Garuda - Garba Rujukan Digital

Article Per Year (5 Year)

All

Jurnal Teknologi Reaktor Nuklir Tri Dasa Mega

tridam
Website

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 210 Documents

1 2 3 4 5

NUCLIDES COMPOSITION OF EXPERIMENTAL POWER REACTOR (RDE) SPENT FUEL Kristina Kristina; Amir Hamzah; Muhammad Subekti; Menik Ariani
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 22, No 1 (2020): February 2020
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

The management of spent fuel is an issue of safety for Indonesia in the phase of designing RDE. Several studies regarding spent fuel are limited by geometrical characteristics and number of nuclides library. Therefore, different methodologies utilizing MCNPX2.6.0 were applied to get better information for further research. In this study, a single fuel pebble containing UO2, was burned using 5 cycles of multi-pass loading scheme for 1080 days to obtain the same energy as RDE’s core, which is about 79.90 GWd/MTU. The multiplication factor k-inf decreased at each cycle and stopped at 1.14575. The calculation results in the nuclides composition of the spent fuel after 1080 days of burning and 5 years of cooling containing 241 nuclides consist of 21 actinides and 220 nonactinides. Actinides with the highest activity of 8.96 Ci is with mass of 0.0867 g, whose half-life time is 14 years long. Nonactinides with the highest activity of 4.47 Ci is with mass of 0.0514 g, whose half-life time is 30.17 years long. The total activity of spent fuel pebble is 22.9 Ci with total mass of 5.28 g. The mass and activity data of each nuclide contained in the spent pebble will be used in the future research for performing safety analysis of the spent fuel storage tank.Keywords: Nuclides composition, Pebble, Spent fuel, RDE, MCNPX

THE EFFECT OF BEACH ENVIRONMENT AND SEA WATER ON NICKEL CORROSION RATE AS A COLLIMATOR MATERIAL FOR THE APPLICATION OF BORON NEUTRON CAPTURE THERAPY Hardi Hidayat; Budi Setyahandana; Yohannes Sardjono; Yulwido Adi
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 3 (2019): October 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

The purpose of this study is to determine the value of corrosion rate influenced by coastal environment and seawater to nickel as a collimator base material for the application of boron neutron capture therapy (BNCT). In this research, the authors used 99.9% pure nickel as the reference material. Corrosion testing was carried out to determine the rate of corrosion of nickel as a base material for BNCT. After the specimens were formed, the test specimens were then corroded for 12 weeks, with various conditions such as indoor, outdoor environment, static seawater, and moving seawater. The results of this study indicated that in corrosion testing with indoor condition, the corrosion rate values are 0.61-1.00 mpy. For outdoor condition, the corrosion rate is 0.89-1.34 mpy. Meanwhile, at static seawater conditions, the corrosion rate is 0.97-1.24 mpy. Lastly, for moving seawater condition, the corrosion rate is 1.64-1.91 mpy. The results showed that corrosion resistance was relatively the same for all nickel exposed to corrosion in the coastal environment. Therefore, in regards to corrosion resistance, using nickel as a collimator base material for BNCT applications is considered as safe.Keywords: BNCT, Nickel, Corrosion, Coastal Environtment, Sea Water

THE RADIOACTIVITY ESTIMATION OF THE IRRADIATED 13 MEV CYCLOTRON’S CONCRETE SHIELD Isdandy Rezki Febrianto; Puradwi Ismu Wahyono; Suharni Suharni
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 22, No 1 (2020): February 2020
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

The Center for Accelerator Science and Technology (PSTA) planned to install K500 concrete shield in its 13 MeV cyclotron facility (DECY-13). However, fast neutrons that are generated by this cyclotron could activate materials of the concrete. It may harm the radiation workers. In this work, we conducted simulations using ORIGEN2 and PHITS computer code to estimate the formed radioactivity and the neutron flux distribution in the DECY-13 cyclotron's concrete shield. Based on the simulation, the induced radioactivity is 2.3478 × 109 Bq, while its gamma dose rate is 22.09 µSv/m2h. The most contributed isotopes are Th-233, Ho-166, Al-28, Mn-56 and Si-31. This dose is quite high. Neutron fluxes in the rear of the simulated concrete shield are also still prominent. Accordingly, it is necessary to attach neutron shielding materials which do not generate high-intensity gamma-ray. The formed radioactivity is high; but it appears from the short half-life isotopes such as Th-233, Ho-166, Al-28, Mn-56 and Si-31. Its activity will diminish quickly after the cyclotron is off. Hence, it will be safe for radiation workers.Keywords: Radioactivity, Concrete Shield, 13 MeV Cyclotron, Neutron Irradiation, DECY-13, PHITS

PERFORMANCE ANALYSIS OF RDE ENERGY CONVERSION SYSTEM IN VARIOUS REACTOR POWER CONDITION Ignatius Djoko Irianto; Sukmanto Dibyo; Sriyono Sriyono; Djati H Salimy; Rahayu Kusumastuti; Marliyadi Pancoko
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 3 (2019): October 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Reaktor Daya Eksperimental (RDE) is an experimental power reactor based on High Temperature Gas-cooled Reactor (HTGR) technology with thermal power of 10 MW. As an experimental power reactor, RDE is designed for electricity generation and provides thermal energy for experimental purposes. RDE energy conversion system is designed with cogeneration configuration in the Rankine cycle. To ensure the effectiveness of its cogeneration, the outlet temperature of the RDE is set at 700°C and steam generator outlet temperature is around 530°C. Analysis of the performance of the energy conversion system in various power levels is needed to determine the RDE operating conditions. This research is aimed to study the performance characteristics of RDE energy conversion systems in various reactor power conditions. The analysis was carried out by simulating thermodynamic parameter calculations on the RDE energy conversion system and the overall cooling system using the ChemCad program package. The simulation is carried out by increasing the reactor power from 0 MW to 10 MW at constant pressure and constant mass flow rate. The simulation results show that the steam fraction at the steam generator outlet increases starting from 3 MW reactor power and reaches saturated steam after the thermal power level of 7.5 MW. From the results, it can be concluded that with constant mass flow rate and operating pressure, optimal turbine power is obtained after the reactor thermal power reached 7.5 MW.Keywords: RDE, Energy Conversion System, Performance, Reactor Power, ChemCad

DOSE ESTIMATION OF THE BNCT WATER PHANTOM BASED ON MCNPX COMPUTER CODE SIMULATION Amanda Dhyan Purna Ramadhani; Susilo Susilo; Irfan Nurfatthan; Yohannes Sardjono; Widarto Widarto; Gede Sutresna Wijaya; Isman Mulyadi Triatmoko
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 22, No 1 (2020): February 2020
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Cancer is a malignant tumor that destroys healthy cells. Cancer treatment can be done by several methods, one of which is BNCT. BNCT uses 10B target which is injected into the human body, then it is irradiated with thermal or epithermal neutrons. Nuclear reaction will occur between boron and neutrons, producing alpha particle and lithium-7. The dose is estimated by how much boron and neutron should be given to the patient as a sum of number of boron, number of neutrons, number of protons, and number of gamma in the reaction of the boron and neutron. To calculate the dose, the authors simulated the reaction with Monte Carlo N Particle-X computer code. A water phantom was used to represent the human torso, as 75% of human body consists of water. Geometry designed in MCNPX is in cubic form containing water and a cancer cell with a radius of 2 cm. Neutron irradiation is simulated as originated from Kartini research reactor, modeled in cylindrical form to represent its aperture. The resulting total dose rate needed to destroy the cancer cell in GTV is 2.0814×1014 Gy.s (76,38%) with an irradiation time of 1,4414×10-13 s. In PTV the dose is 5.2295×1013 Gy.s (19,19%) with irradiation time of 5.7367×10-13 s. In CTV, required dose is 1.1866×1013 Gy.s (4,35%) with an irradiation time of 2.5283×10-12 s. In the water it is 1.9128×1011 Gy.s (0,07%) with an irradiation time of 1,5684×10-10 s. The irradiation time is extremely short since the modeling is based on water phantom instead of human body.Keywords: BNCT, Dose, Cancer, Water Phantom, MCNPX

ROUTING DESIGN ON THE PRIMARY COOLING PIPING SYSTEM IN PLATE-TYPE CONVERTED TRIGA 2000 REACTOR BANDUNG Veronica Indriati Sri Wardhani; Henky Poedjo Rahardjo; Rasito Tursinah
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 3 (2019): October 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

In 2015, research activities to modify TRIGA 2000 Reactor Bandung fuel element from cylindrical to plate-type have been initiated. By using plate-type fuel elements, core cooling process will be altered due to different generated heat distribution. The direction of cooling flow is changed from bottom-to-top natural convection to top-to-bottom forced convection. This change of flow direction requires adjustment on the cooling piping system, in order to produce simple, economical, and safe piping route. This paper will discuss the design of suitable piping routing based on pipe stress and N-16 radioactivity. The design process was carried out in several stages which include thermal-hydraulic data of reactor core to determine the process variables, followed by modeling various pipeline routes. Based on available space and ease of manufacture, four possible alternative routings were determined. Four routings were produced and analyzed to minimize the amount of N-16 radioactivity on the surface of the reactor tank, prolonging the cooling fluid travel time to reach at least five times of N-16 half-life. Subsequent pipe stress analysis using CAESAR II software was conducted to ensure that the piping system will be able to withstand various loads such as working fluid load, pipe weight, along with working temperature and pressure. The results showed that the occurred stresses were still below the safety limit as required in ASME B31.1 Code, indicated that the designed and selected pipeline routing of primary cooling system in the Plate-type Converted TRIGA 2000 Reactor Bandung has met the safety standards.Keywords: TRIGA reactor, Cooling system modification, Pipeline routing design, Pipe stress analysis, N-16 radioactivity

QUANTIFICATION OF ALUMINUM CONTENTS IN COOKED FOODSTUFFS FROM THREE REGIONS IN JAVA USING NEUTRON ACTIVATION ANALYSIS Ahmad Hasan As'ari; Saeful Yusuf; Alfian Alfian
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 22, No 1 (2020): February 2020
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Aluminum is widely available in nature and the third most abundant element on earth. Improper intake of aluminum can increase toxicity and correlate with Alzheimer's disease. One source of aluminum comes from food. In this study, aluminum content in foodstuffs was analyzed using neutron activation analysis. Various foodstuffs were purchased from markets in three regions in Java, namely Bangkalan (East Java), Magelang (Central Java), and Cianjur (West Java) and cooked at a temperature above 80°C until the ready-to-eat condition. The cooked samples were freeze-dried and irradiated in the G.A. Siwabessy research reactor with neutron flux of 5x1013 neutrons.m2.s-1. Post-irradiation samples were analyzed using gamma spectrometry. The results show that the aluminum contents in each foodstuff from one region have a strong correlation with other regions (Pearson correlation coefficient r>0.9, P<0.001), indicating that the distribution of aluminum content does not differ from one region to another. The staple food category has a relatively low aluminum content with an average value of 24 mg/kg and a maximum value of 35 mg/kg. The dish category has higher aluminum content with an average value of 51 mg/kg and a maximum value of 77 mg/kg. The vegetable category has the highest content with an average value of 156 mg/kg and a maximum value of 710 mg/kg owned by caisim. Caisim is interesting for further research because of its ability to store large amounts of several elements. In general, the intake of aluminum sourced from these foods is still below the allowed value.Keywords: Neutron activation analysis, Food safety and security, Alzheimer, Aluminum distribution, Pearson correlation

ESTIMATION OF THE RADIOACTIVE SOURCE TERM FROM RDE ACCIDENT POSTULATION Pande Made Udiyani; Ihda Husnayani; Mohamad Budi Setiawan; Sri Kuntjoro; Hery Adrial; Amir Hamzah
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 3 (2019): October 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

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 removable parameters.The result of source term calculation due to the water ingress accident for Xe-133 noble gas is 8.97E+12 Bq, Cs-137 is 3.59E+07 Bq, and I-131 is 4.34E+10 Bq. As for depressurization accident, the source term activity for Xe-133 is 3.90E+13Bq, Cs-137 is 1.56E+07 Bq, and I-131 is 1.89E+10Bq. The source term calculation results obtained in this work shows a higher number compared to the HTR-10 source term used as a reference. The difference is possibly due to the differences in reactor inventory calculations and the more conservative assumptions for source term calculation.Keywords: RDE, HTGR, Radioactive, Source term, accident

DESIGN OF IRRADIATION FACILITIES AT CENTRAL IRRADIATION POSITION OF PLATE TYPE RESEARCH REACTOR BANDUNG Epung Saepul Bahrum; Wawan Handiaga; Yudi Setiadi; Henky Wibowo; Prasetyo Basuki; Alan Maulana; Mohamad Basit Febrian; Jupiter Sitorus Pane
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 22, No 1 (2020): February 2020
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

One of the results from Plate Type Research Reactor Bandung (PTRRB) research program is PTRRB core design. Previous study on PTRRB has not calculated neutron flux distribution at its central irradiation position (CIP). Distribution of neutron flux at CIP is of high importance especially in radioisotope production. In this study, CIP was modeled as a stack of four to five aluminum tubes (AT), each filled by four aluminum irradiation capsules (AIC). Considering AIC dimension and geometry, there are three possibilities of AT configuration. For irradiation sample, 1.45 gr of molybdenum (Mo) was put into AIC. Neutron flux distribution at Mo sample was calculated using TRIGA MCNP and MCNP software. The calculation was simulated at condition when fresh fuel is loaded into reactor core. Analyses of excess reactivity show that, after installing irradiation AT and Mo sample was put into each configuration, the excess reactivity is less than 10.9 %. The highest calculated thermal neutron flux at Mo sample is 5.08×1013 n/cm2.s at configuration 1. Meanwhile, the highest total neutron flux at Mo sample is located at capsule no. II and III. Thermal neutron flux profile is the same for all configurations. This result will be used as a basic data for PTRRB utilization.Keywords: Central Irradiation Position, Neutron Flux Distribution, MCNP, PTRRB

NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE Alexandre Ezzidi Nakata; Masanori Naitoh; Chris Allison
JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA Vol 21, No 3 (2019): October 2019
Publisher : Pusat Teknologi Dan Keselamatan Reaktor Nuklir (PTKRN)

Two international severe accident benchmark problems have been performed recently by using several existing parametric severe accident codes: The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Plant (BSAF) and the Benchmark of the In-Vessel Melt Retention (IVMR) Analysis of a VVER-1000 Nuclear Power Plant (NPP). The BSAF project was organized by the Nuclear Power Engineering Center (NUPEC) of the Institute of Applied Energy (IAE) in Japan for the three Boiling Water Reactors (BWRs) of the Fukushima NPP. The IVMR Project was organized by the Joint Research Center (JRC) of the European Commission (EC) in Holland (Europe) for a Pressurized Water Reactor (PWR). The obtained results of both projects have shown very large discrepancies between the used severe accident codes for both reactor types BWR and PWR. Consequently, the results for a real plant analysis by these integral codes, may not be correct after the beginning of core melt. Discrepancies of results of ex-vessel phenomena in the containment between the codes are in general larger. Therefore, there is a strong need for a reliable new generation mechanistic severe accident code which can simulate severe accident scenarios from an initiating event till containment failure with better accuracy not only for existing light water reactors but also for new generation IV reactor types. SAMPSON mechanistic ex-vessel modules coupled with SCDAPSIM and a new thermal-hydraulic module ASYST-ISA with particularly newly developed options for the reactor coolant system (RCS) and material properties applicable to new reactor deigns, is proposed as a best etimate new generation severe accident code for several reasons which are described in this paper.Keywords: Severe accident, SAMPSON, SCDAPSIM, ASYST-ISA, Steam explosion, Hydrogen detonation

Page 1 of 21 | Total Record : 210

1 2 3 4 5

Filter by Year

2010 2023

Filter By Issues

All Issue 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

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

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

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

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

Abstract

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