JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
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Experimental Investigation of Natural Circulation Stability Phenomena in a New Loop Heat Pipe Model
Full text linkThe severe accident at the Fukushima Dai-ichi Nuclear Power Plant in Japan in 2011 highlighted the critical need for a passive cooling system to dissipate residual decay heat following the failure of active cooling systems in the nuclear facility. The loop heat pipe (LHP) is a promising technology for such applications. The objective of this research is to understand the natural circulation stability phenomena of new LHP model under varying conditions of filling ratio and heat load. The experimental methodology employed a laboratory-scale LHP model made of copper with an inner diameter of 0.104 m. The experiments were designed with filling ratios of 20%, 40%, 60%, 80%, and 100%, and hot water temperature as the evaporator heat source with variations of 60°C, 70°C, 80°C, and 90°C. The initial operating pressure was 10665.6 Pa, with a 5˚ inclination angle, demineralized water as the working fluid, and cooled by air at a velocity of 2.5 m/s. The results show that the natural circulation within the LHP occurs in two phases and maintained stability, with optimal performance observed at an 80% filling ratio and 90°C. The conclusion of this research indicates that natural circulation stability in the LHP operates well and occurs in two phases, proving that natural circulation in the LHP is effective in heat dissipation
Dispersia-BRIN: A Radiation Dose Gaussian-Air-Dispersion Calculation Program for Radioactive Releases from a Nuclear Facility
No full textThe development of the Dispersia-BRIN program, alongside STDISPERSIA WRDISPERSIA, aims to facilitate swift and effective radiological analysis crucial for rapid decision-making during nuclear accidents. These programs are based on the SIMPACT version 1.0 framework, ensuring easy and rapid analysis. Validation against SIMPACT confirms their accuracy and reliability. Python was chosen as the primary programming language for its simplicity and versatility. Dispersia-BRIN integrates geospatial analysis tools, allowing researchers to visualize results on polar grids and effectively map nuclear facilities and their surroundings. The program calculates the dispersion and impact of radioactive materials by analyzing their interaction with humans, livestock, and plants, measured through dose assessment. This process helps identify high-dose areas, critical populations, and emergency planning zones. A case study of a hypothetical Nuclear Power Plant demonstrated Dispersia-BRIN ability to accurately calculate and visualize radionuclide dispersion, aiding in pinpointing high-radiation areas. In conclusion, Dispersia-BRIN provides a comprehensive tool for radiological analysis, supporting informed decision-making for radiation protection and emergency response, thereby ensuring public health and environmental safety. Its user-friendly design and powerful analytical capabilities make it an invaluable resource for addressing radiological hazard
Insider Intervention Model in the Sabotage Attack Scenario of a Nuclear Reactor Facility
Full text linkThe 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
Investigation of Natural Circulation Flow Under Steady-State Conditions Using a Rectangular Loop
Full text linkPassive 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
Analysis of the Reactivity Coefficient of the PWR Thorium Fuel
Full text linkIn 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
Techno-Economic Assessment and Optimization of a Standalone System in Sebira Island, Indonesia
Full text linkNuclear 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 0.6577. The CO2 emissions generated in the optimal nuclear-PV-battery system are zero, making this system far more viable than other hybrid system schemes
Neutronic Analysis of the NuScale Fuel Assembly using Accident Tolerant Fuel with SiC Coated Alumina Cladding
Full text linkNuclear fuel design is influential on the performance and safety of a nuclear reactor. Accident Tolerant Fuel (ATF) is a novel concept in nuclear fuel technology developed to improve performance and safety of a nuclear power reactor. Different ATF material can impact the neutronic aspect of a nuclear reactor, and thus must be analysed accordingly. This research is a neutronic analysis of an ATF design using alumina (Al2O3) and outer silicon carbide (SiC) coating implemented in NuScale SMR fuel assembly. MCNP6.2 code was utilised to perform neutronic calculation. Standard M5 cladding in NuScale was compared with Al2O3 and Al2O3+ SiC cladding. Analysed parameters were fuel burnup, kinetic parameters, Doppler temperature coefficient, moderator tempereature coefficient, and the evolution of several radionuclides. The results show that there are no significant differences in neutronic performance of Al2O3 clad compared to standard M5 clad. Therefore, Al2O3 cladding is potential for application in PWR fuel
HAZOP-Based Radiological Risk Assessment of Pebble Bed Fuel Handling Systems
No full textThe High-Temperature Gas-Cooled Reactor (HTGR), a promising candidate for Generation IV nuclear reactors, boasts superior inherent passive safety features and a continuous fuel handling system. This system employs multi-pass cycles, utilizing pneumatic and gravitational mechanisms to feed, circulate, and unload the pebble bed fuel element. This paper presents a descriptive analysis assessing the safety risk of the fuel handling system design in HTR-10. The Hazard and Operability Study (HAZOP) methodology is employed to identify hazard parameters, deviation limitations, causes, impacts, and potential risks to the system’s main components. The establishment of probability scales, consequence criteria, risk level ratings, and control activities adheres to the ISO 31000 standard. Primary data were gathered through expert interviews, while secondary data were sourced from design layout documentation, literature reviews, and safety analysis reports. Six main components - the elevator, core, singulator, failed fuel separator, burnup measurement, and distributor - were selected as assessment nodes from the piping and instrumentation diagram. The assessment revealed that each node initially presented a moderate to extreme risk potential (risk level rating C to E). However, after applying the effectiveness index of the designed control, the residual risk for all nodes was reduced to an acceptable limit (risk rating A - very low). Therefore, the fuel handling system design already incorporates adequate control activities to mitigate potential safety risks due to system component failure. As safety risk assessment is dynamic, it should be reviewed periodically or whenever there are design changes at any project stage. This ensures the safety risk magnitude is consistently known and managed effectively