teras reaktor. Tujuan dari penelitian ini adalah untuk menganalisis aktivitas isotop Mo - 99 sebesar 300 Ci hasil dari iradisai target plat LEU yang diiradiasi di teras reaktor RSG - GAS dengan program ORIGEN2. Sebagai masukan untuk program tersebut adalah fluks neutron di posisi LEU yang diiradiasi, lama iradiasi serta massa U - 235 dan U - 238 yang diiradiasi. Selain itu analisis dilakukan berdasarkan hasil pengolahan beberapa target LEU yang telah diiradiasi seb...

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...

Journal of Physics: Conference Series, 2018

Abstract A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-G... more Abstract A preliminary study of minor actinides (MA) transmutation in the high flux profile RSG-GAS research reactor was performed, aiming at an optimal transmutation loading for present nuclear energy development. The MA selected in the analysis includes Am-241 discharged from pressurized water reactors (PWRs) in Asian region. Until recently, studies have been undertaken in various methods to reduce radiotoxicity from actinides in high-level waste. From the cell calculation using computer code SRAC2006, it is obtained that the target Am-241 which has a cross section of the thermal energy absorption in the region (group 8) is relatively large; it will be easily burned in the RSG-GAS reactor. Minor actinides of Am-241 which can be inserted in the fuel (B/T fuel) is 2.5 kg which is equivalent to Am-241 resulted from the partition of spent fuel from 2 units power reactors PWR with power 1000MW(th) operated for one year.

AIP Conference Proceedings, 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.