scholarly journals Uranium Exploration, Deposit and Resources: The Key of Nuclear Power Plant Development Program in Indonesia

2021 ◽  
Vol 2048 (1) ◽  
pp. 012003
Author(s):  
H Syaeful ◽  
I G Sukadana ◽  
Y S B Susilo ◽  
F D Indrastomo ◽  
A G Muhammad ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Development Program ◽  
Indonesian Archipelago ◽  
Uranium Fuel ◽  
Uranium Exploration ◽  
Mineral Production ◽  
Area North ◽  
The 1960S ◽  
Almost All

Abstract Uranium deposit in Indonesia was found in almost all Indonesian Archipelago, mainly in Kalimantan, Sulawesi, Sumatera, Papua, Bangka Belitung and Riau islands. Uranium exploration activities started in the 1960s to recent, conducted in many exploration stages. The exploration in prospects area are completed with drilling activities to delineate the mineralization zone and continued to resources estimation. In Kalan Area, the research had been completed with underground/tunneling mining. The uranium resources are classified into discovered or undiscovered based on exploration stages, and conventional or unconventional based on sources of primary/secondary/by-product mineral production. The resources are calculated from Kalan Area and its surroundings (Kalimantan) with addition of Mamuju Area (West Sulawesi) and Sibolga Area (North Sumatera). Uranium identified resource in Indonesia is 13,503 tU while the undiscovered is 62,330 tU. Meanwhile, categorized by uranium source, the conventional and unconventional resources are 48,388 tU and 27,445 tU respectively. The uranium resources categories should be increased and completed with feasibility study to increase the resources to reserve classification. The exploration, deposit, and resources are the key to ensure the readiness of developing nuclear power plants in Indonesia, where one of them is Experimental Power Reactor (EPR) or Reaktor Daya Eksperimental (RDE) with domestic uranium fuel.

Predictor Displays for Complex, Dynamic Tasks: A Preliminary Investigation

1986 ◽  
Vol 30 (7) ◽  
pp. 684-688 ◽  
Author(s):  
K. B. Bennett ◽  
D. D. Woods ◽  
E. M. Roth ◽  
P. H. Haley
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Precise Measurement ◽  
Preliminary Investigation ◽  
Development Program ◽  
Manual Control ◽  
Control Performance ◽  
Control Task ◽  
Complex Dynamic ◽  
Operator Performance

The operators of nuclear power plants are asked to perform a task that has proven to be particularly difficult: manual control of feedwater during startup. We have initiated a research and development program to address human factors issues related to this task. An analysis of cognitive aspects of the feedwater control task was used to develop a generic part-task simulator. New displays to enhance manual control performance (including a predictor display) were developed with the simulator. The test capability provided by the simulator allowed precise measurement of performance differences associated with these displays in a mixed-fidelity laboratory experiment. The results suggest that the displays reduced the complexity of the task and resulted in improved operator performance.

scholarly journals Experimental Research of Valve Tightness at Different Closure Forces

2018 ◽  
pp. 14-17
Author(s):  
A. V. Koroliov ◽  
P. Y. Pavlyshyn ◽  
I. V. Bandurko
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Valve Closure ◽  
Criterial Equation ◽  
Experimental Data Processing ◽  
Rotational Moment ◽  
Valve Leakage ◽  
Incomplete Closure ◽  
Almost All

Power valves are installed on almost all the pipelines of nuclear power plants performing the functions of regulation and shutting off the flow, so its failure often leads to emergencies. A particularly large number of failures is observed in motor-operated valves. Incorrect setting of the limiting clutch leads either to incomplete closure of the valve or to rod failure. Therefore, the valves are equipment of a nuclear power plant, which often falls into repair shops. Failures leading to an increase of valve leakage are especially dangerous for nuclear power plants. In this case, leakage of high-pressure valves leads to erosion of the sealing surfaces, which only increases the leakage. Thus, it is very important to determine the optimum rotational value when the valve is closed. The lack of conditions for closure force in the standards for valve leakage complicates the issue. A bench that allows working in the air with a pressure up to 3.5 MPa was developed on valve rod to study dependence of valve leakage on the rotational moment. Four independent parameters were measured: air pressure in front of the valve under study, closure force of the valve, volume of air loss through the valve and leakage time. A standard stop valve with a nominal diameter of 15 mm and a nominal pressure of 64 atm was used for the study. The determined dependence of the leakage on torque value allows recommending a gentler mode of valve closure without significantly reducing its tightness. As a result of experimental data processing, a criterial equation is obtained linking a leakage rate, pressure drop on the valve and a rotational moment value. The received criterial equation will allow defining the compromise between valve closure force and permissible leak level according to regulatory requirements. The analysis of the “leakage/rotational moment” diagram showed the possibility to reveal the damaged valves. This possibility may be used during the incoming inspection of the valves supplied to NPP, which should significantly improve the reliability of their operation.

Development of RI-ISI Program for Krško NPP

2013 ◽  
Author(s):  
Scott Kulat ◽  
Robin Graybeal ◽  
Benjamin Montgomery ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Selection Process ◽  
Degradation Mechanism ◽  
Electrical Power ◽  
Implementation Program ◽  
Class 1 ◽  
Process Risk ◽  
Almost All ◽  
The U.S

Risk-informed methodologies for inservice inspections of safety related piping in nuclear power plants were formally established in mid-1990s in the U.S. Since then, they have been adopted and applied by almost all of the U.S. plants. Nowadays, risk-informed inservice inspection (RI-ISI) is considered to be a standard for the operating plants in the U.S. It was not long before the RI-ISI practice started to be “exported” from the U.S. to other countries. By now, RI-ISI had found its way to a number of European and other countries. Among the recent examples is the Krško Nuclear Power Plant (Krško NPP), a two-loop Westinghouse-designed PWR located in Slovenia and owned by Slovenian and Croatian utilities. Krško NPP finished its third inservice inspection (ISI) interval in July 2012 and initiated implementation of the RI-ISI program at the start of the fourth interval. The process used to develop the RI-ISI program conformed to the methodology described in Electrical Power Research Institute (EPRI) Topical Report TR-112657 and included a degradation mechanism evaluation, consequence analysis and risk characterization for ASME Class 1 and Class 2 piping, as well as an element/examination selection process, risk impact assessment and inspection implementation program development. This paper describes the development of the Krško NPP RI-ISI program and the results of its RI-ISI application. A discussion is, also, provided on some aspects relevant for application of RI-ISI approaches developed in the U.S to plants outside of the U.S.

Study on the Coal-Based Synfuels System Coupled with Nuclear Energy

2013 ◽  
Vol 805-806 ◽  
pp. 1429-1433 ◽  
Author(s):  
Jin Li ◽  
Chu Fu Li
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Energy ◽  
Coal Gasification ◽  
Coupled System ◽  
Low Carbon ◽  
Coal Consumption ◽  
Peak Shaving ◽  
Water Electrolyser ◽  
Almost All

Coal-based synfuels plants are facing serious pressure on CO2 emissions reduction. Developing the coal-based synfuels system coupled with nuclear energy is an effective approach to reduce CO2 emissions. This work analyzes CO2 emissions features in the coal-based synfuels system, and further investigates three coupling paths between the coal-based synfuels system and nuclear energy. Subsequently, an inherent-safety and low-carbon coal-based synfuels system coupled with nuclear energy is proposed. In the coupled system, valley nuclear power is provided to conventional water electrolyser for hydrogen/oxygen production, and oxygen and hydrogen are supplied to the coal gasification and fuel synthesis processes, respectively. The simulation results show that the coupled system can reduce about 50% raw coal consumption and almost all CO2 emissions compared to the conventional coal-based synfuels system, meanwhile it can improve the peak shaving capacity of nuclear power plants.

scholarly journals Towards Developing of Oberon System with Specific Requirements of Ergodicity

2020 ◽  
Vol 32 (6) ◽  
pp. 67-78
Author(s):  
Dmitry Victorovich Dagaev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
High Reliability ◽  
Computer Systems ◽  
Critical Systems ◽  
Qualitative Properties ◽  
Runtime Environment ◽  
Almost All ◽  
Time Systems ◽  
Reliability Systems

The ergodic system keeps the time average is the same for almost all initial points. It is important for computer systems to prevent the degradation of the properties of the system over time. Ergodicity is especially required for mission-critical systems in demanding industries. Software development based on the functional safety requirements of the IEC 60880 category A standard is implemented only on newly created software that meets the most stringent requirements for nuclear power plants, it is impossible to use standard operating systems and compilers. For these purposes, a prototype of the runtime environment and application software of the command display system (DSCU) was implemented. The runtime was created based on the Active Oberon A2 system. A2 is a single-user multi-tasking system. Application area - industrial embedded real-time systems, high reliability systems. The DSKU execution environment is implemented by a significant revision of the minimum subset A2 to meet the requirements of the standard. The system of restrictions formed according to the requirements of the standard makes it possible to create computer systems with new properties. The use of these constraints leads to the proof that there is no possibility of the occurrence of the failures they cause and allows us to consider a computer system based on the presumption of non-ergodicity. This «via negative» approach is based on restrictions, the addition of which allows one to obtain new qualitative properties. The more restrictions, the greater the gain in system reliability and stability.

Effect of Impurities on Compressor and Cooler in Supercritical CO2 Cycles

2018 ◽  
Vol 141 (1) ◽  
Author(s):  
Ladislav Vesely ◽  
K. R. V. Manikantachari ◽  
Subith Vasu ◽  
Jayanta Kapat ◽  
Vaclav Dostal ◽  
...  
Keyword(s):  
Critical Point ◽  
Carbon Capture ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Waste Heat ◽  
Operating Conditions ◽  
Working Fluid ◽  
Compressor Inlet ◽  
Almost All

With the increasing demand for electric power, the development of new power generation technologies is gaining increased attention. The supercritical carbon dioxide (S-CO2) cycle is one such technology, which has relatively high efficiency, compactness, and potentially could provide complete carbon capture. The S-CO2 cycle technology is adaptable for almost all of the existing heat sources such as solar, geothermal, fossil, nuclear power plants, and waste heat recovery systems. However, it is known that optimal combinations of operating conditions, equipment, working fluid, and cycle layout determine the maximum achievable efficiency of a cycle. Within an S-CO2 cycle, the compression device is of critical importance as it is operating near the critical point of CO2. However, near the critical point, the thermo-physical properties of CO2 are highly sensitive to changes of pressure and temperature. Therefore, the conditions of CO2 at the compressor inlet are critical in the design of such cycles. Also, the impurity species diluted within the S-CO2 will cause deviation from an ideal S-CO2 cycle as these impurities will change the thermodynamic properties of the working fluid. Accordingly, the current work examines the effects of different impurity compositions, considering binary mixtures of CO2 and He, CO, O2, N2, H2, CH4, or H2S on various S-CO2 cycle components. The second part of the study focuses on the calculation of the basic cycles and component efficiencies. The results of this study will provide guidance and define the optimal composition of mixtures for compressors and coolers.

Advanced Construction Methods for New Nuclear Power Plants

2010 ◽  
Author(s):  
Dae-Yul Jung ◽  
Yoon-Kee Kang ◽  
Chang-Hyung You
Keyword(s):  
Construction Projects ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Critical Path ◽  
Development Program ◽  
3D Cad ◽  
Cad System ◽  
Construction Methods ◽  
Construction Schedule

This paper covers advanced construction technologies that are generally used for nuclear power plants and presents advanced construction methods that can contribute to an efficient and short construction schedule. The construction schedule is driven by the activities of the critical path. The advanced construction methods consist of Modularization, Improvement of Mechanical Rebar splices, Application of 3D-CAD system for information and control, and the installation of RVI and RCL at the same time. Some methods have been applied in actual Nuclear power projects and others have been developing under the research and development program. It incorporates the experiences and insights from recent nuclear construction projects all over the world.

A Novel More Reliable and Extensible Architecture of Instrumentation and Control Systems

2018 ◽  
Author(s):  
Shuqiao Zhou ◽  
Chao Guo ◽  
Duo Li ◽  
Xiaojin Huang
Keyword(s):  
Communication Technology ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Traditional Architecture ◽  
Digital Instrumentation ◽  
Control Units ◽  
Computing Performance ◽  
Almost All ◽  
And Control

Digital instrumentation and control (I&C) systems are widely used in many industrial areas. In the recent years, the digitalization process for nuclear power plants has also been moving on rapidly. Full digital I&C systems are now adopted in almost all new constructed nuclear power plants. The architecture of a digital I&C system plays a pivotal role for the safety, reliability and security of the whole nuclear power plant. Moreover, for the advanced small modular reactors, both the reliability and extensibility of I&C systems are especially required. Therefore, in this paper we propose a new architecture of the digital I&C systems based on the developed computing performance and communication technology. The control units and the data servers in the new proposed architecture are decentralized and working in a mutually redundant and distributed computing/storage way. Thus the architecture is with a flexible extensibility. Moreover, other control units or data servers can take over the functions of a certain number of failed ones. This characteristic benefits the system’s reliability significantly. The reliability of the new architecture is theoretically evaluated and the results demonstrate that it is much higher than that of the traditional architecture of I&C systems.

Development of a High-Temperature Molten Lithium Pump

1966 ◽  
Vol 88 (1) ◽  
pp. 13-21
Author(s):  
R. W. Kelly ◽  
G. M. Wood ◽  
J. J. Milich ◽  
C. Ferguson ◽  
D. V. Manfredi
Keyword(s):  
High Temperature ◽  
Liquid Metal ◽  
Nuclear Power ◽  
High Performance ◽  
Power Plants ◽  
High Efficiency ◽  
High Reliability ◽  
Development Program ◽  
Development Programs ◽  
Molten Lithium

The circulation of the liquid-metal heat-transport fluids used in high-performance, mobile, nuclear power plants requires high-temperature pumps. These pumps must be capable of moderately high efficiency over a very long lifetime and have small size, low weight, and high reliability. As an initial phase of a lithium pump development program and to provide pumps for companion development programs, a 195-gpm pump was designed and successfully developed. Extensive testing of pump components, as well as water and liquid-metal tests of complete pump assemblies, was accomplished to meet the program objectives of high performance and high reliability for the required long operating lifetime. Several successful lithium tests of 10,000-hr duration were accomplished with the lithium development pumps and a pump used in a companion heat-exchanger development program.

APR1000: Further Improvements of the OPR1000 Design

2010 ◽  
Author(s):  
Insik Kim ◽  
Sung-Goo Chi ◽  
Keun-Bae Yoo
Keyword(s):  
Cyber Security ◽  
Nuclear Power ◽  
Power Plants ◽  
Sea Water ◽  
Power Reactor ◽  
Development Program ◽  
Nuclear Industry ◽  
Design Features ◽  
Design Requirements ◽  
Under Construction

Korean nuclear industry has demonstrated a success of KSNP (Korean Standard Nuclear Power Plant) program through operations of its six units starting from Ulchin 3&4 since 1998, and four more units are currently under construction with its new name of OPR1000 (Optimized Pressurized Reactor 1000). During the past 20 years of OPR1000 constructions and operations, Korean nuclear industry has put continuous effort to secure its better confidence in safety and performance. As a result, OPR1000 is showing the highest capacity factor and the lowest unplanned trip records among the worldwide operating nuclear power plants. The goal of APR1000 (Advanced Power Reactor 1000) program launched recently is to develop an improved version of OPR1000 design. It is required to adopt selected advanced design features of APR1400 (Advanced Power Reactor 1400) and to fulfill the new and strengthened design requirements including longer plant lifetime, colder head reactor, longer refueling interval, enhanced cyber security, etc. It is also essential to consider regional specific conditions such as electric frequency, sea water temperature and seismic conditions, etc. In this paper, the overall APR1000 development program is introduced, and design requirements as well as advanced design features (ADFs) for APR1000 are discussed comparing with those for OPR1000.

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