Search

Thermocouples are generally used to provide real-time temperature indications in instrumented tests performed at material and test reactors. Passive temperature monitors, such as Silicon Carbide (SiC) and melt wires, may be included in such tests as an independent technique of detecting peak temperatures experienced during irradiation. In less expensive static (drop-in) capsule tests, which have no leads attached for real-time data transmission, melt wires, and SiC temperature monitors (TMs) are essentially the only possibility for peak temperature indication. A melt wire involves placing materials (wires) of a known composition and melting temperature in a test. An inventory is maintained at Material Science Laboratory (MSL) for melt wires ranging in temperatures from 30°C to 1500°C. Unfortunately, melt wires are limited in that it can only detect whether a single temperature is or is not exceeded (melt wire melted or not). SiC TMs, which can also be used to detect peak irradiation temperatures, are advantageous because a single monitor can allow to determine the peak temperature reached within a relatively broad range (100 – 1200°C) resulting in accuracies within ±20°C. Irradiation temperature is determined by measuring a property change after isochronal annealing or during a continuously monitored annealing process using specialized equipment at MSL. Recent research has produced a passive monitor known as sublime temperature monitor. This passive sensor has the capability of recording temperature gradients and pinpointing exactly where a temperature is located along that gradient. Long measurement lengths are achieved with very high accuracy in the location of desired temperature measurements (±2 mm over a 1 m span); however, this sensor has not been deployed in a nuclear reactor. This article will focus only on passive temperature sensors currently being researched and implemented under the Advanced Sensors and Instrumentation (ASI) program at Idaho National Laboratory (INL). [ABSTRACT FROM AUTHOR]

The paper is focused on the identification of critical issues with the primary circuit components based on the operational experience with the importance of safe long-term operation (LTO) of VVER 440 and VVER1000 nuclear reactors. Results from the actual running H2020-Europe project DELISA-LTO are presented in cumulative form in detail. Our approach is to combine the development of the simulation tools, experimental work (material analyses), specific irradiations and in-service and/or non-destructive inspection techniques to develop the effective "early warning" tool for the assessment of the system integrity for the LTO of the current VVER reactors. [ABSTRACT FROM AUTHOR]

Currently, there are 5 nuclear reactors in operation in Slovakia at two locations: Jaslovske Bohunice and Mochovce. At the time of the construction of the first reactors, both locations were assessed as having no increased seismic hazard. Nevertheless, since the start of NPP operation several reassessments were made for both sites ensuring that the current site assessment is performed in accordance with the current state-of-the-art knowledge. Consequently, the seismic resistance of all relevant structures, systems, and components was during operation continuously increased. Actual reinforcement for Jaslovske Bohunice NPP is set to peak ground acceleration PGA=0.344·g and for Mochovce NPP to value PGA=0.15·g. [ABSTRACT FROM AUTHOR]

Oxide dispersion strengthened steels (ODS) are key alloys used in nuclear installations. Steels with chromium content of up to 10 wt. % are suitable for the construction of technological devices in the primary circuit of nuclear power plants. Chromium has anti-corrosive properties due to formation of a passivation layer, which results in lower activation of the material by neutrons. Macroscopic properties are determined by their microstructure, and therefore the description of the microstructure is important. For this purpose, transmission Mössbauer spectroscopy and atom probe tomography were applied. In this way, the physical and/or chemical environments of the resonant atoms can be described. The obtained spectral parameters reach saturation values from which the solubility limit of chromium in iron can be determined. In Cr-rich phase, the solubility limit can be estimated from the spectral parameters of a singlet located in the centre of a Mössbauer spectrum which corresponds to the alloy annealed for the longest time. The suggested procedures are subsequently applied to the case studies of stainless steels suitable for the construction of various components of the III+/IVth generation of nuclear reactors. Presence of trace elements in the investigated samples was determined by neutron activation analysis. [ABSTRACT FROM AUTHOR]

Severe accidents in nuclear reactor have warned the nuclear scientists to put more attention about reactor safety. To decrease the risk when the reactor melts down, it is proposed a core catcher system. The behavior of the liquid is one of the essential information that must be provided, especially inside the channels of core catcher. Dynamics of water passing through a 90° horizontal bend of 90° is examined in the context of core catcher as a part of nuclear reactor. In this study, the dynamics is simulated using Moving Particle Semi-Implicit method. The simulation results are then compared to the experiments. The simulation results show that the water depth patterns obtained in this study are similar to experiment results. In addition, the travel time of water agrees with the experimental results. This shows the capability of Moving Particle Semi-Implicit method as an alternative numerical method for liquid transport. For further investigation, the chosen number of particles is sufficient for the purpose of predicting how fast the water front will reach each position along the channel. [ABSTRACT FROM AUTHOR]

Nuclear Engineering and Engineering Physics Department – Faculty of Engineering – Universitas Gadjah Mada developes new nuclear reactor to produce medical radioisotopes (primarily Mo-99). The reactor is AHR (Aqueous Homogeneous Reactor) type namely GAMA-AHR. The GAMA-AHR uses fissile (seed) fuel of 3.5 % mole uranyl nitrate solution in H2O and fertile (blanket) fuel and of 4 % mole thorium nitrate solution in H2O. The GAMA-AHR uses light water as primary coolant, neutron reflector as well as radiation shielding. The aim of this design is to avoid utilization of graphite reflector, thus simplify the design and allows totally passive post shutdown cooling systems. The GAMA-AHR is designed with 200 kWth thermal power and with Mo-99 production capability more than 2000 six-day Ci per week with near continuous operation. The analysis of Mo-99 production capability is performed by solving Mo-99 material balance equation for each process stages. The Mo-99 material balance formulation includes Mo-99 generation by fission reaction, by related its parent decay, Mo-99 depletion due to decay and neutron absorption and Mo-99 accumulation in reactor fuel solution. And also Mo-99 extraction, decay and accumulation in extraction apparatus, delay tank, purification apparatus and final processing apparatus, i.e. packaging and calibrating apparatus. Mole balance of other Mo isotopes is formulated and solved to calculate the purify degree of Mo-99. The calculation results shows that GAMA-AHR with 200 kWth thermal power has equilibrium standard delivered Mo-99 production capability of 2531 six-day Ci (delivered) per week. The equilibrium production capability does not depend on extraction delay time. The peak Mo-99 production capability however, depend on extraction delay. The longer time extraction delay time the higher the peak Mo-99 production capability. All of Mo radioisotope contaminants have far shorter half-life then the half-life of Mo-99. Delay time and processing time significantly eliminate these contaminants. [ABSTRACT FROM AUTHOR]

In order to ensure the continuity and safe operation of Non-Reactor Nuclear Installation, a programmed system is needed to maintain process equipment and supporting infrastructure. Maintenance activities consist of preventive action carried out by the operator regularly, and curative action carried out when the equipment is damaged or malfunctions. Repair activities are carried out by personnel who are experts in their fields. Since preventive maintenance is routine, its implementation can be scheduled. On the other hand, curative maintenance cannot be scheduled because of its unpredictable nature. EFEI received an operating license in 2001 for its fabrication and quality control (FFL and QC) facilities, while for the conversion facility (PCP) an operating permit was obtained in 2013. Both facilities still hold operating licenses today. Currently EFEI is around 35 years old, so it is certain that the systems, structures and components (SSCs) are ageing. The effects of ageing need to be managed properly as they can have an impact on safety. Therefore, we need an ageing management program that can be implemented effectively (targeted and timely) and efficiently. Changes that occur as a function of time are the ageing of components and equipment in the installation. Ageing can reduce the availability and reliability of the systems, structures, and components (SSCs). This will affect the overall operating performance of the installation. If this ageing occurs in the SSCs which is related to safety functions, it will put the workers, public and environment around the installation in danger. Therefore, it is necessary to know the causes of degradation due to aging so that the effects of aging can be managed properly. [ABSTRACT FROM AUTHOR]

Until now, the price of non-solid fuels (oil and gas) has increased, there is a big opportunity for a shift towards the use of biomass fuels. Besides its unlimited availability, biomass is also a well-known fuel. For its use, of course, a device called a stove or stove is needed, which has been designed for biomass fuel. Even though Riau Province has an electrification ratio of 92.43%, out of 12 districts and cities in Riau, there are three more districts that have a low electrification ratio, namely Pelalawan Regency, an area with an electrification ratio of 52.75%, Indragiri Hilir (Inhil) was 69.01% and Rokan Hulu (Rohul) was 64.91%. In this research, the TEG module is used as a source of electricity generation. The working system of the TEG module is to convert heat energy or, more precisely, the difference in temperature directly into electrical energy. This research as a whole consists of the design stage, the development stage, the testing phase and the evaluation stage. The design stage aims to design a stove, the development stage aims at the stove production process, the testing stage aims to retrieve research data and the evaluation stage aims to analyze the results of the study. The results obtained from this study are the design and dimensions of the updraft type semi gasification biomass stove with a reactor diameter of 20 cm, a height of 30 cm, and the dimensions of the primary hole 12 cm x 9 cm and the secondary hole 0.1 cm of 54 holes arranged 18 holes horizontally. and 3 holes vertically. The rate of fuel demand is 3.27 kg / hour and the highest voltage output is 9.24 volts in the 55th minute with a temperature difference (ΔT) of 36.5 °C, while the highest current is 0.343 amperes in the 20th minute with a difference in temperature (ΔT) 28.75 °C. [ABSTRACT FROM AUTHOR]

The heat pipe is a device that can transfer heat very effectively and passively. Heat pipes are widely used as thermal management tools in various fields, including nuclear reactors. Although heat pipes can have various geometries and shapes, heat pipes consist of three parts: the evaporator, adiabatic, and condenser. The adiabatic will pass the working fluid in gas from the evaporator to the condenser and pass the working fluid in a liquid from the condenser to the evaporator. There are very limited studies on the effect of adiabatic length on thermal performance. Therefore, research has been carried out on the effect of adiabatic length for heat pipes that have a fixed evaporator and condenser. The research was conducted by numerical simulation on a straight wick-less heat pipe using the RELAP5 computer program. The heat pipe is made of a tube of 1-inch stainless steel, and the working fluid is water. The evaporator and condenser sections have the same length of 50 cm, while the adiabatic sections have variations of 10, 20, 30, 40, 50, 100, 150, and 200 cm. The heat pipe is placed on a heat source of atmospheric water at 80 ○C as well as saturated steam with temperature variations from 80 to 120 ○C. The heat sink is cooling water at a temperature of 29 ○C. The study results show that the thermal performance of the heat pipe is proportional to the adiabatic length when the heat source temperature is much greater than 110 ○C and vice versa when the heat source temperature is much lower than 110 ○C. The adiabatic length does not affect the performance of the heat pipe when the heat source temperature is around 110○C. However, the increase or decrease in heat transfer capability is minimal, so it can be concluded that the adiabatic length has little effect on the thermal performance of the wick-less straight heat pipe. [ABSTRACT FROM AUTHOR]

The Nuclear Power Plant Accident in Fukushima, Japan in 2011 became a valuable lesson in the safety of NPP development in the world. In particular, passive safety systems began to assist the reactor cooling system when the active system failed. However, it has a lower active cooling system to effectiveness compared. The passive cooling systems can be relied upon to assist cooling that they do not require a power source to drive the work system and can work naturally in circulation. The safety system technology is a passive safety system, also developed by National Nuclear Energy Agency (BATAN). It cools and dissipates heat passively for nuclear heat pipe technology. The Loop Heat Pipe (LHP) type dissipates heat passively. Therefore, it is necessary to carry out the clearance function to determine whether a nuclear technology is feasible or not to be widely used and commercially valid. This study intended to measure the level of technology readiness to avoid technological failures through technology audits. This research method used is measuring Technology Readiness Level (TRL) and ensuring that the LHP nuclear technology used has met mature requirements and meets quality standard requirements. The results of the LHP technology audit with TRL measurements show TRL 4 of Data Acquisition System (DAS), LHP Design Program, System Integration sub-systems. The cause of problems about the preparation of hardware production not implemented, market research and laboratory research to select the fabrication process and the production process not been reviewed by the manufacturing department. Therefore, it is necessary to implement of prepare hardware production, market research, and laboratory research. Also, the manufacturing department review the fabrication and production process to increase the measurement of TRL [ABSTRACT FROM AUTHOR]

Holmium-166 (166Ho) is one of theranostic radionuclides that was potential for use in nuclear medicine applications. This radiolanthanide can be produced in nuclear reactor through direct or indirect methods. Radioisotope of 166Ho produced from indirect method is a carrier free, so that it has a high specific activity which can be separated from irradiated target. The preliminary study for separation of 166Ho through indirect method from irradiated natural dysprosium oxide (Dy2O3) target at TRIGA 2000 Bandung reactor has been carried out. The separation of 166Ho radionuclide from Dy/Ho matrix of irradiated Dy2O3 target was obtained using extraction chromatography method with LN resin which has a particle size of 50-100 µm as a stationary phase, while nitric acid solution was used as a mobile phase. The purpose of this study is to obtain the separation condition of 166Ho from irradiated natural Dy2O3 target with various amounts of LN resin and concentrations of nitric acid solutions at room temperature. The irradiated of natural Dy2O3 target contained 166Ho, 166Dy and 157Dy radionuclides of 75.10%, 24.56% and 0.34%, respectively. The separation condition with 1.4 N HNO3 (fraction 1-53) and 3 N HNO3 (fraction 54-80) solutions using 3 g of LN resin obtained 166Ho radionuclide with a yield of 40.35±1.23%. The radioisotope of 166Ho obtained from the separation has a high radionuclide purity of 100 %. [ABSTRACT FROM AUTHOR]

Sources of electricity generation in Indonesia still rely on fossils as the main source in meeting domestic needs. In 2020, the Indonesian National Medium-Term Development Plan (RPJMN) 2020-2024 targets a total power plant generated is 14.5 GW. The National Research and Innovation Agency (BRIN) program in the nuclear field is to develop a National Research Priority (PRN) for nuclear energy. The output of nuclear energy program in 2024 is the prototype of the first nuclear power plant in Indonesia. In accordance with the energy grade in West Kalimantan, the Small Modular Reactor (SMR) of the 160MWt PWR was chosen. Preparation for the construction of nuclear power plants requires licensing documents and technology assessments. Study on dispersion analysis and radiation safety assessment of the nuclear reactor is conveyed. The aim of the study is to justify that the radioactive releases from the nuclear reactor is in accordance with the regulations. It also to ensure that the public, and the environment receive the lowest possible radiation dose in accordance with the principle as low as reasonably achievable (ALARA). The study includes the calculation of the reactor core inventory, using software Origen 2.1. The source term released into the environment along with meteorological data is used to calculate radiological consequences using PC-Cream code. The meteorological and spatial data were taken from data gathered by Mempawah weather station (116km from Bengkayang site). The 160MWt PWR-type SMR is postulated to be deployed in the prospective site of Bengkayang, West Kalimantan. Data analysis of the dispersion and the radiological consequences have been obtained for the normal operation of one unit of SMR in the candidate site. The results showed that the dose received surrounding the area was far below the community dose limit value set by the Indonesian Nuclear Energy Regulatory Agency (BAPETEN). [ABSTRACT FROM AUTHOR]

The Kartini research reactor core cooling uses natural convection. Primary cooling system is used to cool coolant by circulating coolant in reactor tank to a heat exchanger dissipating the heat to the secondary cooling system and back to the reactor tank. The coolant is injected to the reactor tank through a diffuser located above the core in order to dilute the N-16 radionuclide. However, it turns out that the primary coolant flow affected the natural circulation in the reactor core. This study was conducted to study the effect of primary coolant flow on natural circulation in the reactor core at various reactor power, as well as to determine the characteristics of natural circulation in the core. The experimental study was focused in the hot channel (ring B). The reactor was operated with a power of 50 kW, 75 kW, and 100 kW. At each power, the flow rate of the primary coolant changed from 100% to 0% flow rate with decrement of 20%. The instrumentation to measure coolant temperature were used to provide the axial coolant temperature distribution in the hot channel. From this temperature distribution, the flow pattern in the channel was predicted. An analytical sub-channel analysis was performed to calculate the natural circulation flow rate and DNBR. The experimental results showed that in all cases the primary flow influenced the flow in the core; the natural circulation was seemed balanced by forced circulation coming from primary flow. It was also observed that at low power, the temperature of the cooling fluid tends to be relatively stable compared to high power. When the primary flow rate is 0%, natural convection occurs in the reactor core with the velocity (u) and DNBR of 9.22 cm/s and 9.02, 9.24 cm/s and 6.02, and 11.55 cm/s and 4.51 at the power of 50 kW, 75 kW and 100 kW, respectively. The values of DNBR are still larger than safety requirement. It is concluded that the flow in the core of Kartini reactor is mixed convection due to primary coolant forced flow and buoyancy of natural circulation. During the condition of fully natural circulation, the thermal-hydraulic safety limit is still kept. [ABSTRACT FROM AUTHOR]

The construction of nuclear installations is an activity that has the potential to be related to safety. Therefore, organizations appointed to set up the construction of nuclear power plants need to implement the Regulation of the Head of Bapeten Number 4 of 2018 concerning Safety Provisions in Carefully Evaluation of Nuclear Installation Locations. Regulation of the Head of Bapeten Number 4 of 2018 states that the Organization or Applicant is required to establish the latest standard of function in site evaluation activities supposed to meet the provisions of the Indonesian National Standard. Currently, the existing standard used for site determination of nuclear installations is SNI 18-2034-1990 regarding Guidelines for Nuclear Reactor Site Determination. Standard issued in 1990. It is no longer following the provisions of the current regulations and conditions. So, it is necessary to conduct studies and standards for renewal. The research method uses qualitative methods with gap analysis. The gap analysis in this study focuses on SNI 18-2034-1990 regarding Guidelines for Determining the Location of Nuclear Reactors against the Regulation of the Head of Bapeten Number 4 of 2018 concerning Safety Provisions for Evaluation of Nuclear Installation Locations. This aims to identify existing standards prepared by the Regulation of the Head of Bapeten Number 4 of 2018 concerning Safety Provisions for Evaluation of Nuclear Installation Locations. The results show that the recommendations for changes in SNI 18-2034-1990 include: 1) standard table of contents writing format; 2) the concept of nuclear development guidelines include: a) data collection; b) metrology; c) assessment; d) design parameter values; 3) General prologue; 4) Change the numbering system up to 4 digits; 5) Include specific requirements related to seismic, volcanic, geotechnical and foundation aspects, meteorology, hydrology, human-caused events. [ABSTRACT FROM AUTHOR]

Small-scale Modular Reactor (SMR) is a type of PWR reactor that is currently developing. The NuScale reactor with a power of 160 MWt which is included in the SMR category provides space for researchers to develop the fuel assembly. This research was conducted with the aim of Optimization the arrangement of the NuScale reactor fuel assembly pattern using the MCNP 6 code with the ENDF/B-VII library. There are several types of fuel assembly in NuScale reactors. One of the fuel assemby that is focused on in this research is a fuel assembly that contains fresh UO2− Gd2O3 with an enrichment of 4.55% and has a Gd content of 8%. In order to optimize the arrangement pattern of the UO2−Gd2O3 fuel assembly, the infinity criticality and reactivity coefficients were calculated using the arrangement patterns of 32, 24, 16 and 8 Burnable Poisson. From the results of the research, the optimum number of Burnable Poissons in the fuel assembly is 24 pins for 8% Gd2O3. It is hoped that the results of the research can be used as complementary materials for complete neutronic calculations, safety parameters and analysis on the NuScale nuclear reactor core. [ABSTRACT FROM AUTHOR]

In the frontier, outermost, and underdeveloped area (3T) in Indonesia, one of the primary problems that must be addressed is the severely limited access to electricity. In order to fulfil the electricity demand in 3T areas, a simple, cheap, safe, and clean power generation system is required. Hydride microreactor is a conceptual micro nuclear reactor design proposed to address the issue. This research is aimed to analyze the neutronic properties of hydride microreactor fueled with a combination of uranium hydride (UH3) and thorium carbide (ThC) fuel rods. The analysis was performed for core criticality, temperature coefficient of reactivity, and control rod worth. The UH3 rods use low-enriched uranium whose enrichment was varied between 11-15%, while the number of ThC fuel rods were varied between 3 to 6 rods. The calculation was done using Serpent-2 code with ENDF/B-VII.0 library. The result shows that 13% LEU enrichment and 6 ThC rods allow the reactor to operate for more than 15 years while fulfilling the passive safety criteria. [ABSTRACT FROM AUTHOR]

The development of the current situation and condition of offices in the area of the Center for Science and Technology Research (PUSPIPTEK) in Serpong has changed with the merger of 4 (four) government research institutions initially, then later added with other Ministries/Institutions (K/L). Likewise, personnel/employees from Ministries/Institutions who served as functional researchers are transferred to a new government agency for research and innovation, namely the National Research and Innovation Agency (BRIN) and they work in one area PUSPIPTEK, Serpong, so that many personnel/staff in and out of PUSPIPTEK area. As it is known that in the PUSPIPTEK area which is a strategic area of vital objects there is a nuclear area in which nuclear installations/facilities already exist and operate such as nuclear reactors, nuclear fuel fabrication facilities, radioactive waste installations and others, will be a potential security threat to nuclear facilities/installations. The Licensee are required to pay increased attention to ensuring adequate nuclear security in nuclear reactors and other nuclear facilities, in addition to paying attention to safety. To anticipate threats that could harm nuclear installations/facilities in the form of theft, and/or unauthorized transfer or sabotage of nuclear materials, it is necessary to establish and to implement a physical protection system. This implementation is a form of real role to take preventive action against possible threats in the Serpong Nuclear Area (KNS) which must be carried out by the licensee as stated in Government Regulation Number 54 Year 2012 on Safety and Security for Nuclear Installations. In this paper, the presentation uses a qualitative descriptive method through literature studies and studies on the suitability of nuclear security/physical protection regulations. Then, from the results of the description of the conformity study of these regulations, it can be recommended a policy strategy related to physical protection and the basic threats to design level facilities in the KNS. The limitations of this paper only describe the KNS locus. [ABSTRACT FROM AUTHOR]

Reactor power calibration is an important procedure in nuclear reactor operation. Power calibration is carried out to ensure that the power measurement results in the reactor instrumentation system are in accordance with the standard. Power calibration in the Kartini reactor is carried out by measuring the increase in the temperature of the reactor water tank. Initially, an analog thermometer is used to measure the temperature. Along with the development of the Internet Reactor Laboratory (IRL), a digital data acquisition system for the reactor water tank temperature is required to send the data into IRL computer. To achieve this goal, several stages of development were carried out. First, a data acquisition system hardware is made using three DS18B20 temperature sensors and one microcontroller. Then, a temperature measurement program is made based on one wire communication. Furthermore, a data acquisition system was made and tested from the microcontroller to the Internet Reactor Laboratory computer wirelessly. Finally, the system was tested by comparing the measurement results to a calibrated thermometer. The test results show that the data acquisition system has good measurement accuracy. The average difference between measurements with a calibrated thermometer is 1.8% and the power calibration has fair accuracy with the 0,970% error. [ABSTRACT FROM AUTHOR]

The G.A Siwabessy Multipurpose Reactor (RSG-GAS) is one of three research reactors in Indonesia used for radioisotope production, experimental activities, and training. One of the risks of a nuclear reactor, including a research reactor, is the possibility of radioactive release into the environment. PSA Level 1 quantifies the core damage frequency (CDF) for a postulated initiating event (PIE). IAEA has published the safety series SSR-3 that provides examples of selected PIE for research reactors, but each reactor may have other PIE depending on the characteristics of the design. PIEs for RSG-GAS are defined solely based on the manufacturer's data and have never been systematically evaluated. Hence, it is necessary to systematically evaluate those pre-defined PIEs. This study aims to evaluate the PIEs of the RSG-GAS secondary cooling system using the HAZOP method. It is important because the possibility of the reactor temperature could increase if the secondary cooling system fails to remove heat from the core, and, hence, can cause radioactive substances released into the environment if the secondary cooling system leakage and the protective measures are failed. This study evaluated 7 possible events and the results show that all events have low risk, low consequence, and/or rare frequency. Therefore, those 7 events did not meet the PIE criteria. Nevertheless, we still recommend that RSG-GAS pay attention to those identified events and provide additional procedures to periodically check the radioactivity concentration in the secondary water when PA-01/PA-02 CR 001 is activated. [ABSTRACT FROM AUTHOR]

The nuclear reactor is designed following the concept of defence in depth (DiD). To ensure the adequacy of the concept is applied in the design, the fundamental safety function can be used as a reference. The fundamental safety function consists of three items: to control reactivity, remove heat from the reactor and confine the radioactive material. The design assessment can be done by determining the failure of the fundamental safety function using probabilistic analysis. PWR is a type of reactor that is mainly operated at this time. One new type of PWR is AP1000, which started commercial operation in 2018. Therefore, its safety level still needs to be assessed continuously. This research aims to evaluate the fundamental safety function by probabilistic analysis and assess the DiD concept capability in the AP1000. The analysis is done by making the fault tree of the fundamental safety function, whereas the assessment of defence in depth is carried out using matrix DiD level against the fundamental safety function. The first fundamental safety function, the second fundamental safety function, and the third fundamental safety function consist of six, five, and five intermediate events, which are analyzed till basic events. Furthermore, it is carried the failure combination of each level of fundamental safety function. The probability data of component failure or human error used in the analysis are collected from the international atomic energy agency (IAEA), Westinghouse, Nuclear Regulatory Commission (NUREG), and other published documents. The results show that the failure frequency for all fundamental safety functions is 4.36E-10 per year. The combination failure frequency of fundamental safety function for first and second, first and third also second and third, is 2.84 E-06, 2.31E-07, and 2.90E-07 per year. In contrast, the frequency of critical subsystems for fundamental safety functions is 5.18E-04, 9.11E-04, and 1.16E-04 per year. Based on the analysis, it was concluded that the failure frequency of the fundamental safety function for AP1000 is still within IAEA criteria, which are smaller than the frequency of infrequent faults and limiting faults classification. The AP1000 design followed the concept of defence in depth. [ABSTRACT FROM AUTHOR]

Nuclear energy has an important role in achieving the target of zero carbon emission in 2060. In line with that, development of competent nuclear human resources through an effective and efficient infrastructures is important. Recent development of virtual reality technology might be a method to result an effective and efficient infrastructure by a feature of an immersive simulation-based tool for nuclear reactor education and training. This research has a focus on developing a virtual reality-based 3D visualization of a PWR for enhancing the recent existing internet Kartini reactor laboratory which has been used for distance learning of nuclear reactor physics and operation. Free software of Unity and Blender 3D is used to create a virtual environment of a PWR and to create 3D model of the power plant. A virtual reality machine of Oculus Quest 2 is used to access the 3D virtual reality model and the results show that the students can feel a real-like experience through the 3D virtual reality model. It enhances effectiveness of the internet Kartini reactor laboratory in imparting a nuclear power reactor plant knowledge. However, further development of a simulation code for calculation of the PWR's physical characteristics and its interfacing with the virtual reality 3D model are necessary to obtain the complete virtual system of the power reactor. [ABSTRACT FROM AUTHOR]

In preparation for the utilization of green energy sources and to ensure the sustainability of the national energy system in Indonesia, the nuclear power plant is an option for green energy technology. Since 2014, BATAN (National Nuclear Energy Agency) has taken steps to develop a High Temperature Gas Cooled Reactor (HTGR) technology through the Experimental Power Reactor (EPR) program which is a Non-Commercial Power Reactor. HTGR technology was chosen for EPR because of its inherent safety system, higher power efficiency, and cogeneration reactor. The safety system is the most critical system of a nuclear power plant, so it should be designed with the highest reliability requirements. This study aims to support EPR development activities, especially in developing a Programmable Logic Controller (PLC)-based Reactor Protection System (RPS). RPS is part of the nuclear reactor safety system that functions to keep the reactor operating within normal limits. If any operating parameters exceed the safety limit, the RPS will automatically trip the reactor. The RPS is designed by considering the Design Basic Accident and General Design Criteria of the reactor safety system. In this research, the design of the RPS architecture uses a redundancy logic of 2 out of 3. The RPS output signal will initiate the actuation of the control rod drop and activate the engineered safety features. The result of RPS design includes the RPS design criteria, determination of RPS input signal, RPS architecture modeling, and PLC-based RPS implementation. From the test results, it is found that this RPS can carry out the reactor protection functions in case of abnormal operation contained in the DBA. The measurement of RPS response time indicates this RPS meets the design requirements regarding RPS response time. [ABSTRACT FROM AUTHOR]

Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactors which is projected to be built in Indonesia. The reactor is fueled with molten salt also serving as the primary coolant. The reactor system has unique fuel management where it continuously feeds the fuel and removes some of the fission products while the reactor is operating. The so-called "feed and bleed" fuel management enables the core to have nearly zero excess reactivity with high neutron economy, as the fission product neutron poisons, e.g. Xe-135, are removed from the core. Due to efficient fuel utilization, the core produces much less spent radioactive fuel as compared to LWR. In addition, removing gaseous fission products significantly reduces the source term in case of a severe accident. Most of the current depletion codes do not have a feature enabling the "feed and bleed" fuel management, which at the same time updates the core neutron spectrum. Hence, the software is being developed without modifying the source of the depletion code, such as the MCNP code. The software enables the current depletion codes to perform feeding, bleeding, and updating the fuel composition at requested user time steps. It runs the input file, extracts the requested isotopes from the output file, creates a new input file with new fuel composition, and runs the new input file. This cycling process runs as many requested number of time steps. The purpose of the research is to develop software that meets the need for online continuous feed and bleed fuel management of MSR without too much intervention by the user. The software is written using Python programming language. [ABSTRACT FROM AUTHOR]

Transuranic (TRU) elements are often considered both a proliferation hazard and radiation hazard if not properly managed. Instead of directly disposing it, incinerating TRU in a nuclear reactor can be beneficial in increasing its usefulness as a nuclear fuel whilst simultaneously reducing the aforementioned perceived risks. This study analyses the efficiency of TRU incineration in Passive Compact Molten Salt Reactor (PCMSR) using various types of TRU, namely Reactor Grade Plutonium (RGPu), Reactor Grade Plutonium with Minor Actinide (RGPuMA), Weapon Grade Plutonium (WGPu), and Weapon Grade Plutonium with Minor Actinide (WGPuMA). Key parameters such as keff, temperature coefficient of reactivity (TCR), and transmutation efficiency (TE) were calculated using MCNP6 radiation transport code. The reactor was run in converter mode, ignoring online fuel reprocessing apart from gaseous fission products and non-soluble metals. The result show that for all cases, PCMSR can effectively transmute TRU whilst maintaining negative TCR, fulfilling inherent safety criteria. [ABSTRACT FROM AUTHOR]

The Indonesian government is currently developing a program to achieve Net Zero Emissions (NZE) targets, one of which is the De-dieselization Program. This program replaces Diesel Power Plants (DPPs) to supply the electricity in remote areas with no electricity grid. Microreactors are an alternative solution for replacing DPPs in Indonesia. Microreactors, which have several advantages, including low social and financial risks, will be the answer to the problem of introducing NPPs in Indonesia. Currently, there are many DPPs spread over many locations, and a screening process for the readiness of the microreactor NPP site has been carried out. The initial screening process is to determine the feasibility of 200 diesel locations included in the Phase I List of DPP Conversion on the Electricity Supply Business Plan (RUPTL) of PLN 2021 to 2030, with a capacity greater than or equal to 5 Mwe. The DPP's capacity selection is based on the minimum capacity of the currently available microreactor NPP design. Then the further screening process uses three site rejection criteria, namely surface faulting, seismicity, and volcanoes. The screening result shows there are 70 locations where site aspect screening is carried out. Then, the results of the site aspect screening process show that there are 7 potential locations to be converted by utilizing microreactor NPP. Those potential DPP locations are in Tanjung Batu, Padang Tikar, Semitau, Kangean, Bawean, Salakan/Sambiut, and Lelang/Banggai. [ABSTRACT FROM AUTHOR]

This study aims to consider the fulfillment of the Net Zero Emission (NZE) roadmap targets, identify several obstacles that will be faced, and then provide recommendations for energy stakeholders. The study uses a literature review method, refers to the related research results, and focuses on the role of nuclear energy by considering four aspects: project management, technical, regulatory, and infrastructure readiness. (a) There are two scenario projects to achieve the 35 GWe target by 2060: (i) if construction starts in 2025 and with a large capacity Nuclear Power Plant (NPP), then at least 1 (one) unit must be built every year with a capacity of 1,200 MWe each; (ii) if the development of NPP after 2025, then to achieve the target NPP portion to meet the NZE roadmap, the construction of NPP must be more than 1 unit per year. (b) It is necessary to choose the proven NPP technology and should find an alternative site for the other 18 GWe. (c) Land-based and LWR technology development Regulations are in the completion stage. However, other technologies do not yet exist. (d) The government must immediately establish NEPIO. [ABSTRACT FROM AUTHOR]

Nuclear reactor operation produces spent nuclear fuel and shall be appropriately managed. The policy for managing nuclear-spent fuel in Indonesia follows an open cycle of nuclear material. Nuclear spent fuel is classified as high-level radioactive waste prohibited from reprocessing but shall be disposed of or returned to the country of origin. In decommissioning the TRIGA 2000 reactor in Bandung, the question arises of how to manage the nuclear-spent fuel from the decommissioned reactor. This study is needed to prepare any options that the applicable laws and regulations can carry out. Several aspects that are considered in eachoption are presented in this paper, including technical aspects, safety, security, and safeguards. This study resulted in many options that can be selected that are: (1) nuclear-spent fuel is transported and stored to the TC-ISSF (wet type)using a transport cask, (2) transported and stored using a dual-purpose cask (dry type), (3) repatriated to the country of origin. Each option also discussed what requirements are prepared and fulfilled according to regulations and technical requirements. This study is helpful for policymakers to determine the choice of nuclear-spent fuel management from the decommissioning activities of the TRIGA 2000 reactor in Bandung. [ABSTRACT FROM AUTHOR]

A characterization and analysis with SEM and EDS have been carried out on Zr-4 material after sputtering with titanium target, applications for nuclear power reactor cladding, Pressurized Water Reactor (PWR) and Pressurized Heavy Water Reactor (PHWR) types. Samples Zr-4 which has a length and width of about 1 cm and a thickness of 10 mm, the surface is sputtered with Ti target with voltage variation of 4 kV and 5 kV and the time is also varied for 2, 3 and 4 hours. It was found that the longer the sputtering process on the Zr-4 material, the thicker the Ti layer. The minimum voltage must be 4 kV. The voltage in the sputtering process does not change much to the Ti coating on the sample surface. This research will be continued with a corrosion test which is applied to the cladding material for the nuclear power reactor type PWR and PHWR. [ABSTRACT FROM AUTHOR]

Zircaloy-4 is a material used in the nuclear industry as a cladding material for the PWR (pressurized water reactor) type nuclear reactor. The corrosion behavior of the anodized layer on Zr-4 substrate in 2000 ppm LiOH was analyzed. The anodization process with various voltages has been carried out at 20 and 25 V. Electrolytes for the anodizing process consisted of (NH4)2SO4 and NH4F. The effects of the anodizing voltage on the microstructure, phase composition, elemental distribution, and corrosion resistance of the coatings were investigated by scanning electron microscopy (SEM-EDS), X-ray diffraction (XRD), and potentiodynamic polarization. The results showed that the morphology of the surface changed after anodizing treatment at the voltage of 20 and 25 V. Elemental composition of all coated samples consisted of Zr and O with different concentrations. Anodizing treatment improved its corrosion resistance and mechanics, especially hardness. The morphology of the oxide layer was zirconia tubular with random directions. An increase in anodization time from 5 to 20 minutes could increase the thickness of coatings. All the coatings showed a similar phase structure with the main phase of monoclinic ZrO2 and less tetragonal ZrO2 phase. All oxide films showed improved corrosion resistance in 2000 ppm LiOH, as indicated by lower corrosion current density and passive corrosion potential compared to the zirconium substrate. The corrosion resistance of coatings was better than that of Zr-4 bare. The corrosion potential changed from –0.512 V of bare Zr-4 alloy to -0.214 V of Zr-4 alloy coated. The oxide layer fabricated in 20 and 25 V shows lower corrosion current density, which displays better corrosion resistance. Corrosion current density of Zr-4 substrate and anodized at 20 V for 5, 10, and 20 minutes were 2.39x10−1; 2.1x10−1; 1.5x10−1 and 1.8x10−1 μA/cm2, respectively. Corrosion current density of Zr-4 anodized at 25 V for 5, 10, 20 and 30 minutes were 4.16x10−2; 1.94x10−1and 8.44x10−2 μA/cm2, respectively. The best corrosion resistance was obtained by anodizing at 25 V for 5 minutes, which is 5.16x10−2 μA/cm2. [ABSTRACT FROM AUTHOR]

There are various non-destructive methods presented to measure residual stresses, among which the ultrasonic technique has received the greatest attention in investigating the physical and mechanical properties of materials, detecting of the defects, and resolving the residual stresses for various materials and structures. In recent years, the ultrasonic technique has shown reliable assessment of residual stresses in several fields such as gears, aerospace industry, welded structures, nuclear reactors, shipbuilding, aerospace industry, and construction. This research is concentrated on the ultrasonic technique as a safe test for measuring the residual stresses due to its reliability, accuracy and convenient to use with freedom of radiation hazards. This research expressed the principles of ultrasonic technique with respect to importance, field of application, advantages, disadvantages, the microstructure effect, and the reciprocal relation between direction of ultrasonic wave. [ABSTRACT FROM AUTHOR]

In the present scenario of thermal management system, the cooling efficiency is one of the crucial factors to enhance the performance and life of the heat generating devices. In single phase mode it is very difficult to achieve large heat recovery. So, two phase mode is best suitable method to recover large quantity of heat from the heat sources. Nuclear reactor, electronics cooling, air conditioning system etc., are employing two phase heat transfer for efficient heat recovery. To analyze the heat transfer, the passive technique is more suitable because of less energy consumption. Instead of using conventional fluids for heat transfer, employing nanofluid is an emerging technique in the current heat transfer technology. In the present investigation, a hybrid nanofluid is prepared for 0.6% volume concentration from Al2O3 and SiO2 at equal volume fraction. An experimental investigation was conducted in pool boiling with novel hybrid nanofluid, the safety and performance characteristic were analyzed. [ABSTRACT FROM AUTHOR]