JURNAL TEKNOLOGI REAKTOR NUKLIR TRI DASA MEGA
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ROUTING DESIGN ON THE PRIMARY COOLING PIPING SYSTEM IN PLATE-TYPE CONVERTED TRIGA 2000 REACTOR BANDUNG
Full text linkIn 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 radioactivit
A PRELIMINARY ASSESSMENT OF THE EFFECTS OF DROUGHT ON WATER SUSTAINABILITY INDICATORS FOR NUCLEAR POWER GENERATION IN MONGOLIA
Full text linkAs the effects of climate change are being felt all over the world, sustainability indicators such as water withdrawn per kilowatt-hour, are becoming more important in the decision-making process for large infrastructure projects. In Mongolia, we are deciding whether to use nuclear as a main power source. However, local droughts in Mongolia can be quite severe, occurring every 4-5 years and several countries have shown droughts to interrupt their power plant operations. This study collects data and conducts analyses to estimate sustainability indicators for a nuclear power plant life cycle and extends these analyses to understand how an event such as a drought would affect such indicators. The first part of this study is to provide background information regarding life cycle water use from power generation facilities. Our study focused on the APR-1400 nuclear power plant. If we account for drought frequency in Mongolia, the life cycle water withdrawal is estimated to be approximately 7,611 L/MWh for the nuclear power plant.Keywords: nuclear, sustainability, water, drough
MODELING OF OPERATOR’S ACTIONS ON A NUCLEAR EMERGENCY CONDITION USING MULTILEVEL FLOW MODELING
Full text linkIn nulear emergency condition, after determining the initiating event and the type of the anomaly, operators should take counteractions to control the reactor to mitigate the accident and to bring back the plant to the safe condition. The actions should based on emergency operating procedures. In order to minimize the human error related to the actions, some necessary information is needed. Such kind of information is the consequence of the actions, which can be derived by modeling the counteractions. Multilevel flow modeling (MFM), a functional modeling, is chosen to model the counteraction with the consideration that it is based on cause-effect relations and consequence reasoning, it provides realization relationship which corresponds physical components with their functions, and it provides comprehensive diagnosis based on human perspective of the system objectives. The counteractions are represented by the control functions in the MFM. This paper discusses how to model the counteractions and the consequences of the actions to the system components, which are necessary to enhance situation awareness and to reduce human errors.Keywords: Operator actions, emergency operating procedures, multilevel flow modeling, control function, nuclear safety, human erro
REQUIREMENT ANALYSIS OF COMPUTER-BASED INSTRUMENTATION AND CONTROL SYSTEM FOR REAKTOR DAYA EKSPERIMENTAL
Full text linkDeveloping and licensing of digital Instrumentation and control (I&C) system for nuclear power plant (NPP) are challenging especially for the new construction since digital technology are composite with a very high complexity of many integrated systems. National Nuclear Energy Agency of Indonesia (BATAN), who design Reaktor Daya Eksperimental (RDE), should prepare the documents to meet the licensing requirements of national regulator in this case Nuclear Energy Regulatory Agency of Indonesia (BAPETEN). BAPETEN’s chairman regulation No.6 year of 2012 is the first national requirement which state requirements related to design of computer-based system concerning on safety of power reactor that should be followed. Since BAPETEN only denotes requirements without state which code and standards to be used, therefore BATAN can add references from International Nuclear Energy Agency (IAEA) guidelines. In this paper, requirement document traceability is developed to determine which code and standards should be used to verify and validate the I&C computer-based system of RDE. The hierarchy of regulatory and utility requirements are developed to guide the design basis documentation. Developing requirements analysis of computer-based I&C system RDE are completed after determining the design requirements from the utility and regulatory requirements. This methodology will help the design engineers to follow the utility requirements by concerning to the production, and follow the regulatory requirements concerning the safety aspect.Keywords: Computer-based system, I&C System, Requirements analysis, Licensing, RD
REACTOR OPERATIONAL EXPERIENCE REVIEW AND ANALYSIS BASED ON UN-INTENDED REACTOR TRIP DATA
Full text linkTo enhance the safety and reliability of a new reactor, human factors should be integrated into its design process. The experimental power reactor (RDE) currently being developed in Indonesia needs to include human factors in the design process. One approach to incorporate human factors into design is by considering reactor operational experience data. This paper reviews and analyses the operational experience data of RSG-GAS reactor. The operational experience data of RSG-GAS reactor with 40,435 hours of total operation time spanning from 2003 to 2013 was used as a base in the study. In depth analysis on human factors was applied to the primary cooling system using Human Factors Analysis and Classification System-HFACS method. An amount of 289 un-intended trips were found in the observation data period. Most of un-intended trip were caused by external factors (38%). A review on the primary and secondary cooling system operational data showed that 3.11% of un-intended reactor trip occurrence causes were associated with human failure. Most suspected human failure/human error corresponds to the pump maintenance task which is classified as A action category. Analysis on the cooling system based on HFACS showed that the challenges to the human factors are related to unsafe acts, preconditions of unsafe acts, and unsafe supervision. The result reaffirm that human factors should be treated appropriately in the design of reactor equipment and operation procedure as well.Keywords: reactor operation experience, research reactor, human factors, reactor tri
SIMULATION OF FEED WATER TEMPERATURE DECREASE ACCIDENT IN NUSCALE REACTOR
Full text linkStudy on thermal hydraulic behavior of the NuScale reactor during secondary system malfunction that causes a feed water temperature decrease has been conducted using RELAP5 code. This study is necessary to investigate the performance of safety system and design in dealing with an accident. The method used involves simulation of reactor transient through numerical modeling and calculation in RELAP5 code covering primary and secondary system, including the decay heat removal system (DHRS). The investigation focuses on the flow and heat transfer characteristics that occurs during the transient. The calculation result shows that at the beginning, core power increases up to trip set point of 200 MW which is driven by positive feedback reactivity of coolant overcooling and automatic control rod bank adjustment. Meanwhile, the core exit coolant temperature increases up to 600 K. and primary system circulation flow rate speeds up to 556 kg/s. After that, the reactor trips and power drops sharply, followed by opening of DHRS valves and closing of steam line and feed water isolation valves. The simulation shows that, the DHRS are capable to transfer decay heat to the reactor pool and as a result the primary system temperature and pressure decreases. The reactor could stay in safe shutdown state afterward.Keywords: NuScale, RELAP5, feed water, decay heat, simulation SIMULASI KECELAKAAN PENURUNAN TEMPERATUR AIR UMPAN DI REACTOR NUSCALE. Studi tentang perilaku termalhidraulik reaktor NuScale saat terjadi kerusakan sistem sekunder yang menyebabkan penurunan suhu air umpan telah dilakukan dengan menggunakan kode RELAP5. Penelitian ini penting untuk menyelidiki kinerja disain dan sistem keselamatan reaktor dalam menghadapi kecelakaan. Metoda yang digunakan melibatkan simulasi transien reaktor melalui pemodelan dan kalkulasi numerik dengan RELAP5 yang meliputi sistem primer dan sekunder serta sistem pembuangan panas peluruhan (DHRS). Investigasi berfokus pada aliran dan karakteristik perpindahan panas yang terjadi selama transien. Hasil perhitungan menunjukkan bahwa pada awalnya, terjadi peningkatan daya teras hingga mencapai titik seting pemadaman (trip) 200 MW, sebagai akibat dari umpan balik reaktivitas positif dari pendinginan fluida sistem primar dan respon otomatis penaikan batang kendali. Sementara itu, suhu keluaran teras meningkat menjadi 600 K serta laju aliran sirkulasi sistem primer meningkat menjadi 556 kg/s. Setelah itu, reaktor padam dimana daya menurun tajam dan diikuti pembukaan katup DHRS dan penutupan katup pada jalur uap dan air umpan. Simulasi ini menunjukkan bahwa, DHRS mampu membuang panas ke kolam reaktor, dimana suhu serta tekanan sistem primer menurun. Reaktor tetap dalam keadaan shutdown aman sesudahnya.Kata kunci: NuScale, RELAP5, air umpan, panas peluruhan, simulas