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Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega
Published by Badan Tenaga Nuklir Nasional
ISSN : 1411240X     EISSN : 25279963     DOI : -
Core Subject : Science,
Energy Materials Science & Nanotechnology
Jurnal Teknologi Reaktor Nuklir "TRI DASA MEGA" adalah forum penulisan ilmiah tentang hasil kajian, penelitian dan pengembangan tentang reaktor nuklir pada umumnya, yang meliputi fisika reaktor, termohidrolika reaktor, teknologi reaktor, instrumentasi reaktor, operasi reaktor dan lain-lain yang menyangkut reaktor nukli. Frekuensi terbit tiga (3) kali setahun setiap bulan Februari, Juni dan Oktober.
Arjuna Subject : -
Articles 225 Documents
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CRITICAL HEAT FLUX NANOFLUIDS MEASUREMENTS SYSTEM USING ARDUINO Santiko Tri Sulaksono; Sudjatmi Kustituantini Alfa; Dani Gustaman Syarif
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 1 (2021): FEBRUARY 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.1.6005

Abstract

Crtical heat flux (CHF) is an important characteristic of nanofluids. The CHF measurements were carried out in nanofluid research at the Center for Applied Nuclear Science and Technology. These measurements are done manually using a variable power supply and a multimeter. However, it was difficult to record the voltage and current due to the sudden break of the wire. In this study, Arduino was used to measure CHF automatically. The voltage is applied to the wire and increases automatically along with the measurement of the voltage and current in the wire. The results of the voltage and current measurements were compared with a multimeter and were not significantly different. It can be concluded that the CHF measurement system using arduino can be used to measure nanofluid CHF.
TRANSIENT ANALYSIS OF SIMULTANEOUS LOFA AND RIA IN RSG-GAS REACTOR AFTER 32 YEARS OPERATION Muhammad Darwis Isnaini; Iman Kuntoro; Muhammad Subekti
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 22, No 3 (2020): OCTOBER 2020
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2020.22.3.5944

Abstract

During the operation of the research reactor RSG-GAS, there are many design parameters should be verified based on postulated accidents. Several design basis accidents (DBA) such as loss of flow accident (LOFA) and reactivity-initiated accident (RIA) also have been conducted separately. This paper discusses about possibility of simultaneous accidents of LOFA and RIA. The accident analyses carry out calculation for transient condition during RIA, LOFA, and postulated accident of simultaneous LOFA-RIA. This study aims to conduct a safety analysis on simultaneous LOFA and RIA, and investigate the impact on safety margins. The calculations are conducted by using the PARET code. The maximum temperature of the center fuel meat at nominal power of 30 MW and steady state conditions is 126.10°C and MDNBR of 2.94. At transients condition, the maximum center fuel meat temperature for LOFA, RIA and simultaneous LOFA-RIA are consecutively 132.99°C, 135.67°C and 138.21°C, and the time of reactor trip are 3.2593s, 3.6494s and 2.7118s, respectively. While the MDNBR for LOFA, RIA and simultaneous LOFA-RIA are respectively at transient condition are 2.88, 2.58 and 2.63, respectively. It is shown that, simultaneous LOFA-RIA has the fastest trip time. In this case, the low flow trip occurs first in advance to over power trip.  From these results, it can be concluded that the RSG-GAS has adequate safety margin against transient of simultaneous LOFA-RIA.Keywords: RSG-GAS, Simultaneous, LOFA, RIA, PARET
EVALUATION OF EQUILIBRIUM CORE OPERATION OF THE RSG-GAS REACTOR Iman Kuntoro; Surian Pinem; Tagor Malem Sembiring; Dwi Haryanto; Sigit Purwanto
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 1 (2021): FEBRUARY 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.1.6150

Abstract

The Indonesian Multipurpose Reactor, RSG-GAS reactor will accomplish its first lifetime in December 2020. The reactor has been operated in safe and reliable manner for about 33 years since it commenced in operation in 1987 to serve radioisotopes production, NAA, neutron beam experiments, material irradiation, and reactor physics experimental activities as well as training. The paper is intended to evaluate its in-core fuel management that is the conformance between the theory and implementation of the equilibrium core. Evaluation of the reactor operation parameter was carried out for core numbers 91 – 100. The data show that excess reactivity, shutdown reactivity and control rod reactivity have no significant difference at each core. The result shows that the BATAN-FUEL accurately determine the equilibrium core and its fuel loading pattern.This in-core fuel management of the RSG-GAS reactor supports the safety of reactor operation.
INFORMATION PROCESSING IN THE REACTOR PROTECTION SYSTEMS OF HIGH TEMPERATURE GAS-COOLED REACTORS Tulis Jojok Suryono; Sudarno Sudarno; Sigit Santoso
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 22, No 3 (2020): OCTOBER 2020
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2020.22.3.5947

Abstract

Reactor protection systems (RPS) transform process variable signals from the sensors into initiation and actuation signals to trip the reactor if the signal's value exceeds the predefined trip setpoints of the RPS. Information on the current value of the process variables signals and the trip setpoint should be displayed properly on the visual display unit (VDU) in order to maintain the situation awareness of the operators in main control rooms (MCR). In addition, it is also helpful for them to investigate the cause of an accident after the reactor trip and to mitigate the accident based on the appropriate emergency operating procedures. This paper investigates how the information is processed in the RPS of Experimental Power Reactor (EPR) based on high temperature reactor (HTR) technology, and how the information is displayed on the human machine interface (HMI) of the MCR of the EPR. It is conducted by classifying the RPS into three layers based on its components and their functions, followed by the investigation of the type and the information processing in each layer. The results show that the form of the information has been changed throughout the RPS, started from the sensors and until it is displayed on the VDU. The results of the investigation are necessary to understand the concept of RPS, especially for new operators, and to prepare the mitigation actions based on the process variable that cause the reactor trip.Keywords: Experimental power reactor, Reactor protection system, Human machine interface, Information processing, Situation awareness
RADIATION DOSE OPTIMIZATION OF BREAST CANCER WITH PROTON THERAPY METHOD USING PARTICLE AND HEAVY ION TRANSPORT CODE SYSTEM Milah Fadhilah Kusuma Fasihu; Andang Widi Harto; Isman Mulyadi Triatmoko; Gede Sutrisna Wijaya; Yohannes Sardjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6290

Abstract

Radiotherapy is one of the cancer treatments conducted by giving a high dose to the tumor target and minimizing the dose exposed in the healthy organs. One of the methods is proton therapy. Proton therapy is usually used in several breast cancer cases by minimizing the damage in the surrounding tissues due to having good precision. In this study, proton therapy in breast cancer will be simulated. This study aims to identify the optimal dose in breast cancer therapy using proton therapy and to identify the dose exposed in the healthy organs surrounding cancer. This study is PHITS program simulation-based to model the geometry and the components of breast cancer and the surrounding organs. The source of radiation used is proton which is the output of proton therapy with proton/sec firing intensity. The variation in beam modelling towards the dose profile of the tumor used is uniform and pencil beam. The proton energy used is 70 MeV up to 120 MeV. The result of this study shows that the dose from using pencil beam scanning technic of proton therapy for breast cancer is 50.3997 Gy (W) with the total amount of fraction is 25 and the result of dose below the threshold dose in the healthy organs is the skin gets 4.4.0553 Gy per fraction, the left breast gets 0,0011 Gy per fraction, the right breast gets 2.6469 Gy per fractions, the right lung gets 0.0125 Gy per fraction, the left lung gets 0.029 Gy per fraction, the rib gets 0.0179 Gy per fraction, and the heart gets 0.0077 Gy per fraction.
PRELIMINARY ASSESSMENT OF ENGINEERED SAFETY FEATURES AGAINST STATION BLACKOUT IN SELECTED PWR MODELS Andi Sofrany Ekariansyah; Surip Widodo; Susyadi Susyadi; Hendro Tjahjono
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6204

Abstract

The 2011 Fukushima accident did not prevent countries to construct new nuclear power plants (NPPs) as part of the electricity generation system. Based on the IAEA database, there are a total of 44 units of PWR type NPPs whose constructions are started after 2011. To assess the technology of engineered safety features (ESFs) of the newly constructed PWRs, a study has been conducted as described in this paper, especially in facing the station blackout (SBO) event. It is expected from this study that there are a number of PWR models that can be considered to be constructed in Indonesia from the year of 2020. The scope of the study is PWRs with a limited capacity from 900 to 1100 MWe constructed and operated after 2011 and small-modular type of reactors (SMRs) with the status of at least under licensing. Based on the ESFs design assessment, the passive core decay heat removal has been applied in the most PWR models, which is typically using steam condensing inside heat exchanger within a water tank or by air cooling. From the selected PWR models, the CPR-1000, HPR-1000, AP-1000, and VVER-1000, 1200, 1300 series have the capability to remove the core decay heat passively. The most innovative passive RHR of AP-1000 and the longest passive RHR time period using air cooling in several VVER models are preferred. From the selected SMR designs, the NuScale design and RITM-200 possess more advantages compared to the ACP-100, CAREM-25, and SMART. NuScale represents the model with full-power natural circulation and RITM-200 with forced circulation. NuScale has the longest time period for passive RHR as claimed by the vendor, however the design is still under licensing process. The RITM-200 reactor has a combination of passive air and water-cooling of the heat exchanger and is already under construction.  
ENVIRONMENTAL CONSEQUENCES OF ROUTINE RELEASES FROM SMALL MEDIUM REACTOR AT BABEL SITE Udiyani Made Pande; Muhamad Budi Setiawan; Anik Purwaningsih; Nursinta Adi Wahanani; Muksin Aji Setiawan; Amir Hamzah; Hery Adrial; Jupiter Sitorus Pane
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6239

Abstract

Radiation protection and safety documents for routine conditions are required to support the licensing requirements for nuclear power plant site. This research is focused in the assessment and analysis of the results of PWR safety study related to the routine release of radioactivity from the SMR subsystems and components of the 100 MWe-type PWR along with its consequences in the site. The core inventory calculation was done using  ORIGEN2 software, applying release parameters from the existing analysis and calculation results. The radiological consequences were calculated by the PC-CREAM program package. Environmental and meteorological data were obtained using Arc-GIS and spatial analysis. The Bangka Belitung (Babel) site was used as the specific footprint. Analyzing PC-CREAM output data the radiological consequences of routine operation of 3 100 MWe PWR modules on Sebagin site (South Bangka) and Muntok site (West Bangka) in 16 sectors and within a radius of 20 km were concluded. The calculation results for the Sebagin site is that the maximumdose within a radius of 500 m (exclusion zone) is 1.15E+02 µSv/year. For a radius beyond 500 m, the maximum dose is 4.71E+01 µSv/year. Whereas for Muntok site (West Bangka), the maximum dose in the exclusion area (<500m) is 9.47E+00 µSv/year, and outside exclusion area (>500m) is 3.10E+00 µSv/year. The individual dose for the Babel site in the exclusion area is below the dose constraint for non-radiation service workers as the general public of 0.3 mSv/year or 300 µSv/year, while the maximum dose for outside exclusion is also below the constraint as stipulated in BAPETEN Regulation No 4 Year 2013 on Radiation Protection and Safety.
CALCULATION OF RADIOACTIVE SOURCE TERM RELEASE FROM FLEXBLUE NPP Muhammad Budi Setiawan; Pande Made Udiyani
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6254

Abstract

One of the National Research Programs (PRN) in the energy sector of the Indonesian Ministry of Research and Technology for the period of 2020-2024 is small modular reactor (SMR) nuclear power plant (NPP) assessment. The France’s Flexblue is a PWR-based SMR submerged reactor with a power of 160 MWe. The Flexblue reactor module was built on the ocean site and easily provided the supply of reactor modules, in accordance with the conditions of Indonesia as an archipelagic country. Therefore, it is necessary to know the release of fission products (source term), which is necessary for the study of the radiation safety of a nuclear reactor. This paper aims to examine the source term in normal operating conditions and abnormal normal operating conditions, as well as postulated accidents. Based on the Flexblue reactor core parameter data, the calculation of the reactor core inventory uses the ORIGEN2 software is previously evaluated. The source term calculation uses a mechanistic approach and a graded approach. The normal source term is calculated assuming the presence of impurities on the fuel plate, due to fabrication limitations. Meanwhile, the abnormal source term is postulated in the LOCA event. The core reactor inventory and source term is divided into 8 radionuclide groups which are Noble gasses group (Xe, Kr); Halogen (I); Akali Metal (Cs, Rb); Tellurium Group (Te, Sb, Sc); Barium-Strontium Group (Ba, Sr); Noble Metals (Ru, Rh, Pd, Mo, Tc, Co); Lanthanides group (La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am) and Cerium Group (Ce, Pu , Np).
ANALYSES OF NEUTRON ABSORBER MATERIALS ON THE SAFETY PARAMETERS IN THE RSG-GAS REACTOR Lily Suparlina; Tukiran Surbakti; Surian Pinem; Purwadi Purwadi
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 2 (2021): June 2021
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.2.6278

Abstract

Shutdown system in RSG-GAS reactor is using neutron absorber. There are 3 kinds of absorber material in research reactors including Ag-In-Cd alloy, B4C, and Hf. In this works, analyses of different neutron absorbers on the main safety core parameters in the RSG-GAS research reactor are selected for analyses. Their integral effects on the main neutronic core parameters important to safety issues are investigated. These parameters are core excess reactivity, shutdown margin, total reactivity worth of control rods, PPF and neutron flux . The RSG-GAS core silicide fuel is selected as the case study to verify calculations. A three-dimensional, four-group diffusion model is selected for core calculations. The well-known WIMSD-5B and Batan-3DIFF reactor codes are used to carry out these calculations. It is found that the largest shutdown margin is gained using the B4C; also the lowest PPF is gained using the Hf material. The maximum point power densities belong to the inside fuel regions surrounding the CIP (centre irradiation position), surrounded by control fuel elements, and the peripheral fuel elements beside the berrylium reflector. The greatest and least fluctuation of the point power densities are gained by using B4C and Ag-In-Cd alloy, respectively.
PARTICLE SWARM OPTIMIZATION BASED PROBABILISTIC NEURAL NETWORK FOR CLASSIFICATION OF SEVERE ACCIDENT OF NUCLEAR REACTOR Yoyok Dwi Setyo Pambudi
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 23, No 3 (2021): October (2021)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2021.23.3.6247

Abstract

Due to its danger and complexity, the identification and prediction of major severe accident scenarios from an initiating event of a nuclear power plant remains a challenging task. This paper aims to classify severe accident at the Advanced Power Reactor (APR) 1400, which includes the loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR) using a standard  probabilistic neural network (PNN)  and Particle Swarm Optimization Based Probabilistic Neural Network (PSO PNN). The algorithm has been implemented in MATLAB.  The experiment results showed that supervised PNN PSO could classify severe accident of nuclear power plant better than the standar PNN.

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