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INDONESIA
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
 18   19   20   21   22 
ANALYSIS OF THE PPF VALUE DEPENDENCE ON THE FUEL BURNUP Lily Suparlina; Purwadi Purwadi; Kunihiko Nabeshima
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 2 (2022): June 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.2.6616

Abstract

The RSG-GAS reactor has been operated in a safe and reliable manner for about 35 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. PPF value is necessary to determine by calculation because it is impossible to determine by experiment and also has a strong relation to the operation safety. The paper is intended to analyze the PPF values of the RSG-GAS reactor core as a function of burn up. The analysis is using WIMSD-5B/BATAN-3DIFF computer codes calculation. The result shows that the PPF values are significantly different for each burn-up or energy in MWD. The result also shows that the BATAN-3DIFF code accurately determines the PPF values of the RSG-GAS reactor core and supports the safety of reactor operation.
ANALYSIS OF THORIUM PIN CELL BURN UP OF THE PWR USING WIMS CODE Jonny Haratua Panggabean; Santo Paulus Rajagukguk; Syaiful Bakhri
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 2 (2022): June 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.2.6626

Abstract

A thorium-fueled benchmark comparison was made in this study between state-of-the-art codes, WIMSD-5B code to MOCUP (MCNP4B + ORIGEN2), and CASMO-4 for burnup calculations as part of efforts to examine the possible benefits of using thorium in PWR fuel. WIMSD-5B calculations employ the same model as a reference, MOCUP, and CASMO, however, there are some variances in methodology and cross-section libraries. On a PWR pin cell model, eigenvalue and isotope concentrations were examined up to high burnup. The eigenvalue comparison as a function of burnup is good, with a maximum difference of less than 5% and an average absolute difference of less than 1%. The isotope concentration comparisons outperform a set of ThO2-UO2 fuel benchmarks and are comparable to a set of uranium fuel benchmarks previously published in the literature. As a function, the eigenvalue comparison The actinide and fission product data sources for a typical thorium fuel are reported in the WIMSD-5B burnup calculations. The reasons for discrepancies in coding are examined and explored.Keywords: Thorium, PWR Fuel, Burn up, Pin Cell, WIMSD-5B   
RISK ASSESSMENT ON THE DECOMMISSIONING STAGE OF INDONESIAN TRIGA 2000 RESEARCH REACTOR Ratih Luhuring Tyas; Deswandri Deswandri; Dinnia Intaningrum; Julwan Hendry Purba
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 2 (2022): June 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.2.6632

Abstract

Decommissioning is the final stage of a nuclear reactor. In preparing the decommissioning plan, one of the important elements that need to be considered is safety assessment. During decommissioning, there are many complex tasks to be done where the radiological and non-radiological hazards arise and can significantly affect not only the workers but also the general public and the environment. Indonesia has no experience with nuclear reactor decommissioning, so it is necessary to study various experiences of decommissioning activities in the world. This study proposes a framework to implement the safety assessment on the decommissioning of the TRIGA 2000 research reactor. The framework was developed on desk-based research and analysis. The proposed framework involves the facility and decommissioning activities, hazard identification, hazard analysis, hazard evaluation, hazard or risk control, and independent review.
ASSESSMENT OF OPERATION SAFETY OF THE RSG-GAS REACTOR TO SERVE RADIOISOTOPE TARGET IRRADIATION Iman Kuntoro; Lily Suparlina; Purwadi Purwadi
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 2 (2022): June 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.2.6634

Abstract

The RSG-GAS multipurpose reactor is operated to serve the utilization in the field of radioisotope production and NAA, material research. The reactor actually has power of 30 MW thermal, but upon considerations of efficiency and of most users requirements, the reactor is mostly operated at the power of 15 MW thermal, 5 days a week to produce a primary radioisotope from target of 2 grams U-235. To guarantee the safe operation and optimum utilization, a safety procedure was established. The paper is intended to assesst the operation safety in serving radioisotope target irradiation at its cycle operation. Assessment was carried out for core numbers 102 – 105. The result shows that excess reactivity and shutdown margin reactivity are safe to provide the target irradiation in the core for each cycle operation. 
ASSESSMENT OF RADIOLOGICAL IMPACTS FROM POSTULATED ACCIDENT CONDITIONS OF HTGR: A CASE STUDY IN SERPONG NUCLEAR AREA Muhammad Budi Setiawan; Ihda Husnayani; Heni Susiati
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 2 (2022): June 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.2.6613

Abstract

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 collective doses. The results showed that the highest dose will be received by the community within a radius of 250 m to the south from the reactor. It was also found that these accidents only cause minor radiological impacts which do not meet the criteria for any countermeasures (iodine thyroid blocking, sheltering, evacuation, food ban, decontamination, and relocation).
COLLISION CASCADE AND PRIMARY RADIATION DAMAGE IN SILICON CARBIDE: A MOLECULAR DYNAMICS STUDY Ihda Husnayani; Muzakkiy Putra Muhammad Akhir
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 3 (2022): October 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.3.6702

Abstract

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 960.000 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 defects were larger than the number of Si point defects. The number of stable point defect was found to be temperature-dependent. It was also obtained that the recovery of point defects was larger at high temperature (>800 C).
PREDICTION OF AP1000’S NUCLEAR REACTOR PRESSURE VESSEL TEMPERATURE DURING NORMAL OPERATION Muhammad Darwis Isnaini; Elfrida Saragi; Veronika Indriati Sri Wardani
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 3 (2022): October 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.3.6684

Abstract

Modeling of thermal-hydraulic calculations for the AP1000 core to predict the reactor pressure vessel (RPV) temperature has been carried out. The reactor’s primary coolant system transfers the heat produced in the reactor fuel during reactor operation to the steam generator. Part of the heat will also be transferred from the coolant to the reactor vessel and the pipe. This paper presents the calculation result of the RPV temperature prediction during AP1000 normal operation. Calculations were performed using COBRA-EN code for analyzing the core thermal hydraulics and using analytics for predicting the RPV temperature. These methods were carried out with the aim to predict the RPV temperature as well as at steady state nominal power conditions, at the function of flow, and at power fluctuation conditions. The calculation results at nominal power 3400 MWt (100% heat generated in fuel was assumed) and thermal design flow with 10% tube plugging (TDF2) of 48,443.7 ton/hr, for the minimum system pressure of 15.1 MPa, nominal system pressure of 15.513 MPa, and design system pressure of 17.133 MPa, show that the core outlet coolant temperature is 326.96°C, 327.01°C, and 327.22°C, and the RPV temperature is 303.65°C, 303.87°C, and 306.67°C, and the minimum departure from nucleate boiling ratio (MDNBR) is 3.21, 3.29, and 3.01, respectively. During reactor operation at a fixed nominal power of 3400 MWt, nominal system pressure, and under the condition of flow fluctuation, the maximum RPV temperature is shown to be 303.87°C.
ESTIMATION OF NEUTRON AND PROMPT PHOTON DOSE RATE DISTRIBUTION IN TMSR-500 USING MCNP6 Luqman Satria Pradana; Utari Utari; Suharyana Suharyana; Azizul Khakim
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 3 (2022): October 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.3.6692

Abstract

Thorium Molten Salt Reactor-500 (TMSR-500), one of the Generation IV nuclear reactors, is designed by Thorcon International, Pte. Ltd, which is projected to be built in Indonesia. The reactor core is radially surrounded by B4C shielding, but not the upper part. As the silo hall sits above the reactor core and is accessible by reactor personnel, the dose rate must be calculated in the area to ensure the workers receive an annual dose below the acceptable limit. The dose rate from neutrons and photons as the result of fission reactions are the only sources to be calculated in this research, without taking the source from fission products into account. This research aims to obtain the dose rate distribution of neutrons and prompt photons using Monte Carlo code MCNP6. The reactor was assumed to operate at a nominal thermal power of 557 MWth. Dose rate calculation was obtained from flux Tally F4 and converted into dose rate using Dose Energy Dose Function (DEDF) factor. Conversion factors of flux to the dose were based on ICRP-21 and ANSI/ANS-6.1.1 1977. The result of the calculations showed that the distribution of neutron and prompt photon fluxes does not reach the silo hall.
NEUTRONIC ANALYSIS OF THE VVER-1200 LATTICE CELL FUEL USING WIMSD-5B CODE Santo Paulus Rajagukguk; Syaiful Bakhri; Ana Muliyana; Juniastel Rajagukguk
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 3 (2022): October 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.3.6697

Abstract

The calculation of safety parameters in nuclear reactors has an important influence on nuclear reactor control and safety. The infinite multiplication factor, reactivity coefficients, and power peaking factor parameters are the most important safety parameters for determining reactor status. The aim of the present study is to analyze the behavior of the nuclear safety parameters for the VVER-1200 core in a normal state of reactor operation. A lattice cell fuel model of the VVER-1200 reactor core was performed using WIMSD-5B. The cross-section library data based on the ENDF/B-VIII.0 was used. The investigated parameters were the value of infinite multiplication factor with different pitch, temperature, enrichment, and boron concentration.  The calculation also investigated the reactivity coefficient parameters. The verification of WIMS model VVER-1200 was performed by comparing the results of the WIMSD-5B code with VVER-1200 data in the SAR document, and it was implied that they are in good agreement. The calculated values of reactivity coefficients illustrated a safe behavior.
MICROCONTROLLER ATMEGA328P TIMER/COUNTER FOR SINGLE CHANNEL GAMMA SPECTROSCOPY Santiko Tri Sulaksono; Putu Sukmabuana; Nanda Nagara
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 3 (2022): October 2022
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.3.6699

Abstract

Soil contamination may occur in the upcoming decommissioning activities of the TRIGA2000 Reactor. Measurement of contaminant radioactivity, which can be performed using single-channel spectroscopy, is required in soil decontamination processes. This research develops a timer/counter system for single-channel spectroscopy using a microcontroller. The performance of the ATMega328P microcontroller Timer/Counter on Arduino has been tested for single-channel spectroscopy. Microcontroller's Timer/Counter1 is used as a counter while Timer/Counter2 is used as a timer. Tests include the linearity test, comparative test, and chi-square test. The test results show that the ATMega328P microcontroller Timer/Counter works well and can be used as the end of a single-channel spectroscopic system.

Page 20 of 23 | Total Record : 225

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