JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
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NUCLIDES COMPOSITION OF EXPERIMENTAL POWER REACTOR (RDE) SPENT FUEL
Full text linkThe 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, MCNP
DOSE ESTIMATION OF THE BNCT WATER PHANTOM BASED ON MCNPX COMPUTER CODE SIMULATION
Full text linkCancer 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, MCNP
EDITORIAL BOARD
No full textThis section consists of Cover Page, Editorial Board Page, Peer Reviewer Page, Table of Content Page and Preface Pag
SAFETY ANALYSIS OF NEUTRON INTERACTION WITH MATERIAL PRACTICUM MODULE FOR THE KARTINI INTERNET REACTOR LABORATORY
Full text linkKartini Research Reactor, which is situated in Yogyakarta, is a 100 kW TRIGA (Training, Research, and Isotope Production by General Atomic)-type reactor mainly used for educational and training purposes. A system for remote learning on nuclear reactor physics named the Internet Rector Laboratory has been developed and is fully operational since 2019. To enrich its curriculum, a new practicum module has been developed, that can be immediately implemented and does not require any additional equipment or materials. To ensure safety in reactor kinetics and radiation protection, a safety analysis on the implementation of the practicum module has been conducted using MCNP and ORIGEN utilizing the current conditions of the reactor regarding its fuel burnup and control rod positions at a certain power level. Based on the results of the analysis, the practicum is safe to perform from a neutronic and radiation protection perspective. Given the long half-life and the large amount of radiation exposure that comes from activation products of iron, it is recommended that only cadmium, boron, graphite, and aluminum are allowed to be irradiated during the practicum.Keywords: Internet Reactor Laboratory, Activation Product, Radiation Protection, Reactor Safet
ACKNOWLEDGMENT
No full textThis Section consists of Abstract Collection Page, Keywords Index Page and Acknowledgment Pag
ESTIMATION OF THE RADIOACTIVE SOURCE TERM FROM RDE ACCIDENT POSTULATION
Full text linkThe 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, acciden
THE EFFECT OF BEACH ENVIRONMENT AND SEA WATER ON NICKEL CORROSION RATE AS A COLLIMATOR MATERIAL FOR THE APPLICATION OF BORON NEUTRON CAPTURE THERAPY
Full text linkThe 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 Wate
PERFORMANCE ANALYSIS OF RDE ENERGY CONVERSION SYSTEM IN VARIOUS REACTOR POWER CONDITION
Full text linkReaktor 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, ChemCa
THE STUDY OF ATMOSPHERIC DISPERSION MODEL ON ACCIDENT SCENARIO OF RESEARCH REACTOR G. A. SIWABESSY USING HOTSPOT CODES AS A NUCLEAR EMERGENCY DECISION SUPPORT SYSTEM
Full text linkG.A. Siwabessy Multipurpose Reactor (RSG-GAS) is a research reactor with thermal power of 30 MW located in the Serpong Nuclear Area (KNS), South Tangerang, Banten, Indonesia. Nuclear emergency preparedness of RSG-GAS needs to be improved by developing a decision support system for emergency response. This system covers three important aspects: accident source terms estimation, radioactive materials dispersion model into the atmosphere and radiological impact visualization. In this paper, radioactive materials dispersion during design basis accident (DBA) is modeled using HotSpot, by utilizing site-specific meteorological data. Based on the modelling, maximum effective dose and thyroid equivalent dose of 1.030 mSv and 26 mSv for the first 7 days of exposure are reached at distance of 1 km from the release point. These values are below IAEA generic criteria related to risk reduction of stochastic effects. The results of radioactive dispersion modeling and radiation dose calculations are integrated with Google Earth Pro to visualize radiological impact caused by a nuclear accident. Digital maps of demographic and land use data are overlayed on Google Earth Pro for more accurate impact estimation to take optimal emergency responses.Keywords: G.A. Siwabessy research reactor, Nuclear emergency, Atmospheric dispersion model, Decision support system, HotSpot code
NEED OF A NEXT GENERATION SEVERE ACCIDENT CODE
Full text linkTwo 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 detonatio