cover
Article Per Year (5 Year)
All
Home Page
OAI Link
Editorial Team
Contact
Reviewer
Google Scholar
Contact Name
-
Contact Email
-
Phone
-
Journal Mail Official
jurtdm@batan.go.id
Editorial Address
Pusat Teknologi dan Keselamatan Reaktor Nukir (PTKRN) Badan Tenaga Nuklir Nasional (BATAN) Gedung 80 Kawasan Puspiptek Setu - Tangerang Selatan Banten - Indonesia (15310)
Location
Kota adm. jakarta selatan,
Dki jakarta
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
 19   20   21   22   23 
Computational Fluid Dynamics Simulation of Temperature Distribution and Flow Characterization in a New Loop Heat Pipe Model Restiawan, Muhammad Mika Ramadhani; Kusuma, Mukhsinun Hadi; Rozi, Khoiri; Kiono, Berkah Fajar Tamtomo; Yunus, Muhammad; Wirza, Alif Rahman; Pambudi, Yoyok Dwi Setyo; ButarButar, Sofia Loren; Giarno, Giarno; Hatmoko, Sumantri
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 26, No 2 (2024): June 2024
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2024.7054

Abstract

The loop heat pipe (LHP) is considered for passive cooling systems in nuclear installations. A combined approach of simulation and experimentation is essential for achieving comprehensive knowledge of the LHP. Research on the LHP using Computational Fluid Dynamics (CFD) is necessary to understand phenomena that are challenging to ascertain experimentally. This study investigates the temperature distribution and flow characterization in a new LHP model. The method used in this research is simulation using CFD Ansys fluent software. In the simulation, the LHP has an inner diameter of 0.1016 m. This LHP features a wick made from a collection of capillary pipes without a compensation chamber. Demineralized water is used as the working fluid with a filling ratio of 100% of evaporator volume. The hot water temperature in the evaporator section is set at 70°C, 80°C, and 90°C. The temperature on the outer surface of the condenser pipe is determined using experimental temperature inputs. An inclination angle of 5° and an initial pressure of 12,100 Pa was applied to LHP. The CFD simulation results show that the temperature distribution profile under steady-state conditions in the  loop heat pipe appears almost uniform. The temperature difference between the evaporator and condenser remains consistent. The flow of working fluid in the LHP is driven by buoyancy forces and fluid flow, allowing the working fluid in the LHP to flow in two phases from the evaporator to the condenser and then condensate from the condenser back to the evaporator. In conclusion, the temperature distribution and flow patterns in the LHP are consistent with common phenomena observed in heat pipes. This modeling can be used to determine the profiles of temperature distribution and flow in LHP of the same dimensions under various thermal conditions.
Insider Intervention Model in the Sabotage Attack Scenario of a Nuclear Reactor Facility Andiwijayakusuma, Dinan; Asmoro, Teguh; Mardhi, Alim; Setiadipura, Topan
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 26, No 1 (2024): February 2024
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2024.7008

Abstract

The Physical Protection System (PPS) at nuclear facilities aims to prevent intrusions into nuclear facilities that cause sabotage attacks or illegal theft of nuclear material. The outsider, the insider or a collaboration of both can carry out this intrusive action. In this study, we modelled the insider collaborating with the outsider to carry out nuclear facility attacks using sabotage attack scenarios. The modelling takes the form of insider intervention on two parameters protection elements:the time delay () and the probability of detection (). Insider intervention in delay protection elements might have fatal consequences and drastically reduce the effectiveness of PPS performance. Therefore, PPS designers need to pay more attention to the delay element to anticipate the negative impact of insider intervention on the effectiveness of the PPS.
The Study of Multiaxial Loading and Damage to the Structure and Materials in the PWR Steam Generator of Nuclear Reactor Subhan, Muhammad; Priyana Soemardi, Tresna; Setiadipura, Topan; Yogatama Sulistyo, Farisy; Subiyah, Hana
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 25, No 3 (2023): October 2023
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2023.6963

Abstract

In Pressurized Water Reactor (PWR) Nuclear Power Plants (NPPs), the steam generator is crucial for transferring heat from the primary to secondary cooling systems, vital for steam production to drive turbines, and central to nuclear power safety. This study explores recent research on multi-axial loading, structural integrity, and material durability in PWR steam generators, shedding light on key factors affecting these systems. Common corrosion-related degradation in steam generators often arises from design, material, and water chemistry factors. However, the shift to All Volatile Treatment (AVT), the development of advanced material alloys, and enhanced water quality control in primary and secondary systems have significantly reduced instances of steam generator degradation. These findings promise to enhance the reliability and safety of steam generators in future nuclear applications. 
Investigation of Natural Circulation Flow Under Steady-State Conditions Using a Rectangular Loop Roswandi, Iwan; Dimas, Dimas; Gunawan, Hyundianto Arif; Budiman, Arif Adtyas; Amelia, Almira Citra; Sanda, Sanda; Tjahjono, Hendro; Juarsa, Mulya
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 26, No 2 (2024): June 2024
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2024.7055

Abstract

Passive safety systems have garnered significant attention, particularly in situations where active systems fail. The comprehension of natural circulation phenomena plays a vital role in the advancement of passive cooling systems in nuclear power plants. The objective of this study is to examine the flow patterns under steady state conditions and assess the Grashof number. The experimental approach involved maintaining temperature differences of 60°C, 70°C, 80°C, and 90°C for a duration of 3 hours, with 3 replications. Alterations in temperature have an impact on the physical properties of water, such as density, viscosity, and specific heat. The calculations indicate that the minimum Grashof number occurs at 60°C (2.49×1012), while the maximum is observed at 90°C (9.42×1012), with an R2 value of 0.96533. Turbulent flow patterns were observed during each temperature fluctuation, which aligns with previous research on the Ress value of Grm/NG.
Techno-Economic Assessment and Optimization of a Standalone System in Sebira Island, Indonesia Farah, Laili; Akhmad, Yus Rusdian; Saryadi, Rezky Mahardika; Mardha, Amil; Mudjiono, Mudjiomo; Nuryanti, Nuryanti; Anzhar, Kurnia; Handayani, Airine Hijrah
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 26, No 1 (2024): February 2024
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2024.7020

Abstract

Nuclear power is known as a baseload generator in central power networks, but its implementation is too large-scale for microgrid applications. Nuclear power as a source of electricity is considered for microgrid applications due to its ability to produce emission-free energy. This research discusses the techno-economic analysis and optimization of a hybrid energy system design on Sebira Island, Indonesia, using a multi-year model in HOMER Pro software. Two scenarios were created: diesel-PV-battery and the second scenario, nuclear-PV-battery, with the baseline system being a diesel generator (DG) only. The research results show that with the optimal use of the nuclear-PV-battery system, the levelized cost of electricity (COE) is $0.128. This value is lower compared to the first scenario with a COE of $0.6577. The CO2 emissions generated in the optimal nuclear-PV-battery system are zero, making this system far more viable than other hybrid system schemes.
Transmutation of Transuranic Elements as Solid Coating in Molten Salt Reactor Fuel Channel Dwijayanto, R. Andika Putra; Miftasani, Fitria; Harto, Andang Widi
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 25, No 3 (2023): October 2023
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2023.6880

Abstract

The accumulation of spent nuclear fuel (SNF) is presently considered as a hindrance of the massive deployment of nuclear power plant, especially regarding the transuranic (TRU) elements. Eliminating TRU through transmutation is one of the most feasible alternative as a technical solution to solve the issue. This study explores the possibility of TRU transmutation using molten salt reactor (MSR) in a heterogeneous configuration, where a solid TRU is coated inside the fuel channel filled with liquid salt fuel. Such configuration is proposed to allow higher TRU loading into fluoride salt mixture without compromising the safety of the reactor. TRU coating was applied in consecutively outward radial fuel channel layers with coating thicknesses of 2.5 mm and 5 mm. Calculation was performed using MCNP6.2 radiation transport code and ENDF/B-VII.0 neutron cross section library. From the results, TRU coating with smaller thickness and positioned closer to the centre of the core exhibit higher transmutation efficiency due to exposure to higher neutron flux. Highest transmutation efficiency was achieved at 67.93% after 160 days of burnup. This shows a potential of achieving highly efficient TRU using heterogeneous configuration in MSR core.
Experimental Study of The Influences of Inclination Angle and Heat Load on Loop Heat Pipe Thermal Performance Pramesywari, Afifa; Kusuma, Mukhsinun Hadi; Kiono, Berkah Fajar Tamtomo; Rozi, Khoiri; Giarno, Giarno; Pambudi, Yoyok Dwi Setyo; Hatmoko, Sumantri; Emara, Haura
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 26, No 2 (2024): June 2024
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2024.7013

Abstract

The utilization of nuclear power brings out a lot of benefits in fulfilling human power needs, however, the thermal incident caused by the failure of an active cooling system because of an earthquake followed by the tsunami such as on the Nuclear Power Plant at Fukushima Dai-Ichi Japan could be taken for lesson learn to keep improve nuclear installation operation safety aspects. Loop heat pipe (LHP) as an alternative cooling system technology could be utilized to handle thermal problems on nuclear installations. This research aims to know the influence of the inclination angle and heat load on the LHP thermal performance. The experimental investigation was performed with varying the inclination angle of 0°, 2.5°, and 5°, and heat load given at 60°C, 70°C, 80°C, and 90°C. LHP was used demineralized water working fluid with a 100% filling ratio. LHP was vacuumed on 2.666,4 Pa. The cooling air velocity in the condenser given by 2,5 m/s. The result of this experiment showed that LHP has the best thermal performance with the lowest thermal resistance of 0.0043°C/W. This result was obtained when the LHP operated with a 5° inclination angle and hot water as the heat load of 90°C. The conclusion from this research is showing better LHP thermal performance as the inclination angle increase on LHP because the steam speed that formed bigger, and condensate flows back to the evaporator faster
Advancements in Accident Tolerance Fuel: A New Horizon in Nuclear Safety Ngarayana, I Wayan; Kurniawan, Rusbani; Rachman, Agus Nur; Nugraha, Eka Djatnika; Ekaranti, Egnes; Andani, Ika Wahyu
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 26, No 1 (2024): February 2024
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2024.7017

Abstract

Accident Tolerant Fuels (ATFs) represent a significant advancement in nuclear safety, offering the potential to mitigate the risks associated with nuclear reactor accidents. This paper provides a comprehensive overview of the development and current state of ATF technology, tracing its evolution and highlighting key technological milestones. Through an analysis of various case studies, we examine the practical application and performance of ATFs in real-world scenarios. Despite the promising capabilities of ATFs, their development and deployment are not without challenges. We delve into the technical, regulatory, and economic hurdles that must be overcome to realize the full potential of ATFs. Looking ahead, we explore the prospects of ATFs, discussing potential advancements and their implications for the nuclear industry. The findings of this paper underscore the transformative role of ATFs in enhancing nuclear reactor safety and charting a new horizon in nuclear technology.
Study of Alternative Radiation Material Shielding for Gamma Radiation using Monte Carlo Simulation Urfa, Gusti Atika; Wianto, Totok; Manik, Tetti Novalina; Nasrulloh, Amar Vijai
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 25, No 3 (2023): October 2023
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2023.6925

Abstract

Lead as the most commonly used material for radiation shielding but possessing toxic properties. This research aims to identify alternative, lead-free, and non-toxic materials for gamma radiation shielding through Monte Carlo simulations. Bismuth Oxide (Bi2O3), Barium Oxide (BaO), Tungsten Trioxide (WO3), Tungsten Dioxide (WO2), and Molybdenum Trioxide (MoO3) were selected as potential substitutes for lead. Pure lead (Pb) and Lead Oxide (PbO) were used for comparison. The simulation were performed using Particle Heavy Ion Tracking System (PHITS) software, with a gamma energy of 662 keV. The result of the simulation shows that the linear attenuation coefficient values for Pb and PbO were 0.902 mm-1 and 0.74 mm-1, respectively. Meanwhile, the simulation results of those simulated materials that are closest to Pb and PbO are Bi2O3 and WO2 with an attenuation coefficient of 0.71 mm-1. This simulation shows that for non-lead materials, BiO2 and WO2 have potential as alternative of non-lead radiation shielding.
Analysis of the Reactivity Coefficient of the PWR Thorium Fuel Rajagukguk, Santo Paulus; Purwadi, Purwadi; Bakhri, Syaiful
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 26, No 2 (2024): June 2024
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Show Abstract | Download Original | Original Source | Check in Google Scholar | DOI: 10.55981/tdm.2024.7032

Abstract

In design, control, and safety—especially in PWR reactors—the Reactivity Coefficient parameter is crucial. The validation of every new library for an accurate parameter prediction is then crucial. The purpose of this work is to determine the value of the reactivity coefficient at BOC and EOC using the WIMDS code based on ENDF/B-VIII.0 nuclear data files. The PWR-1175 MWe experiment critical reactors, which use Th-UO2 (thorium) fuel pellets, are a set of light water-moderated lattice experiments that are used for this purpose. The study is applied to the new cross-section libraries for WIMSD-5B and WIMSD-5B with ENDF/B-VIII.0 lattice code. The results showed that the fuel temperature reactivity coefficients for the PWR reactor at BOC and EOC using new libraries are – 4.07 pcm/K and – 2.72 pcm/K, respectively. Moderator Temperature Reactivity Coefficient at BOC and EOC are -1.8E-03 pcm/K and 3.73 pcm/K, respectively. Compared to the experimental data of the reactor core, the difference is in the range of 5.0 %. It can be concluded that for the PWR using thorium fuel as a model, all reactivity coefficients are negative and it is a good design for the safety of operation.

Page 22 of 23 | Total Record : 225

 19   20   21   22   23 

Filter by Year

2010 2024


Reset
Filter By Issues
All Issue Vol 26, No 2 (2024): June 2024 Vol 26, No 1 (2024): February 2024 Vol 25, No 3 (2023): October 2023 Vol 25, No 2 (2023): June 2023 Vol 25, No 1 (2023): February 2023 Vol 24, No 3 (2022): October 2022 Vol 24, No 2 (2022): June 2022 Vol 24, No 1 (2022): February (2022) Vol 23, No 3 (2021): October (2021) Vol 23, No 2 (2021): June 2021 Vol 23, No 1 (2021): FEBRUARY 2021 Vol 22, No 3 (2020): OCTOBER 2020 Vol 22, No 2 (2020): June 2020 Vol 22, No 1 (2020): February 2020 Vol 21, No 3 (2019): October 2019 Vol 21, No 2 (2019): JUNI 2019 Vol 21, No 1 (2019): February 2019 Vol 20, No 3 (2018): Oktober 2018 Vol 20, No 2 (2018): JUNI 2018 Vol 20, No 1 (2018): Februari 2018 Vol 19, No 3 (2017): Oktober 2017 Vol 19, No 2 (2017): Juni 2017 Vol 19, No 1 (2017): Februari 2017 Vol 18, No 3 (2016): Oktober 2016 Vol 18, No 2 (2016): Juni 2016 Vol 18, No 1 (2016): Februari 2016 Vol 17, No 3 (2015): Oktober 2015 Vol 17, No 2 (2015): Juni 2015 Vol 17, No 1 (2015): Pebruari 2015 Vol 16, No 3 (2014): Oktober 2014 Vol 16, No 2 (2014): Juni 2014 Vol 16, No 1 (2014): Pebruari 2014 Vol 15, No 3 (2013): Oktober 2013 Vol 15, No 2 (2013): Juni 2013 Vol 15, No 1 (2013): Pebruari 2013 Vol 14, No 3 (2012): Oktober 2012 Vol 14, No 2 (2012): Juni 2012 Vol 14, No 1 (2012): Pebruari 2012 Vol 13, No 3 (2011): Oktober 2011 Vol 13, No 2 (2011): Juni 2011 Vol 13, No 1 (2011): Pebruari 2011 Vol 12, No 3 (2010): Oktober 2010 Vol 12, No 2 (2010): Juni 2010 Vol 12, No 1 (2010): Pebruari 2010 More Issue