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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
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MAP OF RADIOISOTOPE PRODUCTION AND BATAN RESEARCH REACTOR UTILIZATION Endiah Puji Hastuti; Iman Kuntoro; Suwoto Suwoto; Syarip Syarip; Prasetyo Basuki; Tukiran Surbakti; Geni Rina Sunaryo; Sudarmono Sudarmono
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.2021.23.3.6288

Abstract

Currently, Indonesia through BATAN is operating three research reactors, namely the RSG-GAS reactor with the power of 30 MWt at Puspiptek south Tangerang (the first criticality in 1987), the TRIGA 2000 reactor with the power of 2 MW in Bandung which the first criticality in 1965 with the power of 250 kW, was increased to 1 MW in 1971, and further upgraded to 2 MW in 2000. Beside that, there is Kartini reactor with a power of 100 kW located in Yogyakarta (first criticality in 1979). These reactors are quite old, and in accordance with Bapeten regulations, have carried out the first periodic safety review, to obtain a reactor license for the next 10 years of operation. In line with this, one of BATAN's current national research programs is to increase the production of radioisotopes and radiopharmaceuticals, where reactors play a very important role in the production of certain isotopes. In tracing the data obtained from operational reports related to irradiation requests from reactor users, namely PTRR, PSTNT, and PT INUKI for radioisotope production, which has been carried out in the last 5 years, May 2015 until 25 August 2020, show that the irradiation request at RSG-GAS is still not optimal. In term of the utilization of RSG-GAS, it can still be optimized, which in this case needs to be balanced with post-irradiation processing capabilities. Meanwhile, from the results of tracing and data collection, it can be shown that at this time the reactors are still operating. The utilization activities of the reactors complement each other according to their age and facilities.
STUDY ON THERMAL CHARACTERISTICS OF U-SHAPED HEAT PIPE Mukhsinun Hadi Kusuma; Anhar Riza Antariksawan; Giarno Giarno; Dedy Haryanto; Surip Widodo
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.2021.23.3.6334

Abstract

The latest accident in Japan's nuclear power station became a valuable experience to start engaging passive cooling systems (PCS) more aggressively to improve safety aspects in nuclear power reactors being studied in Indonesia. This investigation is related to the U-shaped heat pipe (UHP) research as PCS of water in the cooling tank (CT). The objective of this research is to study the thermal characteristics of UHP as PCS in the CT. The experiment on small-scale UHP and simulation with RELAP5 code has been conducted to understand the performance of UHP. The experiment results of the small-scale UHP model will be used as a basic understanding of simulating and designing a UHP with big scaling. The study result showed the highest thermal performance of UHP was obtained when it operated on the higher temperature of heat load and higher air cooling velocity. The more UHPs inserted into the cooling pool, the more heat that can be discharged into the environment. This result also shows promising use of UHP for CT PCS. The use of UHP as PCS can enhance the safety aspect of the nuclear reactor, especially in station blackout event.
CRITICALITY SAFETY ANALYSIS OF THE DRY CASK DESIGN WITH AIR GAPS FOR RDNK SPENT PEBBLE FUELS STORAGECriticality Safety Analysis of the Dry Cask Design with Air Gaps for RDNK Spent Pebble Fuels Storage Pungky Ayu Artiani; Yuli Purwanto; Aisyah Aisyah; Ratiko Ratiko; Jaka Rachmadetin; Kuat Heriyanto
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.2021.23.3.6355

Abstract

Reaktor Daya Non-Komersial (RDNK) with a 10 MW thermal power has been proposed as one of the technology options for the first nuclear power plant program in Indonesia. The reactor is a High Temperature Gas-Cooled Reactor-type with spherical fuel elements called pebbles. To support this program, it is necessary to prepare dry cask to safely store the spent pebble fuels that will be generated by the RDNK. The dry cask design has been proposed based on the Castor THTR/AVR but modified with air gaps to facilitate decay heat removal. The objective of this study is to evaluate criticality safety through keff  value of the proposed dry cask design for the RDNK spent fuel. The keff  values were calculated using MCNP5 program for the dry cask with 25, 50, 75, and 100% of canister capacity. The values were calculated for dry casks with and without air gaps in normal, submerged, tumbled, and both tumbled and submerged conditions. The results of calculated keff  values for the dry cask with air gaps at 100% of canister capacity from the former to the latter conditions were 0.127, 0.539, 0.123, and 0.539, respectively. These keff values were smaller than the criticality threshold value of 0.95. Therefore, it can be concluded that the dry cask with air gaps design comply the criticality safety criteria in the aforementioned conditions.
PRELIMINARY DEVELOPMENT OF RADIONUCLIDES RELEASE OF INDIVIDUAL DOSE CODE PROGRAM FOR RADIATION MONITORING PURPOSES Jupiter Sitorus Pane; Pande Made Udiyani; Muhammad Budi Setiawan; Surip Widodo; I Putu Susila
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.2021.23.3.6240

Abstract

Environmental radiation monitoring is one of the important efforts in protecting society and the environment from radiation hazards, both natural and artificial. The presence of three nuclear research reactors and plans to build a nuclear power plant reactor prompted Indonesia to prepare a radiation monitoring system for safety and security (SPRKK). The goal of the study is to provide an appropriate method for developing radiation monitoring system to support the development of nuclear power plant in the near future.  For this preliminary study, the author developed a code program using Gaussian distribution model approach for predicting radionuclide release and individual dose acceptancy by human being within 16 wind directions sectors and up to 50 km distance. The model includes estimation of source term from the nuclear installation, release of radionuclides source into air following Gaussian diffusion model, some of the release deposit to the land and entering human being through inhalation, direct external exposure, and resuspension, and predicted its accepted individual dose. This model has been widely used in various code program such as SimPact and PC-Cosyma. For this study, the model will be validated using SimPact code program. The model has been successfully developed with less than 5% deviation.   Further study will be done by evaluating the model with real measuring data from research reactor installation and prepare for interfacing with real time radiation data acquisition and monitoring as part of radiation monitoring system during normal and accident condition.
MEASURED AND CALCULATED INTEGRAL REACTIVITY OF CONTROL RODS IN RSG-GAS FIRST CORE Wahid Luthfi; Surian Pinem; Donny Hartanto; Lily Suparlina; Dwi Haryanto
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6593

Abstract

The control rod worth is one of the important parameters for the operation of a nuclear reactor. Proper measurement and calculation of the control rod worth are essential for the safe reactor operation under normal and transient conditions that are initiated by a postulated event such as a stuck rod, control rods ejection, etc. This paper presents calculation results of integral reactivity of the RSG-GAS research reactor first core and its comparison with the experimental data. Calculations were performed using the continuous energy transport code Serpent 2 with ENDF/B-VIII.0 nuclear data. Integral reactivity measurement was done by compensating method with control rod bank, regulating rod, and reactivity meter. Calculations are carried out for each method used in control rod measurement data with an aim to validate calculated results to experimental data. Compared with the measured experiment data, there are no significant differences in calculation results of integral reactivity. The maximum difference of the control rod's total reactivity is 1.26% compared to the measurement carried out by compensating method with regulating rod.
ANALYSIS OF COGENERATION ENERGY CONVERSION SYSTEM DESIGN IN IPWR REACTOR Ign. Djoko Irianto; Sriyono Sriyono; Sukmanto Dibyo; Djati Hoesen Salimy; Tukiran Surbakti; Rahayu Kusumastuti
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6414

Abstract

The acceleration of national development, especially in the industrial sector, requires an adequate national energy supply. There are various types of energy sources which include conventional energy sources as well as new and renewable energy sources including nuclear energy. The problem is how to utilize these energy sources into energy that is ready to be utilized. BATAN as a research and development institution in the nuclear field has taken the initiative to contribute to the development of technology for providing electricity and other thermal energy, particularly reactor technology as a power plant and a provider of thermal energy. This research aims to analyze the design of the IPWR type SMR reactor cogeneration energy conversion system. The IPWR reactor cogeneration energy conversion system which also functions as a reactor coolant is arranged in an indirect cycle configuration or Rankine cycle. Between the primary cooling system and the secondary cooling system is mediated by a heat exchanger which also functions as a steam generator. The analysis was carried out using ChemCAD computer software to study the temperature characteristics and performance parameters of the IPWR reactor cogeneration energy conversion system. The simulation results show that the temperature of saturated steam coming out of the steam generating unit is around 505.17 K. Saturated steam is obtained in the reactor power range between 40 MWth to 100 MWth. The results of the calculation of the energy utilization factor (EUF) show that the IPWR cogeneration configuration can increase the value of the energy utilization factor up to 91.20%.
DOSE DISTRIBUTION ANALYSIS OF PROTON THERAPY FOR MEDULLOBLASTOMA CANCER WITH PHITS 3.24 Moh. Miftakhul Dwi Fianto; Yohannes Sardjono; Andang Widi Harto; Isman Mulyadi Triatmoko; Gede Sutresna Wijaya; Yaser Kasesaz
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6581

Abstract

One of the developments in particle therapy is proton radiation therapy. Meanwhile, a limited number of available proton therapy facilities makes research related to proton therapy difficult. Therefore, there is a need for alternative proton therapy simulations using programs other than those in proton therapy facilities. This research was aimed to simulate medulloblastoma brain cancer which children often experience.The program used in this research was PHITS version 3.24. The human body was modeled with the revised ORNL-MIRD phantom for a 10-year-old child. The therapy scheme was a whole posterior fossa boost of 19.8 Gy. The proton passive scattering was simulated by passing a uniform proton beam through the aperture and compensator with energy variations. The proton pencil beam scanning was simulated with small cylindrical beams with a radius of 0.5 cm, which were adjusted to the planning target volume with layers variations.The total duration to give the prescription dose was 550 seconds with passive scattering and 605 seconds with pencil beam scanning. In passive scattering, the OAR(s) with the most significant percentage of absorbed dose were the skin, cranium, and muscle, i.e., 8.22 ± 0.15 %, 5.51 ± 0.05 % and 1.39 ± 0,04 % respectively to their maximum tolerated dose, while in the pencil beam scanning, the OAR(s) with the most significant percentage of absorbed dose were the skin, cranium, and muscle, i.e., 5.42 ± 0.08 %, 4.43 ± 0.05 % and 0.51 ± 0.05 % respectively to their maximum tolerated dose. Dose distribution in passive scattering was relatively better than in pencil beam scanning in terms of dose homogeneity using dose sampling analysis at some points within the planning target volume.
ANALYSIS OF FUEL TEMPERATURE REACTIVITY COEFFICIENT OF THE PWR USING WIMS CODE Santo Paulus Rajagukguk; Syaiful Bahkri; Tukiran Surbakti
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6329

Abstract

The Fuel Temperature Reactivity Coefficient (FTRC) is an important parameter in design, control, and safety, particularly in PWR reactor. It is then very important to validate any new library for an accurate prediction of this parameter. The objective of this work is to determine the value of the FTRC parameter using the new WIMDS library based on ENDF/BVIII.0 nuclear data files. For this purpose, it is used a set of light water moderated lattice experiments as the PWR-1175 MWe experiment critical reactors, the reactor using UO2 fuel pellet. The analysis is used with WIMSD-5B lattice code with original cross-section libraries and WIMSD-5B with ENDF/B-VIII.0 new cross-section libraries. The results showed that the fuel temperatures reactivity coefficients for the PWR reactor using original libraries is – 3.10 pcm/K with enrichment of 3.1% but for ENDF/B-VlII.0 libraries is – 3.00 pcm/K. Compared to the experimental data of the reactor core, the difference is in the range of 6.9 % for ENDF/B-VIII.0 libraries. It can be concluded that for the reactor, it is better to use ENDF/B-VIII.0 libraries because the original library is not accurate anymore.
Bagian Depan Vol.24 No. 1 (2022): February 2022 Editor in Chief
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar

Abstract

This section consists of Cover Page, Editorial Board Page, Peer Reviewer Page, Table of Content Page and Preface Page
PREDICTION OF REMAINING USEFUL LIFE FOR COMPONENTS IN SSC OF RSG-GAS BASED ON RELIABILITY ANALYSIS Entin Hartini; Endiah Puji Hastuti; Geni Rina Sunaryo; Aep Saepudin; Sri Sudadiyo; Amir Hamzah; Mike Susmikanti
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 24, No 1 (2022): February (2022)
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.17146/tdm.2022.24.1.6400

Abstract

In the maintenance system, efforts are needed to improve the effectiveness of the maintenance system and organization. For effective maintenance planning it is necessary to have a good understanding of the reliability and component availability of the system. For this reason, it is necessary to determine the remaining component life using Remaining Useful Life (RUL), so that maintenance tasks can be planned effectively. The purpose of this study is to determine the remaining life of the safety A component from SSC RSG-GAS based on reliability analysis. The method used in this paper is a statistical approach to estimating RUL. The Weibull hazard model is determined for modeling the hazard function so that it can be integrated in the reliability analysis. The model is verified using data from the safety A component from the SSC RSG-GAS. The results obtained from the analysis are useful for estimating the remaining useful lives of these components which can then be used to plan for effective maintenance and help control unplanned outages. The results obtained can be used for maintenance development and preventive repair planning.

Page 19 of 23 | Total Record : 225

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