WorldWideScience

Sample records for developing nuclear plant

  1. Development of nuclear power plant Risk Monitor

    International Nuclear Information System (INIS)

    Yang Xiaoming; Sun Jinlong; Ma Chao; Wang Lin; Gu Xiaohui; Bao Zhenli; Qu Yong; Zheng Hao

    2014-01-01

    Risk Monitor is a tool to monitor the real-time risk of a nuclear power plant for risk management and comprehensive decision-making, which has been widely used all over the world. The nuclear power plant Risk Monitor applies the real-time risk model with low-complicacy that could reflect the plant's actual configuration, automatically reads the plant's configuration information from the engineering system through the developed interface, and efficiently analyzes the plant's risk Dy the intelligent parallel-computing method in order to provide the risk basement for the safety management of nuclear power plant. This paper generally introduces the background, architecture, functions and key technical features of a nuclear power plant Risk Monitor, and validates the risk result, which could well reflect the plant's risk information and has a significant practical value. (authors)

  2. Engineering development in nuclear power plant construction

    International Nuclear Information System (INIS)

    Guenther, P.

    1979-01-01

    Proceeding from the up-to-now experience in the erection of nuclear power stations, especially of the first and second unit of the Greifswald nuclear power plant, the following essential aspects of the development of constructional engineering are discussed: (1) constructional features and criteria, (2) organizational management, (3) current status and problems in prelimary operations, and (4) possibilities of further expenditure reductions in constructing nuclear power stations

  3. Nuclear plant analyzer development at INEL

    International Nuclear Information System (INIS)

    Laats, E.T.; Russell, K.D.; Stewart, H.D.

    1983-01-01

    The Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission (NRC) has sponsored development of a software-hardware system called the Nuclear Plant Analyzer (NPA). This paper describes the status of the NPA project at the INEL after one year of development. When completed, the NPA will be an integrated network of analytical tools for performing reactor plant analyses. Development of the NPA in FY-1983 progressed along two parallel pathways; namely, conceptual planning and software development. Regarding NPA planning, and extensive effort was conducted to define the function requirements of the NPA, conceptual design, and hardware needs. Regarding software development conducted in FY-1983, all development was aimed toward demonstrating the basic concept and feasibility of the NPA. Nearly all software was developed and resides on the INEL twin Control Data Corporation 176 mainframe computers

  4. SWOT of nuclear power plant sustainable development

    International Nuclear Information System (INIS)

    Abbaspour, M.; Ghazi, S.

    2008-01-01

    SWOT Analysis is a Useful tool that can he applied to most projects or business ventures. In this article we are going to examine major strengths, weaknesses, opportunities and threats of nuclear power plants in view of sustainable development. Nuclear power plants have already attained widespread recognition for its benefits in fossil pollution abatement, near-zero green house gas emission, price stability and security of energy supply. The impressive new development is that these virtues are now a cost -free bonus, because, in long run, nuclear energy has become an inexpensive way to generate electricity. Nuclear energy's pre-eminence economically and environmentally has two implications for government policy. First, governments should ensure that nuclear licensing and safety oversight arc not only rigorous but also efficient in facilitating timely development of advanced power plants. Second, governments should be bold incentivizing the transformation to clean energy economics, recognizing that such short-term stimulus will, in the case of nuclear plants, simply accelerate desirable changes that now have their own long-term momentum. The increased competitiveness of nuclear power plant is the result of cost reductions in all aspects of nuclear economics: Construction, financing, operations, waste management and decommissioning. Among the cost-lowering factors are the evolution to standardized reactor designs, shorter construction periods, new financing techniques, more efficient generation technologies, higher rates of reactor utilization, and longer plant lifetimes. U.S World Nuclear Association report shows that total electricity costs for power plant construction and operation were calculated at two interest rates. At 10%, midrange generating costs per kilowatt-hour are nuclear at 4 cents, coal at 4.7 cents and natural gas at 5.1 cent. At a 5% interest rate, mid-range costs per KWh fall to nuclear at 2.6 cents, coal at 3.7 cents and natural gas at 4.3 cents

  5. Development of nuclear plant Operation Management System

    Energy Technology Data Exchange (ETDEWEB)

    Koide, I.; Okada, T.; Ishida, K. [Chubu Electric Power Co. Inc., Nagoya (Japan)

    1998-09-01

    Recently it has become more important to detect a change in operational characteristics and to take appropriate corrective actions before it deteriorates to an incident in nuclear power plants. Therefore, aiming at earlier detection of a tendency change, swifter corrective actions and more effective application of operational data, we have developed Operation Management System which automatically acquires, accumulates and observes operational data of Hamaoka Nuclear Power Station through cycles. (author)

  6. Development of nuclear plant Operation Management System

    International Nuclear Information System (INIS)

    Koide, I.; Okada, T.; Ishida, K.

    1998-01-01

    Recently it has become more important to detect a change in operational characteristics and to take appropriate corrective actions before it deteriorates to an incident in nuclear power plants. Therefore, aiming at earlier detection of a tendency change, swifter corrective actions and more effective application of operational data, we have developed Operation Management System which automatically acquires, accumulates and observes operational data of Hamaoka Nuclear Power Station through cycles. (author)

  7. Nuclear plant analyzer development and analysis applications

    International Nuclear Information System (INIS)

    Laats, E.T.

    1984-10-01

    The Nuclear Plant Analyzer (NPA) is being developed as the US Nuclear Regulatory Commission's (NRC's) state of the art safety analysis and engineering tool to address key nuclear plant safety issues. This paper describes four applications of the NPA in assisting reactor safety analyses. Two analyses evaluated reactor operating procedures, during off-normal operation, for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The third analysis was performed in support of a reactor safety experiment conducted in the Semiscale facility. The final application demonstrated the usefulness of atmospheric dispersion computer codes for site emergency planning purposes. An overview of the NPA and how it supported these analyses are the topics of this paper

  8. Development of a nuclear plant analyzer (NPA)

    International Nuclear Information System (INIS)

    De Vlaminck, M.; Mampaey, L.; Vanhoenacker, L.; Bastenaire, F.

    1990-01-01

    A Nuclear Plant Analyzer has been developed by TRACTABEL. Three distinct functional units make up the Nuclear Plant Analyser, a model builder, a run time unit and an analysis unit. The model builder is intended to build simulation models which describe on the one hand the geometric structure and initial conditions of a given plant and on the other hand command control logics and reactor protection systems. The run time unit carries out dialog between the user and the thermal-hydraulic code. The analysis unit is aimed at deep analyzing of the transient results. The model builder is being tested in the framework of the International Standard Problem ISP-26, which is the simulation of a LOCA on the Japanese ROSA facility

  9. Development of robots for nuclear power plants

    International Nuclear Information System (INIS)

    Sasaki, Masayoshi

    1982-01-01

    In nuclear power plants, the reduction of maintenance time, the reduction of radiation exposure and man-power saving are increasingly required. To achieve these purposes, various remote-controlled devices, such as robots in a broad sense, have been earnestly developed. Of these, three machines for replacing, four devices for inspection, two systems for cleaning, and two equipment for processing are tabulated in this paper. Typical eight machines or equipment are briefly introduced, mainly describing their features or characteristics. Those are: a remotely handling machine for control rod drive mechanism, an automatic refueling machine, an automatic ultrasonic flaw detection system replacing for a manually operated testing system for the welded parts of primary cooling system, an automatic cask washing machine for decontamination, a floor-type remote inspection vehicle for various devices operating inside power plants, a monorail-type remote inspection vehicle for inspection in spaces where floor space is short, and a remote-controlled automatic pipe welding machine for welding operations in a radioactive environment such as replacing the piping of primary cooling system. Most of these devices serves for radiation exposure reduction at the same time. Existing nuclear power plant design assumes direct manual maintenance, which limits the introduction of robots. Future nuclear power plants should be designed on the assumption of automatic remote-controlled tools and devices being used in maintenance work. (Wakatsuki, Y.)

  10. Nuclear plant analyzer development and analysis applications

    International Nuclear Information System (INIS)

    Laats, E.T.

    1984-01-01

    The Nuclear Plant Analyzer (NPA) is being developed as the U.S. Nuclear Regulatory Commission's (NRC's) state of the art safety analysis and engineering tool to address key nuclear plant safety issues. The NPA integrates the NRC's computerized reactor behavior simulation codes such as RELAP5 and TRAC-BWR, both of which are well-developed computer graphics programs, and large repositories of reactor design and experimental data. Utilizing the complex reactor behavior codes as well as the experiment data repositories enables simulation applications of the NPA that are generally not possible with more simplistic, less mechanistic reactor behavior codes. These latter codes are used in training simulators or with other NPA-type software packages and are limited to displaying calculated data only. This paper describes four applications of the NPA in assisting reactor safety analyses. Two analyses evaluated reactor operating procedures, during off-normal operation, for a pressurized water reactor (PWR) and a boiling water reactor (BWR), respectively. The third analysis was performed in support of a reactor safety experiment conducted in the Semiscale facility. The final application demonstrated the usefulness of atmospheric dispersion computer codes for site emergency planning purposes. An overview of the NPA and how it supported these analyses are the topics of this paper

  11. Developments in nuclear power plant water chemistry

    International Nuclear Information System (INIS)

    Fruzetti, K.; Wood, C.J.

    2007-01-01

    This paper illustrates the changing role of water chemistry in current operation of nuclear power plants. Water chemistry was sometimes perceived as the cause of materials problems, such as denting in PWR steam generators and intergranular stress corrosion cracking in BWRs. However, starting in the last decade, new chemistry options have been introduced to mitigate stress corrosion cracking and reduce fuel performance concerns. In BWRs and PWRs alike, water chemistry has evolved to successfully mitigate many problems as they have developed. The increasing complexity of the chemistry alternatives, coupled with the pressures to increase output and reduce costs, have demonstrated the need for new approaches to managing plant chemistry, which are addressed in the final part of this paper. (orig.)

  12. Development of Dukovany nuclear power plant

    International Nuclear Information System (INIS)

    Kouklik, I.

    2000-01-01

    The presentation evaluates the present situation of the Dukovany nuclear power plant operation. It analyses the present at the plant, and also in both the near and distant localities. Based on these analyses the conclusions and targets that are necessary for keeping the plant's safe operation are made. From the current situation the report about the results of the year 1999 in the operation and financial field, stress on the safety index, is presented. Further, there is a list of finished and semi-finished activities that are taking place within the harmonization of the Dukovany NPP. At the same time the report presents the terms of licensing and the going of implementation of the 'KAPA' project - the complex application of the Atomic Energy Act. In the part, comparison to the environment, there is an example of a comparison with the results of a plant in the EU. Further the report compares the advantages and disadvantages of the WWER reactors with other light water reactors. In the last part of the presentation there is a brief opinion of the author on the steps that are awaiting us in the future. First, the 'Harmonization program' of Dukovany NPP is explained, its philosophy and principles. The basic idea is to take the right steps and strategic decisions, with the vision for the operation till the year 2025 at least. There is also a list of the most important events that are planned to increase the safety. Further the report goes briefly over the current situation in the filed of nuclear energy in the period of joining the European Union and also mentions the risks that result from this process. It concludes with the prognosis for the further position of the Dukovany nuclear power plant. (author)

  13. New developments in nuclear power plant construction

    International Nuclear Information System (INIS)

    Bivens, A.C.

    1983-01-01

    Specific examples of construction activities are presented which demonstrate that excellent results have been achieved in the areas of cost, schedule and quality. Examples of innovation and development are given that would be particularly applicable to future work either for new plants or for plants not yet completed. (author)

  14. Development of a decommissioning plan for nuclear power plant 'Krsko'

    International Nuclear Information System (INIS)

    Tankosic, Djurica; Fink, Kresimir

    1991-01-01

    Nuclear Power Plant 'Krsko' (NEK), is the only nuclear power plant in Yugoslavia, is a two-loop, Westinghouse-design, pressurized water reactor rated at 632 MWe. When NEK applied for an operating license in 1981, it did not have to explain how the plant would be decommissioned and decommissioning provisions were not part of the licensing process. Faced with mounting opposition to nuclear power and a real threat that the plant would be shut down, the plant management developed a Mission Plan for resolving the decommissioning problem. The Mission Plan calls for a preliminary decommissioning plan to be prepared and submitted to the local regulatory body before the end of 1992

  15. Appliance of software engineering in development of nuclear power plant

    International Nuclear Information System (INIS)

    Baek, Y. W.; Kim, H. C.; Yun, C.; Kim, B. R.

    1999-01-01

    Application of computer technology in nuclear power plant is also a necessary transformation as in other industry fields. But until now, application of software technology was not wide-spread because of its potential effect to safety in nuclear field. It is an urgent theme to develop evaluation guide and regulation techniques to guarantee safety, reliability and quality assurance. To meet these changes, techniques for development and operation should be enhanced to ensure the quality of software systems. In this study, we show the difference between waterfall model and software life-cycle needed in development of nuclear power plant and propose the consistent framework needed in development of instrumentation and control system of nuclear power plant

  16. Appliance of software engineering in development of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Y. W.; Kim, H. C.; Yun, C. [Chungnam National Univ., Taejon (Korea, Republic of); Kim, B. R. [KINS, Taejon (Korea, Republic of)

    1999-10-01

    Application of computer technology in nuclear power plant is also a necessary transformation as in other industry fields. But until now, application of software technology was not wide-spread because of its potential effect to safety in nuclear field. It is an urgent theme to develop evaluation guide and regulation techniques to guarantee safety, reliability and quality assurance. To meet these changes, techniques for development and operation should be enhanced to ensure the quality of software systems. In this study, we show the difference between waterfall model and software life-cycle needed in development of nuclear power plant and propose the consistent framework needed in development of instrumentation and control system of nuclear power plant.

  17. Development of nuclear power plants database system, (2)

    International Nuclear Information System (INIS)

    Izumi, Fumio; Ichikawa, Michio

    1984-06-01

    A nuclear power plant data base system has been developed. The data base involves a large amount of safety design informations for nuclear power plants on operating and planning stage in Japan. The informations, if necessary, can be searched for at high speed by use of this system. The present report is an user's guide for access to the informations utilizing display unit of the JAERI computer network system. (author)

  18. Nuclear Power Plants and Sustainable Development on a Liberalized Market

    International Nuclear Information System (INIS)

    Androcec, I.; Stanic, Z.; Tomsic, Z.

    2002-01-01

    Finding a way to generate electricity so as to satisfy the terms of sustainable development of the entire society is the only way which will secure safe energy future. If we talk about energy in the context of sustainable development, one of the most important element is environmental protection. Since CO 2 emissions stemming from electricity generation have predominant impact on climate change, one of the options for reducing emissions is the use of fuels without carbon, such as e.g. nuclear fuel. The future of nuclear power plants was considered in view of: nuclear fuel supply; potential impact of fuel cycle on environment, power plant operation, decommissioning and secondary products from electricity generation; and the entire nuclear power plant economy. Nuclear power plants were also examined in the context of the Kyoto Protocol stipulating reduction of greenhouse gases emissions. Nuclear power plants can not reduce CO 2 emissions in a short-term because they already operate with maximum output, but in a long-run they can play a significant role. This paper is aiming to analyse the role of nuclear power plants in long term environmental sustainability in electricity sector reform (liberalisation, deregulation, privatisation) in small or medium sized power supply systems. Nuclear power plants are associated with certain environmental aspects which will be taken into account. A comparison will be made through externalities with other energy resources, especially fossil fuels, which are prevailing energy resources, considering possible use of nuclear power plants in the countries with small and medium-size grids. It will be given an example of the role of NPP Krsko on air emissions reduction in Croatia. (author)

  19. Nuclear Plant Analyzer development at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Laats, E.T.

    1986-10-01

    The Nuclear Plant Analyzer (NPA) is a state-of-the-art safety analysis and engineering tool being used to address key nuclear power plant safety issues. Under the sponsorship of the US Nuclear Regulatory Commission (NRC), the NPA has been developed to integrate the NRC's computerized reactor behavior simulation codes such as RELAP5, TRAC-BWR and TRAC-PWR, with well-developed computer color graphics programs and large repositories of reactor design and experimental data. An important feature of the NPA is the capability to allow an analyst to redirect a RELAP5 or TRAC calculation as it progresses through its simulated scenario. The analyst can have the same power plant control capabilities as the operator of an actual plant. The NPA resides on the dual Control Data Corporation Cyber 176 mainframe computers at the Idaho National Engineering Laboratory and Cray-1S computers at the Los Alamos National Laboratory (LANL) and Kirtland Air Force Weapons Laboratory (KAFWL)

  20. Development of the nuclear plant analyzer for Korean standard Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Shin Hwan; Kim, Hyeong Heon; Song, In Ho; Hong, Eon Yeong; Oh, Yeong Taek [Korea Power Engineering Company Inc., Yongin (Korea, Republic of)

    2000-12-15

    The purpose of this study is to develop an NPA for the Ulchin Nuclear Power Plant Unit 3 and 4, the first KSNP type plant. In this study, the process model simulating the overall plant systems, GUI and simulation executive which provide the functions of an engineering simulator were developed, and the NPA was completed by integrating them. The contents and the scope of this study are as follows : main feedwater system, auxiliary feedwater system, Chemical and Volume Control System(CVCS), Safety Injection System(SIS), Shutdown Cooling System(SCS), electric power supply system, Core Protection Calculator(CPC), various plant control system, development of the graphics screens for each system, real-time simulation, simulation control for the enhancement of functional capabilities, user friendly GUI, collection of the design and operating data, establishment of the NPA database, integration of the GUI and simulation control program with process model, collection of the data for the verification and validation of the developed NPA, collection of the plant test data, collection and review of the results of other computer codes, verification of the simulation accuracy by comparing the NPA results with the actual plant data, validation of the simulation capability of the NPA, comparison against available data from other analysis suing different computer codes.

  1. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Kim, Yun Jae; Choi, Jae Boong [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2002-03-15

    This project focuses on developing reliable life evaluation technology for nuclear power plant components, and is divided into two parts, development of a life evaluation system for nuclear pressure vessels and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered in this project: defect assessment method for steam generator tubes, development of fatigue monitoring system, assessment of corroded pipes, domestic round robin analysis for constructing P-T limit curve for RPV, development of probabilistic integrity assessment technique, effect of aging on strength of dissimilar welds, applicability of LBB to cast stainless steel, and development of probabilistic piping fracture mechanics.

  2. Development of bus duct inspection robot at nuclear power plant

    International Nuclear Information System (INIS)

    Hamada, Mamoru; Hoshi, Teruaki; Komura, Yoshinari

    2017-01-01

    Under the present situation, nuclear power plant has some places which are inspected with difficulty or not inspected due to narrowness or physical restriction, when carrying out periodical inspection. The subject of our research and development is to improve the accuracy of inspection and also to save labor (liberation from distress work of the worker) by applying a robot technology to the periodical inspection of the nuclear power plant. As a specific example, we report that developed robot can inspect inside the narrow space of Isolated Phase Bus ducts, which connect between a turbine generator and the main transformer. (author)

  3. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin [Sungkyunkwan Univ., Seoul (Korea, Republic of); Kwon, J. D. [Yeungnam Univ., Gyeongsan (Korea, Republic of); Kang, K. J. [Chonnam National Univ., Gwangju (Korea, Republic of)] (and others)

    2001-03-15

    This research focuses on development of reliable life evaluation technology for nuclear power plant (NPP) components, and is divided into two parts, development of life evaluation systems for pressurized components and evaluation of applicability of emerging technology to operating plants. For the development of life evaluation system for nuclear pressure vessels, the following seven topics are covered: development of expert systems for integrity assessment of pressurized components, development of integrity evaluation systems of steam generator tubes, prediction of failure probability for NPP components based on probabilistic fracture mechanics, development of fatigue damage evaluation technique for plant life extension, domestic round robin analysis for pressurized thermal shock of reactor vessels, domestic round robin analysis of constructing P--T limit curves for reactor vessels, and development of data base for integrity assessment. For evaluation of applicability of emerging technology to operating plants, on the other hand, the following eight topics are covered: applicability of the Leak-Before-Break analysis to Cast S/S piping, collection of aged material tensile and toughness data for aged Cast S/S piping, finite element analyses for load carrying capacity of corroded pipes, development of Risk-based ISI methodology for nuclear piping, collection of toughness data for integrity assessment of bi-metallic joints, applicability of the Master curve concept to reactor vessel integrity assessment, measurement of dynamic fracture toughness, and provision of information related to regulation and plant life extension issues.

  4. Development of nuclear power plant diagnosis technique using neural networks

    International Nuclear Information System (INIS)

    Horiguchi, Masahiro; Fukawa, Naohiro; Nishimura, Kazuo

    1991-01-01

    A nuclear power plant diagnosis technique has been developed, called transient phenomena analysis, which employs neural network. The neural networks identify malfunctioning equipment by recognizing the pattern of main plant parameters, making it possible to locate the cause of an abnormality when a plant is in a transient state. In a case where some piece of equipment shows abnormal behavior, many plant parameters either directly or indirectly related to that equipment change simultaneously. When an abrupt change in a plant parameter is detected, changes in the 49 main plant parameters are classified into three types and a characteristic change pattern consisting of 49 data is defined. The neural networks then judge the cause of the abnormality from this pattern. This neural-network-based technique can recognize 100 patterns that are characterized by the causes of plant abnormality. (author)

  5. Nuclear Plant Analyzer development at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Laats, E.T.; Beelman, R.J.; Charlton, T.R.; Hampton, N.L.; Burtt, J.D.

    1985-01-01

    The Nuclear Plant Analyzer (NPA) is a state-of-the-art safety analysis and engineering tool being used to address key nuclear power plant safety issues. The NPA has been developed to integrate the NRC's computerized reactor behavior simulation codes such as RELAP5, TRAC-BWR, and TRAC-PWR, with well-developed computer graphics programs and large repositories of reactor design and experimental data. An important feature of the NAP is the capability to allow an analyst to redirect a RELAP5 or TRAC calculation as it progresses through its simulated scenario. The analyst can have the same power plant control capabilities as the operator of an actual plant. The NPA resides on the dual CDS Cyber-176 mainframe computers at the INEL and is being converted to operate on a Cray-1S computer at the LANL. The subject of this paper is the program conducted at the INEL

  6. Research and development of advanced robots for nuclear power plants

    International Nuclear Information System (INIS)

    Tsukune, Hideo; Hirukawa, Hirohisa; Kitagaki, Kosei; Liu, Yunhui; Onda, Hiromu; Nakamura, Akira

    1994-01-01

    Social and economic demands have been pressing for automation of inspection tasks, maintenance and repair jobs of nuclear power plants, which are carried out by human workers under circumstances with high radiation level. Since the plants are not always designed for introduction of automatic machinery, sophisticated robots shall play a crucial role to free workers from hostile environments. We have been studying intelligent robot systems and regarded nuclear industries as one of the important application fields where we can validate the feasibility of the methods and systems we have developed. In this paper we firstly discuss on the tasks required in nuclear power plants. Secondly we introduce current status of R and D on special purpose robots, versatile robots and intelligent robots for automatizing the tasks. Then we focus our discussions on three major functions in realizing robotized assembly tasks under such unstructured environments as in nuclear power plants; planning, vision and manipulation. Finally we depict an image of a prototype robot system for nuclear power plants based on the advanced functions. (author) 64 refs

  7. New developments in seismic surveillance of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Anesa, Franco; Vavassori, Maurizio [Instrumentation and Systems Department, ISMES, Bergamo (Italy); Casirati, Mario [Dynamic Department, ISMES, Bergamo (Italy)

    1988-07-01

    In the present article, new trends and development of ISMES techniques and methods relating to the seismic surveillance of nuclear power plants, are described. The experience gained in this field is described, and particular attention is paid to the new trend in surveys and instrumentation design criteria (from both the hardware and software point of view). (author)

  8. New developments in seismic surveillance of nuclear power plants

    International Nuclear Information System (INIS)

    Anesa, Franco; Vavassori, Maurizio; Casirati, Mario

    1988-01-01

    In the present article, new trends and development of ISMES techniques and methods relating to the seismic surveillance of nuclear power plants, are described. The experience gained in this field is described, and particular attention is paid to the new trend in surveys and instrumentation design criteria (from both the hardware and software point of view). (author)

  9. Development on database for foreign nuclear power plants

    International Nuclear Information System (INIS)

    Okuda, Yasunori; Yanagi, Chihiro

    1999-01-01

    The Nuclear Information Project in Institute of Nuclear Technology, Institute of Nuclear Safety Systems, Inc. (INSS) has been carrying out two activities related to technical information about nuclear power plants. The first is collection and analysis of accidents and incidents (troubles) of nuclear power plants in U.S.A. and West Europe and making draft of action proposals. The second is collection of main laws, government ordinances, regulatory guides, standard and domestic and international technical news connected with nuclear power plants. This report describes these two data bases about nuclear power plants details. (author)

  10. Toshiba's developments on construction techniques of nuclear power plants

    International Nuclear Information System (INIS)

    Hayashi, Y.; Itoh, N.

    1987-01-01

    Reliable and economic energy supplies are fundamental requirements of energy policies in Japan. To accomplish these needs, nuclear power plants are being increased in Japan. In recent years, construction cost increases and schedule extensions have affected the capital cost of nuclear energy, compared with fossil power plants, due to lower costs of oil and coal. On the other hand, several severe regulations have been applied to nuclear power plant designs. High-quality and cooperative engineering and harmonized design of equipment and parts are strongly required. Therefore, reduced construction costs and scheduling, as well as higher quality and reliability, are the most important items for nuclear industry. Toshiba has developed new construction techniques, as well as design and engineering tools for control and management, that demonstrate the positive results achieved in the shorter construction period of 1100-MW(electric) nuclear power plants. The normal construction period so far is 64 months, whereas the current construction period is 52 months. (New construction techniques are partially applied). In future years, the construction period will be lowered to 48 months. (New construction techniques are fully applied). A construction period is defined as time from the start of rock inspection to the start of commercial operation

  11. Development of instructors for nuclear power plant personnel training

    International Nuclear Information System (INIS)

    2004-06-01

    In 1996 the IAEA published Technical Reports Series No. 380, Nuclear Power Plant Personnel Training and its Evaluation, A Guidebook, which provides guidance with respect to development, implementation and evaluation of training programmes. The IAEA Technical Working Group on Training and Qualification of Nuclear Power Plant Personnel recommended that an additional publication be prepared to provide further details concerning the development of instructors for NPP personnel training. The quality of nuclear power plant personnel training is strongly dependent on the availability of competent instructors. Instructors must have a comprehensive practical as well as theoretical understanding of all aspects of the subjects being taught and the relationship of the subject to nuclear plant operation. Instructors should have the appropriate knowledge, skills and attitudes (KSAs) in their assigned areas of responsibility. They should thoroughly understand all aspects of the contents of the training programmes and the relationship between these contents and overall plant operation. This means that they should be technically competent and show credibility with the trainees and other plant personnel. In addition, the instructors should be familiar with the basics of adult learning and a systematic approach to training, and should have adequate instructional and assessment skills. This TECDOC provides practical guidance on various aspects of instructor selection, development and deployment, by quoting actual examples from different countries. It highlights the importance of having an appropriate training policy, especially considering the various organisational arrangements that exist in different utilities/countries. This should result in: plant performance improvement, improved human performance, meeting goals and objectives of the business (quality, safety, productivity), and improving training programs. This publication is available in two formats - as a conventional printed

  12. Development of support system for nuclear power plant piping

    International Nuclear Information System (INIS)

    Horino, Satoshi

    1987-01-01

    Ishikawajima-Harima Heavy Industries Co., Ltd. has advanced the development of Integrated Nuclear Plant Piping System (INUPPS) for nuclear power plants since 1980, and continued its improvement up to now. This time as its component, a piping support system (PISUP) has been developed. The piping support system deals with the structures such as piping supports and the stands for maintenance and inspection, and as for standard supporting structures, it builds up automatically the structures including the selection of optimum members by utilizing the standard patterns in cooperation with the piping design system including piping stress analysis. As for the supporting structures deviating from the standard, by amending a part of the standard patterns in dialogue from, structures can be built up. By using the data produced in this way, this system draws up consistently a design book, production management data and so on. From the viewpoint of safety, particular consideration is given to the aseismatic capability of nuclear power plants, and piping is fundamentally designed regidly to avoid resonance. It is necessary to make piping supports so as to have sufficient strength and rigidity. The features of the design of piping supports for nuclear power plant, the basic concept of piping support system, the constitution of the software and hardware, the standard patterns and the structural patterns of piping support system and so on are described. (Kako, I.)

  13. Developing new methodology for nuclear power plants vulnerability assessment

    International Nuclear Information System (INIS)

    Kostadinov, Venceslav

    2011-01-01

    new methodology and solution methods for vulnerability assessment can help the overall national energy sector to identify and understand the terrorist threats to and vulnerabilities of its critical infrastructure. Moreover, adopted methodology could help national regulators and agencies to develop and implement a vulnerability awareness and education programs for their critical assets to enhance the security and a safe operation of the entire energy infrastructure. New methods can also assist nuclear power plants to develop, validate, and disseminate assessment and surveys of new efficient countermeasures. Consequently, concise description of developed new quantitative method and adapted new methodology for nuclear regulatory vulnerability assessment of nuclear power plants are presented.

  14. Development and application of the plant condition monitoring system for nuclear power plants

    International Nuclear Information System (INIS)

    Ono, S.

    2014-01-01

    To achieve the stable operation of nuclear power plants, we developed the plant condition monitoring system based on the heat and mass balance calculation. In this system, it is a significant feature to adopt the sophisticated heat balance model based on the actual plant data to find the symptoms of anomalies in the turbine system from heat balance changes. (author)

  15. Infrastructure development assistance modeling for nuclear power plant

    International Nuclear Information System (INIS)

    Park, J. H.; Hwang, K.; Park, K. M.; Kim, S. W.; Lee, S. M.

    2012-01-01

    The purpose of this paper is to develop a model, a general frame to be utilized in assisting newcomer countries to start a nuclear power program. A nuclear power plant project involves technical complexity and high level of investment with long duration. Considering newcomers are mostly developing countries that lack the national infrastructure, key infrastructure issues may constitute the principal constraints to the development of a nuclear power program. In this regard, it is important to provide guidance and support to set up an appropriate infrastructure when we help them with the first launch of nuclear power plant project. To date, as a sole nuclear power generation company, KHNP has been invited many times to mentor or assist newcomer countries for their successful start of a nuclear power program since Republic of Korea is an exemplary case of a developing country which began nuclear power program from scratch and became a major world nuclear energy country in a short period of time. Through hosting events organized to aid newcomer countries' initiation of nuclear power projects, difficulties have been recognized. Each event had different contents according to circumstances because they were held as an unstructured and one-off thing. By developing a general model, we can give more adequate and effective aid in an efficient way. In this paper, we created a model to identify necessary infrastructures at the right stage, which was mainly based on a case of Korea. Taking into account the assistance we received from foreign companies and our own efforts for technological self-reliance, we have developed a general time table and specified activities required to do at each stage. From a donor's perspective, we explored various ways to help nuclear infrastructure development including technical support programs, training courses, and participating in IAEA technical cooperation programs on a regular basis. If we further develop the model, the next task would be to

  16. Infrastructure development assistance modeling for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, J. H.; Hwang, K.; Park, K. M.; Kim, S. W.; Lee, S. M. [Korea Hydro and Nuclear Power Co., LTD, 23, 106 gil, Yeongdong-daero, Gangnam-gu, 153-791 (Korea, Republic of)

    2012-07-01

    The purpose of this paper is to develop a model, a general frame to be utilized in assisting newcomer countries to start a nuclear power program. A nuclear power plant project involves technical complexity and high level of investment with long duration. Considering newcomers are mostly developing countries that lack the national infrastructure, key infrastructure issues may constitute the principal constraints to the development of a nuclear power program. In this regard, it is important to provide guidance and support to set up an appropriate infrastructure when we help them with the first launch of nuclear power plant project. To date, as a sole nuclear power generation company, KHNP has been invited many times to mentor or assist newcomer countries for their successful start of a nuclear power program since Republic of Korea is an exemplary case of a developing country which began nuclear power program from scratch and became a major world nuclear energy country in a short period of time. Through hosting events organized to aid newcomer countries' initiation of nuclear power projects, difficulties have been recognized. Each event had different contents according to circumstances because they were held as an unstructured and one-off thing. By developing a general model, we can give more adequate and effective aid in an efficient way. In this paper, we created a model to identify necessary infrastructures at the right stage, which was mainly based on a case of Korea. Taking into account the assistance we received from foreign companies and our own efforts for technological self-reliance, we have developed a general time table and specified activities required to do at each stage. From a donor's perspective, we explored various ways to help nuclear infrastructure development including technical support programs, training courses, and participating in IAEA technical cooperation programs on a regular basis. If we further develop the model, the next task

  17. Development of automatic inspection robot for nuclear power plants

    International Nuclear Information System (INIS)

    Yamada, K.; Suzuki, K.; Saitoh, K.; Sakaki, T.; Ohe, Y.; Mizutani, T.; Segawa, M.; Kubo, K.

    1987-01-01

    This robot system has been developed for automatic inspection of nuclear power plants. The system configuration is composed of vehicle that runs on monorail, the sensors on the vehicle, an image processer that processes the image information from the sensors, a computer that creates the inspection planning of the robot and an operation panel. This system has two main features, the first is the robot control system. The vehicle and the sensors are controlled by the output data calculated in the computer with the three dimensional plant data. The malfunction is recognized by the combination of the results of image processing, information from the microphone and infrared camera. Tests for a prototype automatic inspection robot system have been performed in the simulated main steam piping room of a nuclear power plant

  18. Objectives for the development of advanced nuclear plants

    International Nuclear Information System (INIS)

    1993-01-01

    The scope of this report is to reiterate the broad objectives for the development of advanced nuclear plants, to set forth some related subobjectives and to propose some universal goals for the development programmes. The majority of these can be categorized under the headings of enhancing safety, improving reliability and gaining better economics. These categories are used in the report followed by additional categories considered to be important within the global framework intended. Additional broad objectives appear unlikely but more subobjectives may become evident as time progresses and the need arises to express them in the intended more global framework. The goals also may change. The scope is therefore a set of objectives for development of advanced nuclear plants. The objectives are believed to be universally acceptable; they have been reviewed on that basis. 13 refs

  19. Development of real options model for nuclear power plants

    International Nuclear Information System (INIS)

    Ono, Kenji

    2004-01-01

    As the Japanese electricity market is deregulated, it becomes more important for electric utilities to recognize their financial risks and to adopt strategic and scientific decision making methodology. We have developed two models for valuation of Japanese nuclear power plants to support utilities' decision making. One is a net present value (NPV) model using discounted cash flow analysis method. Another is a real options model. This model is based on strict financial technology theory and can calculate value of early retirement, life extension and new unit addition options of nuclear units under electricity price uncertainty. This can also derive an optimal period for retirement, life extension and new unit addition. (author)

  20. Development of Inventory Optimization System for Operation Nuclear Plants

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Se-Jin; Park, Jong-Hyuk; Yoo, Sung-Soo; Lee, Sang-Guk [Korea Electric Power Research Institutes, Taejon (Korea, Republic of)

    2006-07-01

    Inventory control of spare parts plays an increasingly important role in operation management. This is why inventory management systems such as manufacturing resources planning(MRP) and enterprise resource planning(ERP) have been added. However, most of these contributions have similar theoretical background. This means the concepts and techniques are mainly based on mathematical assumptions and modeling inventory of spare parts situations. Nuclear utilities in Korea have several problems to manage the optimum level of spare parts though they used MRP System. Because most of items have long lead time and they are imported from United States, Canada, France and so on. We developed the inventory optimization system for Operation Nuclear Plants to resolve these problems. In this paper, we report a data flow process, data load and inventory calculation process. The main contribution of this paper is development of inventory optimization system which can be used in domestic power plants.

  1. Development of Inventory Optimization System for Operation Nuclear Plants

    International Nuclear Information System (INIS)

    Jang, Se-Jin; Park, Jong-Hyuk; Yoo, Sung-Soo; Lee, Sang-Guk

    2006-01-01

    Inventory control of spare parts plays an increasingly important role in operation management. This is why inventory management systems such as manufacturing resources planning(MRP) and enterprise resource planning(ERP) have been added. However, most of these contributions have similar theoretical background. This means the concepts and techniques are mainly based on mathematical assumptions and modeling inventory of spare parts situations. Nuclear utilities in Korea have several problems to manage the optimum level of spare parts though they used MRP System. Because most of items have long lead time and they are imported from United States, Canada, France and so on. We developed the inventory optimization system for Operation Nuclear Plants to resolve these problems. In this paper, we report a data flow process, data load and inventory calculation process. The main contribution of this paper is development of inventory optimization system which can be used in domestic power plants

  2. Development of construction technology for nuclear power plants

    International Nuclear Information System (INIS)

    Schnellenbach, G.

    1975-01-01

    Stringent safety requirements for nuclear power plants create new technological problems in their construction. The development is influenced by the mode of operation of the respective reactors which leads to different construction styles of the reactor buildings. Accomodation of extraordinary load cases such as earthquake, airplane crash, and blast due to chemical explosions, requires additional treatment. Significant new problems arise for the prestressed concrete reactor pressure vessels and concrete containments, prestressed or reinforced. (orig.) [de

  3. Development of halogen-free cables for nuclear power plants

    International Nuclear Information System (INIS)

    Yamamoto, Mitsuo; Ito, Kazumi; Yaji, Takeo; Yoshida, Shin; Sakurai, Takako; Matsushita, Shigetoshi.

    1990-01-01

    On the occasion where serious fire accidents were experienced in the past, the need for making flame-retardant wire and cable incombustible took place and has since been generalizing. Various sorts of flame-retardant cables have already been developed and been actually used. From the viewpoint of avoiding the interference with the evacuation and fire-fighting activity in case of fire or the secondary accidents such as corrosion of the distributing panel, etc., the demand for non-halogen flame-retardant cable has rapidly been increasing in recent years in some fields of general industries, because this specific cable would generate the least amount of toxic smoke or corrosive gas even when it should burn. Similar demand has been increasing also for the cable used for nuclear power plants. In this field, earnest desire has been made for the development of non-halogen flame-retardant cable having specific environmental resistance specially required at nuclear power plants in addition to the properties and capacities required in general industries. The authors have continued examinations on the anti-environmental properties of the materials for cable such as long heat resistance, radiation resistance, steam resistance and succeeded in completing various sorts of non-halogen flame-retardant cable for nuclear power plants. In this report, we will introduce various features of the cable we have developed this time as well as the long-term reliability of non-halogen flame-retardant materials. (author)

  4. New nuclear plant development - balancing localization with competitiveness

    Energy Technology Data Exchange (ETDEWEB)

    Caplan, M.; Thompson, T.S. [MZ Consulting Inc., ON (Canada)

    2014-07-01

    Nuclear power plants are large infrastructure projects that require government support and approval. This paper will highlight and contrast the larger, mostly government-desired, nuclear program localization objectives with the more utility-specific requirements for successful project implementation. Governments are concerned about sustainable industrial development, particularly manufacturing, and job creation while utilities are focused on delivering reliable electricity to consumers at the lowest cost. Numerous countries emphasize local content as a key requirement when procuring a station. For countries like China and Korea that have large programs, their strategy has been to localize to the point of having their own indigenous design. However, developing a workable localization strategy that truly benefits the local economy for others including existing nuclear markets like Canada, the UK, South Africa and Brazil as well as in newly developing markets such as Vietnam and Malaysia is more challenging. These countries may not look to indigenize a new design, rather they would localize elements of the nuclear program that best fit their strengths. The paper will discuss the issues related to developing successful localization and industrialization strategies in a changing nuclear world. (author)

  5. New nuclear plant development - balancing localization with competitiveness

    International Nuclear Information System (INIS)

    Caplan, M.; Thompson, T.S.

    2014-01-01

    Nuclear power plants are large infrastructure projects that require government support and approval. This paper will highlight and contrast the larger, mostly government-desired, nuclear program localization objectives with the more utility-specific requirements for successful project implementation. Governments are concerned about sustainable industrial development, particularly manufacturing, and job creation while utilities are focused on delivering reliable electricity to consumers at the lowest cost. Numerous countries emphasize local content as a key requirement when procuring a station. For countries like China and Korea that have large programs, their strategy has been to localize to the point of having their own indigenous design. However, developing a workable localization strategy that truly benefits the local economy for others including existing nuclear markets like Canada, the UK, South Africa and Brazil as well as in newly developing markets such as Vietnam and Malaysia is more challenging. These countries may not look to indigenize a new design, rather they would localize elements of the nuclear program that best fit their strengths. The paper will discuss the issues related to developing successful localization and industrialization strategies in a changing nuclear world. (author)

  6. Development of a SPV management program for nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Eun-Chan; Na, Jang-Hwan; Lee, Doo-Young; Oh, Seong-Jong; Jerng, Dong-Wook

    2009-01-01

    The Single Point Vulnerability (SPV) is a characteristic of a component whose failure results in plant transients. KHNP (Korea Hydro and Nuclear Power Co.) has evaluated major systems with critical components and developed a SPV management program to reduce the trip frequency and to raise plant's availability. This study includes a comprehensive methodology for SPV evaluation and its results. This methodology consists of the qualitative evaluation focused on the SPV component list preparation and the quantitative evaluation through FMEA (Failure Mode Effect Analyses) and FTA (Fault Tree Analyses) of critical systems. The qualitative evaluation reduced differences of the SPV lists between the identically designed plants and established strategies for the improvement of the SPV component reliability. The quantitative evaluation identified additional SPV components and developed the fault tree model for a Trip Monitor which showed logic relationships of channel components in the trip-related systems. (author)

  7. Development of an intelligent annunciation system for nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Chang-Gi; Che, Myoung-Eun

    1997-01-01

    Yonggwang Nuclear Units 1 and 2 have developed an intelligent annunciation system to replace the existing obsolete system and to enhance operator support. The new annunciation system, which is currently operating at both units, uses the distributed control technology to enhance reliability and to provide versatile function to operations and maintenance personnel. The hardware and software configuration is based on redundancy so that a component failure would not initiate system malfunction. The data base of the new system provides, through a touch screen, an automatic alarm response procedure for selected alarms, which increases availability of information for plant operation. Other KEPCO nuclear units and the fossil plants are considering installing the new system. (author). 6 figs, 2 tabs

  8. Development of an intelligent annunciation system for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang-Gi; Che, Myoung-Eun [Instrumentation and Control, Yonggwang Nuclear Units 1 and 2, Korea Electric Power Corp. (Korea, Republic of)

    1997-09-01

    Yonggwang Nuclear Units 1 and 2 have developed an intelligent annunciation system to replace the existing obsolete system and to enhance operator support. The new annunciation system, which is currently operating at both units, uses the distributed control technology to enhance reliability and to provide versatile function to operations and maintenance personnel. The hardware and software configuration is based on redundancy so that a component failure would not initiate system malfunction. The data base of the new system provides, through a touch screen, an automatic alarm response procedure for selected alarms, which increases availability of information for plant operation. Other KEPCO nuclear units and the fossil plants are considering installing the new system. (author). 6 figs, 2 tabs.

  9. Development of advanced RFID application system for nuclear power plant

    International Nuclear Information System (INIS)

    Onda, Kimiharu; Wakabayashi, Eisuke; Arai, Ryota; Shigemi, Ryosuke; Muro, Keiro; Yuda, Shinya

    2008-01-01

    In late years there comes to be close request for traceability of the information such as production control, construction and maintenance record and work history of nuclear power plants. On the other hand, the Radio-frequency identification (RFID) technology that can specify a product and personnel by an electric wave has raised the functionality and versatility as the base technology that can support ubiquitous information society around the mass production industry. In such a background, this article described the developed system, which applied the RFID to nuclear power plants in the areas of production control of the piping manufacture, the construction management and condition monitoring for maintenance works in order to improve their quality and reliability. (T. Tanaka)

  10. Impacts of nuclear power plant developments on community service capacity

    International Nuclear Information System (INIS)

    Krannich, R.S.

    1978-03-01

    With the passage of the National Environmental Policy Act in 1969 a legislative mandate was established to assess ''socioeconomic'' as well as environmental consequences of large-scale development projects. However, the developing literature base in the area of socioeconomics has exhibited a pronounced tendency to stress social and economic pathologies associated with the so-called ''boom town'' syndrome. While boom growth and associated problems do appear to provide relevant conceptualizations of the socioeconomic impacts of energy resource development projects in geographically isolated sectors of the western United States, the argument is presented that such a perspective is generally inappropriate when assessing the consequences of nuclear power plant developments. Survey data on 21 nuclear generating facilities are analyzed in order to provide a comparative perspective on socioeconomic consequences and factors which may influence their relative importance

  11. Development trends for diagnostic systems in nuclear power plants

    International Nuclear Information System (INIS)

    Kunze, U.; Pohl, U.

    1998-01-01

    Monitoring systems used in nuclear power plants have made remarkable progress over the past four or five years. Development has followed the trends and changes in philosophy for the purpose of monitoring systems in nuclear power plants: They are no longer expected to fulfill only safety tasks, the plant personnel require information on which to base condition-oriented maintenance. A new generation of monitoring and diagnostic systems has been developed by Siemens recently. This new generation, called Series '95, is PC-based. An overview is given for the KUeS '95 loose parts diagnostic system, the SUeS '95 vibration monitoring system, the FLUeS leak detection system and the SIPLUG valve diagnostics system. The objectives behind the development of these new systems are both safety-related and economic. The new systems improve the reliability and quality of monitoring techniques and incorporate better detection and diagnostic capabilities. Progress has also been made in automation of the systems so as to reduce routine work, give higher sensitivity for the monitoring task and reduce the scope of maintenance. (author)

  12. Development of chemistry management for Onagawa Nuclear Power Plant

    International Nuclear Information System (INIS)

    Honda, Kazuaki; Sato, Junichi; Maeda, Katsuji; Nagasawa Katsumi; Hashiura, Sintaro

    2000-01-01

    Onagawa nuclear power plant developed a system for chemistry management of prevention and preservation of power plant. It is able to early detection of data change for prevention of plant. The system supports management of chemical custodian and consists of four parts such as management of water quality of plant , management of liquid waste, management of vapor waste and estimation of performance of chemical equipment. The system has three functions: management of operation, estimation of omen and examination of origin. The function of management of operation supports the routine inspection of chemical custodian by increasing efficiency of analytical and process data collection, practical use of data,, accuracy of data and rapid analysis. Estimation function of omen observes data fetched via online during 24 hr, indicates the small primary change and determines the origins. Examination function of origin supports their quick correspondences at accident and certificates the detailed origins. Histories of development of the system, business systemization, system construction, system functions are explained.The diagram of background of system development, system construction, management functions, verification of analytical data, automatic continuos monitoring diagram, screen of detection of abnormal phenomena, classified diagram of origins for change of water quality in reactor were developed. (S.Y.)

  13. Development of remote maintenance technology for nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Kawahara, Akira; Saito, Masayuki; Kawamura, Hironobu; Yamade, Atsushi; Sugiyama, Sen; Sugiyama, Sakae.

    1986-01-01

    In the plants for reprocessing spent nuclear fuel containing fission products, due to the facts that the facilities are in high radiations fields, and the surfaces of equipments are contaminated with radioactive substances, the troubles of process equipments are directly connected to the remarkable drop of the rate of operation of the facilities. Therefore, the development of various remote maintenance techniques has been carried out so far, but this time, Hitachi Ltd. got a chance to take part in the repair of spent fuel dissolving tanks in the Tokai Reprocessing Plant of Power Reactor and Nuclear Fuel Development Corp. and the development of several kinds of remote checkup equipment related to the repair work. Especially in the repair of the dissolving tanks, a radiation-withstanding checkup and repair apparatus which has high remote operability taking the conditions of radioactive environment and the restriction of the repaired objects in consideration was required, and a dissolving tank repairing robot composed of six kinds has been developed. The key points of the development were the selective use of high radiation-withstanding parts and materials, small size structure and the realization of full remote operability. The full remote maintenance apparatus of this kind is unique in the world, and applicable to wide fields. (Kako, I.)

  14. The development of robotic system for the nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Ho; Lee, Jae Kyung; Kim, Ki Ho; Jung, Seung Ho; Kim, Chang Hoi; Kim, Byung Soo; Hwang, Suk Yeoung; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Kim, Woong Ki; Park, Soon Yong

    1996-07-01

    This project focuses on the development of a heavy-duty telerobotic system (HDTS) and a light-duty mobile robotic system (LDMRS) for use in nuclear power plants. HDTS has been developed for performing tasks such as the installation and removal of nozzle dam inside of water chamber of steam generator. HDTS that is remotely controlled could eliminate or significantly reduce human exposure to hazardous nuclear environment. HDTS has four major subsystems : a 6 degree of freedom master-slave manipulator, a guiding device, a monitoring device and a remote control center. Functional connections of each subsystems has resulted in HDTS that exhibits a high level of dexterity and a broad range of capabilities. LDMRS has been developed to be used in emergency response applications such as monitoring and mapping radiation areas, handling radioactive materials and performing decontamination tasks. LDMRS equipped with four-omnidirectional planetary wheels is capable of ascending and descending stairs by employing a automatic stair climbing algorithm. A force-reflective algorithm developed enables LDMRS to be navigated flat surface with zero turning radius without collision by giving an operator a sense of force. The significance of developments is in providing both HDTS and LDMRS that can be operated from remote locations to perform tasks such as the maintenance of nozzle dam and the video surveillance of the nuclear facilities efficiently and without endangering human workers. This report describes the mechanical design, features, control system, and capabilities of both HDTS and LDMRS. (author). 59 refs., 38 tabs., 132 figs.

  15. The development of robotic system for the nuclear power plant

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Lee, Jae Kyung; Kim, Ki Ho; Jung, Seung Ho; Kim, Chang Hoi; Kim, Byung Soo; Hwang, Suk Yeoung; Seo, Yong Chil; Lee, Young Kwang; Lee, Yong Bum; Kim, Woong Ki; Park, Soon Yong.

    1996-07-01

    This project focuses on the development of a heavy-duty telerobotic system (HDTS) and a light-duty mobile robotic system (LDMRS) for use in nuclear power plants. HDTS has been developed for performing tasks such as the installation and removal of nozzle dam inside of water chamber of steam generator. HDTS that is remotely controlled could eliminate or significantly reduce human exposure to hazardous nuclear environment. HDTS has four major subsystems : a 6 degree of freedom master-slave manipulator, a guiding device, a monitoring device and a remote control center. Functional connections of each subsystems has resulted in HDTS that exhibits a high level of dexterity and a broad range of capabilities. LDMRS has been developed to be used in emergency response applications such as monitoring and mapping radiation areas, handling radioactive materials and performing decontamination tasks. LDMRS equipped with four-omnidirectional planetary wheels is capable of ascending and descending stairs by employing a automatic stair climbing algorithm. A force-reflective algorithm developed enables LDMRS to be navigated flat surface with zero turning radius without collision by giving an operator a sense of force. The significance of developments is in providing both HDTS and LDMRS that can be operated from remote locations to perform tasks such as the maintenance of nozzle dam and the video surveillance of the nuclear facilities efficiently and without endangering human workers. This report describes the mechanical design, features, control system, and capabilities of both HDTS and LDMRS. (author). 59 refs., 38 tabs., 132 figs

  16. Development of web based performance analysis program for nuclear power plant turbine cycle

    International Nuclear Information System (INIS)

    Park, Hoon; Yu, Seung Kyu; Kim, Seong Kun; Ji, Moon Hak; Choi, Kwang Hee; Hong, Seong Ryeol

    2002-01-01

    Performance improvement of turbine cycle affects economic operation of nuclear power plant. We developed performance analysis system for nuclear power plant turbine cycle. The system is based on PTC (Performance Test Code), that is estimation standard of nuclear power plant performance. The system is developed using Java Web-Start and JSP(Java Server Page)

  17. Development of RCM analysis software for Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ho; Choi, Kwang Hee; Jeong, Hyeong Jong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A software called KEPCO RCM workstation (KRCM) has been developed to optimize the maintenance strategies of Korean nuclear power plants. The program modules of the KRCM were designed in a manner that combines EPRI methodologies and KEPRI analysis technique. The KRCM is being applied to the three pilot system, chemical and volume control system, main steam system, and compressed air system of Yonggwang Units 1 and 2. In addition, the KRCM can be utilized as a tool to meet a part of the requirements of maintenance rule (MR) imposed by U.S. NRC. 3 refs., 4 figs. (Author)

  18. Development of RCM analysis software for Korean nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ho; Choi, Kwang Hee; Jeong, Hyeong Jong [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A software called KEPCO RCM workstation (KRCM) has been developed to optimize the maintenance strategies of Korean nuclear power plants. The program modules of the KRCM were designed in a manner that combines EPRI methodologies and KEPRI analysis technique. The KRCM is being applied to the three pilot system, chemical and volume control system, main steam system, and compressed air system of Yonggwang Units 1 and 2. In addition, the KRCM can be utilized as a tool to meet a part of the requirements of maintenance rule (MR) imposed by U.S. NRC. 3 refs., 4 figs. (Author)

  19. A study on the development and application of expert system for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Hee Gon; Kim, Seong Bok [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1995-12-31

    It is a final report of the research that is a study on the development and application of expert system for nuclear power plants and development of the schemes computing environments and user interfaces for the expert system, which is a systematic and efficient development of expert system for nuclear power plants in the future. This report is consisted of -Development trends of expert system for nuclear power plants. -Classification of expert system applications for nuclear power plants. -Systematic and efficient developments schemes of expert system for nuclear power plants, and -Suitable computing environments and user interfaces for the expert systems. (author). 113 refs., 85 figs.

  20. A study on the development and application of expert system for nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Hee Gon; Kim, Seong Bok [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center

    1996-12-31

    It is a final report of the research that is a study on the development and application of expert system for nuclear power plants and development of the schemes computing environments and user interfaces for the expert system, which is a systematic and efficient development of expert system for nuclear power plants in the future. This report is consisted of -Development trends of expert system for nuclear power plants. -Classification of expert system applications for nuclear power plants. -Systematic and efficient developments schemes of expert system for nuclear power plants, and -Suitable computing environments and user interfaces for the expert systems. (author). 113 refs., 85 figs.

  1. Development of recycling techniques for nuclear power plant decommissioning waste

    International Nuclear Information System (INIS)

    Ishikura, Takeshi; Oguri, Daiichiro; Abe, Seiji; Ohnishi, Kazuhiko

    2003-01-01

    Recycling of concrete and metal waste will provide solution to reduce waste volume, contributing to save the natural resources and to protect the environment. Nuclear Power Engineering Corporation has developed techniques of concrete and metal recycling for decommissioning waste of commercial nuclear power plants. A process of radioactive concrete usage for mortar solidification was seen to reduce concrete waste volume by 2/3. A concrete reclamation process for high quality aggregate was confirmed that the reclaimed aggregate concrete is equivalent to ordinary concrete. Its byproduct powder was seen to be utilized various usage. A process of waste metal casting to use radioactive metal as filler could substantially decrease the waste metal volume when thinner containers are applied. A pyro-metallurgical separation process was seen to decrease cobalt concentration by 1/100. Some of these techniques are finished of demonstration tests for future decommissioning activity. (author)

  2. Development of Nuclear Plant Specific Analysis Simulators with ATLAS

    International Nuclear Information System (INIS)

    Jakubowski, Z.; Draeger, P.; Horche, W.; Pointner, W.

    2006-01-01

    The simulation software ATLAS, based on the best-estimate code ATHLET, has been developed by the GRS for a range of applications in the field of nuclear plant safety analysis. Through application of versatile simulation tools and graphical interfaces the user should be able to analyse with ATLAS all essential accident scenarios. Detailed analysis simulators for several German and Russian NPPs are being constructed on the basis of ATLAS. An overview of the ATLAS is presented in the paper, describing its configuration, functions performed by main components and relationships among them. A significant part of any power plant simulator are the balance-of-plant (BOP) models, not only because all the plant transients and non-LOCA accidents can be initiated by operation of BOP systems, but also because the response of the plant to transients or accidents is strongly influenced by the automatic operation of BOP systems. Modelling aspects of BOP systems are shown in detail, also the interface between the process model and BOP systems. Special emphasis has been put on the BOP model builder based on the methodology developed in the GRS. The BOP modeler called GCSM-Generator is an object oriented tool which runs on the online expert system G2. It is equipped with utilities to edit the BOP models, to verification them and to generate a GCSM code, specific for the ATLAS. The communication system of ATLAS presents graphically the results of the simulation and allows interactively influencing the execution of the simulation process (malfunctions, manual control). Displays for communications with simulated processes and presentation of calculations results are also presented. In the framework of the verification of simulation models different tools are used e.g. the PC-codes MATHCAD for the calculation and documentation, ATLET-Input-Graphic for control of geometry data and the expert system G2 for development of BOP-Models. The validation procedure and selected analyses results

  3. Development of supplier evaluation model applying in nuclear power plants

    International Nuclear Information System (INIS)

    Wang Yonggang; Fang Chunfa

    2006-01-01

    It is essential for the safe and stable operations of Nuclear Power Plants that various resources in the supply chain are effectively managed. Supplier is a significant resource of nuclear entities serving as an extension of the operation process. Scientific and radiation evaluation of the performance of suppliers is of vital importance to an effective and high quality supply chain. This paper establishes an advance and practical supplier evaluation system that is applicable for the operational nuclear power plants, based on the analysis of the current operation status of Daya Bay Nuclear Power Station against its targeted objectives, the acquisition of relevant practices home and abroad and the benchmarking with advanced peers, in order to enhance the core competence of nuclear power plant. (authors)

  4. Developing and assessing accident management plans for nuclear power plants

    International Nuclear Information System (INIS)

    Hanson, D.J.; Johnson, S.P.; Blackman, H.S.; Stewart, M.A.

    1992-07-01

    This document is the second of a two-volume NUREG/CR that discusses development of accident management plans for nuclear power plants. The first volume (a) describes a four-phase approach for developing criteria that could be used for assessing the adequacy of accident management plans, (b) identifies the general attributes of accident management plans (Phase 1), (c) presents a prototype process for developing and implementing severe accident management plans (Phase 2), and (d) presents criteria that can be used to assess the adequacy of accident management plans. This volume (a) describes results from an evaluation of the capabilities of the prototype process to produce an accident management plan (Phase 3) and (b), based on these results and preliminary criteria included in NUREG/CR-5543, presents modifications to the criteria where appropriate

  5. Angra 1 nuclear power plant full scope simulator development project

    Energy Technology Data Exchange (ETDEWEB)

    Selvatici, Edmundo; Castanheira, Luiz Carlos C.; Silva Junior, Nilo Garcia da, E-mail: edsel@eletronuclear.gov.br, E-mail: lccast@eletronuclear.gov.br, E-mail: nilogar@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (SCO/ELETRONUCLEAR), Angra dos Reis, RJ (Brazil). Superintendencia de Coordenacao da Operacao; Zazo, Francisco Javier Lopez; Ruiz, Jose Antonio, E-mail: jlopez@tecnatom.es, E-mail: jaruiz@tecnatom.es [Tecnatom S.A., San Sebastian de los Reyes, Madrid (Spain)

    2015-07-01

    Specific Full Scope Simulators are an essential tool for training NPP control room operators, in the formation phase as well as for maintaining their qualifications. In the last years availability of a Plant specific simulator has also become a Regulator requirement for Nuclear Power Plant operation. By providing real-time practical training for the operators, the use of a simulator allows improving the operator's performance, reducing the number of unplanned shutdowns and more effective response to abnormal and emergency operating conditions. It can also be used, among other uses, to validate procedures, test proposed plant modifications, perform engineering studies and to provide operation training for the technical support staff of the plant. The NPP site, in Angra dos Reis-RJ, Brazil, comprises the two units in operation, Unit 1, 640 MWe, Westinghouse PWR and Unit 2, 1350 MWe, KWU/Areva PWR and one unit in construction, Unit 3, 1405 MWe, KWU/Areva PWR, of the same design of Angra 2. Angra 2 has had its full scope simulator from the beginning, however this was not the case of Angra 1, that had to train its operators abroad, due to lack of a specific simulator. Eletronuclear participated in all the phases of the project, from data supply to commissioning and validation. The Angra 1 full scope simulator encompasses more than 80 systems of the plant including the Primary system, reactor core and associated auxiliary systems, the secondary system and turbo generator as well as all the Plant operational and safety I and C. The Angra 1 Main Control Room panels were reproduced in the simulator control room as well as the remote shutdown panels that are outside the control room. This paper describes the project for development of the Angra 1 NPP Full Scope Simulator, supplied by Tecnatom S.A., in the period of Feb.2012 to Feb.2015. (author)

  6. Angra 1 nuclear power plant full scope simulator development project

    International Nuclear Information System (INIS)

    Selvatici, Edmundo; Castanheira, Luiz Carlos C.; Silva Junior, Nilo Garcia da

    2015-01-01

    Specific Full Scope Simulators are an essential tool for training NPP control room operators, in the formation phase as well as for maintaining their qualifications. In the last years availability of a Plant specific simulator has also become a Regulator requirement for Nuclear Power Plant operation. By providing real-time practical training for the operators, the use of a simulator allows improving the operator's performance, reducing the number of unplanned shutdowns and more effective response to abnormal and emergency operating conditions. It can also be used, among other uses, to validate procedures, test proposed plant modifications, perform engineering studies and to provide operation training for the technical support staff of the plant. The NPP site, in Angra dos Reis-RJ, Brazil, comprises the two units in operation, Unit 1, 640 MWe, Westinghouse PWR and Unit 2, 1350 MWe, KWU/Areva PWR and one unit in construction, Unit 3, 1405 MWe, KWU/Areva PWR, of the same design of Angra 2. Angra 2 has had its full scope simulator from the beginning, however this was not the case of Angra 1, that had to train its operators abroad, due to lack of a specific simulator. Eletronuclear participated in all the phases of the project, from data supply to commissioning and validation. The Angra 1 full scope simulator encompasses more than 80 systems of the plant including the Primary system, reactor core and associated auxiliary systems, the secondary system and turbo generator as well as all the Plant operational and safety I and C. The Angra 1 Main Control Room panels were reproduced in the simulator control room as well as the remote shutdown panels that are outside the control room. This paper describes the project for development of the Angra 1 NPP Full Scope Simulator, supplied by Tecnatom S.A., in the period of Feb.2012 to Feb.2015. (author)

  7. Next Generation Nuclear Plant Materials Research and Development Program Plan

    International Nuclear Information System (INIS)

    G.O. Hayner; R.L. Bratton; R.N. Wright

    2005-01-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R and D) Program is responsible for performing R and D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R and D Program includes the following elements: (1) Developing a specific approach, program plan and other project management

  8. Next Generation Nuclear Plant Materials Research and Development Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    G.O. Hayner; R.L. Bratton; R.N. Wright

    2005-09-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Project is envisioned to demonstrate the following: (1) A full-scale prototype VHTR by about 2021; (2) High-temperature Brayton Cycle electric power production at full scale with a focus on economic performance; (3) Nuclear-assisted production of hydrogen (with about 10% of the heat) with a focus on economic performance; and (4) By test, the exceptional safety capabilities of the advanced gas-cooled reactors. Further, the NGNP program will: (1) Obtain a Nuclear Regulatory Commission (NRC) License to construct and operate the NGNP, this process will provide a basis for future performance based, risk-informed licensing; and (2) Support the development, testing, and prototyping of hydrogen infrastructures. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. The NGNP Materials R&D Program includes the following elements: (1) Developing a specific approach, program plan and other project management tools for

  9. Next Generation Nuclear Plant Research and Development Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    None

    2005-01-01

    The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: (1) Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission (2) Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented

  10. Development of safety review advisory system for nuclear power plants

    International Nuclear Information System (INIS)

    Kim, M. W.; Lee, H. C.; Park, S. O.; Park, W. J.; Lee, J. I.; Hur, K. Y.; Choi, S. S.; Lee, S. J.; Kang, C. M.

    2001-01-01

    For the development of an expert system supporting the safety review of nuclear power plants, the application program was implemented after gathering necessary theoretical background and practical requirements. The general and the detail functional specifications were established, and they were investigated by the safety review experts at KINS. Safety Review Advisory System (SRAS), the windows application on client-server environment was developed according to the above specifications. Reviewers can do their safety reviewing regardless of speciality or reviewing experiences because SRAS is operated by the safety review plans which are converted to standardized format. When the safety reviewing is carried out by using SRAS, the results of safety reviewing are accumulated in the database and may be utilized later usefully, and we can grasp safety reviewing progress. Users of SRAS are categorized into three groups, administrator, project manager, and reviewer. Each user group has appropriate access capability. The function and some screen shots of SRAS are described in this paper

  11. Development of automatic pipe welder for nuclear power plant

    International Nuclear Information System (INIS)

    Iwamoto, Taro; Ando, Shimon; Omae, Tsutomu; Ito, Yoshitoshi; Araya, Takeshi.

    1978-01-01

    Numerous pipings are installed in nuclear power plants, and of course, the reliability of these pipings are very important to preserve the safety of the plants. These pipings undergo periodic inspection yearly, and when some defects are found or some reconstructions to superior systems are made, field welding in the plants is required. When the places to be welded are in containment vessels, the works must be carried out in radiation environment. In order to maintain the highest quality of welding and to reduce the radiation exposure of workers, many skilled workers are required. This automatic pipe welder was developed to solve these problems, aiming at carrying out welding works by remote control at the safe places outside containment vessels. Especially in order to obtain the highest quality of welding, it was not perfectly automated, but the man-machine system so as to enable to utilize the delicate sense of workers was adopted. The visual and contact detecting systems to monitor welding works, remote control system, computer control, light, small and easily installed welding head, grinding and supersonic flow detecting equipments, the power source of transistor switching type, air cooling equipment, and the function for setting welding conditions according to algorithm were added to the welding machine. The outline and main components of this automatic pipe welder are explained. (Kako, I.)

  12. Development of a dynamical systems model of plant programmatic performance on nuclear power plant safety risk

    International Nuclear Information System (INIS)

    Hess, Stephen M.; Albano, Alfonso M.; Gaertner, John P.

    2005-01-01

    Application of probabilistic risk assessment (PRA) techniques to model nuclear power plant accident sequences has provided a significant contribution to understanding the potential initiating events, equipment failures and operator errors that can lead to core damage accidents. Application of the lessons learned from these analyses has resulted in significant improvements in plant operation and safety. However, this approach has not been nearly as successful in addressing the impact of plant processes and management effectiveness on the risks of plant operation. The research described in this paper presents an alternative approach to addressing this issue. In this paper we propose a dynamical systems model that describes the interaction of important plant processes on nuclear safety risk. We discuss development of the mathematical model including the identification and interpretation of significant inter-process interactions. Next, we review the techniques applicable to analysis of nonlinear dynamical systems that are utilized in the characterization of the model. This is followed by a preliminary analysis of the model that demonstrates that its dynamical evolution displays features that have been observed at commercially operating plants. From this analysis, several significant insights are presented with respect to the effective control of nuclear safety risk. As an important example, analysis of the model dynamics indicates that significant benefits in effectively managing risk are obtained by integrating the plant operation and work management processes such that decisions are made utilizing a multidisciplinary and collaborative approach. We note that although the model was developed specifically to be applicable to nuclear power plants, many of the insights and conclusions obtained are likely applicable to other process industries

  13. Development of Core Monitoring System for Nuclear Power Plants (I)

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.H.; Kim, Y.B.; Park, M.G; Lee, E.K.; Shin, H.C.; Lee, D.J. [Korea Electric Power Research Institute, Daejeon (Korea, Republic of)

    1997-12-31

    1.Object and Necessity of the Study -The main objectives of this study are (1)conversion of APOLLO version BEACON system to HP-UX version core monitoring system, (2)provision of the technical bases to enhance the in-house capability of developing more advanced core monitoring system. 2.Results of the Study - In this study, the revolutionary core monitoring technologies such as; nodal analysis and isotope depletion calculation method, advanced schemes for power distribution control, and treatment of nuclear databank were established. The verification and validation work has been successfully performed by comparing the results with those of the design code and measurement data. The advanced graphic user interface and plant interface method have been implemented to ensure the future upgrade capability. The Unix shell scripts and system dependent software are also improved to support administrative functions of the system. (author). 14 refs., 112 figs., 52 tabs.

  14. Development of nuclear power plant automated remote patrol system

    International Nuclear Information System (INIS)

    Nakayama, R.; Kubo, K.; Sato, K.; Taguchi, J.

    1984-01-01

    An Automated Remote Patrol System was developed for a remote inspection, observation and monitoring of nuclear power plant's components. This automated remote patrol system consists of; a vehicle moving along a monorail; three rails mounted in a monorail for data transmission and for power supply; an image fiber connected to a TV camera; an arm type mechanism (manipulator) for moving image fiber; a computer for control and data processing and operator's console. Special features of this Automated Remote Patrol System are as follows: The inspection vehicle runs along horizontal and vertical (up/down) monorails. The arm type mechanism (manipulator) on the vehicle is used to move image fiber. Slide type electric collectors are used for data transmission and power supply. Time-division multiplexing is adapted for data transmission. Voice communication is used for controlling mechanisms. Pattern recognition is used for data processing. The experience that has been obtained from a series of various tests is summarized. (author)

  15. Development of SC structure modularization in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Mun, Taeyoup

    2008-01-01

    New Focus on NPP are Rising Concerns on Global Warming, Potential energy crisis (geo-political), Improved reliability and safety of nuclear power plant, Advent of Generation 3+ NPP technology and Economical Energy Resource. New NPPs are 6 units in Korea and 23 in Asia being built, 32 units being planned in China by 2020 (150 by 2050), 10 units being planned in US by 2020 and IAEA expects $200 billions on NPP construction next 25 years (up to 30% of total world energy). □ SC(Steel Plate Concrete) structure · Steel Plate is used as a Structural Element instead of Reinforcing Bars in RC · SC structure consists of Steel Plate with Headed Studs. Connected by Tie-bars - The Primary Purpose of Tie-bars is to Stiffen and Hold Together the Plates during Construction Process - Headed Studs are Welded to the Inside of Steel Plate for composite action □ Benefits of SC Structure · Shorten Construction Duration for Re bar, Forming and Scaffolding Works · Minimize Site Labors · Improve the Construction Quality · Enable Construction Sites to be kept Clean □ SC Modularization · Fit for Modular Construction for Structural Features · Fit for Modular Construction for Structural Features · Inattentively Effective for Integrated Modules · Pre-fabrication, Pre-assembly and Modularization □ Project Overview · Project Name: Development of SC structure for Modularization in NPP · Project Type: Electric Power Industry R and D (Ministry of Knowledge Economy) · Duration: Sep. 2005 ∼ Aug. 2008 (36 Months) · Research Team and Scopes - Project Management: Korea Hydro and Nuclear Power Company (KHNP) - Development of Code and Standards for SC Structure: Korea Society of Steel Construction (KSSC) Korea Electric Power Research Institute (KEPRI) - Development of SC Structural Analysis and Design: Korea Power Engineering Company (KOPEC) - Development of Construction Techniques for SC Modularization: KHNP, Korea Institute of Nuclear Safety(KINS), KOPEC □ Performance

  16. Development of Northeast Asia Nuclear Power Plant Accident Simulator.

    Science.gov (United States)

    Kim, Juyub; Kim, Juyoul; Po, Li-Chi Cliff

    2017-06-15

    A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Korea's nuclear emergency response staff for training and potentially operational support in Korea's national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  17. Next Generation Nuclear Plant Research and Development Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    P. E. MacDonald

    2005-01-01

    The U.S Department of Energy (DOE) is conducting research and development (R&D) on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core could be either a prismatic graphite block type core or a pebble bed core. Use of a liquid salt coolant is also being evaluated. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission Demonstrate safe and economical nuclearassisted production of hydrogen and electricity. The DOE laboratories, led by the INL, will perform R&D that will be critical to the success of the NGNP, primarily in the areas of: High temperature gas reactor fuels behavior High temperature materials qualification Design methods development and validation Hydrogen production technologies Energy conversion. The current R&D work is addressing fundamental issues that are relevant to a variety of possible NGNP designs. This document describes the NGNP R&D planned and currently underway in the first three topic areas listed above. The NGNP Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is presented in Section 2, the NGNP Materials R&D Program Plan is presented in Section 3, and the NGNP Design Methods Development and Validation R&D Program is presented in Section 4. The DOE-funded hydrogen

  18. Next Generation Nuclear Plant Materials Research and Development Program Plan

    Energy Technology Data Exchange (ETDEWEB)

    G. O. Hayner; E.L. Shaber

    2004-09-01

    The U.S Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production without greenhouse gas emissions. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed, thermal neutron spectrum reactor that will produce electricity and hydrogen in a state-of-the-art thermodynamically efficient manner. The NGNP will use very high burn-up, low-enriched uranium, TRISO-coated fuel and have a projected plant design service life of 60 years.

  19. Application and development analysis of nuclear power plant modular construction

    International Nuclear Information System (INIS)

    Fang Xiaopeng

    2015-01-01

    Modular Construction is currently one of the major development trends for the nuclear power plant construction technology worldwide. In the first-of-a-kind AP1000 Nuclear Power Project practiced in China, the large-scale structural modules and mechanical modules have been successfully fabricated, assembled and installed. However, in the construction practice of the project, some quality issues are identified with the assembly and installation process of large-scale structural modules in addition to the issue of incomplete supply of mechanical modules, which has failed to fully demonstrate the features and merits of modular construction. This paper collects and consolidates the issues of modular construction of AP1000 first of a kind reactor, providing root cause analysis in the aspects of process design, quality control, site construction logic, interface management in the process of module fabrication, assembly and installation; modular construction feasibility assessment index is proved based on the quantification and qualitative analysis of the impact element. Based on the modular construction feasibility, NPP modular construction improvement suggestions are provided in the aspect of modular assembly optimization definition, tolerance control during the fitting process and the construction logic adjustment. (author)

  20. Selection, competency development and assessment of nuclear power plant managers

    International Nuclear Information System (INIS)

    1998-06-01

    This publication provides information on proven methods and good practices with respect to the selection, development and assessment of nuclear power plant (NPP) managers. The report is organized into four sections, a glossary, two appendices, and several annexes. The Introduction (Section 1) provides the framework for the report. Section 2 describes how appropriate management competencies can be used for the selection, development and assessment of NPP managers, including: -Selection which includes recruitment, promotion and succession management. -Management development programmes including formal training, job rotation, on the job training, mentoring, and outside assignments. -Assessment of individual performance. Section 3 describes a systematic process for identifying the competencies needed by NPP managers. This section culminates in a set of suggested core competencies for NPP managers which are further expanded in Appendix A. The annexes included provide specific examples of competency-based management selection, development, and assessment programmes in several Member States. -Annex A is one method to organize and display competencies. -Annex B is an example of using competencies for selection of first line managers. -Annex C is an example of using management competencies for succession management. -Annexes -H are examples of management development programmes. -Annexes I and J are examples of management assessment programmes. A glossary of terms is provided at the end of the report to explain the use of some key terms explain the use of some key terms

  1. Development of a nuclear power plant system analysis code

    International Nuclear Information System (INIS)

    Sim, Suk K.; Jeong, J. J.; Ha, K. S.; Moon, S. K.; Park, J. W.; Yang, S. K.; Song, C. H.; Chun, S. Y.; Kim, H. C.; Chung, B. D.; Lee, W. J.; Kwon, T. S.

    1997-07-01

    During the period of this study, TASS 1.0 code has been prepared for the non-LOCA licensing and reload safety analyses of the Westinghouse and the Korean Standard Nuclear Power Plants (KSNPP) type reactors operating in Korea. TASS-NPA also has been developed for a real time simulation of the Kori-3/4 transients using on-line graphical interactions. TASS 2.0 code has been further developed to timely apply the TASS 2.0 code for the design certification of the KNGR. The COBRA/RELAP5 code, a multi-dimensional best estimate system code, has been developed by integrating the realistic three-dimensional reactor vessel model with the RELAP5 /MOD3.2 code, a one-dimensional system code. Also, a 3D turbulent two-phase flow analysis code, FEMOTH-TF, has been developed using finite element technique to analyze local thermal hydraulic phenomena in support of the detailed design analysis for the development of the advanced reactors. (author). 84 refs., 27 tabs., 83 figs

  2. Development of life evaluation technology for nuclear power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae [Sungkwunkwan Univ., Seoul (Korea, Republic of); Lee, Hyang Beom [Soongsil Univ., Seoul (Korea, Republic of); Shin, Young Kil [Kunsan National Univ., Gunsan (Korea, Republic of); Chung, Hyun Jo [Wonkwang Univ., Iksan (Korea, Republic of); Park, Ik Keun; Park, Eun Soo [Seoul National University of Technology, Seoul (Korea, Republic of)

    2001-03-15

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing.

  3. Development of life evaluation technology for nuclear power plant components

    International Nuclear Information System (INIS)

    Song, Sung Jin; Kim, Young Hwan; Shin, Hyun Jae; Lee, Hyang Beom; Shin, Young Kil; Chung, Hyun Jo; Park, Ik Keun; Park, Eun Soo

    2001-03-01

    Retaining reliabilities of nondestructive testing is essential for the life-time maintenance of nuclear power plant. In order to Improve reliabilities of ultrasonic testing and eddy current testing, the following five subjects were carried out in this study: development of BEM analysis technique for ECT of SG tube, development of neural network technique for the intelligent analysis of ECT flaw signals of SG tubes, development of RFECT technology for the inspection of SG tube, FEM analysis of ultrasonic scattering field and evaluation of statistical reliability of PD-RR test of ultrasonic testing. As results, BEM analysis of eddy current signal, intelligent analysis of eddy current signal using neural network, and FEM analysis of remote field eddy current testing have been developed for the inspection of SG tubes. FEM analysis of ultrasonic waves in 2-dimensional media and evaluation of statistical reliability of ultrasonic testing with PD-RR test also have been carried out for the inspection of weldments. Those results can be used to Improve reliability of nondestructive testing

  4. Financing of nuclear power plants in developing countries

    International Nuclear Information System (INIS)

    Krymm, R.

    1976-01-01

    An estimation of the developing countries' nuclear energy demand until the year 2000 and the corresponding investment costs is followed by a survey of the main parameters of financing. The influence of financing on the competitive position of nuclear energy is pointed out. In the annex formulas for evaluating the modes of financing are developed. (UA) [de

  5. Development of stainless steels for nuclear power plant - Advanced nuclear materials development -

    International Nuclear Information System (INIS)

    Hong, Jun Hwa; Ryu, Woo Seog; Chi, Se Hwan; Lee, Bong Sang; Oh, Yong Jun; Byun, Thak Sang; Oh, Jong Myung

    1994-07-01

    This report reviews the status of R and D and the material specifications of nuclear components in order to develop the stainless steels for nuclear applications, and the technology of computer-assisted alloy design is developed to establish the thermodynamic data of Fe-Cr-Ni-Mo-Si-C-N system which is the basic stainless steel systems. High strength and corrosion resistant stainless steels, 316LN and super clean 347, are developed, and the manufacturing processes and heat treatment conditions are determined. In addition, a martensitic steel is produced as a model alloy for turbine blade, and characterized. The material properties showed a good performance for nuclear applications. (Author)

  6. Remote maintenance system technology development for nuclear fuel cycle plants

    International Nuclear Information System (INIS)

    Kashihara, Hidechiyo

    1984-01-01

    The necessity of establishing the technology of remote maintenance, the kinds of maintenance techniques and the change, the image of a facility adopting remote maintenance canyon process, and the outline of the R and D plan to put remote maintenance canyon process in practical use are described. As the objects of development, there are twin arm type servo manipulator system, rack system, remote tube connectors, solution sampling system, inspection system for in-cell equipment, and large plugs for wall penetration. The outline of those are also reported. The development of new remote maintenance technology has been forwarded in the Tokai Works aiming at the application to a glass solidification pilot plant and a FBR fuel recycling test facility. The lowering of the rate of utilization of cells due to poor accessibility and the increase of radiation exposure of workers must be overcome to realize nuclear fuel cycle technology. The maintenance technology is classified into crane canyon method, direct maintenance cell method, remote maintenance cell method and remote maintenance canyon method, and those are described briefly. The development plan of remote maintenance technology is outlined. (Kako, I.)

  7. Development and future perspective of nuclear power plants. Current status and future prospect of world nuclear power plants

    International Nuclear Information System (INIS)

    Kobayashi, Masaharu

    2013-01-01

    Fukushima Daiichi NPS accidents occurred on 11 March 2011 brought about great effects on nuclear development not only in Japan but also in the world. In Japan restart of operation of periodically inspected nuclear power plants (NPPs) could not be allowed except Oi NPPs two units and most parties except Liberal Democratic Party (LDP) pledged to possibly phasing out nuclear power at House of Councillors election in July and public opinion was mostly against nuclear power after the accident. LDP clearly stated that, with the inauguration of new government last December, Japan would not pursuing the policy of the prior government of possibly phasing out nuclear power by the 2030s, but would instead make a 'zero-base' review of energy policy. Germany decided to close eight reactors immediately and remaining nine by the end of 2022. For many countries, nuclear power would play an important role in achieving energy security and sustainable development goals. In 2011 NPPs 6 units started operation with 2 units under construction, and in 2012 NPPs 3 units started operation with 7 units under construction. At present there are now over 400 NPPs operating in 31 countries and world trend seemed nuclear development was continued and number of countries newly deploying NPPs was increasing as much as eighteen. This article presented current status and future prospect of world NPPs in details. Japan would like to share its experiences and information obtained from the accident with the world and also promote NPPs overseas to meet the world's expectations. (T. Tanaka)

  8. The Role of Nuclear Power Plants in the Development of a Country

    International Nuclear Information System (INIS)

    Cintra do Prado, L.

    1966-01-01

    Nuclear power plants oan contribute to the economic development of a country in two ways - as generators of power and as stimuli for the growth of new national industries. Some countries build nuclear power plants because other sources of cheap power are lacking. Other countries engage in nuclear activities such as the production of nuclear components and the purification of reactor materials before there is an immediate need for nuclear power in an attempt to plan for future national requirements and to create a nuclear export industry. A country can also import all or nearly all the components needed for the construction of nuclear power plants. In this case the importing country enjoys the benefits of facilities generating power without the advantages of a national nuclear industry. Industrialised countries generally find this solution undesirable save in exceptional circumstances, for example in the initial stages of a nuclear programme or in the case of specific nuclear components. Developing countries can also aim at a phased introduction of nuclear energy and try to combine the advantages of nuclear power production and the development of a national nuclear industry. Thus, in addition to the inherent advantages of nuclear plants as sources of electricity there are various ways in which the establishment of such plants can stimulate industry, namely through the use of nuclear and other ores utilizable in reactors, the promotion of ore-exploration activities, and the development of transformation industries

  9. The Role of Nuclear Power Plants in the Development of a Country

    Energy Technology Data Exchange (ETDEWEB)

    Cintra do Prado, L. [Brazilian Nuclear Energy Commission (Brazil)

    1966-07-01

    Nuclear power plants oan contribute to the economic development of a country in two ways - as generators of power and as stimuli for the growth of new national industries. Some countries build nuclear power plants because other sources of cheap power are lacking. Other countries engage in nuclear activities such as the production of nuclear components and the purification of reactor materials before there is an immediate need for nuclear power in an attempt to plan for future national requirements and to create a nuclear export industry. A country can also import all or nearly all the components needed for the construction of nuclear power plants. In this case the importing country enjoys the benefits of facilities generating power without the advantages of a national nuclear industry. Industrialised countries generally find this solution undesirable save in exceptional circumstances, for example in the initial stages of a nuclear programme or in the case of specific nuclear components. Developing countries can also aim at a phased introduction of nuclear energy and try to combine the advantages of nuclear power production and the development of a national nuclear industry. Thus, in addition to the inherent advantages of nuclear plants as sources of electricity there are various ways in which the establishment of such plants can stimulate industry, namely through the use of nuclear and other ores utilizable in reactors, the promotion of ore-exploration activities, and the development of transformation industries.

  10. Development of new construction technologies for nuclear power plant

    International Nuclear Information System (INIS)

    Kawamata, Susumu; Itoh, Daisuke; Ichizono, Katsuyuki

    1995-01-01

    In order to proceed rationally with the construction of a nuclear power station, the followings are the subject to solve beside reducing the construction cost: shortening the construction period, improving the quality, and securing safety in the construction work. As a measure to solve the matters, we correctly construct the plant mainly applying 'all-weather proof construction method' and 'pre-assembled large-block construction method'. Furthermore, as the plant construction control system, we perform a construction work applying the design change management system, pre-assessment for safety, and the whole facility check. As a result of our effort, it was attained that when we compare the matured plant with the first unit, the construction cost is reduced by 30%, the construction period is shortened by 12 months, site manpower is decreased by 30%, and the plant had no sudden shutdowns even during the trial operation period. (author)

  11. Development of radiological performance indicators for nuclear power plants

    International Nuclear Information System (INIS)

    Lee, B.S.; Jung, K.H.; Lee, S.H.; Jang, S.Y.

    2000-01-01

    The purpose of this work was to improve the regulatory approach to check the licensee's compliance with regulation regarding radiation protection in operating nuclear power plants (NPPs). The current domestic inspection program for NPPs requires inspectors to conduct compliance-inspection for the systems/equipment and the procedures of NPPs. In this work, we have developed a set of draft radiological performance indicators (PIs) to assess radiation safety in NPPs. The development of PIs was based on the concept that the licensees' implementation of the radiation protection program in NPPs should be able to achieve the goal of radiation protection which the International Commission on Radiological Protection (ICRP) has recommended as ICRP 60 (1991). We selected and/or developed the radiological performance indicators considering the radiation exposure network (source-environment-receptor) for NPPs. The PIs intend to be applied only to normal exposure due to normal operations including transient operational conditions, but not to potential exposure due to accidents. Also, we have chosen the receptor as workers who are occupationally exposed to radiation as well as the members of public who are exposed to radiation from radioactive effluents. The PIs intend to track the past performance rather than to expect the future performance. Finally, the individual PIs do not verify the root cause of the trend of performance; however, they provide the basis for deciding whether the procedures and work management have been properly implemented. Currently a set of 21 draft PIs has been developed for the exposure network in NPPs. For the receptor, the PIs are divided into worker individual dose, worker collective dose and public individual dose. For the environment, the PIs are related to the dose rates of controlled areas, radioactive material concentrations in controlled areas, radioactive contamination in controlled areas and at exit points, and radioactive effluent

  12. Development of SC structure modularization in Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Mun, Taeyoup [Korea Hydro and Nuclear Power Co., Ltd., Seoul (Korea, Republic of)

    2008-04-15

    New Focus on NPP are Rising Concerns on Global Warming, Potential energy crisis (geo-political), Improved reliability and safety of nuclear power plant, Advent of Generation 3+ NPP technology and Economical Energy Resource. New NPPs are 6 units in Korea and 23 in Asia being built, 32 units being planned in China by 2020 (150 by 2050), 10 units being planned in US by 2020 and IAEA expects $200 billions on NPP construction next 25 years (up to 30% of total world energy). {open_square} SC(Steel Plate Concrete) structure {center_dot} Steel Plate is used as a Structural Element instead of Reinforcing Bars in RC {center_dot} SC structure consists of Steel Plate with Headed Studs. Connected by Tie-bars - The Primary Purpose of Tie-bars is to Stiffen and Hold Together the Plates during Construction Process - Headed Studs are Welded to the Inside of Steel Plate for composite action {open_square} Benefits of SC Structure {center_dot} Shorten Construction Duration for Re bar, Forming and Scaffolding Works {center_dot} Minimize Site Labors {center_dot} Improve the Construction Quality {center_dot} Enable Construction Sites to be kept Clean {open_square} SC Modularization {center_dot} Fit for Modular Construction for Structural Features {center_dot} Fit for Modular Construction for Structural Features {center_dot} Inattentively Effective for Integrated Modules {center_dot} Pre-fabrication, Pre-assembly and Modularization {open_square} Project Overview {center_dot} Project Name: Development of SC structure for Modularization in NPP {center_dot} Project Type: Electric Power Industry R and D (Ministry of Knowledge Economy) {center_dot} Duration: Sep. 2005 {approx} Aug. 2008 (36 Months) {center_dot} Research Team and Scopes - Project Management: Korea Hydro and Nuclear Power Company (KHNP) - Development of Code and Standards for SC Structure: Korea Society of Steel Construction (KSSC) Korea Electric Power Research Institute (KEPRI) - Development of SC Structural Analysis and Design

  13. Human resource development for the new nuclear power plant unit in Armenia

    International Nuclear Information System (INIS)

    Gevorgyan, A.; Galstyan, A.; Donovan, M.

    2008-01-01

    This paper presents a discussion of a study to define the programs for development of the human resource infrastructure needed for a new nuclear power plant unit in the Republic of Armenia. While Armenia has a workforce experienced in operation and regulation of a nuclear power plant (NPP), a significant portion of the current Armenia Nuclear Power Plant (ANPP) workforce is approaching retirement age and will not be available for the new plant. The Government of Armenia is performing a human resource infrastructure study in cooperation with the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO), sponsored by the JAEA. The study of Human Resource Development for Armenia uses the INPRO methodology for assessment of human resources. The results of this study will provide the basis for decisions on human resource development programs for nuclear power in Armenia and provide a model for countries with the limited resources that are working to develop nuclear energy in the future. (authors)

  14. Nuclear power plants

    International Nuclear Information System (INIS)

    1985-01-01

    Data concerning the existing nuclear power plants in the world are presented. The data was retrieved from the SIEN (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: nuclear plants, its status and type; installed nuclear power plants by country; nuclear power plants under construction by country; planned nuclear power plants by country; cancelled nuclear power plants by country; shut-down nuclear power plants by country. (E.G.) [pt

  15. Development of Cyber Security Scheme for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hong, S. B.; Choi, Y. S.; Cho, J. W. (and others)

    2009-12-15

    Nuclear I and C system is considered to be safe on the cyber threat because of the use of exclusive communication network and operating system. But the trend of open architecture and standardization on the equipment of I and C system, it is not safe on the cyber threat such as hacking and cyber terror. It is needed to protect nuclear I and C systems by the cyber attack, Countermeasures of the cyber security is required a lot of time and endeavors because there are many factors on the environment of cyber security and cyber attack. For the nuclear cyber security, we should make structural framework and eliminate cyber vulnerabilities by the analysis of cyber environment. The framework for the cyber security includes planning, embodiment of security technologies, security audit, security management and security maintenance. In this report, we examined IT security technology and the trend of standard in the industrial I and C system, and proposed a method to construct cyber security for the nuclear power plant. We analysed the threat of cyber security, vulnerability and cyber risk, then we present a method for the cyber security structure and the countermeasures.

  16. Development of Cyber Security Scheme for Nuclear Power Plant

    International Nuclear Information System (INIS)

    Hong, S. B.; Choi, Y. S.; Cho, J. W.

    2009-12-01

    Nuclear I and C system is considered to be safe on the cyber threat because of the use of exclusive communication network and operating system. But the trend of open architecture and standardization on the equipment of I and C system, it is not safe on the cyber threat such as hacking and cyber terror. It is needed to protect nuclear I and C systems by the cyber attack, Countermeasures of the cyber security is required a lot of time and endeavors because there are many factors on the environment of cyber security and cyber attack. For the nuclear cyber security, we should make structural framework and eliminate cyber vulnerabilities by the analysis of cyber environment. The framework for the cyber security includes planning, embodiment of security technologies, security audit, security management and security maintenance. In this report, we examined IT security technology and the trend of standard in the industrial I and C system, and proposed a method to construct cyber security for the nuclear power plant. We analysed the threat of cyber security, vulnerability and cyber risk, then we present a method for the cyber security structure and the countermeasures

  17. Nuclear Power Plants. Revised.

    Science.gov (United States)

    Lyerly, Ray L.; Mitchell, Walter, III

    This publication is one of a series of information booklets for the general public published by the United States Atomic Energy Commission. Among the topics discussed are: Why Use Nuclear Power?; From Atoms to Electricity; Reactor Types; Typical Plant Design Features; The Cost of Nuclear Power; Plants in the United States; Developments in Foreign…

  18. Development of nuclear standard filter elements for PWR plant

    International Nuclear Information System (INIS)

    Weng Minghui; Wu Jidong; Gu Xiuzhang; Zhang Jinghua

    1988-11-01

    Model FRX-5 and FRX-10 nuclear standard filter elements are used for the fluid clarification of the chemical and volume control system (CVCS), boron recycle system (BRS), spent fuel pit cooling system (SFPCS) and steam generator blowdown system (SGBS) in Qinshan Nuclear Power Plant. The radioactive contaminant, fragment of resin and impurity are collected by these filter elements, The core of filter elements consists of polypropylene frames and paper filter medium bonded by resin. A variety of filter papers are tested for optimization. The flow rate and comprehensive performance have been measured in the simulation condition. The results showed that the performance and lifetime have met the designing requirements. The advantages of the filter elements are simple in manufacturing, less expense and facilities for waste-disposal. At present, some of filter elements have been produced and put in operation

  19. Development of nuclear power plant simulators for Soviet-designed nuclear reactors

    International Nuclear Information System (INIS)

    Kohut, P.; Tutu, N.K.; Cleary, E.J.; Erickson, K.G.; Yoder, J.; Kroshilin, A.

    2001-01-01

    The US Department of Energy (US DOE), under the US government's International Nuclear Safety Program (INSP), is implementing a program of developing and providing simulators for many of the Russian and Ukrainian Nuclear Power Plants (NPPs). Pacific Northwest National Laboratory (PNNL) and Brookhaven National Laboratory (BNL) manage and provide technical oversight of the various INSP simulator projects for DOE. The program also includes a simulator technology transfer process to simulator design organizations in Russia and Ukraine. Training programs, installation of new simulators, and enhancements in existing simulators, are viewed as providing a relatively fast and cost-effective technology transfer that will result in measurable improvement in the safety culture and operation of NPPs. A review of this program, its present status, and its accomplishments are provided in this paper

  20. The problems of plant control on modern stage of development of the nuclear industry

    Directory of Open Access Journals (Sweden)

    Gusev S. S.

    2017-04-01

    Full Text Available the article describes the successful development of nuclear industry at the present stage of its development, the problems of control of nuclear power plants, their classes and the choices of modern science in the field of design of fast neutron reactors as one of the promising directions of development of atomic energy.

  1. Development of an integrated cost model for nuclear plant decommissioning

    International Nuclear Information System (INIS)

    Amos, G.; Roy, R.

    2003-01-01

    A need for an integrated cost estimating tool for nuclear decommissioning and associated waste processing and storage facilities for Intermediate Level Waste (ILW) was defined during the authors recent MSc studies. In order to close the defined gap a prototype tool was developed using logically derived CER's and cost driver variables. The challenge in developing this was to be able to produce a model that could produce realistic cost estimates from the limited levels of historic cost data that was available for analysis. The model is an excel based tool supported by 3 point risk estimating output and is suitable for producing estimates for strategic or optional cost estimates (±30%) early in the conceptual stage of a decommissioning project. The model was validated using minimal numbers of case studies supported by expert opinion discussion. The model provides an enhanced approach for integrated decommissioning estimates which will be produced concurrently with strategic options analysis on a nuclear site

  2. Support system development for evacuation plan decision in nuclear plant disaster

    International Nuclear Information System (INIS)

    Fujita, Masahiko; Takayama, Jun-ichi; Nakayama, Sho-ichiro; Ushiba, Takashi

    2011-01-01

    These days, our interest in nuclear plant accidents has increased, and civic actions for them have also been activated. Therefore, improvement of the disaster prevention planning to nuclear plant accidents is requested. In this study, we developed a microscopic traffic simulation system for evacuation plan near the nuclear plant as a system which supports to examine the disaster prevention planning, and applied the system to Kashiwazaki-Kariwa nuclear plant area. Furthermore, the risk of each region near the nuclear plant disaster from the viewpoint of wind direction and the population was considered, the importance of each evacuation simulation was examined. As a result, we found that the present plan Kashiwazaki-Kariwa made has the problem on evacuation routes and others. (author)

  3. Human factors in maintenance: Development and research in Swedish nuclear power plants

    International Nuclear Information System (INIS)

    Salo, I.; Svensson, Ola

    2001-11-01

    The present report investigated previously completed, ongoing, and planned research and development projects focusing human factors and maintenance work carried out at Swedish nuclear power plants and SKI. In addition, needs for future research and development works were also investigated. Participants from all nuclear power plants and SKI were included in the study. Participants responded to a set of questions in an interview. The interviews also generated a list of future research and development projects

  4. Human factors in maintenance: development and research in Swedish nuclear power plants

    International Nuclear Information System (INIS)

    Salo, I.; Svenson, O.

    2001-11-01

    The report investigated previously completed, ongoing, and planned research and development projects focusing human factors and maintenance work carried out at Swedish nuclear power plants and SKI. In addition, needs for future research and development works were also investigated. Participants from all nuclear power plants and SKI were included in the study. Participants responded to a set of questions in an interview. The interviews also generated a list of future research and development projects. (au)

  5. Human factors in maintenance: Development and research in Swedish nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Salo, I. [Lund Univ. (Sweden). Dept. of Psychology; Svensson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology

    2001-11-01

    The present report investigated previously completed, ongoing, and planned research and development projects focusing human factors and maintenance work carried out at Swedish nuclear power plants and SKI. In addition, needs for future research and development works were also investigated. Participants from all nuclear power plants and SKI were included in the study. Participants responded to a set of questions in an interview. The interviews also generated a list of future research and development projects.

  6. Research and development towards decommissioning of Fukushima Daiichi Nuclear Power Plants

    International Nuclear Information System (INIS)

    Minato, Kazuo

    2013-01-01

    Towards the decommissioning of Fukushima Daiichi Nuclear Power Plants, science-based research and development is important and useful, as well as technology and engineering development. Research and development activities based on radiation chemistry, radiochemistry, thermodynamics, etc., have contributed to safe and efficient decommissioning of the plants. (author)

  7. Results of research and development for nuclear power plants with WWER-1000 type light water reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The conference met in three sessions: 1. Project designing and construction of nuclear power plants; 2. Materials, technologies and applied mechanics; 3. Physics, thermal physics and control. The proceedings contains 82 papers of which only two have not been inputted in INIS. The final resolutions of session 1 related to the reduction of capital costs for newly built units, processing of project documentation, the introduction of step motors manufactured in Czechoslovakia, in-service diagnostics of nuclear power plants, etc. The final recommendations of session 2 dealt with the centralization of the management of research into the reliability, safety and residual life of nuclear installations, with radiation stability of weld metals, repairs of nuclear power plants by patch welding, with welding in nuclear power plants and stress calculations using mathematical methods. Session 3 centred on questions of the safety, reliability and economy of nuclear power plant operation. It was recommended to make a comparison of the results of theoretical calculations with experiments, to concentrate on the automation of measurement, to extend international division of labour and cooperation of CMEA countries, to extend publishing activities in the field of thermal physics, etc. General recommendations were related to the conception of the construction of nuclear power plants in Czechoslovakia, the implementation of original scientific, research and development work, to the question of personnel for nuclear research, the experimental base of the Czechoslovak nuclear programme and to planning and management of technical development. (E.S.)

  8. Development of new nuclear power plants in the Republic of Korea. Annex 4

    International Nuclear Information System (INIS)

    Kim, Jung-Cha; Park, Kee-Cheol

    2002-01-01

    Nuclear power in Korea is one of our major energy sources, which accounting for approximately 50 % of the total share of electric generation using the safest and most stable methods. Based on the outstanding performance of nuclear power generation, Korea plans to construct eight (8) new nuclear power plants to maintain nuclear power as a major contributor to the national energy mix by 2014. In order to ensure that nuclear power plants are safer and more economical than any conventional electric power sources, KEPCO has developed the improved Korea Standard Nuclear Power Plant (KSNP + ) and the Korea Next Generation Reactor (KNGR) by utilizing over 30 years of expertise and learned technologies gained from construction, design, operation of sixteen nuclear units. Recently, KEPCO has developed its own project management tool, the Nuclear Project Control System (NPCS), which integrates schedule, material, cost, drawing and documentation into a computerized system, to be utilized for construction of the nuclear power plants. This paper summarizes KEPCO's various efforts for design improvement of KSNP + and KNGR in terms of performance and economic viability for construction of new nuclear power plants in Korea. (author)

  9. Nuclear design and technical development required for ISER plant

    International Nuclear Information System (INIS)

    Yokoyama, Takashi; Yamano, Naoki.

    1987-01-01

    The report outlines some results of a study carried out by the ISER (intrinsically safe and economical reactor) Investigation Group. In particular, nuclear design concepts are examined in relation to the fuel cycle. Discussion is also made on technical development efforts to be made for realizing an ISER. Calculation of some basic nuclear design parameters is performed and results are used to examine the reactor core and fuel for ISER. As a result, it is indicated that a high-burnup type reactor core should be used on a 4.5-batch replacement, 15 EFPM (effective full power month) scheme to optimize an ISER. For technical development, consideration is made on various tests to be performed with an experimental reactor, called ISER-E, as well as other tests to provide basic data required for demonstrating the inherent safety of ISER. The study also deals with the possibility of the application of currently available light water reactor techniques to experiments with a critical assembly, non-nuclear test loop, and the experimental reactor ISER-E. It is revealed that many of the required experiments can be carried out by using test facilities and light water reactor techniques which are currently, or will be readily, available. It is stressed that international cooperation is necessary to accomplish these tests. (Nogami, K.)

  10. Nuclear power plant construction

    International Nuclear Information System (INIS)

    Lima Moreira, Y.M. de.

    1979-01-01

    The legal aspects of nuclear power plant construction in Brazil, derived from governamental political guidelines, are presented. Their evolution, as a consequence of tecnology development is related. (A.L.S.L.) [pt

  11. Developing a Hierarchical Decision Model to Evaluate Nuclear Power Plant Alternative Siting Technologies

    Science.gov (United States)

    Lingga, Marwan Mossa

    A strong trend of returning to nuclear power is evident in different places in the world. Forty-five countries are planning to add nuclear power to their grids and more than 66 nuclear power plants are under construction. Nuclear power plants that generate electricity and steam need to improve safety to become more acceptable to governments and the public. One novel practical solution to increase nuclear power plants' safety factor is to build them away from urban areas, such as offshore or underground. To date, Land-Based siting is the dominant option for siting all commercial operational nuclear power plants. However, the literature reveals several options for building nuclear power plants in safer sitings than Land-Based sitings. The alternatives are several and each has advantages and disadvantages, and it is difficult to distinguish among them and choose the best for a specific project. In this research, we recall the old idea of using the alternatives of offshore and underground sitings for new nuclear power plants and propose a tool to help in choosing the best siting technology. This research involved the development of a decision model for evaluating several potential nuclear power plant siting technologies, both those that are currently available and future ones. The decision model was developed based on the Hierarchical Decision Modeling (HDM) methodology. The model considers five major dimensions, social, technical, economic, environmental, and political (STEEP), and their related criteria and sub-criteria. The model was designed and developed by the author, and its elements' validation and evaluation were done by a large number of experts in the field of nuclear energy. The decision model was applied in evaluating five potential siting technologies and ranked the Natural Island as the best in comparison to Land-Based, Floating Plant, Artificial Island, and Semi-Embedded plant.

  12. Application of balanced score card in the development of performance indicator system in nuclear power plant

    International Nuclear Information System (INIS)

    Shen Shuguang; Huang Fang; Fang Zhaoxia

    2013-01-01

    Performance indicator, which is one of ten performance monitoring tools recommended by WANO performance improvement model, has become an effective tool for performance improvement of nuclear power plant. At present, performance indicator system has been built in nuclear power plant. However, how to establish the performance indicator system that is reasonable and applicable for plant is still a question to be discussed. Performance indictor is closely tied to the strategic direction of a corporation by a balanced score card, and the performance indicator system is established from the point of performance management and strategic development. The performance indicator system of nuclear power plant is developed by introducing the balanced score card, and can be as a reference for other domestic nuclear power plants. (authors)

  13. Development of distributed computer systems for future nuclear power plants

    International Nuclear Information System (INIS)

    Yan, G.; L'Archeveque, J.V.R.

    1978-01-01

    Dual computers have been used for direct digital control in CANDU power reactors since 1963. However, as reactor plants have grown in size and complexity, some drawbacks to centralized control appear such as, for example, the surprisingly large amount of cabling required for information transmission. Dramatic changes in costs of components and a desire to improve system performance have stimulated a broad-based research and development effort in distribution systems. This paper outlines work in this area

  14. Development and application of emergency operating procedures for nuclear power plants

    International Nuclear Information System (INIS)

    Lin Chengge

    1990-01-01

    The development and application of emergency operating procedures (EOPs) is an important measure to assure the operational safety for nuclear power plants. Event-oriented, symptom-, function- and state-oriented EOPs with their structures, interfaces, development procedures and practical application are described. The ideas and approach can be available for the preparation of EOPs for nuclear power plants which are going to be in service

  15. Development of a computerized nuclear materials control and accounting system for a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Crawford, J.M.; Ehinger, M.H.; Joseph, C.; Madeen, M.L.

    1979-07-01

    A computerized nuclear materials control and accounting system (CNMCAS) for a fuel reprocessing plant is being developed by Allied-General Nuclear Services at the Barnwell Nuclear Fuel Plant. Development work includes on-line demonstration of near real-time measurement, measurement control, accounting, and processing monitoring/process surveillance activities during test process runs using natural uranium. A technique for estimating in-process inventory is also being developed. This paper describes development work performed and planned, plus significant design features required to integrate CNMCAS into an advanced safeguards system

  16. Development of a computerized nuclear materials control and accounting system for a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Crawford, J.M.; Ehinger, M.H.; Joseph, C.; Madeen, M.L.

    1979-01-01

    A computerized nuclear materials control and accounting system (CNMCAS) for a fuel reprocessing plant is being developed by Allied-General Nuclear Services at the Barnwell Nuclear Fuel Plant. Development work includes on-line demonstration of near real-time measurement, measurement control, accounting, and processing monitoring/process surveillance activities during test process runs using natural uranium. A technique for estimating in-process inventory is also being developed. This paper describes development work performed and planned, plus significant design features required to integrate CNMCAS into an advanced safeguards system. 2 refs

  17. Some issues related to the development of nuclear power plant in Indonesia

    International Nuclear Information System (INIS)

    Panggabean, L.M.

    1987-01-01

    Indonesia being a member of ASEAN belongs to the group of developing country. If Indonesia decides to embark on the establishment of nuclear power plan then the country will have no choice but to discuss the following issues: Safe operation of the plant and management of the nuclear waste. Safe operation of the power plant is important not only from the point of view of hazards to human being, or economic loss, or even death, but equally important it is also from the psychological point of view in that not to loose the society's confidence in the overall nuclear power plant programme in the future. The issue of safe operation involves both a safely designed system as well as skilled personnels to execute a well designed operation procedure. The issue of nuclear safety is getting more and more attention lately for various, some due to a deep concern about the quality of the inherent safety of the nuclear power plant to be built, others may just use their emotion to ask question like ''what its''. The issue of nuclear waste is as fundamental as the plant safety. Common people make very little difference between an atomic bomb and waste from a nuclear power plant. Another issue is one of transfer of technology which needs to be tied up with the overall industrialization process, meaning that embarking on nuclear power programme needs to contribute to local industrial activities, at least for some parts or components which can be manufactured locally. (author)

  18. Development of high purity large forgings for nuclear power plants

    International Nuclear Information System (INIS)

    Tanaka, Yasuhiko; Sato, Ikuo

    2011-01-01

    The recent increase in the size of energy plants has been supported by the development of manufacturing technology for high purity large forgings for the key components of the plant. To assure the reliability and performance of the large forgings, refining technology to make high purity steels, casting technology for gigantic ingots, forging technology to homogenize the material and consolidate porosity are essential, together with the required heat treatment and machining technologies. To meet these needs, the double degassing method to reduce impurities, multi-pouring methods to cast the gigantic ingots, vacuum carbon deoxidization, the warm forging process and related technologies have been developed and further improved. Furthermore, melting facilities including vacuum induction melting and electro slag re-melting furnaces have been installed. By using these technologies and equipment, large forgings have been manufactured and shipped to customers. These technologies have also been applied to the manufacture of austenitic steel vessel components of the fast breeder reactors and components for fusion experiments.

  19. Development of high purity large forgings for nuclear power plants

    Science.gov (United States)

    Tanaka, Yasuhiko; Sato, Ikuo

    2011-10-01

    The recent increase in the size of energy plants has been supported by the development of manufacturing technology for high purity large forgings for the key components of the plant. To assure the reliability and performance of the large forgings, refining technology to make high purity steels, casting technology for gigantic ingots, forging technology to homogenize the material and consolidate porosity are essential, together with the required heat treatment and machining technologies. To meet these needs, the double degassing method to reduce impurities, multi-pouring methods to cast the gigantic ingots, vacuum carbon deoxidization, the warm forging process and related technologies have been developed and further improved. Furthermore, melting facilities including vacuum induction melting and electro slag re-melting furnaces have been installed. By using these technologies and equipment, large forgings have been manufactured and shipped to customers. These technologies have also been applied to the manufacture of austenitic steel vessel components of the fast breeder reactors and components for fusion experiments.

  20. Utilizing reliability concepts in the development of IEEE recommended good practices for nuclear plant maintenance

    International Nuclear Information System (INIS)

    Gradin, L.P.

    1986-01-01

    This paper presents information describing the concern for nuclear power plant electrical equipment maintenance and the IEEE Nuclear Power Engineering Committee's method to address that concern. That method includes the creation of Working Group 3.3, ''Maintenance Good Practices'' which is developing specific maintenance good practice documents, supporting technical information exchange, and providing a vehicle to promote practices which can reduce cost and enhance plant safety. The foundation for that effort is the utilization of Reliability concepts

  1. The Community's research and development programme on decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    1982-01-01

    The programme, adopted by the Council of the European Communities, seeks to promote a number of research and development projects as well as the identification of guiding principles. The projects concern the following subjects: long-term integrity of buildings and systems; decontaminations for decommissioning purposes; dismantling techniques; treatment of specific waste materials (steel, concrete and graphite); large transport containers for radioactive waste arising from decommissioning of nuclear power plants in the Community; and influence of nuclear power plant design features on decommissioning

  2. Summary of developments and future projects in nuclear power plant control and instrumentation in the Netherlands

    International Nuclear Information System (INIS)

    Plas, Y. van der

    1990-01-01

    A general view is given on the developments and trends due to instrumentation and control of the two nuclear power plants in the Netherlands around the year 1989. Several projects, under which for classification of systems and components and for emergency operating procedures, are executed in both plants. An OSART mission initiated a project to make possible the periodic test of safety commands during operation. An other large project concerned the replacement of the process presentation system in Nuclear Power Plant Borssele. In the article several other developments due to the application of I and C in existing plants are outlined generally. Since 1974, no new nuclear power plants have been constructed in the Netherlands. (author). 2 figs

  3. Economic aspects of the development of nuclear power and fuel-cycle plants in the USSR

    International Nuclear Information System (INIS)

    Dergachev, N.P.; Kruglov, A.K.; Sedov, V.M.; Shuklin, S.V.

    1977-01-01

    Different possible versions of the construction programme for nuclear power stations and fuel-cycle plants in the USSR are discussed in relation to the target level of installed electrical capacity for 1980 and the predictions for the year 2000. The likely structure of the nuclear power industry is considered and the role of nuclear power stations with fast reactors is discussed, including their effect on the natural uranium supply situation. The effect of the development of fuel-cycle plants and of the organization of the reprocessing of fuel from nuclear power stations on the rate of introduction of fast reactor stations is analysed, and the effect of the technical and economic characteristics of fuel-cycle plants on the economic indices of nuclear power is studied. (author)

  4. The Communities' research and development programme on decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    This is the first progress report of the European Community's programme (1979-1983) of research on the decommissioning of nuclear power plants. It shows the status of the programme on 31 December 1980. The programme seeks to promote a number of research and development projects as well as the identification of guiding principles. The projects concern the following subjects: long-term integrity of buildings and systems; decontamination for decommissioning purposes; dismantling techniques; treatment of specific waste materials: steel, concrete and graphite; large transport containers for radioactive was produced in the dismantling of nuclear power plants; estimation of the quantities of radioactive wastes arising from decommissioning of nuclear power plants in the Community; influence of nuclear power plant design features on decommissioning

  5. Manpower development for safe operation of nuclear power plant. China. Simulator software development. UNDP-Activity: 2.1.8-IAEA-Task-01. Technical report

    International Nuclear Information System (INIS)

    Feng, C.P.

    1994-01-01

    In the frameworks of the project ''manpower development for safe operation of nuclear power plant'' the development of reactor simulator software is described. Qinshan nuclear power plant was chosen as a reference one

  6. Development of hydrogeological modelling approaches for assessment of consequences of hazardous accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Rumynin, V.G.; Mironenko, V.A.; Konosavsky, P.K.; Pereverzeva, S.A.

    1994-07-01

    This paper introduces some modeling approaches for predicting the influence of hazardous accidents at nuclear reactors on groundwater quality. Possible pathways for radioactive releases from nuclear power plants were considered to conceptualize boundary conditions for solving the subsurface radionuclides transport problems. Some approaches to incorporate physical-and-chemical interactions into transport simulators have been developed. The hydrogeological forecasts were based on numerical and semi-analytical scale-dependent models. They have been applied to assess the possible impact of the nuclear power plants designed in Russia on groundwater reservoirs

  7. Nuclear power plants maintenance

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    Nuclear power plants maintenance now appears as an important factor contributing to the competitivity of nuclea energy. The articles published in this issue describe the way maintenance has been organized in France and how it led to an actual industrial activity developing and providing products and services. An information note about Georges Besse uranium enrichment plant (Eurodif) recalls that maintenance has become a main data not only for power plants but for all nuclear industry installations. (The second part of this dossier will be published in the next issue: vol. 1 January-February 1989) [fr

  8. Development of knowledge acquisition methods for knowledge base construction for autonomous nuclear plant

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, M.; Kitamura, Y.; Mizoguchi, R. [Osaka Univ., Ibaraki (Japan). Inst. of Scientific and Industrial Research; Yoshikawa, S.

    1994-03-01

    This report describes a development of an ontology for computerized representation of behaviors and functions of number power plants. This ontology also enables computers to predict plant behaviors under assumed changes of boundary conditions and to derive cause identification suggestions of observed anomalies, as well as to represent functions and behaviors. Technical requirements in this development were recognized as clarified distinguishment of functions from behaviors and efficient plant behavior reasoning, and these have been successfully attained. The purpose of this development is to achieve knowledge sharing between human and automized instrumentation and control system about nuclear plant functions and behaviors, to realize cooperation of human and automized system for safe and cost acceptable nuclear plant operation. (author)

  9. Developing an economic performance system to enhance nuclear power plant competitiveness

    International Nuclear Information System (INIS)

    2002-01-01

    In 1999 about 16% of the world's electricity was produced by nuclear power, and the total worldwide operating experience of nuclear power plants was over 9200 reactor-years. Some 16 countries are dependent on nuclear power for more than 25% of their electricity generation. In some countries deregulation of the electricity market has either happened or is currently ongoing, while in others it is planned for the future. Nevertheless, many countries are already facing open electricity markets and operating costs are under unprecedented pressure, with competition expected to come soon to the nuclear industry worldwide. To a certain extent, however, the industry has already prepared or is currently preparing to face competition. This report is primarily intended for nuclear power plant and utility managers. It discusses the means and principal issues for the development of the nuclear economic performance international system (NEPIS), which should enhance nuclear power plant competitiveness. The following issues are addressed: The major transformations occurring in the electricity generation industry that require reductions in operations and maintenance costs at nuclear utilities; The methods that nuclear plant management use to identify and justify the economic optimum level of a plant and its use of resources; The value of collecting cost and performance data and the analysis techniques that use that data; The cost data required to be collected; The difficulty of collecting data with existing cost accounting systems; The new cost accounting and collection systems that will be required, The cost effectiveness of the overall process. This report also presents the preliminary results of a pilot project that was established to collect cost data on a few nuclear power plants and was used to verify the adequacy of the definitions and terminology set for NEPIS

  10. Experiences in the development of an emergency response facility (ERF) system for a nuclear power plant

    International Nuclear Information System (INIS)

    Seisdedos, A.; Sanchez-Fornie, M.A.

    1985-01-01

    The TMI-2 accident gave rise to a series of new requirements with which Nuclear Power Plants must comply and amongst which the implementation of emergency response facilities, particularly the SPDS, has received special attention. This paper covers the experience and problems encountered in the developing of the engineering necessary for the detailed definition of the ERF in a Nuclear Power Plant in the commercial operation phase. Also, a real example is provided for the case of a plant in the last phase of construction and installation. This will serve to illustrate each of the topics covered. (author)

  11. Reliability database development and plant performance improvement effort at Korea Hydro and Nuclear Power Co

    International Nuclear Information System (INIS)

    Oh, S. J.; Hwang, S. W.; Na, J. H.; Lim, H. S.

    2008-01-01

    Nuclear utilities in recent years have focused on improved plant performance and equipment reliability. In U.S., there is a movement toward process integration. Examples are INPO AP-913 equipment reliability program and the standard nuclear performance model developed by NEI. Synergistic effect from an integrated approach can be far greater than as compared to individual effects from each program. In Korea, PSA for all Korean NPPs (Nuclear Power Plants) has been completed. Plant performance monitoring and improvement is an important goal for KHNP (Korea Hydro and Nuclear Power Company) and a risk monitoring system called RIMS has been developed for all nuclear plants. KHNP is in the process of voluntarily implementing maintenance rule program similar to that in U.S. In the future, KHNP would like to expand the effort to equipment reliability program and to achieve highest equipment reliability and improved plant performance. For improving equipment reliability, the current trend is moving toward preventive/predictive maintenance from corrective maintenance. With the emphasis on preventive maintenance, the failure cause and operation history and environment are important. Hence, the development of accurate reliability database is necessary. Furthermore, the database should be updated regularly and maintained as a living program to reflect the current status of equipment reliability. This paper examines the development of reliability database system and its application of maintenance optimization or Risk Informed Application (RIA). (authors)

  12. The Process of Legal Drafting Regulation in the Development of the Nuclear Power Plant in Indonesia

    OpenAIRE

    Mardha, Amil

    2009-01-01

    THE PROCESS OF LEGAL DRAFTING REGULATION IN THE DEVELOPMENT OF THE NUCLEAR POWER PLANT IN INDONESIA. In Indonesia, the process of legal drafting to establish the regulation is based on the Act No. 10 Year 2004 on the Establishment of Legislation. The process shall comply with the constitutional and institutional requirements of national political and legal system. In drafting the development of the regulation of nuclear energy, BAPETEN has been involving some other agencies or other related g...

  13. Development of management systems for nuclear power plant of Hokuriku Electric Power Company

    International Nuclear Information System (INIS)

    Nakamura, Tatsuaki; Hasunuma, Junichi; Suzuki, Shintaro

    2009-01-01

    Hokuriku Electric Power Company has been operating the Shika Nuclear Power Station that it constructed in Shika city, Ishikawa prefecture, for over 15 years since bringing Unit 1 of this plant online in July 1993. In addition to electricity generation, maintenance and inspection tasks constitute a big part of operating a large-scale nuclear power plant, and in recent years, problems at power stations in the nuclear power industry have led to several revisions of nationally regulated maintenance and inspection systems. This paper describes the background, objectives, development method, and features of the Maintenance Management System and Maintenance History Management System that make effective use of information technology to promote safer and more efficient maintenance work at large-scale nuclear power plants. (author)

  14. Development of advanced I and C in nuclear power plants: ADIOS and ASICS

    International Nuclear Information System (INIS)

    Kim, Jung-Taek; Kwon, Kee-Choon; Hwang, In-Koo; Lee, Dong-Young; Park, Won-Man; Kim, Jung-Soo; Lee, Sang-Jeong

    2001-01-01

    In this paper Automatic Startup Intelligent Control System (ASICS) that automatically controls the PWR plant from cold shutdown to 5% of reactor power and Alarm and Diagnosis-Integrated Operator Support System (ADIOS) that is integrated with alarms, process values, and diagnostic information to an expert system focused on alarm processing are described. Nuclear Power Plant is manually controlled from cold shutdown to 5% according to the general operation procedures for startup operation of nuclear power plant. Alarm information is the primary sources to detect abnormalities in nuclear power plants or other process plants. The conventional hardwired alarm systems, characterized by one sensor-one indicator may lead the control room operators to be confused with avalanching alarms during plant transients. ASICS and ADIOS are designed to reduce the operator burden. The advances in computer software and hardware technology and also in information processing provide a good opportunity to improve the control systems and the annunciator systems of nuclear power plants or other similar process plants. It is very important to test and evaluate the performance and the function of the computer- or software-based systems like ASICS and ADIOS. The performance and the function of ASICS and ADIOS are evaluated with the real-time functional test facility and the results have shown that the developed systems are efficient and useful for operation and operator support

  15. Development of advanced I and C in nuclear power plants: ADIOS and ASICS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jung-Taek E-mail: jtkim@nanum.kaeri.re.kr; Kwon, Kee-Choon; Hwang, In-Koo; Lee, Dong-Young; Park, Won-Man; Kim, Jung-Soo; Lee, Sang-Jeong

    2001-07-01

    In this paper Automatic Startup Intelligent Control System (ASICS) that automatically controls the PWR plant from cold shutdown to 5% of reactor power and Alarm and Diagnosis-Integrated Operator Support System (ADIOS) that is integrated with alarms, process values, and diagnostic information to an expert system focused on alarm processing are described. Nuclear Power Plant is manually controlled from cold shutdown to 5% according to the general operation procedures for startup operation of nuclear power plant. Alarm information is the primary sources to detect abnormalities in nuclear power plants or other process plants. The conventional hardwired alarm systems, characterized by one sensor-one indicator may lead the control room operators to be confused with avalanching alarms during plant transients. ASICS and ADIOS are designed to reduce the operator burden. The advances in computer software and hardware technology and also in information processing provide a good opportunity to improve the control systems and the annunciator systems of nuclear power plants or other similar process plants. It is very important to test and evaluate the performance and the function of the computer- or software-based systems like ASICS and ADIOS. The performance and the function of ASICS and ADIOS are evaluated with the real-time functional test facility and the results have shown that the developed systems are efficient and useful for operation and operator support.

  16. Quality assurance requirements for the reliability of nuclear power plants in developing countries

    International Nuclear Information System (INIS)

    Bhutta, S.M.

    1978-01-01

    Quality Assurance (QA) has not been taken very seriously in developing countries despite their having significant nuclear power programmes. This may affect plant reliability and demands that urgent and serious attention be given to the development and implementation of Quality Assurance programmes according to local conditions. Confusion has been created by the differences in definitions and interpretations of terminologies of Quality Assurance and Control. Problems have been aggravated by the lack of clearly defined responsibilities and accountabilities during the projects execution phases of site selection, data collection, design, equipment fabrication and construction, etc. Therefore, reliability of nuclear power plants in developing countries is relatively low. But whenever some programme of QA has been implemented it has helped to improve the plant performance. This is highlighted in this paper by the practical examples for the experience of Karachi Nuclear Power Plant whose forced outages have been reduced by over a half within a period of 3 years. In view of the benefits, a QA programme for Chashma Nuclear Power Project has also been initiated. In this paper urgency is emphasized for the establishment and implementation of a formal QA programme in the developing countries if the reactor suppliers and purchasers both want to ensure higher reliability of their plants. The best way that management can play its role effectively is by setting up a strong QA organization with local personnel, thus helping to attain self-reliance and higher reliability during plant operation. (author)

  17. The impact of team building and leadership development on nuclear plant performance

    International Nuclear Information System (INIS)

    Fiedler, P.B.; Long, R.L.; Childress, J.R.

    1988-01-01

    Within the nuclear utility industry, the pressures of complex technologies, increasing regulations, and critical public scrutiny create a working environment filled with numerous pressures. The difficult nature of the industry puts a premium on effective teamwork, interdepartmental cooperation, and communication skills. A well-conceived and implemented team building and leadership development program can substantially improve the operating performance of a nuclear plant. This paper describes one such implementation effort at GPU Nuclear Corporation and at the Oyster Creek nuclear generating station (OCNGS) over an 18-month period

  18. Development of advanced secondary chemistry monitoring system for Korea nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Sang Hak; Kim, Chung Tae

    1997-01-01

    Water chemistry control is one of the most important tasks in order to maintain the reliability of plant equipments and extend the operating life of the plant. KEPCO and KOPEC developed a computerized tool for this purpose -ASCMS (advanced secondary chemistry monitoring system) which is able to monitor and diagnose the secondary water chemistry. A prototype system had been installed at KORI 3 nuclear power plant since April 1993 in order to evaluate the system performance. After the implementation of enhancements identified during the testing of the prototype, we have developed the advanced secondary monitoring system, ASCMS which is installed at 5 nuclear power plants and has been under operations since April 1997. The ASCMS comprises PC subsystem designed for data acquisition, data analysis, and data diagnosis. The ASCMS will provide overall information related to steam generator secondary side water chemistry problems and improve plant availability, reduce radiation exposure to workers, and reduce operating and maintenance costs. 6 figs

  19. Economics of Nuclear Power Plant and the development of nuclear power in Viet Nam

    International Nuclear Information System (INIS)

    Thanh, Thuy Nguyen Thi; Song, JinHo; Ha, Kwang Soon

    2015-01-01

    There are many factors affecting the capital costs like: increased plant size, multiple unit construction, improved construct methods, increase the lifetime of plant and so on, and beside is technical to enhancing the safety for NPPs. For the question that whether building a NPP is really economic than other energy resources or not, we will find the answer by comparing the USD per kWh of different energy sources as: nuclear power, coal, oil, hydro natural energy sources. The situation of energy in Vietnam was also mentioned in this paper. Vietnam has an abundant natural resources likes: coal, gas, hydro power etc, but from year 2013 to now Vietnam facing of electricity shortage and to solve the problem, Vietnam Government has chosen nuclear power energy to achieve energy balance between the rate of energy consumption and the ability to energy supply. Eight units will be built in Vietnam and in October 2014 Vietnamese officials have chosen Rosatom's AES-2006 design with VVER-1200/v-491 reactors for country's first nuclear power plant at Ninh Thuan and a second plant should follow based on a partnership with Japan. In this paper, the breakdown of NPP costs is considered. All the costs for building a NPP includes: the investment costs are the largest components (about 60%), fuel costs (15%), O and M costs (25%) and external costs are lower than 1% of the kWh costs. The situation for energy in Vietnam was mentioned with increase annually by 5.5 %, and now the shortage electricity is the big problem in power section. The purpose of this report is to give a general picture to consider the cost of nuclear power. It includes all the costs for building a nuclear power plant like total capital investment costs, production costs, external costs in which the capital investment costs is the largest component of the kWh cost. Nuclear energy Power was chosen to deal with situation of diminishing resources shortages

  20. Economics of Nuclear Power Plant and the development of nuclear power in Viet Nam

    Energy Technology Data Exchange (ETDEWEB)

    Thanh, Thuy Nguyen Thi; Song, JinHo [University of Science and Technology, Daejeon (Korea, Republic of); Ha, Kwang Soon [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    There are many factors affecting the capital costs like: increased plant size, multiple unit construction, improved construct methods, increase the lifetime of plant and so on, and beside is technical to enhancing the safety for NPPs. For the question that whether building a NPP is really economic than other energy resources or not, we will find the answer by comparing the USD per kWh of different energy sources as: nuclear power, coal, oil, hydro natural energy sources. The situation of energy in Vietnam was also mentioned in this paper. Vietnam has an abundant natural resources likes: coal, gas, hydro power etc, but from year 2013 to now Vietnam facing of electricity shortage and to solve the problem, Vietnam Government has chosen nuclear power energy to achieve energy balance between the rate of energy consumption and the ability to energy supply. Eight units will be built in Vietnam and in October 2014 Vietnamese officials have chosen Rosatom's AES-2006 design with VVER-1200/v-491 reactors for country's first nuclear power plant at Ninh Thuan and a second plant should follow based on a partnership with Japan. In this paper, the breakdown of NPP costs is considered. All the costs for building a NPP includes: the investment costs are the largest components (about 60%), fuel costs (15%), O and M costs (25%) and external costs are lower than 1% of the kWh costs. The situation for energy in Vietnam was mentioned with increase annually by 5.5 %, and now the shortage electricity is the big problem in power section. The purpose of this report is to give a general picture to consider the cost of nuclear power. It includes all the costs for building a nuclear power plant like total capital investment costs, production costs, external costs in which the capital investment costs is the largest component of the kWh cost. Nuclear energy Power was chosen to deal with situation of diminishing resources shortages.

  1. Development of new nuclear power plant in Argentina. Annex 9

    International Nuclear Information System (INIS)

    Mutsumi, V.; Fukami, I.

    2002-01-01

    Argentina has started the design of its own nuclear power, CAREM. The CAREM is an indirect cycle reactor with some distinctive features that greatly simplify the reactor and also contribute to a higher level of safety: integrated primary cooling system, primary cooling by natural circulation, self-pressurised primary system and safety systems relying on passive features. (author)

  2. IAEA activities in technology development for advanced water-cooled nuclear power plants

    International Nuclear Information System (INIS)

    Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

    2003-01-01

    As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

  3. Development of a standard equipment management model for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Hee Seung; Ju, Tae Young; Kim, Jung Wun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    Most utilities that have achieved high performance have introduced a management model to improve performance and operate plants safely. The Nuclear Energy Institute has developed and updated its Standard Nuclear Performance Model (SNPM) in order to provide a summary of nuclear processes, cost definitions, and key business performance measures for business performance comparison and benchmarking. Over the past decade, Korea Hydro and Nuclear Power Co. (KHNP) has introduced and implemented many engineering processes such as Equipment Reliability (ER), Maintenance Rule (MR), Single Point Vulnerability (SPV), Corrective Action Program (CAP), and Self Assessment (SA) to improve plant performance and to sustain high performance. Some processes, however, are not well interfaced with other processes, because they were developed separately and were focused on the process itself. KHNP is developing a Standard Equipment Management Model (SEMM) to integrate these engineering processes and to improve the interrelation among the processes. In this paper, a draft model and attributes of the SEMM are discussed.

  4. Development of a standard equipment management model for nuclear power plants

    International Nuclear Information System (INIS)

    Chang, Hee Seung; Ju, Tae Young; Kim, Jung Wun

    2012-01-01

    Most utilities that have achieved high performance have introduced a management model to improve performance and operate plants safely. The Nuclear Energy Institute has developed and updated its Standard Nuclear Performance Model (SNPM) in order to provide a summary of nuclear processes, cost definitions, and key business performance measures for business performance comparison and benchmarking. Over the past decade, Korea Hydro and Nuclear Power Co. (KHNP) has introduced and implemented many engineering processes such as Equipment Reliability (ER), Maintenance Rule (MR), Single Point Vulnerability (SPV), Corrective Action Program (CAP), and Self Assessment (SA) to improve plant performance and to sustain high performance. Some processes, however, are not well interfaced with other processes, because they were developed separately and were focused on the process itself. KHNP is developing a Standard Equipment Management Model (SEMM) to integrate these engineering processes and to improve the interrelation among the processes. In this paper, a draft model and attributes of the SEMM are discussed

  5. Nuclear power plant analyzer

    International Nuclear Information System (INIS)

    Stritar, A.

    1986-01-01

    The development of Nuclear Power Plant Analyzers in USA is described. There are two different types of Analyzers under development in USA, the forst in Idaho and Los Alamos national Lab, the second in brookhaven National lab. That one is described in detail. The computer hardware and the mathematical models of the reactor vessel thermalhydraulics are described. (author)

  6. Czechoslovakia's electrical energy industry with special regard on the development of nuclear power plants

    International Nuclear Information System (INIS)

    Paulina, A.

    1979-01-01

    In electric energy production and consumption, Czechoslovakia holds a remarkable place. Its development after the second world war can be divided into four characteristic periods. The author summarizes the features of the past development and points out the tasks of the future in which the extension of nuclear power plant building plays an important role. (author)

  7. Development of standard technical report on using and selecting wireless device in nuclear power plants

    International Nuclear Information System (INIS)

    Koo, I. S.; Hong, S. B.; Cho, J. W.

    2011-11-01

    1. Purpose · Development of IEC technical report on wireless device using in nuclear power plants 2. Contents · IEC technical reports of draft for circulation and final draft for next planary meeting · Case study on experiment of wireless devices 3. Implementation methods · Preparation of first draft with experts group, its circulation, discussions on the results of the circulation · Organizing three teams such as preparation, reviews and experiment 4. Results · Maintenance cost will be reduced with application of the wireless technologies in nuclear power plants · Commercial wireless devices will be developed before standard is issued

  8. Research and Development Technology Development Roadmaps for the Next Generation Nuclear Plant Project

    Energy Technology Data Exchange (ETDEWEB)

    Ian McKirdy

    2011-07-01

    The U.S. Department of Energy (DOE) has selected the high temperature gas-cooled reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for process heat, hydrogen and electricity production. The reactor will be graphite moderated with helium as the primary coolant and may be either prismatic or pebble-bed. Although, final design features have not yet been determined. Research and Development (R&D) activities are proceeding on those known plant systems to mature the technology, codify the materials for specific applications, and demonstrate the component and system viability in NGNP relevant and integrated environments. Collectively these R&D activities serve to reduce the project risk and enhance the probability of on-budget, on-schedule completion and NRC licensing. As the design progresses, in more detail, toward final design and approval for construction, selected components, which have not been used in a similar application, in a relevant environment nor integrated with other components and systems, must be tested to demonstrate viability at reduced scales and simulations prior to full scale operation. This report and its R&D TDRMs present the path forward and its significance in assuring technical readiness to perform the desired function by: Choreographing the integration between design and R&D activities; and proving selected design components in relevant applications.

  9. Nuclear power plant safety

    International Nuclear Information System (INIS)

    Otway, H.J.

    1974-01-01

    Action at the international level will assume greater importance as the number of nuclear power plants increases, especially in the more densely populated parts of the world. Predictions of growth made prior to October 1973 [9] indicated that, by 1980, 14% of the electricity would be supplied by nuclear plants and by the year 2000 this figure would be about 50%. This will make the topic of international co-operation and standards of even greater importance. The IAEA has long been active in providing assistance to Member States in the siting design and operation of nuclear reactors. These activities have been pursued through advisory missions, the publication of codes of practice, guide books, technical reports and in arranging meetings to promote information exchange. During the early development of nuclear power, there was no well-established body of experience which would allow formulation of internationally acceptable safety criteria, except in a few special cases. Hence, nuclear power plant safety and reliability matters often received an ad hoc approach which necessarily entailed a lack of consistency in the criteria used and in the levels of safety required. It is clear that the continuation of an ad hoc approach to safety will prove inadequate in the context of a world-wide nuclear power industry, and the international trade which this implies. As in several other fields, the establishment of internationally acceptable safety standards and appropriate guides for use by regulatory bodies, utilities, designers and constructors, is becoming a necessity. The IAEA is presently planning the development of a comprehensive set of basic requirements for nuclear power plant safety, and the associated reliability requirements, which would be internationally acceptable, and could serve as a standard frame of reference for nuclear plant safety and reliability analyses

  10. Development of halogen-free flame-retardant cable for nuclear power plant

    International Nuclear Information System (INIS)

    Ishii, Nobuhisa; Morii, Akira; Fujimura, Shunichi

    1991-01-01

    Conventional flame-retardant cables release a large amount of corrosive and toxic gases and also smoke during combustion on fire. Cables covered with halogen-free flame-retardant material, containing no halogen in it, have been developed to reduce generation of such gases and smoke, and already used in telecommunication plant, subway and shipboard applications. In the case of nuclear power plant application, cable covering materials should also have radiation resistance and other properties including long-term physical stability. We have developed halogen-free flame-retardant cables for nuclear power plant with sufficient flame retardancy, radiation resistance, and environmental resistance including steam-exposure resistance, all of which are in accordance with Japanese specifications for nuclear cables, and with characteristics as low corrosiveness, low toxicity, and low smoke evolution. (author)

  11. Introduction of Integrity Evaluation Criteria Developing during and after fire for Nuclear Power Plant Buildings

    International Nuclear Information System (INIS)

    Lee, Jin Woo

    2016-01-01

    The first project for nuclear power plant built in Korea to taking account the engineering based approach was started on October 2015 including the whole process such as fire hazard analysis, standard fire modeling, cable tray fire with multi spurious operation, structural fire integrity evaluation, and large area fire induced air craft crash. This paper covers the brief developing scheme and roadmap focusing on structural fire evaluation criteria. The meaningful first step for developing the structural fire integrity in nuclear power plant building is started with the series of fire related sub sections mentioned in earlier section. The recognition and sufficient effort of fire research leads to set up the safe and reliable design of nuclear power plant

  12. Introduction of Integrity Evaluation Criteria Developing during and after fire for Nuclear Power Plant Buildings

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jin Woo [KEPCo, Gimcheon (Korea, Republic of)

    2016-05-15

    The first project for nuclear power plant built in Korea to taking account the engineering based approach was started on October 2015 including the whole process such as fire hazard analysis, standard fire modeling, cable tray fire with multi spurious operation, structural fire integrity evaluation, and large area fire induced air craft crash. This paper covers the brief developing scheme and roadmap focusing on structural fire evaluation criteria. The meaningful first step for developing the structural fire integrity in nuclear power plant building is started with the series of fire related sub sections mentioned in earlier section. The recognition and sufficient effort of fire research leads to set up the safe and reliable design of nuclear power plant.

  13. Antitrust review of nuclear power plants. Recent developments

    International Nuclear Information System (INIS)

    Delaney, J.B.; Honeycutt, T.C.; Messier, M.

    1978-04-01

    This report updates and augments an earlier report that cataloged antitrust license conditions resulting from the Nuclear Regulatory Commission's (NRC's) antitrust review process mandated by the 1970 antitrust amendments from December 1970 to April 1976. An overview of the NRC's antitrust review process relevant to applications for construction permits and operating licenses is presented. Information on the antitrust reviews completed from May 1976 to February 1978 is also given. This information includes the identity of applicants, utility system and unit sizes, the geographic location of the applicant, and a summary of any antitrust license conditions

  14. Success and failure of nuclear power plant Quality Assurance when applied to research and development activities

    Energy Technology Data Exchange (ETDEWEB)

    Michels, R.D.; Stark, W.A.; Eberhart, C.F.

    1986-01-01

    Quality Assurance in some form or fashion has been practiced at research and development laboratories and facilities for quite some time. This quality assurance effort has largely been that of the scientist taking good notes as the experiment or observation is conducted to provide adequate documentation of the research and development effort. With the advent of the national laboratories participating in such projects as the Waste Isolation Project Plant (WIPP), Three Mile Island, and the National Waste Storage Program, a need for planned, systematic quality assurance for research and development activities has come into being. Based on experience evolved from nuclear power plant and facility construction, the government has chosen to utilize for these new projects the criteria developed for use in construction of power plants. While these criteria have worked reasonably well in the construction, maintenance, and operation of those nuclear plants, their application to research and development activities has not been an unqualified success. This paper compares and contrasts four criteria; surveillance, inspection, procurement, and calibration for application in the research and development situation, and nuclear power plants.

  15. Success and failure of nuclear power plant Quality Assurance when applied to research and development activities

    International Nuclear Information System (INIS)

    Michels, R.D.; Stark, W.A.; Eberhart, C.F.

    1986-01-01

    Quality Assurance in some form or fashion has been practiced at research and development laboratories and facilities for quite some time. This quality assurance effort has largely been that of the scientist taking good notes as the experiment or observation is conducted to provide adequate documentation of the research and development effort. With the advent of the national laboratories participating in such projects as the Waste Isolation Project Plant (WIPP), Three Mile Island, and the National Waste Storage Program, a need for planned, systematic quality assurance for research and development activities has come into being. Based on experience evolved from nuclear power plant and facility construction, the government has chosen to utilize for these new projects the criteria developed for use in construction of power plants. While these criteria have worked reasonably well in the construction, maintenance, and operation of those nuclear plants, their application to research and development activities has not been an unqualified success. This paper compares and contrasts four criteria; surveillance, inspection, procurement, and calibration for application in the research and development situation, and nuclear power plants

  16. Technology development on the assessment of structural integrity of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jeong Moon; Choun, Y. S.; Choi, I. K. and others

    1999-04-01

    Nuclear power plants in Korea show drop off in their performance and safety margin as the age of plants increase. The reevaluation of Kori-1 Unit on its performance and safety for life extension is expected in the near future. However, technologies and information related are insufficient to quantitatively estimate them. The final goal of this study is to develop the basic testing and evaluation techniques related with structural integrity of important nuclear equipment and structures. A part of the study includes development of equipment qualification technique. To ensure the structural integrity of structures, systems, and equipment in nuclear power plants, the following 5 research tasks were performed in the first year. - Analysis of dynamic characteristics of reactor internals - Analysis of engineering characteristics of instrumental earthquakes recorded in Korea - Analysis of ultimate pressure capacity and failure mode of containments building - Development of advanced NDE techniques using ultrasonic resonance scattering - Development of equipment qualification technique against vibration aging. These technologies developed in this study can be used to ensure the structural safety of operational nuclear power plants, and for the long-term life management. (author)

  17. Development of EDG Engine Condition Diagnosis Logic in Korean Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Byoung Oh; Choi, Kwang Hee; Lee, Sang Guk

    2012-01-01

    Through benchmarking using the excellent record of the nuclear power plants under operation in the United States and Europe and with the continuous development of nuclear-related technology, the Korea Hydro and Nuclear Power Co., LTD (KHNP) reached an average planned preventive maintenance period of 29.6 days in 2009. In addition, KHNP plans to reduce the planned preventive maintenance period at Korea standard nuclear plants (KSNPs) from 29.6 days to less than 21 days by 2014 through a combination of domestic research and development (R and D) and the introduction of the technical know-how applied in the very best overseas nuclear power plants (NPPs). Accordingly, it is necessary to reduce the inspection and maintenance periods of an emergency diesel generator (EDG), which are currently set in the planned preventive maintenance period. If the condition-based predictive maintenance (CBM) technology is applied to EDG engines, the maintenance period of an EDG will be shortened because engine maintenance is accomplished according to the engine condition under this plan. In this study, in the series of CBM program developments which will be applied to EDG engines, the development results of condition diagnosis logic to be applied to EDG engines for exiting domestic NPPs are introduced

  18. Risk-informed technology developments for nuclear power plants within the ASME in 2000-2001

    International Nuclear Information System (INIS)

    Wesley Rowley, C.; Balkey, K.R.

    2001-01-01

    The purpose of this paper is to provide information on developments within the ASME to support risk-informing NRC regulations for nuclear power plants. This paper builds on a publication at ICONE-8 that discussed ASME risk-informed nuclear power plant initiatives, both in Research and in Codes and Standards, particularly those related to risk-informing Part 50 of the 10 CFR (Code of federal regulations). During the past year, the ASME BNCS formed a Task Force to focus the Society's efforts to support risk-informing 10 CFR Part 50. Key efforts underway that are guided by the task force include finalizing the ASME PRA (probability risk assessment) Standard, developing a Code Case to risk-inform the repair, replacement, and modification activities for ASME components, and developing a Code Case to risk-inform the safety classification of pressure boundary components. Several other initiatives are also under investigation such as introducing risk insights into other ASME nuclear codes and standards supported by appropriate research and technical basis information. Supplementary information will also be provided to update an initial high level plan of ASME risk-informed initiatives for nuclear power plants that was presented at ICONE-8, including plans to communicate these risk-informed technology developments to the public. The authors included and acknowledged contributions from several other cognizant members of the ASME BNCS (board on nuclear codes standards) Task Group on RIP50 in the paper. (authors)

  19. Development of passive condensers for accident localization systems at nuclear power plants in the former USSR

    International Nuclear Information System (INIS)

    Kuznecov, M.V.

    1992-01-01

    The development is summarized of passive condensers for accident localization systems at nuclear power plants (with RBMK and WWER reactors) in the former USSR. Basic properties and criteria defining their availability are described, as are experimental tests and technical solution optimization results. (author) 2 fig

  20. Development of maintenance tools based on tribology for nuclear power plant

    International Nuclear Information System (INIS)

    Iwai, Yoshiro; Honda, Tomomi; Yoshinaga, Shigeki

    2007-01-01

    Many tribological problems occur in the maintenance of nuclear power plants. In this paper, we introduced our developed maintenance tools based on tribology such as a particle counter and wear sensors. Their principals were shown briefly and their usefulness was clarified by the experiments on monitoring of wear. (author)

  1. Development of Internal Dose Assessment Program for Nuclear Power Plant Employees

    Energy Technology Data Exchange (ETDEWEB)

    Song, Myung Jae; Kang, Duck Won; Maeng, Sung Jun; Kim, Hee Geun; Son, Soon Whan; Lim, Young Kee; Son, Joong Kwon; Park, Keyoung Rock [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jang, See Young; Ha, Jong Woo; Suh, Keyoung Won [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Oh, Oak Doo; Lee, Joong Woo; Yoon, Sung Sik [Yonsei University, Seoul (Korea, Republic of)

    1996-12-31

    Internal exposure monitoring based on new concept of radiation protection. Analysis and Performance test of the in vivo systems being operated in nuclear power plants in Korea. Design and fabrication of humanoid phantom for calibration of in vivo system. Development of internal dose evaluation code based on the ICRP 30 dosimetric model. (author). 44 refs., figs.

  2. Implementation of nuclear power plant in a developing country - the case of Turkey

    International Nuclear Information System (INIS)

    Aybers, N.

    1996-01-01

    This paper discusses problems encountered in the integration of the first nuclear power plant in Turkey. Past activities and the current situation is reviewed, special problems associated with the Akkuyu project are discussed, and topics such as optimal size selection, policy for nuclear fuel cycle, choice of reactor type, and regulatory matters are discussed. It is concluded that the introduction of a well-planned, long-term, economically viable nuclear power program into a developing country is a difficult task, which can be accomplished only through effective organization. (author). 7 refs., 4 tabs., 2 figs

  3. Implementation of nuclear power plant in a developing country - the case of Turkey

    International Nuclear Information System (INIS)

    Aybers, N.

    1996-01-01

    This paper discusses problems encountered in the integration of the first nuclear power plant in Turkey. Past activities and the current situation is reviewed, special problems associated with the AKKUYU project are discussed, and topics such as optimal size selection, policy for nuclear fuel cycle, choice of reactor type, and regulatory matters are discussed. It is concluded that the introduction of a well-planned, long-term, economically viable nuclear power program into a developing country is a difficult task, which can be accomplished only through effective organization. (author)

  4. Current developments in mechanized non-destructive testing in nuclear power plants

    International Nuclear Information System (INIS)

    Zeilinger, R.

    2008-01-01

    Nuclear power plants require frequent in-service activities to be carried out conscientiously in areas potentially hazardous to human operators (because of the associated radiation exposure), such as non-destructive testing of pressurized components of the steam system. Locations to be inspected in this way include the reactor pressure vessel, core internals, steam generators, pressurizers, and pipes. The codes to be used as a basis of these inspections demand high absolute positioning and repeating accuracy. These requirements can be met by mechanized test procedures. Accordingly, many new applications of, mostly mobile, robots have been developed over the past few years. The innovative control and sensor systems for stationary and mobile robots now on the market offer a potential for economic application in a large number of new areas in inspection, maintenance and service in nuclear power plants. More progress in this area is expected for the near future. Areva NP founded the new NDT Center, NETEC (Non-destructive Examination Technical Center), as a global technical center for non-destructive materials testing. NETEC is to advance research and development of all basic NDT technologies, robotics included. For many years, intelligeNDT has offered solutions and products for a variety of inspection and testing purposes and locations in nuclear power plants and is involved in continuous further development of the experience collected in nuclear power plants on the spot. (orig.)

  5. Development of fragility descriptions of equipment for seismic risk assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Hardy, G.S.; Campbell, R.D.

    1983-01-01

    Probabilistic risk assessment (PRA) of a nuclear power plant for postulated hazard requires the development of fragility relationships for the plants' safety related equipment. The objective of this paper is to present some general results and conclusions concerning the development of these seismic fragility levels. Participation in fragility-related research and experience gained from the completion of several PRA studies of a variety of nuclear power plants have provided much insight as to the most vulnerable equipment and the most efficient use of resources for development of fragilities. Plants studied had seismic design bases ranging from very simple equivalent static analysis for some of the earlier plants to state-of-the-art complex multimode dyanamic analyses for plants currently under construction. Increased sophistication and rigor in seismic qualification of equipment has resulted for the most part in increased seismic resistance. The majority of equipment has been found, however, to possess more than adequate resistance to seismic loading regardless of the degree of sophistication utilized in design as long as seismic loading was included in the design process. This paper presents conclusions of the authors as to which items of equipment typically require an individual ''plant-specific'' fragility analysis and which can be treated in a generic fashion. In addition, general conclusions on the relative seismic capacity levels and most frequent failure modes are summarized for generic equipment groups

  6. Development of human factors evaluation techniques for nuclear power plants

    International Nuclear Information System (INIS)

    Oh, I.S.; Lee, Y.H.; Lee, J.W.; Sim, B.S.

    1999-01-01

    This paper describes development of an operator task simulation analyzer and human factors evaluation techniques performed recently at Korea Atomic Energy Research Institute. The first is the SACOM (Simulation Analyzer with a Cognitive Operator Model) for the assessment of task performance by simulating control room operation. The latter has two objectives: to establish a human factors experiment facility, the Integrated Test Facility (ITF), and to establish techniques for human factors experiments. (author)

  7. Development of Halogen-free flame-retardant cable for nuclear power plant

    International Nuclear Information System (INIS)

    Ishii, Nobuhisa; Morii, Akira; Fujimura, Shunichi

    1992-01-01

    Conventional flame-retardant cables release a large volume of corrosive and toxic gases as well as smoke while combusted. Cables covered with halogen-free flame-retardant material, containing no halogen in it, have been developed to reduce generation of such gases and smoke, and have already been used in telecommunication service, subway and shipboard applications. However, for cables for nuclear power plant, covering materials should also have radiation resistance and other properties, including long-term physical stability. We have developed halogen-free flame-retardant cables for BWR nuclear power plant with sufficient flame retardancy radiation resistance and environmental resistance including steam-exposure resistance all of which are in accordance with Japanese specifications for BWR nuclear cables and have such characteristics as low corrosiveness, low toxicity and low smoke emission. (author)

  8. Development of operator thinking model and its application to nuclear reactor plant operation system

    International Nuclear Information System (INIS)

    Miki, Tetsushi; Endou, Akira; Himeno, Yoshiaki

    1992-01-01

    At first, this paper presents the developing method of an operator thinking model and the outline of the developed model. In next, it describes the nuclear reactor plant operation system which has been developed based on this model. Finally, it has been confirmed that the method described in this paper is very effective in order to construct expert systems which replace the reactor operator's role with AI (artificial intelligence) systems. (author)

  9. A basic study for development of environmental standard review plan of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Chang Hyun; Cho, Jae Seon; You, Young Woo [Seoul Nationl Univ., Seoul (Korea, Republic of)] (and others)

    1999-12-15

    In this study is performed a basic study to be ready for the development and detail analysis of NUREG-1555 ESRP. As a fundamental research for literature survey and development of draft review plan, review and translation of NUREG-1555 published by NRC, and which is applied to licensing procedure of Nuclear Power Plants are included. These provided the basic information for the developments of the environmental standard review plan.

  10. Public perception toward information and knowledge of nuclear power plant development; A Malaysian case study

    Science.gov (United States)

    Misnon, Fauzan Amin; Rahman, Irman Abd.; Hu, Yeoh Siong; Yasir, Muhamad Samudi

    2018-04-01

    Knowledge has been known as a key element in developing support and perception by the public towards any new policy by the government, including the development of nuclear energy. The success of the policy is mainly dependent on public support which is related to the perception cultivated by the knowledge that is already held by the people. A public survey was conducted between 14 March 2016 to 10 May 2016 focusing on the Malaysian public acceptance and perception towards the implementation of nuclear energy in Malaysia (n=1438). This research was aimed on the research question which is `Does the level of general knowledge and education regarding nuclear energy in Malaysia influence the acceptance and support of the people to develop nuclear technology?'. These finding suggest that open discourse is a must even at the initial stages of developing a Nuclear Power Plant (NPP) program, which will touch on a few points that will assist to society's knowledge and understanding towards nuclear energy from the aspects of management, safety, radioactive waste, impact on economic competitiveness as well as the benefits and risks regarding the development of NPP. By increasing the knowledge regarding nuclear power and radioactive waste, perception towards the pros of NPP as well as the conviction towards its safety can be increased. It is suggested for the parties involved to use the latest approach of information technology as a way to provide the most effective medium of information dissemination to the public.

  11. Direction of Technology Development for Nuclear Power Plants at the O and M Phase

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Insu; Park, Hwanpyo; Kim, Younghyun [Korea Institute of Construction Technology, Goyang (Korea, Republic of)

    2014-05-15

    Recently, Korea has attempted to advance overseas markets by securing competitive power in nuclear power technology. In order to develop and operate overall construction management systems with Korea's own brand equipped with sufficient applicability and competitive power in the market abroad and to ensure equal competitive power with other foreign advanced companies of nuclear power plants, Korea has launched a project called 'Data Centric Integration/Automation Technology for NP Project Management System' since July 2011. This project is divided into two phases: the first phase from 2012 to 2016 realizes EPCS stage, and the second phase from 2017 to 2020 extends to O and M stage. Appropriate technology development planning must be established if 'Data Centric Integration/Automation Technology for NP Project Management System' conducted at the first phase would extend to O and M stage at the second phase. Therefore, this study aimed at drawing out the direction of technology development based on present analysis of process at the operational phase of nuclear power plants in Korea conducted as previous study. This study analyzed current operation and maintenance systems first, analyzing the results of differences between the operation process of nuclear plants in Korea which was suggested at the previous study and the process of the Korea Hydro and Nuclear Power Co., Ltd. (hereafter referred to as 'KHNP') and drawing out the direction of technology development for nuclear power plants at the operational phase from the viewpoint of life cycle.

  12. Direction of Technology Development for Nuclear Power Plants at the O and M Phase

    International Nuclear Information System (INIS)

    Jung, Insu; Park, Hwanpyo; Kim, Younghyun

    2014-01-01

    Recently, Korea has attempted to advance overseas markets by securing competitive power in nuclear power technology. In order to develop and operate overall construction management systems with Korea's own brand equipped with sufficient applicability and competitive power in the market abroad and to ensure equal competitive power with other foreign advanced companies of nuclear power plants, Korea has launched a project called 'Data Centric Integration/Automation Technology for NP Project Management System' since July 2011. This project is divided into two phases: the first phase from 2012 to 2016 realizes EPCS stage, and the second phase from 2017 to 2020 extends to O and M stage. Appropriate technology development planning must be established if 'Data Centric Integration/Automation Technology for NP Project Management System' conducted at the first phase would extend to O and M stage at the second phase. Therefore, this study aimed at drawing out the direction of technology development based on present analysis of process at the operational phase of nuclear power plants in Korea conducted as previous study. This study analyzed current operation and maintenance systems first, analyzing the results of differences between the operation process of nuclear plants in Korea which was suggested at the previous study and the process of the Korea Hydro and Nuclear Power Co., Ltd. (hereafter referred to as 'KHNP') and drawing out the direction of technology development for nuclear power plants at the operational phase from the viewpoint of life cycle

  13. Development of safety assessment of nuclear power plants using indicators; Ydinvoimalaitosten turvallisuuden arvioinnin kehittaeminen tunnuslukujen avulla

    Energy Technology Data Exchange (ETDEWEB)

    Tiippana, P.

    1997-11-01

    The study is based on an indicator system which is under development at the Radiation and Nuclear Safety Authority (STUK). The goal of this study was to define and develop both PSA-based indicators and indicators from failure statistics. As PSA-based indicators the possibility was studied to define and express the risk importance of exemptions from the Technical Specifications, failures, preventive maintenance and other disconnections of devices covered by the Technical Specifications, operating events covered by Guide YVL 1.5 and plant modifications. In this piece of research the applicability of plant specific living PSA-models used for calculation of indicators was examined. The research included both Loviisa and Olkiluoto nuclear power plants in Finland. 47 refs.

  14. Research and development on the human factors technologies for nuclear power plants in Korea

    International Nuclear Information System (INIS)

    Sim, B.S.; Lee, J.W.; Cheon, S.W.

    1996-01-01

    This paper introduces our research project on the development of human factors technologies for nuclear power plants in Korea. The project is divided into two sub-projects. The first sub-project is the development of human factors experimental evaluation techniques, which aims to develop the techniques for experimental design, experimental measurement and data collection/analysis, and to construct an integrated test facility (ITF) suitable for the experimental evaluation of man-machine interfaces (MMIs) at an advanced control room. The second is the development of human behavior analysis techniques, which has two research areas: one is the development of a task simulation analyzer and the other is human error case studies for nuclear power plants. (author)

  15. Developing a computerized aging management system for concrete structures in finnish nuclear power plants

    International Nuclear Information System (INIS)

    Al-Neshawy, F.; Piironen, J.; Sistonen, E.; Vesikari, E.; Tuomisto, M.; Hradil, P.; Ferreira, M.

    2013-01-01

    Finland has four nuclear reactors units in two power plants. The first unit started operation in 1977 and in the early 1980's all four units were in use. During the last few years the aging management of the Nuclear Power Plant's (NPP) concrete structures has grown an important issue because the existing structures are reaching the end of their licensed operating lifetime (about 40 years). Therefore the nuclear power companies are developing aging management systems to avoid premature degradation of NPP facilities and to be able to extend their operating lifetime. This paper is about the development of a computerized ageing management system for the nuclear power plants concrete structures. The computerized ageing management system is built upon central database and implementation applications. It will assist the personnel of power companies to implement the aging management activities at different phases of the lifetime of a power plant. It will provide systematic methods for planning, surveillance, inspection, monitoring, condition assessment, maintenance and repair of structures. (authors)

  16. Development of Intelligent Database Program for PSI/ISI Data Management of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Park, Un Su; Park, Ik Keun; Um, Byong Guk; Park, Yun Won; Kang, Suk Chul

    1998-01-01

    For an effective and efficient management of large amounts of preservice/inservice inspection(PSI/ISI) data in nuclear power plants, an intelligent Windows 95-based data management program was developed. This program enables the prompt extraction of previously conducted PSI/ISI conditions and results so that the time-consuming data management, painstaking data processing and analysis in the past are avoided. The program extracts, and the associated remedies. Furthermore, additional inspection data and comments can be easily added or deleted for subsequent PSI/ISI operation. Although the initial version of the program was applied to Kori nuclear power plant, this program can be equally applied to other nuclear power plant. And also this program can be used to offer the fundamental data for application of evaluation data related to fracture mechanics analysis(FMA), probabilistic reliability assessment(PRA) of PSI/ISI results, performance demonstration initiative(PDI) and risk-informed ISI based on probability of detection(POD) information of ultrasonic examination. Besides, the program can be further developed as a unique PSI/ISI data management expert system that can be apart of PSI/ISI data management expert system that can be a part of PSI/ISI Total Support System(TSS) for Korean nuclear power plants

  17. Development of graphic display program of reactor operating parameters for emergency exercise at nuclear power plants

    International Nuclear Information System (INIS)

    Okuda, Yasunori; Yoshida, Yoshitaka; Gotou, Kazuko

    2001-01-01

    A scenario of nuclear emergency exercise based on the result of accident progress analysis is expected to ensure effective training. Thereupon a new graphic display program for reactor operating parameters has been developed to present real-time of plant process values (parameters), released radioactivities from the plant, and dose rate data around the site calculated by using the accident analysis code MAAP4 and other codes. This system has a trend graph screen displaying reactor operating parameters, an environmental dose rate summary screen indicating dose rate distribution around the site on the map, and a plant parameters summary screen showing important plant parameters on a simplified plant system diagram. One screen can be switched to another any time. It also has a jump-function easily accessing any stage during the exercise scenario in accordance with progress of the exercise. As a result of the application of this system to a real nuclear emergency exercise, it has been verified that this system is quite useful for confirming the parameters when the nuclear emergency exercise starts and the licensee reports the plant conditions to related bodied. (author)

  18. Development of graphic display program of reactor operating parameters for emergency exercise at nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Okuda, Yasunori; Yoshida, Yoshitaka [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan); Gotou, Kazuko [Kansai Electric Power Co., Inc., Osaka (Japan)

    2001-09-01

    A scenario of nuclear emergency exercise based on the result of accident progress analysis is expected to ensure effective training. Thereupon a new graphic display program for reactor operating parameters has been developed to present real-time of plant process values (parameters), released radioactivities from the plant, and dose rate data around the site calculated by using the accident analysis code MAAP4 and other codes. This system has a trend graph screen displaying reactor operating parameters, an environmental dose rate summary screen indicating dose rate distribution around the site on the map, and a plant parameters summary screen showing important plant parameters on a simplified plant system diagram. One screen can be switched to another any time. It also has a jump-function easily accessing any stage during the exercise scenario in accordance with progress of the exercise. As a result of the application of this system to a real nuclear emergency exercise, it has been verified that this system is quite useful for confirming the parameters when the nuclear emergency exercise starts and the licensee reports the plant conditions to related bodied. (author)

  19. Development of safety culture by improving risk communication in a nuclear power plant

    International Nuclear Information System (INIS)

    Sugiman, Toshio; Yoshikawa, Hidekazu

    2004-01-01

    Safety culture is conceptually examined from the viewpoint of activity theory proposed by Engestroem. The theory is instrumental in broadening our scope of views to the extent that a particular (group of person(s) and his/her (their) environment are regarded as a part of activity of a larger collectivity consisting of people and their physical and institutional environments. Overwhelming orientation toward an unmanned plant is reflected in object → outcome and mediating artifacts in the structure of activity in a nuclear power plant while it is contradicted with another orientation toward the improvement of employee's ability in community, rules, and division of labor. Three possible ways to transform a currently dominant activity into a new form were suggested depending on a preliminary study of our research project headed by the seconds author. First, community of a nuclear power plant may as well be expanded to the extent that nuclear experts outside the plant, especially those working for a plant/equipment manufacturing company, can share information with workers inside the plant through internet system that has been developed in our project. Second, community of workplace may as well be reexamined concerning division of labor between a supervisor and subordinates. Among all, leader behaviors of a supervisor that were effective to reduce psychological stress of subordinates, which, in turn, make positive contribution to safety culture, were identified by a questionnaire survey. Last, a minority of workers who tend to take risks for radiation exposure more seriously than the majority and share similar risk cognition with ordinary citizens may as well be utilized as linking pin that brings a warning signal from both a minority of nuclear experts outside the plant and ordinary citizens to the majority of workers in the plant who tend to be devoted to implementing everyday job. (author)

  20. Nuclear Plant Data Bank

    International Nuclear Information System (INIS)

    Booker, C.P.; Turner, M.R.; Spore, J.W.

    1986-01-01

    The Nuclear Plant Data Bank (NPDB) is being developed at the Los Alamos National Laboratory to assist analysts in the rapid and accurate creation of input decks for reactor transient analysis. The NPDB will reduce the time and cost of the creation or modification of a typical input deck. This data bank will be an invaluable tool in the timely investigation of recent and ongoing nuclear reactor safety analysis. This paper discusses the status and plans for the NPDB development and describes its anticipated structure and capabilities

  1. Some problems on domestic technology development from a point of fabricator of nuclear power plant. [Japan

    Energy Technology Data Exchange (ETDEWEB)

    Watamori, T [Hitachi Ltd., Tokyo (Japan)

    1976-06-01

    During past 20 years, the nuclear power industry in Japan has introduced foreign technology, digested it in a short period, and continued to research and develop domestic technology. Now, 95% of the machinery and equipments for nuclear power generation with light water reactors can be produced domestically, and some technologies are going to be exported. However, the nuclear power industry is still in a severe environment. The progress of the development of nuclear power plants passed the periods of organizational preparation, the construction of research reactors, the import of foreign technologies and reactors for practical use, and the construction of domestically produced reactors for practical use. The supplying capacity of the nuclear power industry in Japan reached 6 units of 1,000 MW yearly, but in order to meet the long term plan of nuclear power generation, this capacity must be further enhanced. The problems in the promotion of domestic production are the establishment of independent technologies, the promotion of standardization, the strengthening of business basis, the upbringing of relating enterprises, and the acceleration of national projects. Since the energy crisis, the trend of filling up energy demand with nuclear power generation became conspicuous, but for the expansion of export, the problems of safety guarantee, nuclear fuel cycle, and financial measures must be solved with government aid.

  2. Some problems on domestic technology development from a point of fabricator of nuclear power plant

    International Nuclear Information System (INIS)

    Watamori, Tsutomu

    1976-01-01

    During past 20 years, the nuclear power industry in Japan has introduced foreign technology, digested them in short period, and continued to research and develop domestic technology. Now, 95% of the machinery and equipments for nuclear power generation with light water reactors can be produced domestically, and some technologies are going to be exported. However, the nuclear power industry is still in severe environment. The progress of the development of nuclear power plants passed the periods of organizational preparation, the construction of research reactors, the import of foreign technologies and reactors for practical use, and the construction of domestically produced reactors for practical use. The supplying capacity of the nuclear power industry in Japan reached 6 units of 1,000 MW yearly, but in order to meet the long term plan of nuclear power generation, this capacity must be further enhanced. The problems in the promotion of domestic production are the establishment of independent technologies, the promotion of standardization, the strengthening of business basis, the upbringing of relating enterprises, and the acceleration of national projects. Since the energy crisis, the trend of filling up energy demand with nuclear power generation became conspicuous, but for the expansion of export, the problems of safety guarantee, nuclear fuel cycle, and financial measures must be solved with government aid. (Kako, I.)

  3. Study on the safety evaluation method development for D and D of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sin, S. W.; Kang, G. D.; Kim, H. S.; Son, J. G.; Choi, Y. J.; Lee, K. J.; Koh, E. O.; Kim, K. D.; Ha, J. H. [Korea Hydro and Nuclear Power Co. Ltd., Seoul (Korea, Republic of)

    2004-02-15

    The final goal of decommissioning of Nuclear Power Plant is to remove or to isolate the radioactivity existing at the nuclear facilities which are confronted with termination of operation. The regulatory guidance and technologies of decommissioning have been developed strategically in some advanced countries. They have already stepped into the application stage from the research and development stage. The contents and the scope of this study is as follows, preparation of standard evaluation plan and establishment of standard safety evaluation system for D and D.

  4. Study on the safety evaluation method development for D and D of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sin, S. W.; Kang, G. D.; Kim, H. S.; Son, J. G.; Choi, Y. J.; Lee, K. J.; Koh, E. O.; Kim, K. D.; Ha, J. H.

    2004-02-01

    The final goal of decommissioning of Nuclear Power Plant is to remove or to isolate the radioactivity existing at the nuclear facilities which are confronted with termination of operation. The regulatory guidance and technologies of decommissioning have been developed strategically in some advanced countries. They have already stepped into the application stage from the research and development stage. The contents and the scope of this study is as follows, preparation of standard evaluation plan and establishment of standard safety evaluation system for D and D

  5. Space nuclear power plant technology development philosophy for a ground engineering phase

    International Nuclear Information System (INIS)

    Buden, D.; Trapp, T.J.; Los Alamos National Lab., NM)

    1985-01-01

    The development of a space qualified nuclear power plant is proceeding from the technical assessment and advancement phase to the ground engineering phase. In this new phase, the selected concept will be matured by the completion of activities needed before protoflight units can be assembled and qualified for first flight applications. This paper addresses a possible philosophy to arrive at the activities to be performed during the ground engineering phase. The philosophy is derived from what we believe a potential user of nuclear power would like to see completed before commitment to a flight development phase. 5 references

  6. Space nuclear power plant technology development philosophy for a ground engineering phase

    International Nuclear Information System (INIS)

    Buden, D.; Trapp, T.J.

    1985-01-01

    The development of a space qualified nuclear power plant is proceeding from the Technical Assessment and Advancement Phase to the Ground Engineering Phase. In this new phase, the selected concept will be matured by the completion of activities needed before protoflight units can be assembled and qualified for first flight applications. This paper addresses a possible philosophy to arrive at the activities to be performed during the Ground Engineering Phase. The philosophy is derived from what we believe a potential user of nuclear power would like to see completed before commitment to a flight development phase

  7. Nuclear plant scram reduction

    International Nuclear Information System (INIS)

    Wiegle, H.R.

    1986-01-01

    The Nuclear Utility Management and Human Resources Committee (NUMARC) is a confederation of all 55 utilities with nuclear plants either in operation or under construction. NUMARC was formed in April 1984 by senior nuclear executives with hundreds of man-years of plant experience to improve (plant) performance and resolve NRC concerns. NUMARC has adopted 10 commitments in the areas of management, training, staffing and performance. One of these commitments is to strive to reduce automatic trips to 3 per year per unit for calendar year 1985 for plants in commercial operation greater than 3 years (with greater than 25% capacity factor). This goal applies to any unplanned automatic protection system trips at any time when the reactor is critical. Each utility has committed to develop methods to thoroughly evaluate all unplanned automatic trips to identify the root causes and formulate plans to correct the root causes thus reducing future unplanned scrams. As part of this program, the Institute of Nuclear Power Operations (INPO) collects and evaluates information on automatic reactor trips. It publishes the results of these evaluations to aid the industry to identify root causes and corrective actions

  8. Development of a remote controlled robot system for monitoring nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Woo, Hee Gon; Song, Myung Jae; Shin, Hyun Bum; Oh, Gil Hwan; Maeng, Sung Jun; Choi, Byung Jae; Chang, Tae Woo [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Lee, Bum Hee; Yoo, Jun; Choi, Myung Hwan; Go, Nak Yong; Lee, Kee Dong; Lee, Young Dae; Cho, Hae Kyeng; Nam, Yoon Suk [Electric and Science Research Center, (Korea, Republic of)

    1996-12-31

    It`s a final report of the development of remote controlled robot system for monitoring the facilities in nuclear power plant and contains as follows, -Studying the technologies in robot developments and analysing the requirements and working environments - Development of the test mobile robot system - Development of the mobile-robot - Development of the Mounted system on the Mobile robot - Development of the Monitoring system - Mobil-robot applications and future study. In this study we built the basic technologies and schemes for future robot developments and applications. (author). 20 refs., figs.

  9. The Planning of a Small Pilot Plant for Development Work on Aqueous Reprocessing of Nuclear Fuels

    Energy Technology Data Exchange (ETDEWEB)

    Sjoeborg, T U; Haeffner, E; Hultgren, Aa

    1963-10-15

    A shielded volume (42 m{sup 3}) in the hot laboratory at Kjeller, Norway, has been used for the installation of a small pilot plant intended for studies on nuclear fuel reprocessing. During the first period of operation (1963) a plutonium separation method (the Silex process) developed at AB Atomenergi will be studied. This document is a description of the project during the stage of technical planning and chemical process development.

  10. Development of web-based integrity evaluation system for primary components in a nuclear power plant

    International Nuclear Information System (INIS)

    Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J.; Choi, S.N.; Jang, K.S.; Hong, S.Y.

    2004-01-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

  11. Development of web-based integrity evaluation system for primary components in a nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.M.; Kim, J.C.; Choi, J.B.; Kim, Y.J. [SAFE Research Center, Sungkyunkwan Univ., Suwon (Korea); Choi, S.N.; Jang, K.S.; Hong, S.Y. [Korea Electronic Power Research Inst., Daejeon (Korea)

    2004-07-01

    A nuclear power plant is composed of a number of primary components. Maintaining the integrity of these components is one of the most critical issues in nuclear industry. In order to maintain the integrity of these primary components, a complicated procedure is required including periodical in-service inspection, failure assessment, fracture mechanics analysis, etc. Also, experts in different fields have to co-operate to resolve the integrity issues on the basis of inspection results. This integrity evaluation process usually takes long, and thus, is detrimental for the plant productivity. Therefore, an effective safety evaluation system is essential to manage integrity issues on a nuclear power plant. In this paper, a web-based integrity evaluation system for primary components in a nuclear power plant is proposed. The proposed system, which is named as WEBIES (web-based integrity evaluation system), has been developed in the form of 3-tier system architecture. The system consists of three servers; application program server, user interface program server and data warehouse server. The application program server includes the defect acceptance analysis module and the fracture mechanics analysis module which are programmed on the basis of ASME sec. XI, appendix A. The data warehouse server provides data for the integrity evaluation including material properties, geometry information, inspection data and stress data. The user interface program server provides information to all co- workers in the field of integrity evaluation. The developed system provides engineering knowledge-based information and concurrent and collaborative working environment through internet, and thus, is expected to raise the efficiency of integrity evaluation procedures on primary components of a nuclear power plant. (orig.)

  12. Developments in operator assistance techniques for nuclear power plant control and operation

    International Nuclear Information System (INIS)

    Poujol, A.; Papin, B.; Beltranda, G.; Soldermann, R.

    1989-01-01

    This paper describes an approach which has been developed in order to improve nuclear power plants control and monitoring in normal and abnormal situations. These developments take full advantage of the trend towards the computerization of control rooms in industrial continuous processes. This research program consists in a thorough exploration of different information processing techniques, ranking from the rather simple visual synthetization of informations on graphic displays to sophisticated Artificial Intelligence (AI) techniques. These techniques are put into application for the solving of man-machine interface problems in the different domains of plant operation

  13. Global development of advanced nuclear power plants, and related IAEA activities

    International Nuclear Information System (INIS)

    2006-09-01

    Renewed interest in the potential of nuclear energy to contribute to a sustainable worldwide energy mix is underlining the IAEA's statutory role in fostering the peaceful uses of nuclear energy, in particular the need for effective exchanges of information and collaborative research and technology development among Member States on advanced nuclear power technologies deployable in the near term as well as in the longer term. For applications in the medium to longer term, with rising expectations for the role of nuclear energy in the future, technological innovation has become a strong focus of nuclear power technology developments by many Member States. To meet Member States' needs, the IAEA conducts activities to foster information exchange and collaborative research and development in the area of advanced nuclear reactor technologies. These activities include coordination of collaborative research, organization of international information exchange, and analyses of globally available technical data and results, with a focus on reducing nuclear power plant capital costs and construction periods while further improving performance, safety and proliferation resistance. In other activities, evolutionary and innovative advances are catalyzed for all reactor lines such as advanced water cooled reactors, high temperature gas cooled reactors, liquid metal cooled reactors and accelerator driven systems, including small and medium sized reactors. In addition, there are activities related to other applications of nuclear energy such as seawater desalination, hydrogen production, and other process heat applications. This brochure summarizes the worldwide status and the activities related to advanced nuclear power technology development and related IAEA activities. It includes a list of the collaborative research and development projects conducted by the IAEA, as well as of the status reports and other publications produced

  14. The development of advanced robotics for the nuclear industry -The development of robotic system for the nuclear power plants-

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung Hoh; Kim, Chang Heui; Kim, Byung Soo; Hwang, Suk Yong; Kim, Kee Hoh; Lee, Jae Kyung; Lee, Yung Kwang; Suh, Yong Chil; Lee, Yong Bum; Kim, Woong Kee; Park, Soon Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The Advanced Robotics Department of the Korea Atomic Energy Research Institute (KAERI) is developing a Dexterous Manipulation System (DMS) and a teleoperated mobile robot, identified as KAEROT/ml, for use in nuclear power plants. The DMS is being developed for performing tasks such as the opening and closing of nozzle dam inside water chamber of steam generator. The DMS has two major subsystems; a master-slave 6 degrees of freedom (dof) manipulator and a support device. The master-slave arms are designed dissimilar kinematically and dynamically, and their functions are performed by a bilateral force-reflecting force control. The slave is a hydraulically powered arm with a 3 dof end effector, and is mounted on the top of the support device for nozzle dam operation. The support device guides the slave arm into the water chamber and supports it during its operation. The DMS can be operated either in teleoperated or supervisory control modes. The KAEROT/ml is designed to be used in emergency response applications such as monitoring and mapping radiation areas, handling radioactive materials and performing decontamination tasks. The KAEROT/ml equipped with four-omnidirectional planetary wheels has a 6 dof joint-controlled arm and is capable of ascending and descending stairs and navigating flat surface with zero turning radius. This report describes the mechanical design, features, modeling and control system of both the DMS and the KAEROT/ml. 209 figs, 49 pix, 69 tabs, 62 refs. (Author).

  15. The development of advanced robotics for the nuclear industry -The development of robotic system for the nuclear power plants-

    International Nuclear Information System (INIS)

    Kim, Seung Hoh; Kim, Chang Heui; Kim, Byung Soo; Hwang, Suk Yong; Kim, Kee Hoh; Lee, Jae Kyung; Lee, Yung Kwang; Suh, Yong Chil; Lee, Yong Bum; Kim, Woong Kee; Park, Soon Yong

    1995-07-01

    The Advanced Robotics Department of the Korea Atomic Energy Research Institute (KAERI) is developing a Dexterous Manipulation System (DMS) and a teleoperated mobile robot, identified as KAEROT/ml, for use in nuclear power plants. The DMS is being developed for performing tasks such as the opening and closing of nozzle dam inside water chamber of steam generator. The DMS has two major subsystems; a master-slave 6 degrees of freedom (dof) manipulator and a support device. The master-slave arms are designed dissimilar kinematically and dynamically, and their functions are performed by a bilateral force-reflecting force control. The slave is a hydraulically powered arm with a 3 dof end effector, and is mounted on the top of the support device for nozzle dam operation. The support device guides the slave arm into the water chamber and supports it during its operation. The DMS can be operated either in teleoperated or supervisory control modes. The KAEROT/ml is designed to be used in emergency response applications such as monitoring and mapping radiation areas, handling radioactive materials and performing decontamination tasks. The KAEROT/ml equipped with four-omnidirectional planetary wheels has a 6 dof joint-controlled arm and is capable of ascending and descending stairs and navigating flat surface with zero turning radius. This report describes the mechanical design, features, modeling and control system of both the DMS and the KAEROT/ml. 209 figs, 49 pix, 69 tabs, 62 refs. (Author)

  16. Development of a smart type motor operated valve for nuclear power plants

    Science.gov (United States)

    Kim, Chang-Hwoi; Park, Joo-Hyun; Lee, Dong-young; Koo, In-Soo

    2005-12-01

    In this paper, the design concept of the smart type motor operator valve for nuclear power plant was described. The development objective of the smart valve is to achieve superior accuracy, long-term reliability, and ease of use. In this reasons, developed smart valve has fieldbus communication such as deviceNet and Profibus-DP, auto-tuning PID controller, self-diagnostics, and on-line calibration capabilities. And also, to achieve pressure, temperature, and flow control with internal PID controller, the pressure sensor and transmitter were included in this valve. And, temperature and flow signal acquisition port was prepared. The developed smart valve will be performed equipment qualification test such as environment, EMI/EMC, and vibration in Korea Test Lab. And, the valve performance is tested in a test loop which is located in Seoul National University Lab. To apply nuclear power plant, the software is being developed according to software life cycle. The developed software is verified by independent software V and V team. It is expected that the smart valve can be applied to an existing NPPs for replacing or to a new nuclear power plants. The design and fabrication of smart valve is now being processed.

  17. Development of on-line operator aid system (OASYSTM) for nuclear power plants

    International Nuclear Information System (INIS)

    Kang, Ki Sig; Choi, Seong Soo; Kim, Han Gon; Chang, Soon Heung; Jeong, Hee Kyo; Yi, Chul Un

    1994-01-01

    In this paper, the development of On - line operator Aid SYStem (OASYS TM ) are discussed by focusing attention on the importance of the operator's role for nuclear power plants. The OASYS TM is under development to support the operator's decision making process and to enhance the safety of a nuclear power plant by providing the plant operators with timely and proper guideline during abnormal and emergency conditions. The OASYS TM has capabilities for the signal validation/management, the alarm processing, the failure diagnosis using abnormal operating procedures, and the dynamic tracking of emergency operating procedures using function restoration guidelines and optimal recovery guidelines with a series of complex logic steps covering a broad spectrum of event sequences. The proposed system is being implemented on a SUN-4/75 Workstation using C and Quintus TM prolog language. For verification studies a full-scope real-time simulator is being used. Test results show that the OASYS TM is capable of diagnosing a plant failure quickly and providing an operator guideline with fast response time. After verification the OASYS TM will be installed in the simulator II of Kori nuclear training center

  18. Development of Air Dispersion Modeling for Future Nuclear Plant in Malaysia

    International Nuclear Information System (INIS)

    Mohd Nahar Othman

    2011-01-01

    The impact development of Nuclear power plant in Malaysia, can be very negative to the nearby population, causing public restlessness and consequently affecting the image of the authorities in the countries. The precise source of the pollution from the nuclear power plant must be determined, pollution emission level and the meteorological conditions are needed to predict and established the ambient air to the save level at the perimeter fence of the plant and address it with respect to the radiological ambient standards. Upon modeling using an established package as well as site measurements, the radiological level at the perimeter fence of the plant must deduced and lower than the normal ambient level. Based on this issue, a modeling study was made in vicinity of Malaysian Nuclear Agency TRIGA reactor in the area of Bangi, Selangor to evaluate the possibility of movement of air around the area and their impact. This paper will address and discuss the modeling base on the data getting from Meteorological Department such as measurement of wind speed, temperature, humidity, ambient air radiological concentration and ect. The purpose of Air Dispersion Modeling is to establish the critical ambient emission level, as well as radiological modeling. The focus will be made on exploring the use of Ausplume modeling to develop correlations between the radiological concentrations, chemical compositions and ambient model for emission controls. (author)

  19. Development of regulatory requirements/guides for desalination unit coupled with nuclear plant

    International Nuclear Information System (INIS)

    Jo, Jong Chull; Yune, Young Gill; Kim, Woong Sik

    2005-10-01

    The basic design of System-integrated Modular Advanced Reactor (SMART), a small-to-medium sized integral type pressurized water reactor (PWR) with the capacity of 330MWth, has been developed in Korea. In order to demonstrate the safety and performance of the SMART design, 'Development Project of SMART-P (SMART-Pilot Plant)' has been being performed as one of the 'National Mid and Long-term Atomic Energy R and D Programs', which includes design, construction, and start-up operation of the SMART-P with the capacity of 65MWth, a 1/5 scaled-down design of the SMART. At the same time, a study on the development of regulatory requirements/guides for the desalination unit coupled with nuclear plant has been carried out by KINS in order to prepare for the forthcoming SMART-P licensing. The results of this study performed from August of 2002 to October of 2005 can be summarized as follows: (1) The general status of desalination technologies has been survey. (2) The design of the desalination plant coupled with the SMART-P has been investigated. (3) The regulatory requirements/guides relevant to a desalination unit coupled with a nuclear plant have been surveyed. (4) A direction on the development of domestic regulatory requirements/guides for a desalination unit has been established. (5) A draft of regulatory requirements/guides for a desalination unit has been developed. (6) Expert technical reviews have been performed for the draft regulatory requirements/guides for a desalination unit. The draft regulatory requirements/guides developed in this study will be finalized and can be applied directly to the licensing of the SMART-P and SMART. Furthermore, it will be also applied to the licensing of the desalination unit coupled with the nuclear plant

  20. Development of Web-Based Common Cause Failure (CCF) Database Module for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Hyun-Gyo; Hwang, Seok-Won; Shin, Tae-young

    2015-01-01

    Probabilistic safety assessment (PSA) has been used to identify risk vulnerabilities and derive the safety improvement measures from construction to operation stages of nuclear power plants. In addition, risk insights from PSA can be applied to improve the designs and operation requirements of plants. However, reliability analysis methods for quantitative PSA evaluation have essentially inherent uncertainties, and it may create a distorted risk profiles because of the differences among the PSA models, plant designs, and operation status. Therefore, it is important to ensure the quality of the PSA model so that analysts identify design vulnerabilities and utilize risk information. Especially, the common cause failure (CCF) has been pointed out as one of major issues to be able to cause the uncertainty related to the PSA analysis methods and data because CCF has a large influence on the PSA results. Organization for economic cooperation and development /nuclear energy agent (OECD/NEA) has implemented an international common cause failure data exchange (ICDE) project for the CCF quality assurance through the development of the detailed analysis methodologies and data sharing. However, Korea Hydro and Nuclear Power company (KHNP) does not have the basis for the data gathering and analysis for CCF analyses. In case of methodology, the Alpha Factor parameter estimation, which can analyze uncertainties and estimate an interface factor (Impact Vector) with an ease, is ready to be applied rather than the Multi Greek Letter (MGL) method. This article summarizes the development of the plant-specific CCF database (DB) module considering the raw data collection and the analysis procedure based on the CCF parameter calculation method of ICDE. Although the portion affected by CCF in the PSA model is quite a large, the development efforts of the tools to collect and analyze data were insufficient. Currently, KHNP intends to improve PSA quality and ensure CCF data reliability by

  1. Development of Web-Based Common Cause Failure (CCF) Database Module for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun-Gyo; Hwang, Seok-Won; Shin, Tae-young [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2015-05-15

    Probabilistic safety assessment (PSA) has been used to identify risk vulnerabilities and derive the safety improvement measures from construction to operation stages of nuclear power plants. In addition, risk insights from PSA can be applied to improve the designs and operation requirements of plants. However, reliability analysis methods for quantitative PSA evaluation have essentially inherent uncertainties, and it may create a distorted risk profiles because of the differences among the PSA models, plant designs, and operation status. Therefore, it is important to ensure the quality of the PSA model so that analysts identify design vulnerabilities and utilize risk information. Especially, the common cause failure (CCF) has been pointed out as one of major issues to be able to cause the uncertainty related to the PSA analysis methods and data because CCF has a large influence on the PSA results. Organization for economic cooperation and development /nuclear energy agent (OECD/NEA) has implemented an international common cause failure data exchange (ICDE) project for the CCF quality assurance through the development of the detailed analysis methodologies and data sharing. However, Korea Hydro and Nuclear Power company (KHNP) does not have the basis for the data gathering and analysis for CCF analyses. In case of methodology, the Alpha Factor parameter estimation, which can analyze uncertainties and estimate an interface factor (Impact Vector) with an ease, is ready to be applied rather than the Multi Greek Letter (MGL) method. This article summarizes the development of the plant-specific CCF database (DB) module considering the raw data collection and the analysis procedure based on the CCF parameter calculation method of ICDE. Although the portion affected by CCF in the PSA model is quite a large, the development efforts of the tools to collect and analyze data were insufficient. Currently, KHNP intends to improve PSA quality and ensure CCF data reliability by

  2. State of the art and further development of reinforced concrete wall cells for nuclear power plant construction

    International Nuclear Information System (INIS)

    Uhlemann, E.; Wartenberg, J.

    1985-01-01

    Reinforced concrete wall cells have been developed for nuclear power plant construction by the USSR and GDR. In this article, a new type of these cells, which will be used for constructing auxiliary equipment of the Stendal nuclear power plant, is described

  3. Development of the simulation system IMPACT for analysis of nuclear power plant severe accidents

    International Nuclear Information System (INIS)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

    1997-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system IMPACT for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT's distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data

  4. Good practices with respect to the development and use of nuclear power plant procedures

    International Nuclear Information System (INIS)

    1998-12-01

    This technical publication is intended to provide information to nuclear power plant managers concerning good practices in respect to the development and use of NPP procedures. It is a part of an ongoing project on nuclear power plant instrumentation and control and is based on lessons learned from NPPs and utilities. The overall objective of this project is to provide systematic guidance on current and emerging instrumentation and control technologies including: protection and automation systems, control rooms, operator support systems, training simulators and human factors. It is intended to be applicable to NPPs in Member States that are in operation, under construction or being commissioned. The good practices described are consistent with the IAEA Safety Series recommendations with respect to the development and use of NPP procedures including the four categories: administrative procedures, operating procedures (normal, alarm response, abnormal and emergency procedures), maintenance and technical support procedures, and testing and surveillance procedures

  5. Development of computer program for safety of nuclear power plant against tsunami

    International Nuclear Information System (INIS)

    Jin, S. B.; Choi, K. R.; Lee, S. K.; Cho, Y. S.

    2001-01-01

    The main objective of this study is the development of a computer program to check the safety of nuclear power plants along the coastline of the Korean Peninsula. The computer program describes the propagation and associated run-up process of tsunamis by solving linear and nonlinear shallow-water equations with finite difference methods. The computer program has been applied to several ideal and simplified problems. Obtained numerical solutions are compared to existing and available solutions and measurements. A very good agreement between numerical solutions and existing measurement is observed. The computer program developed in this study can be to check the safety analysis of nuclear power plants against tsunamis. The program can also be used to study the propagation of tsunamis for a long distance, and associated run-up and run-down process along a shoreline. Furthermore, the computer program can be used to provide the proper design criteria of coastal facilities and structures

  6. Development of a general-purpose mobile robot for use in nuclear power plants

    International Nuclear Information System (INIS)

    Martinez, A.; Yague, M.A.; Linares, F.

    1993-01-01

    In recent years, the Space Division of CONSTRUCCIONES AERONAUTICAS (CASA) and EQUIPOS NUCLEARES (ENSA) have participated in several national and international robotics programs in the respective space and nuclear areas. In mid-1992, they decided to jointly undertake the development of a mobile inspection and maintenance robot for Nuclear Power Plants. The success of such a multidisciplinary project was ensured by the way both companies complement each other and by their previous development. Work was begun on the feasibility study and specifications, for which technical meetings were held with personnel from the Medical and Health Physics Association of the utilities (AMYS) and several Nuclear Power Plants. The result of these conversations was a preliminary system design along with the specifications with which the system must comply. With these results, a report and job plan were prepared for construction of two prototypes and submitted to the INI (National Institute of Industry Shareholder of both CASA and ENSA), which decided to finance this second Phase of Development by charging it to the Group's Research Development Fund

  7. Beloyarsk Nuclear Power Plant

    International Nuclear Information System (INIS)

    1997-01-01

    The Beloyarsk Nuclear Power Plant (BNPP) is located in Zarechny, approximately 60 km east of Ekaterinberg along the Trans-Siberian Highway. Zarechny, a small city of approximately 30,000 residents, was built to support BNPP operations. It is a closed city to unescorted visitors. Residents must show identification for entry. BNPP is one of the first and oldest commercial nuclear power plants in Russia and began operations in 1964. As for most nuclear power plants in the Russian Federation, BNPP is operated by Rosenergoatom, which is subordinated to the Ministry of Atomic Energy of the Russian Federation (Minatom). BNPP is the site of three nuclear reactors, Units 1, 2, and 3. Units 1 and 2, which have been shut-down and defueled, were graphite moderated reactors. The units were shut-down in 1981 and 1989. Unit 3, a BN-600 reactor, is a 600 MW(electric) sodium-cooled fast breeder reactor. Unit 3 went on-line in April 1980 and produces electric power which is fed into a distribution grid and thermal power which provides heat to Zarechny. The paper also discusses the SF NIKIET, the Sverdiovsk Branch of NIKIET, Moscow, which is the research and development branch of the parent NIKEIT and is primarily a design institute responsible for reactor design. Central to its operations is a 15 megawatt IVV research reactor. The paper discusses general security and fissile material control and accountability at these two facilities

  8. Model developments for quantitative estimates of the benefits of the signals on nuclear power plant availability and economics

    International Nuclear Information System (INIS)

    Seong, Poong Hyun

    1993-01-01

    A novel framework for quantitative estimates of the benefits of signals on nuclear power plant availability and economics has been developed in this work. The models developed in this work quantify how the perfect signals affect the human operator's success in restoring the power plant to the desired state when it enters undesirable transients. Also, the models quantify the economic benefits of these perfect signals. The models have been applied to the condensate feedwater system of the nuclear power plant for demonstration. (Author)

  9. Development of wall thinning screening system and its application to a commercial nuclear power plant

    International Nuclear Information System (INIS)

    Ryu, Kyung Ha; Hwang, Il Soon; Kim, Ji Hyun

    2013-01-01

    Highlights: • Wall loss screening system (WalSS) has been developed based on ES-DCPD method. • Screening criteria was established based on the thinning of the actual shape that occur in the power plant. • With the criteria, the WalSS gives priority of the need for inspection. • This technique was successfully applied to commercial nuclear power plant. - Abstract: A new non-destructive evaluation (NDE) method has been developed for metal pipes for the detection wall thinning. The method has been showed to be suitable for applications to electric power generation plants where flow accelerated corrosion (FAC) of carbon steel piping is a significant cause of increased maintenance and plant personnel casualty. The wall thinning screening system (WalSS) was developed in two major phases. In the first phase, the equipotential switching direct current potential drop (ES-DCPD) method was developed for piping wall (Ryu et al., 2008a, 2010). In the second phase, in this paper, a quantitative detection criteria was developed. The relative ES-DCPD change of 3.8% has been defined as the screening criteria for wall thinning schematization. This criteria means that the component with measured ES-DCPD change greater than 3.8% is called for a more comprehensive examination. In the criteria development, all variables were taken into consideration based on commercial plant piping inspection data such as initial thickness distributions, wall thinning shape and nominal thickness. The developed WalSS based on ES-DCPD was applied to a moisture separator reheater (MSR) drain line of a commercial nuclear power plant (NPP) during a scheduled overhaul. The measured ES-DCPD change was 2.16%, which is lower than the ES-DCPD criteria, identifying the pipe having adequate wall thickness. This is confirmed by site thickness inspection using ultrasonic technique (UT)

  10. Nuclear power plants

    International Nuclear Information System (INIS)

    Margulova, T.Ch.

    1976-01-01

    The textbook focuses on the technology and the operating characteristics of nuclear power plants equiped with pressurized water or boiling water reactors, which are in operation all over the world at present. The following topics are dealt with in relation to the complete plant and to economics: distribution and consumption of electric and thermal energy, types and equipment of nuclear power plants, chemical processes and material balance, economical characteristics concerning heat and energy, regenerative preheating of feed water, degassing and condenser systems, water supply, evaporators, district heating systems, steam generating systems and turbines, coolant loops and pipes, plant siting, ventilation and decontamination systems, reactor operation and management, heat transfer including its calculation, design of reactor buildings, and nuclear power plants with gas or sodium cooled reactors. Numerous technical data of modern Soviet nuclear power plants are included. The book is of interest to graduate and post-graduate students in the field of nuclear engineering as well as to nuclear engineers

  11. Analysis of safety information for nuclear power plants and development of source term estimation program

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Choi, Seong Soo; Park, Jin Hee

    1999-12-01

    Current CARE(Computerized Advisory System for Radiological Emergency) in KINS(Korea Institute of Nuclear Safety) has no STES(Source Term Estimation System) which links between SIDS(Safety Information Display System) and FADAS(Following Accident Dose Assessment System). So in this study, STES is under development. STES system is the system that estimates the source term based on the safety information provided by SIDS. Estimated source term is given to FADAS as an input for estimation of environmental effect of radiation. Through this first year project STES for the Kori 3,4 and Younggwang 1,2 has been developed. Since there is no CARE for Wolsong(PHWR) plants yet, CARE for Wolsong is under construction. The safety parameters are selected and the safety information display screens and the alarm logic for plant status change are developed for Wolsong Unit 2 based on the design documents for CANDU plants

  12. Material and welding development of anchor plates to build nuclear power plant by blue arc process

    International Nuclear Information System (INIS)

    Gibelli, C.E.

    1986-01-01

    To build nuclear power plants, anchor plates are plenty used. These anchor plates serve as a system with the purpose to fix many heavy components or a simple stair. Considering the necessity of element fabrication fastly, with reasonable economy and quality, the arc study welding process (blue arc) was used. A special development of the material concept as well as a welding procedure and a subsuppliers qualification of the raw material was necessary. (Author) [pt

  13. Development of technology for next generation reactor - Research of evaluation technology for nuclear power plant -

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyun; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1993-09-01

    For development of next generation reactor, a project for evaluation technology for nuclear power plant is performed. Evaluation technology is essential to next generation reactor for reactor safety and system analysis. For design concept, detailed evaluation technologies are studied as follows: evaluation of safety margin, evaluation of safety facilities, evaluation of measurement and control technology; man-machine interface. Especially for thermal efficiency, thermal properties and chemical composition of inconel 690 tube, instead of inconel 600 tube, are measured for steam generator. (Author).

  14. Development of a Flexible Computerized Management Infrastructure for a Commercial Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ali, Syed Firasat; Hajek, Brian K.; Usman, Shoaib

    2006-01-01

    The report emphasizes smooth transition from paper-based procedure systems (PBPSs) to computer-based procedure systems (CBPSs) for the existing commercial nuclear power plants in the U.S. The expected advantages and of the transition are mentioned including continued, safe and efficient operation of the plants under their recently acquired or desired extended licenses. The report proposes a three-stage survey to aid in developing a national strategic plan for the transition from PBPSs to CBPSs. It also includes a comprehensive questionnaire that can be readily used for the first stage of the suggested survey

  15. Development of the on-site power supply in German nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Simon, M. [Gesellschaft fuer Reaktorsicherheit - GRS mbH, Schwertnergasse 1, D-5000 Koeln 1, Cologne (Germany)

    1986-02-15

    The design of the on-site power supply is different in German Nuclear Power Plants, depending on age and size of the plant. The cause for this is the evolution of the safety requirements. The general development of the design of safety Systems, which resulted in a strict separation of redundant trains is also reflected in the design of the emergency power system and even the complete on-site power supply System. This will be demonstrated by different examples. The advantages of this design with respect to the availability of on-site power will be explained and verified by means of operating experience. (author)

  16. Development of a Flexible Computerized Management Infrastructure for a Commercial Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Ali, Syed Firasat; Hajek, Brian K.; Usman, Shoaib

    2006-05-01

    The report emphasizes smooth transition from paper-based procedure systems (PBPSs) to computer-based procedure systems (CBPSs) for the existing commercial nuclear power plants in the U.S. The expected advantages and of the transition are mentioned including continued, safe and efficient operation of the plants under their recently acquired or desired extended licenses. The report proposes a three-stage survey to aid in developing a national strategic plan for the transition from PBPSs to CBPSs. It also includes a comprehensive questionnaire that can be readily used for the first stage of the suggested survey.

  17. The IAEA's activities in safeguarding nuclear materials and in developing internationally acceptable safety codes and guides for nuclear power plants

    International Nuclear Information System (INIS)

    Rometsch, Rudolf; Specter, Herschel

    1977-01-01

    Promoting the peaceful use of nuclear energy and aiming at the international sharing of its benefits are objectives that guide the activities of the Agency. But this promotional work is carried out on condition that security and safety are provided for. All Agency assistance involving nuclear facilities will be subjected to standards of safety or other standards, which are proposed by a State the Agency finds essentially equivalent. Safeguards are always applied on the basis of agreement. States party to NPT are obligated to negotiate and conclude with the Agency agreements which cover all their peaceful nuclear activities. Safeguards agreements concluded outside NPT are applied to specific supplies of facilities, equipment and material. To assist countries in laying down their nuclear safety regulations the Agency's program for the developing of codesof practice and safety guides for nuclear power plants draws up guidelines for governmental organizations, siting, design, operation and quality assurance. Codes are the fundamental documents laying down the objectives of each field of nuclear safety

  18. Development of reliability-based safety enhancement technology; development of organization concept model in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Hyun; Kim, Ju Youl; Kim, Yoon Ik; Yang, Hui Chang; Lee, Yong Sik; Kim, Se Hyung [Seoul National University, Seoul (Korea)

    2002-03-01

    The influences of organizational factors on safety of nuclear power plants are mentioned in the early 1970s and noticed after being focused on in the accident report of TMI in 1979. These needs let us implement this research and the purposes of this research are to assess the organizational influences and to develop the organizational conceptual model to establish the basis of identifying the organizational factors, using this model to contribute to enhance safety and economics in nuclear power plants. Eventually research on the organizational influences is expected to have two effects, which are to improve safety through identifying potential causes of accidents and to elevate economics as a new approach to more efficient operation of nuclear power plants. In this study, recent studies were surveyed on the organizational conceptual model, the identification of organizational factors, assessment of organizational influences and evaluation methods of organizational factors and organizational influences among the overseas and domestic researches. In addition specific characteristics of domestic nuclear power plants were tried to identify through plant visit and an evaluation method of organizational influences on component maintenance and human performance were developed and presented. 71 refs., 40 figs., 18 tabs. (Author)

  19. Development of high-reliability control system for nuclear power plants

    International Nuclear Information System (INIS)

    Asami, K.; Yanai, K.; Hirose, H.; Ito, T.

    1983-01-01

    In Japan, many nuclear power generating plants are in operation and under construction. There is a general awareness of the problems in connection with nuclear power generation and strong emphasis is put on achieving highly reliable operation of nuclear power plants. Hitachi has developed a new high-reliability control system. NURECS-3000 (NUclear Power Plant High-REliability Control System), which is applied to the main control systems, such as the reactor feedwater control system, the reactor recirculation control system and the main turbine control system. The NURECS-3000 system was designed taking into account the fact that there will be failures, but the aim is for the system to continue to function correctly; it is therefore a fault-tolerant system. It has redundant components which can be completely isolated from each other in order to prevent fault propagation. The system has a hierarchical configuration, with a main controller, consisting of a triplex microcomputer system, and sub-loop controllers. Special care was taken to ensure the independence of these subsystems. Since most of the redundant system failures are caused by common-mode failures and the reliability of redundant systems depends on the reliability of the common-mode parts, the aim was to minimize these parts. (author)

  20. Proceedings of the NEA workshop on development priorities for NDE of concrete structures in nuclear plants

    International Nuclear Information System (INIS)

    1998-01-01

    The first session's objectives of this conference were to identify the perspectives of national regulators and plant operators on what is required of NDE. The second session objectives were to provide opportunity for NDE practitioners to share experience and views on the status of development of key NDE techniques: tomographic imaging for investigation of concrete structures, four examples of modern NDE techniques applied to the investigation of nuclear and non-nuclear concrete structures and a vision of future improvements, investigating concrete structures by 3D Radar imaging and imaging using mechanical impact, synopsis NDT of concrete using ultrasonics and radar. The third session objectives were to prioritise development of NDE techniques for safety related concrete structures in nuclear installations: key conclusions from earlier sessions, proposed priorities and next steps

  1. Development of small size wall decontamination robot systems in nuclear power plants

    International Nuclear Information System (INIS)

    Fujita, Tsuneaki; Takahashi, Tsuyosi

    2004-01-01

    This paper describes the development of wall decontamination robot systems for nuclear power plants. In nuclear power plants, it is required to reduce maintenance costs, including annual inspection, repairs and so on. Most of such maintenance activities are actually performed after decontamination processes are completed. In particular, the decontamination process of reactor wells is very important for reducing the radiation exposure of human workers. In the past, decontamination of reactor wells used to be done by extra large machine and tools, which caused long working hours and tiresome works. It was one of the reasons maintenance costs couldn't have been easily reduced. There are narrow spaces in the reactor wells that have to be decontaminated by human workers. In order to minimize the radiation exposure to humans, wall decontamination robot systems have been developed. The decontamination robots have rolled brushes and suction mechanisms and are capable of removing contaminants attached to the wall surface of the reactor wells. By making the robots smaller, it is possible to work in narrower spaces. In this paper, the effectiveness of decontamination by the developed robots is shown through experiments in the actual nuclear power plants. (author)

  2. Nuclear power plants: recent developments in Brazil relating to the legal aspect of installation - erection - operation

    International Nuclear Information System (INIS)

    Lima Moreira, Y.M. de.

    1981-10-01

    A Federal statutory system governs the setting up and operation of nuclear power plants in Brazil. This paper describes the general regulatory framework for electric utilities and in particular analyses the licensing procedure for nuclear installations. (NEA) [fr

  3. The critical issue of nuclear power plant safety in developing countries

    International Nuclear Information System (INIS)

    Rosen, M.

    1977-01-01

    A little more than a decade from now, large commercial nuclear power facilities will be in operation in almost 40 countries, of which approximately one-half are presently considered industrially less developed. Ambitious nuclear programmes coupled with minimal and frequently under-staffed regulatory and utility organizations are only one aspect of the difficulties related to the safety of nuclear plants that face these developing countries. Inherent problems of meeting current safety standards and requirements for the significantly non-standard nuclear power plant exports can be compounded by financial considerations that may lead to purchases of reactors of various types, from more than one supplier country and with different safety standards and requirements. An examination of these issues points to the necessity and opportunity for effective action which could include provision for adequate funding for safety considerations in the purchase contract, and for sufficient regulatory assistance and training from the developed countries. The article will introduce the topic, discuss specific examples, and offer some suggestions. (author)

  4. Nuclear power plant outages

    International Nuclear Information System (INIS)

    1998-01-01

    The Finnish Radiation and Nuclear Safety Authority (STUK) controls nuclear power plant safety in Finland. In addition to controlling the design, construction and operation of nuclear power plants, STUK also controls refuelling and repair outages at the plants. According to section 9 of the Nuclear Energy Act (990/87), it shall be the licence-holder's obligation to ensure the safety of the use of nuclear energy. Requirements applicable to the licence-holder as regards the assurance of outage safety are presented in this guide. STUK's regulatory control activities pertaining to outages are also described

  5. Integrated Nuclear Recycle Plant

    International Nuclear Information System (INIS)

    Patodi, Anuj; Parashar, Abhishek; Samadhiya, Akshay K.; Ray, Saheli; Dey, Mitun; Singh, K.K.

    2017-01-01

    Nuclear Recycle Board (NRB), Tarapur proposes to set up an 'Integrated Nuclear Recycle Plant' at Tarapur. This will be located in the premises of BARC facilities. The project location is at coastal town of Tarapur, 130 Km north of Mumbai. Project area cover of INRP is around 80 hectares. The plant will be designed to process spent fuel received from Pressurized Heavy Water Reactors (PHWRs). This is the first large scale integrated plant of the country. INRP will process spent fuel obtained from indigenous nuclear power plants and perform left over nuclear waste disposal

  6. Arguments for the installation of small LWR nuclear power plants in developing countries

    International Nuclear Information System (INIS)

    Lezenik, B.

    1977-01-01

    soft- and hardware according to the latest state of technical developement, different procedures of awarding are selected, depending on the nuclear experience of the electric utility. An analysis and comparison show the advantages and disadvantages of the usual awarding procedures and possibilities of having the national industry take part in the construction of a plant are discussed

  7. Development of remote repair robots for dissolvers in nuclear fuel reprocessing plants

    International Nuclear Information System (INIS)

    Sugiyama, Sen; Kunikata, Michio; Kawamura, Hironobu.

    1985-01-01

    For nuclear facilities, various types of remote maintenance and inspection devices have been developed to reduce radiation exposure to workers, save labor, and improve the operating rate of the plant. Existing robot technology, however, could not be employed when we were recently called upon to inspect and repair pinhole defects which had occurred in the spent fuel dissolvers of the Power Reactor and Nuclear Fuel Development Corporation's Tokai Reprocessing Plant, because the work had to be done in an extremely radioactive environment, conditions too extreme for conventional robots. For this reason, we developed highly radiation-resistant repair robots capable of fully remote-controlled operation inside the barrels of the dissolvers, which have the inconvenient shape of 270 mm inside diameter and 6 m length. The process for developing the six different repair robots and the their functions are described in this paper. This development was sponsored by the Power Reactor and Nuclear Fuel Development Corporation (PNC) under contract with Hitachi, Ltd. (author)

  8. Development of intelligent database program for PSI/ISI data management of nuclear power plant

    International Nuclear Information System (INIS)

    Um, Byong Guk; Park, Un Su; Park, Ik Keun; Park, Yun Won; Kang, Suk Chul

    1998-01-01

    An intelligent database program has been developed under fully compatible with windows 95 for the construction of total support system and the effective management of Pre-/In-Service Inspection data. Using the database program, it can be executed the analysis and multi-dimensional evaluation of the defects detected during PSI/ISI in the pipe and the pressure vessel of the nuclear power plants. And also it can be used to investigate the NDE data inspected repetitively and the contents of treatment, and to offer the fundamental data for application of evaluation data related to Fracture Mechanics Analysis(FMA). Furthermore, the PSI/ISI database loads and material properties can be utilized to secure the higher degree of safety, integrity, reliability, and life-prediction of components and systems in nuclear power plant.

  9. Development of a daily maintenance work management system for nuclear power plants

    International Nuclear Information System (INIS)

    Yoshimura, Sadanori; Maita, Keikichi; Ogawa, Koutaro; Tabata, Nobuyuki; Shibuya, Shinya

    1999-01-01

    Maintenance work plays an important role in keeping the safety and reliability of nuclear power plants. Especially, with the recent trends of reducing the outage length, safe and reliable execution of maintenance work is more earnestly required than ever. Taking this situation into account, a daily maintenance work management system has been developed to support the execution of maintenance work for nuclear power plants. This system supports maintenance personnel in writing and approving the daily work instruction sheet and reduces their workload by providing predetermined lists of items necessary for filling up the sheet and by applying an electronic approval procedure. It also provides information through a computer network to improve the communication among maintenance personnel. For work steps to be treated in the daily work instruction sheet, an analysis was made to identify potential human errors and related counter measures, which were then linked to safety and quality assurance instructions of the sheet. Based on this information, the system automatically presents the safety and quality assurance instructions related to the work content listed in the sheet to improve the appropriateness of the instructions. A prototype of the daily maintenance work management system was developed and a trial use was made for actual periodic maintenance work at a nuclear power plant. This experience shows the usefulness of the system in reducing the workload of maintenance personnel and in providing safer work instructions. (author)

  10. Developing a taxonomy of coordination behaviours in nuclear power plant control rooms during emergencies.

    Science.gov (United States)

    Wang, Dunxing; Gao, Qin; Li, Zhizhong; Song, Fei; Ma, Liang

    2017-12-01

    This study aims to develop a taxonomy of coordination behaviours during emergencies in nuclear power plants (NPPs). We summarised basic coordination behaviours from literature in aviation, health care and nuclear field and identified coordination behaviours specific to the nuclear domain by interviewing and surveying control crew operators. The established taxonomy includes 7 workflow stages and 24 basic coordination behaviours. To evaluate the reliability and feasibility of the taxonomy, we analysed 12 videos of operators' training sessions by coding coordination behaviours with the taxonomy and the inter-rater reliability was acceptable. Further analysis of the frequency, the duration and the direction of the coordination behaviours revealed four coordination problems. This taxonomy provides a foundation of systematic observation of coordination behaviours among NPP crews, advances researchers' understanding of the coordination mechanism during emergencies in NPPs and facilitate the possibility to deepen the understanding of the relationships between coordination behaviours and team performance. Practitioner Summary: A taxonomy of coordination behaviours during emergencies in nuclear power plants was developed. Reliability and feasibility of the taxonomy was verified through the analysis of 12 training sessions. The taxonomy can serve as an observation system for analysis of coordination behaviours and help to identify coordination problems of control crews.

  11. Development and evaluation of tracking method for augmented reality system for nuclear power plant maintenance support

    International Nuclear Information System (INIS)

    Ishii, Hirotake; Bian, Zhiqiang; Sekiyama, Tomoki; Shimoda, Hiroshi; Yoshikawa, Hidekazu; Izumi, Masanori; Kanehira, Yoshiki; Morishita, Yoshitsugu

    2007-01-01

    This study aims at developing an augmented reality system to support maintenance work of nuclear power plants. An accurate and wide-range tracking method is required as a key technology in order to realize the system. In this study, a new tracking method using multi-camera and gyro sensor has been developed in order to enlarge the area where the tracking is available with limited number of markers. Experimental evaluation result shows that the area where the developed method can cover is about 3 times larger than the method using only single camera. (author)

  12. Development and evaluation of tracking method for augmented reality system for nuclear power plant maintenance support

    International Nuclear Information System (INIS)

    Ishii, Hirotake; Bian, Zhiqiang; Sekiyama, Tomoki; Shimoda, Hiroshi; Yoshikawa, Hidekazu; Izumi, Masanori; Kanehira, Yoshiki; Morishita, Yoshitsugu

    2006-01-01

    This study aims at developing an augmented reality system to support maintenance work of nuclear power plants. An accurate and wide-range tracking method is required as a key technology in order to realize the system. In this study, a new tracking method using multi-camera and gyro sensor has been developed in order to enlarge the area where the tracking is available with limited number of markers. Experimental evaluation result shows that the area where the developed method can cover is about 3 times larger than the method using single camera. (author)

  13. Development of module-based simulation system for nuclear power plant

    International Nuclear Information System (INIS)

    Yoshikawa, H.

    1990-01-01

    Module-based Simulation System (MSS) has been developed to realize a new software environment enabling versatile dynamic simulation of a complex nuclear power plant system flexibly. Described in the paper are (i) fundamental methods utilized in MMS and its software systemization, (ii) development of human interface system to help users in generating integrated simulation programs automatically, and (iii) development of an intelligent user support system for helping users in the two phases of automatical semantic diagnosis and consultation to automatic input data setup for the MSS-generated programs

  14. Use of Pilot Plants for Developing Used Nuclear Fuel Recycling Facilities

    Energy Technology Data Exchange (ETDEWEB)

    Phillips, Chris; Arm, Stuart [EnergySolutions LLC (United States); Banfield, Zara; Jeapes, Andrew; Taylor, Richard [National Nuclear Laboratory (United Kingdom)

    2009-06-15

    EnergySolutions and its teaming partners are working with the US Department of Energy (DOE) to develop processes, equipment and facilities for recycling used nuclear fuel (UNF). Recycling significantly reduces the volume of wastes that ultimately will be consigned to the National Geologic Repository, enables the re-use in new fuel of the valuable uranium and plutonium in the UNF, and allows the long-lived minor actinides to be treated separately so they do not become long term heat emitters in the Repository. A major requirement of any new UNF recycling facility is that pure plutonium is not separated anywhere in the process, so as to reduce the nuclear proliferation attractiveness of the facility. EnergySolutions and its team partner the UK National Nuclear Laboratory (NNL) have developed the NUEX process to achieve this and to handle appropriately the treatment of other species such as krypton, tritium, neptunium and technetium. NUEX is based on existing successful commercial UNF recycling processes deployed in the UK, France and imminently in Japan, but with a range of modifications to the flowsheet to keep some uranium with the plutonium at all times and to minimize aerial and liquid radioactive discharges. NNL's long-term experience in developing the recycling and associated facilities at the Sellafield site in the UK, and its current duties to support technically the operation of the Thermal Oxide Reprocessing Plant (THORP) at Sellafield provides essential input to the design of the US NUEX-based facility. Development work for THORP and other first-of-kind nuclear plants employed miniature scale fully radioactive through large scale inactive pilot plants. The sequence of development work that we have found most successful is to (i) perform initial process development at small (typically 1/5000) scale in gloveboxes using trace active materials, (ii) demonstrate the processes at the same small scale with actual irradiated fuel in hot cells and (iii

  15. Elecnuc. Nuclear power plants worldwide

    International Nuclear Information System (INIS)

    1998-01-01

    This small folder presents a digest of some useful information concerning the nuclear power plants worldwide and the situation of nuclear industry at the end of 1997: power production of nuclear origin, distribution of reactor types, number of installed units, evolution and prediction of reactor orders, connections to the grid and decommissioning, worldwide development of nuclear power, evolution of power production of nuclear origin, the installed power per reactor type, market shares and exports of the main nuclear engineering companies, power plants constructions and orders situation, evolution of reactors performances during the last 10 years, know-how and development of nuclear safety, the remarkable facts of 1997, the future of nuclear power and the energy policy trends. (J.S.)

  16. Development of experimental method to simulate the corrosion products in the primary system of nuclear power plant

    International Nuclear Information System (INIS)

    Kim, Sang Hyun; Kim, In Sup; Jang, Chang Heui

    2005-01-01

    Corrosion products are recognized as one of the major sources of occupational radiation exposure for nuclear power plant workers. Numerous studies have been conducted on the primary water chemistry to reduce the amount of crud in the primary circuit to avoid the radioactivity build-up in the plant. However, experiments with crud are restricted in laboratory because the crud is highly radioactive material. The objective of this study is to develop the simulating method of corrosion product in nuclear power plant

  17. Development of a Leading Performance Indicator from Operational Experience and Resilience in a Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Pamela F. Nelson

    2016-02-01

    Full Text Available The development of operational performance indicators is of utmost importance for nuclear power plants, since they measure, track, and trend plant operation. Leading indicators are ideal for reducing the likelihood of consequential events. This paper describes the operational data analysis of the information contained in the Corrective Action Program. The methodology considers human error and organizational factors because of their large contribution to consequential events. The results include a tool developed from the data to be used for the identification, prediction, and reduction of the likelihood of significant consequential events. This tool is based on the resilience curve that was built from the plant's operational data. The stress is described by the number of unresolved condition reports. The strain is represented by the number of preventive maintenance tasks and other periodic work activities (i.e., baseline activities, as well as, closing open corrective actions assigned to different departments to resolve the condition reports (i.e., corrective action workload. Beyond the identified resilience threshold, the stress exceeds the station's ability to operate successfully and there is an increased likelihood that a consequential event will occur. A performance indicator is proposed to reduce the likelihood of consequential events at nuclear power plants.

  18. Development of a seismic damage assessment program for nuclear power plant structures

    Energy Technology Data Exchange (ETDEWEB)

    Koh, Hyun Moo; Cho, Yang Heui; Shin, Hyun Mok [Seoul National Univ., Seoul (Korea, Republic of)] (and others)

    2001-12-15

    The most part of the nuclear power plants operating currently in Korea are more than 20 years old and obviously we cannot pretend that their original performance is actually maintained. In addition, earthquake occurrences show an increasing trend all over the world, and Korea can no more be considered as a zone safe from earthquake. Therefore, need is to guarantee the safety of these power plant structures against seismic accident, to decide to maintain them operational and to obtain data relative to maintenance/repair. Such objectives can be reached by damage assessment using inelastic seismic analysis considering aging degradation. It appears to be more important particularly for the structure enclosing the nuclear reactor that must absolutely protect against any radioactive leakage. Actually, the tendency of the technical world, led by the OECD/NEA, BNL in the United States, CEA in France and IAEA, is to develop researches or programs to assess the seismic safety considering aging degradation of operating nuclear power plants. Regard to the above-mentioned international technical trend, a technology to establish inelastic seismic analysis considering aging degradation so as to assess damage level and seismic safety margin appears to be necessary. Damage assessment and prediction system to grasp in real-time the actual seismic resistance capacity and damage level by 3-dimensional graphic representations are also required.

  19. Development of a seismic damage assessment program for nuclear power plant structures

    Energy Technology Data Exchange (ETDEWEB)

    Koh, Hyun Moo; Cho, Ho Hyun; Cho, Yang Hui [Seoul National Univ., Seoul (Korea, Republic of)] (and others)

    2000-12-15

    Some of nuclear power plants operating currently in Korea have been passed about 20 years after construction. Moreover, in the case of KORI I the service year is over 20 years, so their abilities are different from initial abilities. Also, earthquake outbreak increase, our country is not safe area for earthquake. Therefore, need is to guarantee the safety of these power plant structures against seismic accident, to decide to maintain them operational and to obtain data relative to maintenance/repair. Such objectives can be reached by damage assessment using inelastic seismic analysis considering aging degradation. It appears to be more important particularly for the structure enclosing the nuclear reactor that must absolutely protect against any radioactive leakage. Actually, the tendency of the technical world, led by the OECD/NEA, BNL in the United States, CEA in France and IAEA, is to develop researches or programs to assess the seismic safety considering aging degradation of operating nuclear power plants. Regard to the above-mentioned international technical trend, a technology to establish inelastic seismic analysis considering aging degradation so as to assess damage level and seismic safety margin appears to be necessary. Damage assessment and prediction system to grasp in real-time the actual seismic resistance capacity and damage level by 3-dimensional graphic representations are also required.

  20. The historical development of criteria on the safety of nuclear power plants

    International Nuclear Information System (INIS)

    Talarek, H.D.

    1976-01-01

    Starting from the lump-sum criterion of distance, the criterion of MCA has developed based on the idea of limiting accident consequences with the object in mind of guaranteeing the protection of the population in case of the maximum credible accident in its absolute sense. This claim has proved to be indefensible for the utilization and design of the MCA criterion on nuclear power plants of larger capacities on sites near to conurbation centres. Within the concept of design basis accidents this is not claimed, an imperative formulation was found instead concerning the sort of accident to be considered, including possible demands on its safety design. Using probability methods has lead to a blurring of the relationship with the technological system of nuclear power plants for formulating a criterion for the advantage of risk quantification. If the probability method is to be applied in the licensing procedure concerning nuclear power plants it will be one of the tasks of future safety research to elaborate this relationship clearly. (orig.) [de

  1. Development of Open Test-bed for Autonomous Operation in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Seungmin; Heo, Gyunyoung

    2017-01-01

    Nuclear power plants also recognize the need for automation. However, it is dangerous technology to have a significant impact on human society. In addition, due to the uncertain legal responsibility for autonomous operation, the application and development speed of nuclear energy related automation technology will be significantly decrease compared to other industries. It is argued that the application of AI and automation technology to power plants should not be prematurely applied or not based on the principle of applying proven technology since nuclear power plants are the highest level security operated facilities. As described above, the overall algorithm of the Test Bed is an autonomous operation algorithm (rulebased algorithm, learning-based algorithm, semiautonomous operation algorithm) to judge the entry condition of the procedure through condition monitoring and to enter the appropriate operating procedure. In order to make a test bed, the investigation for the heuristic part of the existing procedures and the heuristic part from the circumstance which is not specified in the procedure is needed. In the learning based and semi-autonomous operation algorithms, using MARS to extract its operating data and operational logs and try out various diagnostic algorithms as described above. Through the completion of these future tasks, the test bed which can compared with actual operators will be constructed and that it will be able to check its effectiveness by improving competitively with other research teams through the characteristics of shared platform.

  2. The design, development and operation of a compact nuclear power plant simulator

    International Nuclear Information System (INIS)

    Lynch, M.F.

    1987-01-01

    This paper discusses the philosophy and technological considerations necessary for constructing and utilizing a plant specific compact nuclear power plant simulator, how it compares to full scope replica simulators, engineering simulators, part task simulators and basic principles training simulators. Included in this discussion are the design process, scope of simulation, the manufacturing process, test programs and experiences with operator training. Items addressed include the applicability and use of a compact simulator, how well it reproduces the actual reference plant, how well the transferral of knowledge is accomplished and what financial considerations need to be evaluated. This paper tries to provide the details on just how this type of machine was designed and developed by Westinghouse for the Swiss Utility, Nordostschweizerische Kraftwerke (NOK) AG

  3. Development of a Multi-Channel Ultrasonic Testing System for Automated Ultrasonic Pipe Inspection of Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lee, Hee Jong; Cho, Chan Hee; Cho, Hyun Joon

    2009-01-01

    Currently almost all in-service-inspection techniques, applied in domestic nuclear power plants, are partial to field inspection technique. These kinds of techniques are related to managing nuclear power plants by the operation of foreign-produced inspection devices. There have been so many needs for development of native in-service-inspection device because there is no native diagnosis device for nuclear power plant inspection yet in Korea. In this research, we developed several core techniques to make an automated ultrasonic pipe inspection system for nuclear power plants. A high performance multi-channel ultrasonic pulser/receiver module, an A/D converter module and a digital main CPU module were developed and the performance of the developed modules was verified. The S/N ratio, noise level and signal acquisition performance of the developed modules showed proper level as we designed in the beginning.

  4. Nuclear Power Plant Control and Instrumentation in Italy: Recent developments and future perspectives

    International Nuclear Information System (INIS)

    Santandrea, E.

    1990-01-01

    The result of the ''referendum'' in Italy in November 1987 led to cut-back nuclear activities in Italy, as far as both construction and operation on nuclear plants are concerned. A new PEN (National Energy Plan), calls, however, for a programme of research and industrial development of reactors with inherent and passive safety features. ENEA is now engaged in a critical analysis and comparison of different concepts, particularly aimed on safety and environmental and social impact. A particular attention is devoted to relevant subsystems and components with high innovation degree. For the Control System and Instrumentation there is the aim to define rules and design criteria, by evaluating the impact of inherent safety goals on the C and I design. The effort on man-machine interface is considered essential to increase safety and efficiency of advanced reactors. In the frame of the new National Energy Plan, the ENEA effort in the CI area is part of a more general Industrial Promotion Program oriented to optimize the plant process, reducing the environmental impact of energy systems. In such an effort, some questions can be solved by knowledges originally developed in the nuclear and aero-spatial field. So, the national industrial promotion program, tends to transfer to industrial application the developed competencies on disciplines like: Software engineering and reliability; engineering of diffused systems and local networks; electromagnetical compatibility; industrial diagnostic; expert systems in industrial applications; modelling and simulators; ''intelligent'' detection systems and sensors; process optimization. (author). 3 tabs

  5. Development of a new generation of ion exchange resin for nuclear and fossil power plant

    International Nuclear Information System (INIS)

    Tsuzuki, Shintaro; Tagawa, Hidemi; Yamashita, Futoshi; Okamoto, Ryutaro

    2008-01-01

    It is required to maintain water quality supplied to steam generator to the water designed based on its water chemistry in order to keep the sound operation of nuclear power plants or fossil power plants. Condensate Polishing Plant (CPP) is installed for removing ions in the water which uses a mixed bed of cation exchange resin and anion exchange resin. We have developed new generation of CPP resin. The product is a unique combination of super high exchange capacity cation exchange resin and high fouling resistant anion exchange resin. The CPP resin has been used in many power plants. Amberjet 1006 was developed as a cation exchange resin with high oxidative stability, high operational capacity and New IRA900CP was developed as an anion exchange resin with high fouling resistant to leachables released out of cation exchange resin by oxidative degradation over the service period. The novel CPP resin was first used in 2000 and has now been used in many power plants in Japan. The CPP resin has been giving excellent quality of water. (author)

  6. Development of a remote monitoring and control system for nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Kim, Jong Hyun; Seong, Poong Hyun

    2002-01-01

    Nuclear Power Plants (NPPs) will be highly connected network enabled system and need to be monitored and controlled round the clock for high safety and availability. Using the network and web enabled tools, NPPs can be monitored remotely by operators at anytime from any place connected to the network via a general web browser. However, there are security and performance issues associated with such tools, as will be further discussed further. We developed a web-based Remote Monitoring and Control System (RMCS) that uses prevalent web technology. This work, as a preliminary study, performed the conceptual design of the web-based RMCS and developed the prototype

  7. Process of liquid radioactive waste treatment in nuclear power plant and development trend

    International Nuclear Information System (INIS)

    Liu Jiean; Wang Xin; Liu Dan; Zhu Laiye; Chen Bin

    2014-01-01

    The popular liquid radioactive waste treatment methods in nuclear power plants (NPP) are Chemical precipitation, evaporation, ion exchange, membrane treatment, chemical coagulation and activated carbon absorption and so on. 'Filter + activated carbon absorption (Chemical coagulation) + ion exchange' has a good prospect for development, as its simple process, high decontamination factor, low energy consumption and smaller secondary wastes. Also the process is used in Sanmen and Haiyang Projects. The severe incident in NPP set an even higher demand on liquid radioactive waste treatment. The new type treatment materials, optimization of the existed treatment, combination of treatment and the mobile treatment facility is the development trend in liquid radioactive waste treatment in NPP. (authors)

  8. Development of Structural Health Monitoring System for pipes in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Eom, H. S.; Choi, Y. C.; Shin, S. H.; Youn, D. B.; Park, J. H.

    2010-01-01

    Structural health monitoring (SHM) has becoming an important issue in the maintenance of various structures such as large steel plates, vessels, and pipes in nuclear power plants. There are important factors to be considered in developing an SHM system. With consideration of these factors, we have developed a computerized multi-channel ultrasonic system that can handle array transducers and generate a high-power pulse for online SHM of the plates and pipes. The proposed system is compact but has all the necessary functions for SHM of important structure such as pipes and plates in a NPP

  9. Development of pilot model of virtual nuclear power plant and its application to radiation management

    International Nuclear Information System (INIS)

    Kang, K. D.; Sin, S. W.

    2002-01-01

    Using Virtual Reality (VR) technique, a real model for radiation controlled area in nuclear power plant was developed and a feasibility study to develop a computational program to estimate radiation dose was performed. For this purpose a pilot model with an dynamic function and bi-directional communication was developed. This model was enhanced from the existing 3-D single-directional communication. In this pilot model, a plant visitor needs a series of security checking process initially. If he(she) enters the controlled area and approaches radiation hazard area, the alarms with warning lamp will be initiated automatically. Throughout the test to connect this model from both domestic and international sites in various time zones it has proven that it showed a sufficient performance. Therefore this model can be applied to broad fields as radiation protection procedures photographic data, on-line dose program

  10. Development of alarm cause tracking system for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Jung Woon; Kim, Jung Taek; Park, Jae Chang; Lee, Hyun Chul; Park, Joong Pal

    2004-05-01

    The proposed system, the ACTS(Alarm Cause Tracking System), in the 1st and 2nd development period(2001. 7 ∼ 2003. 6), tracks and displays the causes of alarms on-line from computerized logic diagrams. And the system highlights the specific procedures related the causes in the procedure of the alarm. In this period(2003. 7 ∼ 2004. 4), we developed the ACTS for Korea standard nuclear power plant. Also, we computerized control logic diagrams and alarm procedures for the ACTS. A long-term target is to apply the ACTS at the real power plant, and a short-term target is to connect the ACTS with the ITF(Intergrated Test Facility) in KAERI site to develop other applications

  11. Development of Soil Derived Concentration Guidance Levels for Decommissioning at Overseas Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Sohn, Wook; Yoon, Suk Bon; Kim, Jeongju [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    In Korea, the criteria are expected to be given in terms of dose as in US and Spain. However, since dose cannot be measured, corresponding measurable concentration limits, so-called Derived Concentration Guidance Levels (DCGLs), should be developed for each radionuclide which is expected to be present in the site. Also, as they serve as a goal of decommissioning and direct dismantling and decontamination methods applicable to the site, DCGLs should be developed in the early phase of decommissioning. This paper describes how each overseas nuclear power plant developed its site-specific Soil DCGLs: what kind of post closure use of the site (scenario) was assumed and how the site-specific Soil DCGLs were calculated based on the scenario assumed for each plant. Through this, it is intended to derive lessons learned which will be instructive for future decommissioning of domestic nuclear power plants including Kori Unit 1. It is very important to have as good under-standing as possible of characteristics of the site by collection of relevant information and data in order to apply a scenario which is most foreseeable and plausible for a site to be decommissioned and to provide site-specific inputs to the calculation of the Soil DCGLs. These efforts will help to have not-overly conservative values for the Soil DCGLs, thus thereby reducing the costs and time needed for performing the decommissioning.

  12. Development of a coordinated control system for BWR nuclear power plant and HVDC transmission system

    International Nuclear Information System (INIS)

    Ishikawa, M.; Hara, T.; Hirayama, K.; Sekiya, K.

    1986-01-01

    The combined use of dc and ac transmissions or so-called hybrid transmission was under study, employing both dc and ac systems to enable stable transmission of 10,000 MW of electric power generated by the BWR nuclear plant, scheduled to be built about 800 km away from the center of the load. It was thus necessary to develop a hybrid power transmission control system, the hybrid power transmission system consisting of a high voltage dc transmission system (HVDC) and an ultrahigh ac transmission system (UHVAC). It was also necessary to develop a control system for HVDC transmission which protects the BWR nuclear power plant from being influenced by any change in transmission mode that occurs as a result of faults on the UHVAC side when the entire power of the BWR plant is being sent by the HVDC transmission. This paper clarifies the requirements for the HVDC system control during hybrid transmission and also during dc transmission. The control method that satisfies these requirements was studied to develop a control algorithm

  13. Development of automated patrol-type monitoring and inspection system for nuclear power plant and application to actual plant

    International Nuclear Information System (INIS)

    Senoo, Makoto; Koga, Kazunori; Hirakawa, Hiroshi; Tanaka, Keiji

    1996-01-01

    An automated patrol-type monitoring and inspection system was developed and applied in a nuclear power plant. This system consists of a monorail, a monitoring robot and an operator's console. The monitoring robot consists of a sensor unit and a control unit. Three kinds of sensor, a color ITV camera, an infrared camera and a microphone are installed in the sensor unit. The features of this monitoring robot are; (1) Weights 15 kg with a cross-sectional dimensions of 152 mm width and 290 mm height. (2) Several automatic monitoring functions are installed using image processing and frequency analysis for three sensor signals. (author)

  14. The Establishment of Regulation for Supporting the Development of the First Nuclear Power Plant in Indonesia

    International Nuclear Information System (INIS)

    Mardha, Amil

    2011-01-01

    To support construction and operation of the first Nuclear Power Plant (NPP), Government of Indonesia through BAPETEN as a national authority in nuclear energy usage has pushed ahead in enhancing the development of regulatory system. BAPETEN is currently focusing its activities to the development of regulations in the form of Chairman Decree, the development of licensing and inspection system, and manpower for NPP regulators. The process of the establishment of national legislation is based on the Act No.10 Year 2004 on The establishment of legislation. The process of legal drafting to establish or to revise a regulation for controlling of the use of nuclear energy shall comply with the constitutional and institutional requirements of national political and legal system. In drafting the regulation of nuclear safety of NPP, BAPETEN has been involving some other agencies or other related government agencies, and also stakeholders such as utility, academic institutions, and publics. In general, in the process of legal drafting, international publications or other country regulations can be a reference. Present paper deals with the legal basis of regulation, the establishment of legislation in Indonesia, the process in legal drafting regulations of NPP, and the current status of NPP regulations. (author)

  15. Occupational dose reduction developments and data collected at nuclear power plants

    International Nuclear Information System (INIS)

    Dionne, B.J.; Baum, J.W.

    1984-01-01

    Occupational dose reduction developments and data collected at nuclear power plants have been described. Written descriptions of repetitive high dose jobs, their collective dose equivalent ranges and list of dose reduction techniques will aid in reducing collective dose equivalents from these dose-reduction targets. Knowing which components contribute to high maintenance or repair dose will aid in reducing routine maintenance collective dose equivalents. The radwaste dose reduction improvements will aid in reducing radwaste operations collective dose equivalent and reduce the number of radwaste workers who exceed their administrative dose limits. The identification and rating of managers' and workers' ALARA incentives will provide the basis for recommendations to improve dose reduction incentives. Lastly, the identification and rating of the key components of an ALARA program will aid in the development and coordination of the nuclear station ALARA programs

  16. Development of vendor independent safety analysis capability for nuclear power plants in Taiwan

    International Nuclear Information System (INIS)

    Tang, J.-R.

    2001-01-01

    The Institute of Nuclear Energy Research (INER) and the Taiwan Power Company (TPC) have long-term cooperation to develop vendor independent safety analysis capability to provide support to nuclear power plants in Taiwan in many aspects. This paper presents some applications of this analysis capability, introduces the analysis methodology, and discusses the significance of vendor independent analysis capability now and future. The applications include a safety analysis of core shroud crack for Chinshan BWR/4 Unit 2, a parallel reload safety analysis of the first 18-month extended fuel cycle for Kuosheng BWR/6 Unit 2 Cycle 13, an analysis to support Technical Specification change for Maanshan three-loop PWR, and a design analysis to support the review of Preliminary Safety Analysis Report of Lungmen ABWR. In addition, some recent applications such as an analysis to support the review of BWR fuel bid for Chinshan and Kuosheng demonstrates the needs of further development of the analysis capability to support nuclear power plants in the 21 st century. (authors)

  17. A study on the development of curriculum of nuclear technology development for training engineering technicians in nuclear plants

    International Nuclear Information System (INIS)

    Lee, Y.S.; Yoon, S.K.; Lee, C.Y.

    1982-01-01

    In this paper, the development of curriculum was studied for Department of Nuclear Technology. In order to make the students suitable for the job as engineering technicians with both theory and practical technique, the basic education in the field related to nuclear energy was emphasized in designing the curriculum. In addition taking the special situation of our department into consideration, we made it a principle to provide them with practical experiences with on-the-job training for 16 weeks. A model curriculum with syllabuses for major subjects, contents of experiments with lists of equipments, and program of on-the-job-training were suggested. (author)

  18. Development of preventive maintenance technology and advanced service equipment for operating nuclear power plants

    International Nuclear Information System (INIS)

    Abe, Kazuhiro; Sumikawa, Masaharu; Hirakawa, Hiromasa; Arakawa, Tadao; Hasegawa, Kunio; Kato, Kanji.

    1990-01-01

    Hitachi Ltd. as a manufacturer of nuclear power plants has carried out the consistent general preventive maintenance activities from the planning of the plants and the design of maintainability in the construction phase to the planning and working of preventive maintenance in the operation and maintenance phase, and exerted efforts to heighten the capacity ratio and reliability of the plants. For the purpose, the steady activities of reliability improvement have been carried out throughout the whole company, and the rationalization of the planning and management of the preventive maintenance with a computer and the development of the robots to which Al is applied have been promoted. As the technology of upgrading the facilities, boron racks, the control rods having long life and so on were developed, and their practical use is advanced. Moreover for the future, the development of the diagnostic technology on material deterioration using superconductive quantum interference devices (SQUID) is in progress. The preventive maintenance activities in Hitachi Ltd., the technical development for the purpose and the upgrading of the plant facilities are reported. (K.I.)

  19. Development of a test bed for operator aid and advanced control concepts in nuclear power plants

    International Nuclear Information System (INIS)

    Turinsky, P.J.; Doster, J.M.; Kim, K.D.; Al-Chalabi, R.M.; Khedro, T.; Sues, R.H.; Yacout, A.M.

    1990-01-01

    A great amount of research and development is currently under way in the utilization of artificial intelligence (AI), expert system, and control theory advances in nuclear power plants as a basis for operator aids and automatic control systems. This activity requires access to the measured dynamic responses of the plant to malfunction, operator- or automatic-control-initiated actions. This can be achieved by either simulating plant behavior or by using an actual plant. The advantage of utilizing an actual plant versus a simulator is that the true behavior is assured of both the power generation system and instrumentation. Clearly, the disadvantages of using an actual plant are availability due to licensing, economic, and risk constraints and inability to address accident conditions. In this work the authors have decided to employ a functional one-ninth scale model of a pressurized water reactor (PWR). The scaled PWR (SPWR) facility is a two-loop representation of a Westinghouse PWR utilizing freon as the working fluid and electric heater rods for the core. The heater rods are driven by a neutron kinetics model accounting for measured thermal core conditions. A control valve in the main steam line takes the place of the turbine generator. A range of normal operating and accident situations can be addressed. The SPWR comes close to offering all the advantages of both a simulator and an actual physical plant in regard to research and development on AI, expert system, and control theory applications. The SPWR is being employed in the development of an expert-system-based operator aid system. The current status of this project is described

  20. Development and recent trend of design of BWR nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kani, J [Tokyo Shibaura Electric Co. Ltd., Kawasaki, Kanagawa (Japan)

    1977-11-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation.

  1. Development and recent trend of disign of BWR nuclear power plants

    International Nuclear Information System (INIS)

    Kani, Jiro

    1977-01-01

    Many improvements have been carried out in BWR nuclear power plants from BWR-1, represented by Dresden No. 1 plant, to the present BWR-6 as the capacity has increased. In Japan, the plants up to BWR-5 have been constructed. In addition, further fine design improvements are being performed in the complete domestic manufacturing of BRWs based on the operational experiences to date. A variety of investigations on the standardization of nuclear power facilities have been progressing under the leadership of Japanese Ministry of International Trade and Industry since 1975. In this standardization, it is intended to forward the plant design taking eight concrete items into consideration, mainly aiming at carrying cut unerringly the maintenance and inspection, reduction of exposure of employees to radiation, and improvements of the rate of operation of plants and equipment reliability. The containment vessel has been developed in three forms, from Mark 1 through 3, adopting the pressure control system consistently since BWR-2. Mark 1 and 2 were constructed in Japan. However, these designs sacrificed the workability and increased radiation exposure during maintenance as a result of placing emphasis on the safety facilities, therefore Toshiba Electric has investigated the advanced Mark 1 type. Its features are the design for improving the work efficiency in a containment vessel, reducing the radiation exposure of workers, shortening plant construction period, and considering the aseismatic capability. In addition, the following themes are being planned as future standardization: (1) electrically driven control rod driving system, (2) improved design of reactor core, and (3) internal pump system as compared with external re-circulation. (Wakatsuki, Y.)

  2. Development of computer-aided design and production system for nuclear power plant

    International Nuclear Information System (INIS)

    Ishii, Masanori

    1983-01-01

    The technically required matters related to the design and production of nuclear power stations tended to increase from the viewpoint of the safety and reliability, and it is indispensable to cope with such technically required matters skillfully for the rationalization of the design and production and for the construction of highly reliable plants. Ishikawajima Harima Heavy Industries Co., Ltd., has developed the computer-aided design data information and engineering system which performs dialogue type design and drawing, and as the result, the design-production consistent system is developed to do stress analysis, production design, production management and the output of data for numerically controlled machine tools consistently. In this paper, mainly the consistent system in the field of plant design centering around piping and also the computer system for the design of vessels and others are outlined. The features of the design works for nuclear power plants, the rationalization of the design and production management of piping and vessels, and the application of the CAD system to other general equipment and improvement works are reported. This system is the powerful means to meet the requirement of heightening quality and reducing cost. (Kako, I.)

  3. Next Generation Nuclear Plant Materials Research and Development Program Plan, Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    G.O. Hayner; R.L. Bratton; R.E. Mizia; W.E. Windes; W.R. Corwin; T.D. Burchell; C.E. Duty; Y. Katoh; J.W. Klett; T.E. McGreevy; R.K. Nanstad; W. Ren; P.L. Rittenhouse; L.L. Snead; R.W. Swindeman; D.F. Wlson

    2007-09-01

    DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Some of the general and administrative aspects of the R&D Plan include: • Expand American Society of Mechanical Engineers (ASME) Codes and American Society for Testing and Materials (ASTM) Standards in support of the NGNP Materials R&D Program. • Define and develop inspection needs and the procedures for those inspections. • Support selected university materials related R&D activities that would be of direct benefit to the NGNP Project. • Support international materials related collaboration activities through the DOE sponsored Generation IV International Forum (GIF) Materials and Components (M&C) Project Management Board (PMB). • Support document review activities through the Materials Review Committee (MRC) or other suitable forum.

  4. Development of probabilistic methods for safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    Schott, H.; Berg, H.P.

    1998-01-01

    Since its introduction by the German Risk Study, Probabilistic Safety Assessment (PSA) has developed in Germany to a valuable tool in regulatory decision-making. Plant specific PSAs of Level 1+ are now conducted for all nuclear power plants in the frame of Periodic Safety Reviews. This paper is devoted to the description or key elements set out in the regulatory guidelines for PSA-Level 1+ and the corresponding technical documents and the further development of PSA methodology in the Federal Republic of Germany. In the course of the next years it is intended to make progress in the modeling of common cause failures, human reliability evaluation, reduction of uncertainties in PSA modeling techniques and data estimation, analysis of low power and shut down states as well as in reaching a mature methodology for inclusion of external events into the analysis. (author)

  5. Nuclear power plant disasters

    International Nuclear Information System (INIS)

    Trott, K.R.

    1979-01-01

    The possibility of a nuclear power plant disaster is small but not excluded: in its event, assistance to the affected population mainly depends on local practitioners. Already existing diseases have to be diagnosed and treated; moreover, these physicians are responsible for the early detection of those individuals exposed to radiation doses high enough to induce acute illness. Here we present the pathogenesis, clinical development and possible diagnostic and therapeutical problems related to acute radiation-induced diseases. The differentiation of persons according to therapy need and prognosis is done on the sole base of the clinical evidence and the peripheral blood count. (orig.) [de

  6. Nuclear power plant diagnostic system

    International Nuclear Information System (INIS)

    Prokop, K.; Volavy, J.

    1982-01-01

    Basic information is presented on diagnostic systems used at nuclear power plants with PWR reactors. They include systems used at the Novovoronezh nuclear power plant in the USSR, at the Nord power plant in the GDR, the system developed at the Hungarian VEIKI institute, the system used at the V-1 nuclear power plant at Jaslovske Bohunice in Czechoslovakia and systems of the Rockwell International company used in US nuclear power plants. These diagnostic systems are basically founded on monitoring vibrations and noise, loose parts, pressure pulsations, neutron noise, coolant leaks and acoustic emissions. The Rockwell International system represents a complex unit whose advantage is the on-line evaluation of signals which gives certain instructions for the given situation directly to the operator. The other described systems process signals using similar methods. Digitized signals only serve off-line computer analyses. (Z.M.)

  7. Nuclear power plant siting

    International Nuclear Information System (INIS)

    Sulkiewicz, M.; Navratil, J.

    The construction of a nuclear power plant is conditioned on territorial requirements and is accompanied by the disturbance of the environment, land occupation, population migration, the emission of radioactive wastes, thermal pollution, etc. On the other hand, a nuclear power plant makes possible the introduction of district heating and increases the economic and civilization activity of the population. Due to the construction of a nuclear power plant the set limits of negative impacts must not be exceeded. The locality should be selected such as to reduce the unfavourable effects of the plant and to fully use its benefits. The decision on the siting of the nuclear power plant is preceded by the processing of a number of surveys and a wide range of documentation to which the given criteria are strictly applied. (B.H.)

  8. The Development of a Snubber Management System for Welds in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Hyun Ju; Cho, Yong-Bae; Kim, Yoo Sung [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    There are snubbers, spring hanger, anchors, rigid supports for the structures which support the static and/or dynamic loads such as thermal load, pressure, impact and vibration from components and pipings of nuclear power plants. Snubbers constrain the displacements generated by loads transmitted to components and systems abruptly. When the loss of function of snubber during normal operation, the thermal load and pressure transmit directly to structures such as pipings and components, and additional loads, which were not considered during the design stage, act on the structures. Therefore, according to regulatory requirement to confirm the stability of supporting system, the inspection for the snubber is reflected to in-service inspection (ISI) and in-service test (IST) plans. In order to comply with the regulatory requirement, KHNP has performed the ISI and IST and inspected the snubbers. As the increment of operating year of nuclear power plants in Korea, the possibility of deterioration of equipment is higher. Therefore, the security related to the integrity of equipment becomes more important. The snubber takes an important role related to the structural integrity equipped on principal pipings. 100% snubbers are inspected during pre-service inspection and 10% snubbers are inspected during in-service inspection as a sample in nuclear power plants in Korea. KHNP has been developed a snubber management system because there was no management tool for snubbers to show the inspection and maintenance results systematically. The inspection and maintenance results of snubbers can be easily reached by plants, head office and CRI. Moreover, the information related to inspection history and condition of snubber can be effectively inquired.

  9. Development of IT-based Cyber Security Technology for Nuclear Power Plant

    International Nuclear Information System (INIS)

    Hong, S. B.; Lee, J. C.; Choi, Y. S.; Choi, Y. R.; Cho, J. W.; Jung, C. E.; Jeong, K. I.; Park, B.; Koo, I. S.

    2009-11-01

    Development and enlargement of the high speed communication network make it possible the user to access online information easily. It generates changing offline activities to online in the economics, expansion of cultural interchanges and convenient life. But it also causes misuse, wiretapping, forgery and alteration of the information via illegal invasion(virus, hacking), and these are derived from the open network characteristic, weakness of the securities of the TCP/IP protocol and information systems. The security of individual and the national foundation facility(industry and government) can be threatened because of these problems, and theses can be used as a instrument of cyber-war. Many cyber security technologies have been developed to corp with the cyber threat. One of the most important national facility is the nuclear power plant and the necessity of the cyber security for the digital I and C of it have been proposed since middle of the 2000. KINS announced the regulation of the cyber security for the digital I and C of the nuclear power plant in 2007. The main concept of the cyber security for it is similar to the IT field that is treated as a leader of the cyber security. Because of the difference of the characteristics between the IT field and the nuclear industry, applying the cyber security technologies developed and used in the IT field to the nuclear industry has some critical constraints. We will analyze these problems and propose a cyber security method based on cryptograph and authentication for the I and C communication network in this report

  10. Development of IT-based Cyber Security Technology for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Hong, S. B.; Lee, J. C.; Choi, Y. S.; Choi, Y. R.; Cho, J. W.; Jung, C. E.; Jeong, K. I.; Park, B.; Koo, I. S

    2009-11-15

    Development and enlargement of the high speed communication network make it possible the user to access online information easily. It generates changing offline activities to online in the economics, expansion of cultural interchanges and convenient life. But it also causes misuse, wiretapping, forgery and alteration of the information via illegal invasion(virus, hacking), and these are derived from the open network characteristic, weakness of the securities of the TCP/IP protocol and information systems. The security of individual and the national foundation facility(industry and government) can be threatened because of these problems, and theses can be used as a instrument of cyber-war. Many cyber security technologies have been developed to corp with the cyber threat. One of the most important national facility is the nuclear power plant and the necessity of the cyber security for the digital I and C of it have been proposed since middle of the 2000. KINS announced the regulation of the cyber security for the digital I and C of the nuclear power plant in 2007. The main concept of the cyber security for it is similar to the IT field that is treated as a leader of the cyber security. Because of the difference of the characteristics between the IT field and the nuclear industry, applying the cyber security technologies developed and used in the IT field to the nuclear industry has some critical constraints. We will analyze these problems and propose a cyber security method based on cryptograph and authentication for the I and C communication network in this report.

  11. Development of full power risk monitoring system for UCN 3 and 4 nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Seung Hwan; Jang, Seung Chul; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae

    2002-03-01

    This report describes full power risk monitoring system (DynaRM) of NPPs which evaluates the current plant risk and monitor the risk change caused by the configuration change due to preventive maintenance or periodic maintenance activities. In order to calculate the core damage frequency, the inoperable equipment's basic event should be submitted to a quantification engine. DynaRM uses a risk monitor model converted from PSA model as the quantification engine, which can respond fast for the various equipment configuration changes. DynaRM is developed to solve these kinds of problems and difficulties. The main two key features of DynaRM are real time risk monitoring and real time maintenance plan decision supporting. With DynaRM, the plant risk can be calculated automatically only with out of service equipment information and equipment maintenance scheduling is easy by modifying the detail schedule plan. We have developed the DynaRM for the Korean standard nuclear power plant and it is currently using at Ulchin 3,4 NPP. The adaptation for the other NPPs is also easy with a little modification since DynaRM was developed by considering it to be used in the other NPPs. Moreover, we also expect DynaRM will be a good advisory tool for the plant risk monitoring and maintenance scheduling

  12. Development of dose assessment code for release of tritium during normal operation of nuclear power plants

    International Nuclear Information System (INIS)

    Duran, J.; Malatova, I.

    2009-01-01

    A computer code PTM H TO has been developed to assess tritium doses to the general public. The code enables to simulate the behavior of tritium in the environment released into the atmosphere under normal operation of nuclear power plants. Code can calculate the doses for the three chemical and physical forms: tritium gas (HT), tritiated water vapor and water drops (HTO). The models in this code consist of the tritium transfer model including oxidation of HT to HTO and reemission of HTO from soil to the atmosphere, and the dose calculation model

  13. Development of nuclear power plant management system for Kyushu Electric Power Co., Inc

    International Nuclear Information System (INIS)

    Nakamura, Kenichi; Akiyoshi, Tatsuo; Tanimoto, Kazuo; Ogura, Kazuhito; Ibi, Yuji; Kawasaki, Michiyuki

    2002-01-01

    The Kyushu Electric Power Co., Ltd. progresses development of the nuclear power plant management system using IT under aims at upgrading of efficiency, level, and reliability on maintenance and administration business under five years planning since 1999 fiscal year. The outline of the system are explained in this paper. As a result of preparation on power station net work and personal computers set in all of company, an environment capable of using these infrastructures and introducing large scale systems on transverse business over every groups of each power station could be established. (G.K.)

  14. Development of database for spent fuel and special waste from the Spanish nuclear power plants

    International Nuclear Information System (INIS)

    Gonzalez Gandal, R.; Rodriguez Gomez, M. A.; Serrano, G.; Lopez Alvarez, G.

    2013-01-01

    GNF Engineering is developing together with ENRESA and with the UNESA participation, the spent fuel and high activity radioactive waste data base of Spanish nuclear power plants. In the article is detailed how this strategic project essential to carry out the CTS (centralized temporary storage) future management and other project which could be emerged is being dealing with, This data base will serve as mechanics of relationship between ENRESA and Spanish NPPS, covering the expected necessary information to deal with mentioned future management of spent fuel and high activity radioactive waste. (Author)

  15. Development of conventional fatigue database for structure materials of nuclear power plant

    International Nuclear Information System (INIS)

    Yang Bing

    2002-01-01

    Management system of the conventional fatigue database for structure materials of nuclear power plant (NPP) is developed. The database included the parameters of design curves, i.e., the stress-life, survival probability-stress-life, strain-life, survival probability-strain-life, stress-strain and survival probability-stress-strain curves, and corresponding information of materials and testing conditions. Two ways, by materials name or by the inter-bounds of material mechanical properties, are constructed to search the database. From the searched information it can be conveniently performed of the conventional fatigue design analysis and reliability assessment of structures

  16. Development programmes in rural communities affected by industrial sites. The case of nuclear plants in Spain

    International Nuclear Information System (INIS)

    Yuncal, A.

    2000-01-01

    A socioeconomic analysis performed for rural communities affected by industrial sites, namely for the case of nuclear power plants in Spain a common proposal for all the European countries is made. Existence of a common European program that could cover the following aspects: Creation of a Committee to enhance the participation in decision making process on different aspects, information policies, security, transport, waste management, investments, economic development; Contract between local authorities, State and companies responsibilities and financing; Legal framework of political organization; common socio-economic program including environment, employment, companies activities, responsibilities, taxes

  17. Development of the multi-function robot for the containment vessel of the nuclear plant

    International Nuclear Information System (INIS)

    Ohmichi, Takeo; Hosaka, Shigetaka; Nishihara, Masatoshi; Ibe, Tomoyoshi; Okino, Akihisa; Nakayama, Junji; Miida, Toshiro; Ishida, Michiyasu

    1985-01-01

    A mobile manipulator has been developed and consummated on an experimental basis to perform various maintenance and repair work in places where it is not accessible to workers during operation of the nuclear power plant. This manipulator provided with legs is capable of climbing stairs, clearing obstacles and adequately moving over flat surfaces, and has an arm to handle various tools with the sensibility and strength beyond the average of men and eyes that have a three dimensional vision. Its feature of size is as small as a man and it can be operated by autonomic and manual control under the severe environment. 5 refs.; 11 figs.; 1 table

  18. Development of effective tool for iterative design of human machine interfaces in nuclear power plant

    International Nuclear Information System (INIS)

    Nakagawa, Takashi; Matsuo, Satoko; Yoshikawa, Hidekazu; Wu, Wei; Kameda, Akiyuki; Fumizawa, Motoo

    2000-01-01

    The authors have developed SEAMAID, which is a Simulation-based Evaluation and Analysis support system for MAn-machine Interface Design (SEAMAID) in the domain of nuclear power plants. The SEAMAID simulated the interaction between an operator and human machine interfaces (HMI), and supports to evaluate the HMI by using the simulation results. In this paper, a case study of evaluation for conventional center control room design was conducted. The authors were confirmed that SEAMAID is a useful tool for improvements of HMI design (J.P.N.)

  19. Materials integrity analysis for application of POSCO developed STS to Korean Nuclear Power Plants

    International Nuclear Information System (INIS)

    Hyun-Young, Ch.; Tae-Eun, J.; Young-Sik, K.

    2009-01-01

    Full text of publication follows: POSCO has developed duplex stainless steel (S32750) and hyper super duplex stainless steels for the purpose of using them in the secondary circulation cooling water system in Korean nuclear power plants. This system supplies seawater to cooling water heat exchanger tubes, related pipes and chlorine injection system. In normal operation, seawater is supplied to heat exchanger through the exit of circulation pump headers and the heat exchanged sea water is extracted to the discharge pipes in circulation cooling water system connected to the circulation water discharge lines. The flow velocity of circulation cooling water system in nuclear power plants is high and damages of components from corrosion are severe. Therefore, this environment makes requiring of using high strength and high corrosion resistant steels. Hyper duplex stainless steel (27Cr-7.0Ni-2.5Mo-3.2W-0.35N) has been newly developed in Korea and is being improved for applying to nuclear power plants. In this study, the physical and mechanical properties and corrosion resistance of currently producing stainless steels and newly developed materials are qualitatively evaluated in comparative to commercial stainless steels in other countries. The properties of weld are analyzed and the best compositions of welding rod are suggested. The optimum weld condition is derived for ensuring HAZ phase ratios and controlling weld cracks. For applying these materials to the seawater heat exchanger, CCT and CPT in weldments are measured using mock-up tube testers that are newly designed for this study. Coupons of candidate materials are introduced in the real system and corrosion resistance of them are analyzed. As results of all experiments, the current CCT and CPT criteria in Korean nuclear power plants are reviewed, and the more actual and strengthened criteria will be suggested. The real scale components made of newly developed hyper super duplex stainless steel will be applied to

  20. Development and utilization of the NRC policy statement on the regulation of advanced nuclear power plants

    International Nuclear Information System (INIS)

    Williams, P.M.; King, T.L.

    1988-06-01

    On March 26, 1985, the US Nuclear Regulatory Commission issued for public comment a ''Proposed Policy for Regulation of Advanced Nuclear Power Plants'' (50 FR 11884). This report presents and discusses the Commission's final version of that policy as titled and published on July 8, 1986 ''Regulation of Advanced Nuclear Power Plants, Statement of Policy'' (51 FR 24643). It provides an overview of comments received from the public, of the significant changes from the proposed Policy Statement to the final Policy Statement, and of the Commission's response to six questions contained in the proposed Policy Statement. The report also discusses the definition for advanced reactors, the establishment of an Advanced Reactors Group, the staff review approach and information needs, and the utilization of the Policy Statement in relation to other NRC programs, including the policies for safety goals, severe accidents and standardization. In addition, guidance for advanced reactors with respect to operating experience, technology development, foreign information and data, and prototype testing is provided. Finally, a discussion on the use of less prescriptive and nonprescriptive design criteria for advanced reactors, which the Policy Statement encourages, is presented

  1. Development of seismic design method for free standing rack and applicability to Japanese nuclear power plant

    International Nuclear Information System (INIS)

    Takaki, Yu; Taniguchi, Katsuhiko; Kishimoto, Junichi; Iwasaki, Akihisa; Nekomoto, Yoshitsugu; Kuga, Tohru; Kameyama, Masashi

    2017-01-01

    Free standing racks which are not anchored to the pool floor nor walls have never been adopted in Japan. Under an earthquake, behaviors of free standing racks are nonlinear and involve a complex combination of motions (sliding, rocking, and twisting) and impacts between a fuel assembly and the fuel cell walls and between a pit floor and rack pedestals. To predict a seismic response of free standing racks, the seismic analysis requires careful considerations of these complex phenomena (sliding, rocking, and twisting), fluid coupling effects and frictional effects. We carried out seismic experiments on the full-scale rack model in both water and dry conditions and obtained the fundamental data about behavior of free standing racks (sliding, and rocking motions). We have developed the nonlinear dynamic analysis method to predict seismic response of free standing racks utilizing the full-scale test result and verified the analysis evaluation method of free standing rack by comparison between analysis results and experimental data. Furthermore, we applied the seismic design method to the free standing rack in the Japanese nuclear plant (Mihama nuclear power station Unit 3), and verified that the free standing rack was applicable to Japanese nuclear plant. (author)

  2. Action research for the development of the organizational climate in nuclear power plants. Review of the 6-year research and development program

    International Nuclear Information System (INIS)

    Yoshida, Michio; Misumi, Jyuji; Misumi, Emiko; Kotani, Fumio; Fukui, Hirokazu; Sakurai, Yukihiro

    1998-01-01

    The Institute of Nuclear Safety System, Incorporated and the Japan Institute for Group Dynamics have conducted action research for the development of the organizational climate in nuclear power plants. First, two types of scales were completed. One is for measuring the leadership behavior of leaders working at nuclear power plants and the other is for measuring the safety consciousness of workers. After having diagnosed the reality of actual nuclear power plants using those scales developed, leadership training courses were developed and implemented successfully. Analyses of the commitment to organization and self-efficacy and the relationship between leadership and personality were conducted as well. (author)

  3. Action research for the development of the organizational climate in nuclear power plants. Review of the 6-year research and development program

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Michio [Kumamoto Univ. (Japan). Faculty of Education; Misumi, Jyuji; Misumi, Emiko; Kotani, Fumio; Fukui, Hirokazu; Sakurai, Yukihiro

    1998-09-01

    The Institute of Nuclear Safety System, Incorporated and the Japan Institute for Group Dynamics have conducted action research for the development of the organizational climate in nuclear power plants. First, two types of scales were completed. One is for measuring the leadership behavior of leaders working at nuclear power plants and the other is for measuring the safety consciousness of workers. After having diagnosed the reality of actual nuclear power plants using those scales developed, leadership training courses were developed and implemented successfully. Analyses of the commitment to organization and self-efficacy and the relationship between leadership and personality were conducted as well. (author)

  4. Development of software for the microsimulator for the KO-RI nuclear power plant unit 2

    International Nuclear Information System (INIS)

    Seok, H.; No, H.C.; Cho, S.J.; Park, S.D.; Jun, H.Y.; Lee, Y.K.

    1994-01-01

    A workstation-based real-time simulator for two-loop pressurized water reactor plants is developed for classroom training in support of a full-scale simulator, on-site transient analysis, and engineering studies. The present simulator consists of three functional modules: plant module, graphic module, and man-machine interaction module. The plant module includes models for the core kinetics, reactor coolant system, steam generator, main steam line, balance of plant, and control and protection system. Each of the models is optimized to obtain the capability of real-time simulation. The graphic module is designed to provide the user with more information at a glance by dynamically displaying schematic diagrams of the systems, symbols indicating the operating status of each component, trend curves, and the main control room. As tools for the man-machine interface, the man-machine interaction model uses a color cathode ray tube monitor, a standard keyboard, and the mouse. The interactive communication module receives parameters from the user via the keyboard and mouse, and transfers them to the plant module so as to enable the user to perform his specific actions. This module provides the user with various initiating events (malfunctions and manual controls) through SYSTEM, CONTROL ROOM, and ACCIDENTS menus, and thus a wide range of nuclear steam supply system transients can be easily simulated. The FISA-2/WS is verified through comparisons with analytical solutions, separated tests and integral tests, and predictions by RETRAN-2 and RELAP5/MOD3

  5. Summary of project to develop handbook of human reliability analysis for nuclear power plant operations

    International Nuclear Information System (INIS)

    Swain, A.D.

    1978-01-01

    For the past two years Alan Swain and Henry E. Guttmann, of the Statistics, Computing, and Human Factors Division, Sandia Laboratories, have been developing a handbook to aid qualified persons to evaluate the effect of human error on the availability of engineered safety systems and features in nuclear power plants. The handbook includes a mathematical model, procedures, derived human failure data, and principles of human behavior and ergonomics. The handbook is expanding the human error analyses which were presented in WASH--1400. The work, under the sponsorship of Probabilistic Analysis Staff, NRC Office of Nuclear Regulatory Research (Dr. M.C. Cullingford, NRC Program Manager), is about half completed. An outline of the handbook contents is given in copies of vugraphs (attached), followed by copies of human performance model abstractors (also attached). A first draft of the handbook is scheduled for NRC review by July 1, 1979

  6. Development of a computerized system for performance monitoring and diagnostics in nuclear power plants

    International Nuclear Information System (INIS)

    Chou, G.H.; Chao, H.J.

    1995-01-01

    An on-line computerized system for thermal performance monitoring and diagnostics has been developed at the Institute of Nuclear Energy Research (INER). It was the product of the ChinShan plant performance Monitoring, Analysis and Diagnostics Expert System (CS-MADES) project sponsored by Taiwan Power Company (TPC). The system can carry out turbine performance monitoring and analysis during normal operation, and yield diagnostic results of component degradation after finding out the missing generation problems. Three subsystems were generated to support the whole system framework. They are Test Data Processing Subsystem (TDPS), On-line Monitoring and Analysis Subsystem (OMAS), and Thermal Performance Diagnostics Expert System (TPDES). Some visible benefits have been gained so far through the prototype system installed at the Chinshan nuclear power station

  7. Development and characteristics of halogen-free flame-retardant cables for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kanemitsuya, K.; Furukawa, K.; Tachibana, T.; Ohara, H.; Ebiike, Y.; Hamachi, K.; Makino, M. (Mitsubishi Cable Industries Ltd., Tokyo (Japan))

    1992-10-01

    Halogen-free flame-retardant (HF-FR) cables for use in nuclear power plants, especially in pressurized water type (PWR) plants, have been developed to advancing the safety and reliability of nuclear power plants. HF-FR cables generate no corrosive gases and minimal amount of toxic gases and smoke during fires, and are accordingly quite safe in comparison with the conventional cables which can cause the secondary calamity by eliminating a large amount of hydrogen halide gas and smoke. HF-FR ethylene-propylene rubber (EPR) and crosslinked polyolefin (XLPO) are used as insulating materials, and HF-FR ethylene-vinyl acetate (EVA) elastomer and polyolefin (PO) are used as jacketing materials. The results of a series of experiments on several types of HF-FR cables have revealed that these cables fully satisfy every requirement, including reliability under LOCA (Loss of Coolant Accidents) simulated conditions, long-term (40y) reliability under thermal and [gamma]-ray exposure, and flame resistivity under vertical tray test (IEEE Std. 383). (author).

  8. Development of computerized supporting system for emergency response in nuclear power plant

    International Nuclear Information System (INIS)

    Lee, Jae Il

    1992-02-01

    In emergency situation of nuclear power plants, effective use of emergency operating procedures (EOPs) is a crucial part of the emergency response process. However, there are several problems in the emergency operating procedures because of the form of the written procedures. They are voluminous and complicate for effective references under high stress situation. Inevitably, it takes time that could be better spent employing measures to control and stabilize to select the correct procedures and apply the decision logic. In this study, a computerized supporting system has been developed to reduce the operator error possibility under emergency situations of nuclear power plant. Using on-line input parameters, the system can determine the status of the critical safety functions and can find appropriate procedures and necessary operator actions automatically. Moreover, the system can help the operator decision making in the core melt accident situation. By tracking the EOP in an on-line mode, most steps concerning checking or verifying plant state are processed automatically without operator participations. Therefore, the interactions between the system and the operator are simplified significantly and the possibility of human error is reduced

  9. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Gi Hong [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2015-12-15

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  10. Research and development for decontamination system of spent resin in Hanbit Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sung, Gi Hong

    2015-01-01

    When reactor coolant leaks occur due to cracks of a steam generator tube, radioactive materials contained in the primary cooling water in nuclear power plant are forced out toward the secondary systems. At this time the secondary water purification resin in the ion exchange resin tower of the steam generator blowdown system is contaminated by the radioactivity of the leaked radioactive materials, so we pack this in special containers and store temporarily because we could not dispose it by ourselves. If steam generator tube leakage occurs, it produces contaminated spent resins annually about 5,000-7,000 liters. This may increase the amount of nuclear waste productions, a disposal working cost and a unit price of generating electricity in the plant. For this reasons, it is required to develop a decontamination process technique for reducing the radioactive level of these resins enough to handle by the self-disposal method. In this research, First, Investigated the structure and properties of the ion exchange resin used in a steam generator blowdown system. Second, Checked for a occurrence status of contaminated spent resin and a disposal technology. Third, identified the chemical characteristics of the waste radionuclides of the spent resin, and examined ionic bonding and separation mechanism of radioactive nuclear species and a spent resin. Finally, we carried out the decontamination experiment using chemicals, ultrasound, microbubbles, supercritical carbon dioxide to process these spent resin. In the case of the spent resin decontamination method using chemicals, the higher the concentration of the drug decontamination efficiency was higher. In the ultrasound method, foreign matter of the spent resin was removed and was found that the level of radioactivity is below of the MDA. In the microbubbles method, we found that the concentration of the radioactivity decreased after the experiment, so it can be used to the decontamination process of the spent resin. In

  11. Development of an Ion Chamber for Monitoring the Containment of a Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Tae-Yung; Kim, Han-Soo; Park, Se-Hwan; Ha, Jang-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-05-15

    Nuclear power plants need many different types of radiation detectors for different purposes. Neutron detectors are installed inside and outside of the core to check the neutron flux. Scintillation detectors are used to check the fission products included in the liquids and gases of plant system. Geiger-Mueller counters are used for the area radiation monitoring. In addition to the above-mentioned detectors, ion chambers are installed to monitor radiation level of the containment. A few ion chambers are located within the reactor containment to monitor radiation level of an accident case. Therefore, the ion chamber should be capable of monitoring high level radiation dose up to 10{sup 7} R/h. Korea Atomic Energy Research Institute (KAERI) developed an ion chamber for monitoring the radiation dose inside the containment.

  12. Development of new CAD system for steel structures of nuclear power plants

    International Nuclear Information System (INIS)

    Morii, Yasuhiro; Kudou, Takashi; Kouno, Kenichi; Yamada, Koutarou

    2000-01-01

    IHI has developed a new Three-Dimensional Computer-Aided Design (3D-CAD) system to improve the design efficiency and quality of the steel structure of nuclear power plants. This system covers every design phase from the initial arrangement of structure to the production design sharing the same database. The system incorporates the design rules and professional expertise of designers, and enable easy and efficient design. The system can easily generate the three-dimensional data for structures, model data for stress analyses and composite arrangement data. The system has already been applied to several plants under construction and has achieved excellent results. The outline of the new CAD system is introduced. (author)

  13. Development and utilization of indicators to summarize and represent performance of nuclear power plants

    International Nuclear Information System (INIS)

    Kawaguchi, Hiroshi; Kakubari, Yukihiro; Kikkawa, Shigeru

    1996-01-01

    We have developed eight performance indicators (PIs) that enable quantitative and overall comprehension of operating performance. Among these eight indicators there are 'capability factor' 'incidents and failures' and 'radiation exposure and radioactive wastes', all used to represent the safety and reliability of a nuclear power plant. Results of analysis and evaluation by means of these PIs are distributed to the regulatory agency and other organizations every year, for the benefit of all involved. We have also been examining a technique that may allow synthesis of these PIs into a single, simpler comprehensive indicator that covers all aspects of plant performance. In this report, we present analysis and evaluation of the PIs, the technique to provide a comprehensive performance indicators and actual application. (author)

  14. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Shiraiwa, Takanori; Watanabe, Atsuo; Miyasawa, Tatsuo

    1984-01-01

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics. (author)

  15. Robotics for nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Shiraiwa, Takanori; Watanabe, Atsuo; Miyasawa, Tatsuo

    1984-10-01

    Demand for robots in nuclear power plants is increasing of late in order to reduce workers' exposure to radiations. Especially, owing to the progress of microelectronics and robotics, earnest desire is growing for the advent of intellecturized robots that perform indeterminate and complicated security work. Herein represented are the robots recently developed for nuclear power plants and the review of the present status of robotics.

  16. Development of Automatic Quality Check Software in Mailbox Declaration For Nuclear Fuel Fabrication Plants

    International Nuclear Information System (INIS)

    Kim, Minsu; Shim, Hye Won; Jo, Seong Yeon; Lee, Kwang Yeol; Ban, Myoung Jin

    2014-01-01

    Short Notice Random Inspection (SNRI) is a new IAEA safeguards inspection regime for bulk handing facility, which utilities random inspection through a mailbox system. Its main objective is to verify 100% of the flow components of the safeguarded nuclear material at such a facility. To achieve the SNRI objective, it is required to provide daily mailbox declaration, by a facility's operator, to the IAEA with regard to information, such as the receipt and shipment of nuclear materials. Mailbox declarations are then later compared with accounting records so as to examine the accuracy and consistency of the facility operator's declaration at the time of the SNRI. The IAEA has emphasized the importance of accurate mailbox declarations and recommended that the ROK initiate its own independent quality control system in order to improve and maintain its mailbox declarations as a part of the SSAC activities. In an effort to improve the transparency of operational activities at fuel fabrication plants and to satisfy IAEA recommendation, an automatic quality check software application has been developed to improve mailbox declarations at fabrication plants in Korea. The ROK and the IAEA have recognized the importance of providing good quality mailbox declaration for an effective and efficient SNRI at fuel fabrication plants in Korea. The SRA developed an automatic quality check software program in order to provide an independent QC system of mailbox declaration, as well as to improve the quality of mailbox declaration. Once the automatic QC system is implemented, it will improve the quality of an operator's mailbox declaration by examining data before sending it to the IAEA. The QC system will be applied to fuel fabrication plants in the first half of 2014

  17. Development of an integrated database management system to evaluate integrity of flawed components of nuclear power plant

    International Nuclear Information System (INIS)

    Mun, H. L.; Choi, S. N.; Jang, K. S.; Hong, S. Y.; Choi, J. B.; Kim, Y. J.

    2001-01-01

    The object of this paper is to develop an NPP-IDBMS(Integrated DataBase Management System for Nuclear Power Plants) for evaluating the integrity of components of nuclear power plant using relational data model. This paper describes the relational data model, structure and development strategy for the proposed NPP-IDBMS. The NPP-IDBMS consists of database, database management system and interface part. The database part consists of plant, shape, operating condition, material properties and stress database, which are required for the integrity evaluation of each component in nuclear power plants. For the development of stress database, an extensive finite element analysis was performed for various components considering operational transients. The developed NPP-IDBMS will provide efficient and accurate way to evaluate the integrity of flawed components

  18. Next Generation Nuclear Plant Project Technology Development Roadmaps: The Technical Path Forward

    Energy Technology Data Exchange (ETDEWEB)

    John Collins

    2009-01-01

    This document presents the Next Generation Nuclear Plant (NGNP) Systems, Subsystems, and Components, establishes a baseline for the current technology readiness status, and provides a path forward to achieve increasing levels of technical maturity.

  19. IAEA-IFRAM joint development of nuclear plant aging management handbooks

    International Nuclear Information System (INIS)

    Yeji Kim, II Soon Hwang

    2015-01-01

    As attentions to the beyond design basis accident have increased following Fukushima accident, the importance of aging management for key system, structure and components of nuclear power plant (NPP) also has been emerged as the key factor for safety and reliability of NPPs. The diverse experiences can serve as the foundation for aging management processes that has major trust in the understanding of materials degradation mechanisms. IAEA and IFRAM have collaborated to produce the nuclear plant aging management handbook in order to help entry-level engineers and graduate students understand big pictures. As the number of old NPPs and new-comer countries have increased, sharing experiences and accumulated knowledge from existing nuclear industries has become increasingly important. From this point of view, the handbook has been completed for publication by IAEA which assembled a panel of international experts to integrate the vast body of background knowledge regarding aging mechanism and the extensive know-hows of expertsinto a primer handbook. The next step is to actively utilize the IAEA handbook to produce its web version. In order to reach the ultimate purpose of nuclear safety culture of IAEA, information the handbook should be made easily accessible ana understandable. In this regards, the conversion of IFRAM handbook to web version is being planned. Careful considerations about how to restructure the content and how to reflect opinions from experts and readers are needed. In addition, this handbook can be extended to a training program for maintenance managers, vendors, research organizations and regulators to assist them in their work on aging management in operating NPPs. Through these two aspects, this IFRAM handbook on aging management of NPPs will support of NPP staff both new-comer and established countries. This paper introduces the purpose, approach, structure, content of already completed IAEA handbook and the status and plan for IFRAM Web

  20. Developments of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2003-03-15

    The objective of this research is to develop an efficient evaluation technology and to investigate applicability of newly-developed technology, such as internet-based cyber platform, to operating power plants. Development of efficient evaluation systems for Nuclear Power Plant components, based on structural integrity assessment techniques, are increasingly demanded for safe operation with the increasing operating period of Nuclear Power Plants. The following five topics are covered in this project: development of assessment method for wall-thinned nuclear piping based on pipe test; development of structural integrity program for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for mam components of NPP; development of internet-based cyber platform and integrity program for primary components of NPP; effect of aging on strength of dissimilar welds.

  1. Nuclear Power Plant Technician

    Science.gov (United States)

    Randall, George A.

    1975-01-01

    The author recognizes a body of basic knowledge in nuclear power plant technoogy that can be taught in school programs, and lists the various courses, aiming to fill the anticipated need for nuclear-trained manpower--persons holding an associate degree in engineering technology. (Author/BP)

  2. CARA Project: development of the advanced ULE fuel element for heavy water nuclear power plants

    International Nuclear Information System (INIS)

    Brasnarof, Daniel O.; Marino, Armando C.; Florido, Pablo C.; Munoz, C.; Bianchi, Daniel R.; Giorgis, Miguel A.

    2006-01-01

    The CARA Project (Spanish acronym of Combustible Avanzado para Reactores Argentinos) is a national fuel element technology development, compatible with our nuclear power plants (Atucha I, Embalse and Atucha II). It takes into account the experience obtained in our nuclear organisations (CNEA-CONUAR-NASA). The goal of the CARA fuel element is the performance improvement for those reactors and the enhancing of their normal operative conditions. The CARA design allows the burnup extension by using 52 rods of the same diameter. Likewise it keeps good thermo-hydraulic behaviour. The fuel bundle can be directly used in nuclear power plants with horizontal channels. By using an additional system it can be installed in the PHWR with vertical channels. The expected profits, by the use of the CARA in our reactors, broadly guaranty the recovery of the fund for its development, due to a reduction of the NPP fuels and back end cost. We estimate a reduction in the generation cost between 20 or 25 % in relation to the present one if we use 0.85 or 0.90% SEU (Slightly Enriched Uranium). The use of the CARA fuel in our reactors will also reduce the amount of spent fuel to be treated. The shortening could be between 17 to 27 % in Atucha I in relation to the present ULE (0.85%), between 38 to 46% for Embalse, and 45 to 53% for Atucha II. The mechanical behaviour and hydraulic compatibility have been verified. Several CARA prototypes were fabricated with a new design of the end plate and with new processes for the welding for the rods. We present in this paper the current status of the CARA fuel element development. (author) [es

  3. The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

    2012-01-01

    This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

  4. Development of CAP code for nuclear power plant containment: Lumped model

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Soon Joon, E-mail: sjhong90@fnctech.com [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul [FNC Tech. Co. Ltd., Heungdeok 1 ro 13, Giheung-gu, Yongin-si, Gyeonggi-do 446-908 (Korea, Republic of); Ha, Sang Jun [Central Research Institute, Korea Hydro & Nuclear Power Company, Ltd., 70, 1312-gil, Yuseong-daero, Yuseong-gu, Daejeon 305-343 (Korea, Republic of)

    2015-09-15

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP.

  5. Development of CAP code for nuclear power plant containment: Lumped model

    International Nuclear Information System (INIS)

    Hong, Soon Joon; Choo, Yeon Joon; Hwang, Su Hyun; Lee, Byung Chul; Ha, Sang Jun

    2015-01-01

    Highlights: • State-of-art containment analysis code, CAP, has been developed. • CAP uses 3-field equations, water level oriented upwind scheme, local head model. • CAP has a function of linked calculation with reactor coolant system code. • CAP code assessments showed appropriate prediction capabilities. - Abstract: CAP (nuclear Containment Analysis Package) code has been developed in Korean nuclear society for the analysis of nuclear containment thermal hydraulic behaviors including pressure and temperature trends and hydrogen concentration. Lumped model of CAP code uses 2-phase, 3-field equations for fluid behaviors, and has appropriate constitutive equations, 1-dimensional heat conductor model, component models, trip and control models, and special process models. CAP can run in a standalone mode or a linked mode with a reactor coolant system analysis code. The linked mode enables the more realistic calculation of a containment response and is expected to be applicable to a more complicated advanced plant design calculation. CAP code assessments were carried out by gradual approaches: conceptual problems, fundamental phenomena, component and principal phenomena, experimental validation, and finally comparison with other code calculations on the base of important phenomena identifications. The assessments showed appropriate prediction capabilities of CAP

  6. Atucha I nuclear power plant transients analysis

    International Nuclear Information System (INIS)

    Castano, J.; Schivo, M.

    1987-01-01

    A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

  7. ASME development of risk-based inspection guidelines for nuclear power plants

    International Nuclear Information System (INIS)

    Gore, B.F.; Balkey, K.R.

    1992-08-01

    A methodology has been developed for use in preparing guidelines for the in-service inspection of nuclear power plant pressure boundary and structural components. This methodology is a further development of a general methodology previously published by the task force for application to any industry. It is unique in that it utilizes probabilistic risk assessment (PRA) information to improve the quantification of risks associated with component ruptures. A procedure has also been recommended for using the resulting quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Characteristics which an inspection strategy must possess in order to maintain target rupture probabilities are being determined by structural risk and reliability analysis (SRRA) calculations. The major features of this methodology are described in this paper, and example applications are briefly discussed. Plans for additional work in developing and applying the methodology are described

  8. Recent developments in methodology for dynamic qualification of nuclear plant equipment

    International Nuclear Information System (INIS)

    Kana, D.D.; Pomerening, D.J.

    1984-01-01

    Dynamic qualification of nuclear plant electrical and mechanical equipment is performed basically under guidelines given in IEEE Standards 323 and 344, and a variety of NRC regulatory guides. Over the last fifteen years qualification methodology prescribed by these documents has changed significantly as interpretations, equipment capability, and imagination of the qualification engineers have progressed. This progress has been sparked by concurrent NRC and industry sponsored research programs that have identified anomalies and developed new methodologies for resolving them. Revisions of the standards have only resulted after a lengthy debate of all such new information and subsequent judgment of its validity. The purpose of this paper is to review a variety of procedural improvements and developments in qualification methodology that are under current consideration as revisions to the standards. Many of the improvements and developments have resulted from recent research programs. All are very likely to appear in one type of standard or another in the near future

  9. Problems of implementation of the first nuclear power plant in developing countries with particular reference to Egypt

    International Nuclear Information System (INIS)

    Effat, K.E.; El-Fouly, M.F.; El-Saiedi, A.F.

    1977-01-01

    Implementation of nuclear power plants generally passes through three main phases; namely, the planning and decision to go to nuclear power, precontractual phase, and the construction phase. Each has its own particular problems that are usually encountered in developing countries, especially in implementation of the first nuclear plant. The decision to go nuclear is tied to the local and international energy market, both in the present and as projected. In the precontractual phase, which involves preparing plant specifications, invitation to tenderers, bids evaluation and negotiations with the selected vendor, problems of a different nature are encountered. Availability of qualified management, engineering, and technical personnel required to participate in design reviews, quality assurance, management, and scheduling in the construction phase is a major problem in developing countries. The paper discusses all these problems with special emphasis on the local conditions in Egypt. Our experience dealing with these issues could be of value to other developing countries

  10. NUCLEAR POWER PLANT

    Science.gov (United States)

    Carter, J.C.; Armstrong, R.H.; Janicke, M.J.

    1963-05-14

    A nuclear power plant for use in an airless environment or other environment in which cooling is difficult is described. The power plant includes a boiling mercury reactor, a mercury--vapor turbine in direct cycle therewith, and a radiator for condensing mercury vapor. (AEC)

  11. Development of Design Information Template for Nuclear Power Plants for Electromagnetic Pulse (EMP) Effect Analysis

    International Nuclear Information System (INIS)

    Kim, Minyi; Ryu, Hosan; Ye, Songhae; Lee, Euijong

    2016-01-01

    An electromagnetic pulse (EMP) is a transient electromagnetic shock wave that has powerful electric and magnetic fields that can destroy electronic equipment. It is generally well-known that EMPs can cause the malfunction and disorder of electronic equipment and serious damages to electric power systems and communication networks. Research is being carried out to protect nuclear power plants (NPPs) from EMP threats. Penetration routes of EMPs can be roughly categorized into two groups, radioactivity and conductivity. The radioactive effect refers to an impact transmitted to the ground from high-altitude electromagnetic pulses (HEMP). Such an impact may affect target equipment through the point of entry (POE) of the concrete structure of an NPP. The conductive effect refers to induced voltage or current coupled to the NPPs cable structure. The induced voltage and current affect the target equipment via connected cables. All these factors must be considered when taking into account EMP effect analysis for NPPs. To examine all factors, it is necessary to fully understand the schemes of NPPs. This paper presents a four type design information template that can be used to analyze the EMP effect in operating nuclear power plants. In order to analyze of the effects of EMPs on operating NPPs, we must consider both the conductive and radioactive effects on the target (system, equipment, structure). For these reasons, not only the equipment information, but also the information about the structure and the external penetration will be required. We are developing a design information template for robust nuclear design information acquisition. We expect to develop a block diagram on the basis of the template

  12. Development of Design Information Template for Nuclear Power Plants for Electromagnetic Pulse (EMP) Effect Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minyi; Ryu, Hosan; Ye, Songhae; Lee, Euijong [KNHP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    An electromagnetic pulse (EMP) is a transient electromagnetic shock wave that has powerful electric and magnetic fields that can destroy electronic equipment. It is generally well-known that EMPs can cause the malfunction and disorder of electronic equipment and serious damages to electric power systems and communication networks. Research is being carried out to protect nuclear power plants (NPPs) from EMP threats. Penetration routes of EMPs can be roughly categorized into two groups, radioactivity and conductivity. The radioactive effect refers to an impact transmitted to the ground from high-altitude electromagnetic pulses (HEMP). Such an impact may affect target equipment through the point of entry (POE) of the concrete structure of an NPP. The conductive effect refers to induced voltage or current coupled to the NPPs cable structure. The induced voltage and current affect the target equipment via connected cables. All these factors must be considered when taking into account EMP effect analysis for NPPs. To examine all factors, it is necessary to fully understand the schemes of NPPs. This paper presents a four type design information template that can be used to analyze the EMP effect in operating nuclear power plants. In order to analyze of the effects of EMPs on operating NPPs, we must consider both the conductive and radioactive effects on the target (system, equipment, structure). For these reasons, not only the equipment information, but also the information about the structure and the external penetration will be required. We are developing a design information template for robust nuclear design information acquisition. We expect to develop a block diagram on the basis of the template.

  13. Development of basic system for sensor calibration support in nuclear power plants

    International Nuclear Information System (INIS)

    Kusumi, Naohiro; Ohga, Yukiharu; Fukuda, Mitsuko; Ishizaki, Yuuichi; Koyama, Mikio; Maeda, Akihiko

    2004-01-01

    It is strongly desirable to reduce maintenance costs and shorten the time of periodic inspections in nuclear power plants. Therefore, it is important to reduce the amount of maintenance work during the inspection. In Japan, sensor calibration is usually performed at every periodic inspection, and the sensor calibration requires a large amount of work. A system for sensor calibration support has been developed to reduce sensor calibration work. The system is composed of two subsystems: a statistical analysis subsystem and a drift detection subsystem, as well as a human-machine interface, which offers support information. The statistical analysis subsystem supports the decision of the sensor calibration intervals based on the statistical analysis of sensor calibration data. There is the possibility that sensor drift increases beyond an allowance value before the sensor calibration intervals determined by the statistical analysis subsystem because of malfunctions, etc. To cope with this, the drift detection subsystem detects the sensor drift online during the plant operation. By combining the statistical analysis subsystem and the drift detection subsystem, a reliable sensor calibration support system is realized. The basic system composed of two subsystems was developed and evaluated using real plant data. The results showed that the sensor calibration intervals can be extended beyond current intervals and that the system is capable of detecting the sensor drift online. (author)

  14. Development of an investment decision tool for aged nuclear power plants under uncertainty

    International Nuclear Information System (INIS)

    Takashima, Ryuta; Nagano, Koji

    2010-01-01

    The nuclear power generating stock in Japan, which provides 30% or more of the nationwide electricity supply in recent years, is heading into the era of plant aging as those units over 40 years of operation emerge in 2010 and onward. Under accumulating uncertainties surrounding investment decisions, one of the key questions is how best to manage the existing capacity by choosing a wisest option among dismantling, refurbishment, and replacement. This report attempts to develop a methodology to analyze investment decision on aged nuclear units based on real-option analysis, and then presents illustrative simulation runs to demonstrate functions of the numerical models. Major findings are summarized as follows: a) When compared with the conventional net present value (NPV) method, the proposed model is capable to deal with uncertainty explicitly by taking volatility in the formula. Also, for the replacement option, one can choose optimal timings of the dismantling the new plant independently. b) As the profitability increases, the model suggests different investment option as optimal, starting from dismantling to replacement, further to refurbishment, and again to replacement as the most basic pattern. Additionally, where two of the strategies have similar values, postponement of both decisions can be optimal. Through these exercises, the proposed methodology proved itself capable for raising valuable implications to investment decisions in managing the production fleet. (author)

  15. Development of management system for nuclear power plant maintenance and information

    International Nuclear Information System (INIS)

    Nomoto, Toshihiro

    1997-01-01

    In Chubu Electric Power Co., Inc., in order to make maintenance works efficient and improve the management, 'Management system for nuclear power plant maintenance and information' was developed, and its operation on full scale was begun in Hamaoka Nuclear Power Station in October, 1996. This system is composed of equipment management system, maintenance and repair management system and work management system. As the features of the system, the dispersion of functions by client/server method, the installation of the server machines for exclusive use in power stations, the adoption of optical communication network and the ensuring of reliability by the doubled system are mentioned. This system is the function dispersion system by client/server method utilizing the in-plant LAN, and has two server computers with double hot standby constitution. The main functions of three subsystems are described. These three subsystems and piping and instrumentation chart management system and whole company work budget system are connected so as to make dealing works quick and efficient. Hereafter, by reflecting the opinions of the users through the operation, further efficient works are to be aimed at. (K.I.)

  16. KWU Nuclear Plant Analyzer

    International Nuclear Information System (INIS)

    Bennewitz, F.; Hummel, R.; Oelmann, K.

    1986-01-01

    The KWU Nuclear Plant Analyzer is a real time engineering simulator based on the KWU computer programs used in plant transient analysis and licensing. The primary goal is to promote the understanding of the technical and physical processes of a nuclear power plant at an on-site training facility. Thus the KWU Nuclear Plant Analyzer is available with comparable low costs right at the time when technical questions or training needs arise. This has been achieved by (1) application of the transient code NLOOP; (2) unrestricted operator interaction including all simulator functions; (3) using the mainframe computer Control Data Cyber 176 in the KWU computing center; (4) four color graphic displays controlled by a dedicated graphic computer, no control room equipment; and (5) coupling of computers by telecommunication via telephone

  17. Nuclear Power Plants (Rev.)

    Energy Technology Data Exchange (ETDEWEB)

    Lyerly, Ray L.; Mitchell III, Walter [Southern Nuclear Engineering, Inc.

    1973-01-01

    Projected energy requirements for the future suggest that we must employ atomic energy to generate electric power or face depletion of our fossil-fuel resources—coal, oil, and gas. In short, both conservation and economic considerations will require us to use nuclear energy to generate the electricity that supports our civilization. Until we reach the time when nuclear power plants are as common as fossil-fueled or hydroelectric plants, many people will wonder how the nuclear plants work, how much they cost, where they are located, and what kinds of reactors they use. The purpose of this booklet is to answer these questions. In doing so, it will consider only central station plants, which are those that provide electric power for established utility systems.

  18. Partner of nuclear power plants

    International Nuclear Information System (INIS)

    Gribi, M.; Lauer, F.; Pauli, W.; Ruzek, W.

    1992-01-01

    Sulzer, the Swiss technology group, is a supplier of components and systems for nuclear power plants. Important parts of Swiss nuclear power stations, such as containments, reactor pressure vessels, primary pipings, are made in Winterthur. Sulzer Thermtec AG and some divisions of Sulzer Innotec focus their activities on servicing and backfitting nuclear power plants. The European market enjoys priority. New types of valves or systems are developed as economic solutions meeting more stringent criteria imposed by public authorities or arising from operating conditions. (orig.) [de

  19. Development of a standard communication protocol for an emergency situation management in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Cheol, E-mail: charleskim@kaeri.re.k [Integrated Risk Assessment Center, Korea Atomic Energy Research Institute, 150, Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Park, Jinkyun; Jung, Wondea [Integrated Risk Assessment Center, Korea Atomic Energy Research Institute, 150, Deokjin-dong, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Kim, Hanjeom; Kim, Yoon Joong [YGN Nuclear Power Division Training Center, Korea Hydro and Nuclear Power Company, 517 Kyemari, Hongnong-eup, Yeongkwang-gun, Chonnam 513-880 (Korea, Republic of)

    2010-06-15

    Correct communication between main control room (MCR) operators is an important factor in the management of emergency situations in nuclear power plants (NPPs). For this reason, a standard communication protocol for the management of emergency situations in NPPs has been developed, with the basic direction of enhancing the safety of NPPs and the standardization of communication protocols. To validate the newly developed standard communication protocol, validation experiments with 10 licensed NPP MCR operator teams was performed. From the validation experiments, it was found that the use of the standard communication protocol required more time, but it can contribute to the enhancement of the safety of NPPs by an operators' better grasp of the safety-related parameters and a more efficient and clearer communication between NPP operators, while imposing little additional workloads on the NPP MCR operators. The standard communication protocol is expected to be used to train existing NPP MCR operators without much aversion, as well as new operators.

  20. Models development for natural circulation and its transition process in nuclear power plant

    International Nuclear Information System (INIS)

    Yu Lei; Cai Qi; Cai Zhangsheng; Xie Haiyan

    2008-01-01

    On the basis of nuclear power plant (NPP) best-estimate transient analysis code RELAP5/MOD3, the point reactor kinetics model in RELAP5/MOD3 was replaced by the two-group, 3-D space and time dependent neutron kinetic model, in order to exactly analyze the responses of key parameters in natural circulation and its transition process considering the reactivity feedback. The coupled model for three-dimensional physics and thermohydraulics was established and corresponding computing code was developed. Using developed code, natural circulation of NPP and its transiton process were calculated and analyzed. Compared with the experiment data, the calculated results show that its high precise avoids the shortage that the point reactor equation can not reflect the reactivity exactly. This code can be a computing and analysis tool for forced circulation and natural circulation and their transitions. (authors)

  1. Status of ASME risk-based inspection guidelines development for nuclear power plants

    International Nuclear Information System (INIS)

    Gore, B.F.; Vo, T.V.; Balkey, K.R.

    1991-01-01

    The general methodology published by the task force is being applied to develop guidelines for the inspection of nuclear power plant components. The methodology has been expanded to include the use of information from probabilistic risk assessments to improve the quantification of risks associated with component ruptures. In addition, a procedure has been recommended for using these quantified risk estimates to determine target component rupture probability values to be maintained by inspection activities. Finally, the method for determining the characteristics which an inspection strategy must possess in order to maintain these target rupture probabilities has been described. This paper presents the major features of this methodology, along with a discussion of the status of efforts to develop and to apply it further

  2. Development of a measure of training program effectiveness for nuclear power plants

    International Nuclear Information System (INIS)

    Connelly, E.M.; Haas, P.; VanHemel, S.B.; Wreathall, J.

    1991-01-01

    The approach taken to develop measures of training program effectiveness recognizes that many subject matter experts (SMEs) with varying viewpoints on training effectiveness are working to insure nuclear power plant (NPP) safety. This paper discusses a model as the effectiveness rating strategies of various SMEs whose ratings are important to training; i.e., build a measure of effectiveness (MOE) representing the rating strategy of a sample of those SMEs. A quantitative model of these assessment strategies permits a broad review of them by ll interested parties and provides the basis for building a broadly accepted training program MOE. Major differences in assessment strategies can be examined to determine whether they are appropriate to the different points of view or are in conflict. This paper describes the basic concepts of building effectiveness measures, the methodology used to build the measures, a summary of the measures developed form NRC and industry sources, and an analysis of the differences among SMEs' measures

  3. Development of an automated operating procedure system using fuzzy colored petri nets for nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Seong, Poong Hyun

    2002-01-01

    In this work, AuTomated Operating Procedure System (ATOPS) is developed. ATOPS is an automation system for operation of a nuclear power plant (NPP) which can monitor signals, diagnose statuses, and generate control actions according to corresponding operating procedures, without any human operator's help. Main functions of ATOPS are anomaly detection function and procedure execution function, but only the procedure execution function is implemented because this work is just the first step. In the procedure execution function, operating procedures of NPP are analyzed and modeled using Fuzzy Colored Petri Nets (FCPN), and executed depending on decision making of the inference engine. In this work, an ATOPS prototype is developed in order to demonstrate its feasibility and it is also validated using FISA-2/WS simulator. The validation is performed for the cases of a loss of coolant accident (LOCA) and a steam generator tube rupture (SGTR). The simulation results show that ATOPS works correctly in the emergency situations

  4. Occupational dose control in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Viktorsson, C.; Lochard, J.; Benedittini, M.; Baum, J.; Khan, T.A.

    1990-01-01

    Reduction in occupational exposure at nuclear power plants is desirable not only in the interest of the health and safety of plant personnel, but also because it enhances the safety and reliability of the plants. This report summarises the current trends of doses to workers at nuclear power plants and the achievements and developments regarding methods for their reduction

  5. Next Generation Nuclear Plant Materials Research and Development Program Plan, Revision 3

    International Nuclear Information System (INIS)

    G.O. Hayner; R.L. Bratton; R.E. Mizia; W.E. Windes; W.R. Corwin; T.D. Burchell; C.E. Duty; Y. Katoh; J.W. Klett; T.E. McGreevy; R.K. Nanstad; W. Ren; P.L. Rittenhouse; L.L. Snead; R.W. Swindeman; D.F. Wlson

    2006-01-01

    This is the 2006 update (Revision 3) of the NGNP Materials Research and Development Program Plan. This law established that the U.S Department of Energy (DOE) Secretary of Energy shall establish a ''Next Generation Nuclear Plant'' (NGNP) project. The NGNP project named in the Act was given the following attributes and guiding principles to manage its development: (1) The NGNP consists of research, development, design (R and DD), construction, and operation of a prototype reactor to generate electricity and hydrogen; (2) The project shall be managed by the DOE Office of Nuclear Energy; (3) The Idaho National Laboratory (INL) shall be the lead DOE laboratory for the NGNP; (4) The INL shall establish collaborations with selected institutions of higher education, other research institutes and international researchers; (5) The INL shall organize an industrial consortium of partners for cost-shared R and DD, construction; (6) The project shall be sited at the INL; (7) The project shall be licensed by the Nuclear Regulatory Commission (NRC) and by July, 2008 the NRC and DOE shall jointly submit a licensing strategy to Congress; (8) The project shall be organized to maximize technical interchange with the nuclear power industry, nuclear power plant construction firms, the chemical process industry and to seek international cooperation, participation and contributions; (9) The Nuclear Energy Research Advisory Committee (NERAC) shall review all program plans for the NGNP; (10) Phase 1 of the project (selection of hydrogen production technology, conduct R and DD and initial design activities) shall be completed no later than September 30, 2011; (11) Phase 2 of the project (continue R and DD, develop final design, apply for a license, construct and start operations) shall be completed by September 30, 2021; and (12) Provision for authorization of appropriations was made. As a result of the direction provided, the INL and the DOE issued an NGNP Preliminary Project Management

  6. Standardization of nuclear power plants in the United States: recent regulatory developments

    International Nuclear Information System (INIS)

    Cowan, B.Z.; Tourtellotte, J.R.

    1992-01-01

    On April 18, 1989, the United States (U.S.) Nuclear Regulatory Commission (NRC) amended the regulations governing the process for licensing nuclear power plants in the United States to provide for issuance of early site permits, standard design certifications and combined construction permits and operating licenses for nuclear power reactors. The new regulations are designed to achieve early resolution of licensing issues and facilitate standardization of nuclear power plants in the United States. The program for design standardization is central to efforts mounted by the U.S. government and industry to ensure that there will be a next generation of nuclear power facilities in the U.S. The most significant changes are provisions for certification of standard designs and for issuance prior to start of construction of combined licenses which incorporate a construction permit and an operating license with conditions. Such certifications and combined licenses must contain tests, inspections and analyses, and acceptance criteria, which are necessary and sufficient to provide reasonable assurance that the facility has been constructed and will operate in accordance with the combined license. A number of significant implementation issues have arisen. In addition a major court case brought by several anti-nuclear groups is pending, challenging NRC authority to issue combined licenses. It is the goal of the U.S. nuclear industry to have the first of the next generation of standardized nuclear power plants ordered, licensed, constructed and on-line by the year 2000. (author)

  7. Development priorities for non-destructive examination of concrete structures in nuclear plant

    International Nuclear Information System (INIS)

    1998-01-01

    The objective of this report is to provide a basis for assessing development priorities for NDE of safety related concrete structures in nuclear plants, taking account of both the benefit and the cost of potential developments in NDE techniques. An OECD/NEA Workshop which considered the requirements for NDE of safety related concrete structures was held in the UK on 12 November 97. NDE techniques have the potential to satisfy at least some of the needs of the nuclear industry. NDE techniques have been used successfully on a variety of reinforced and post-tensioned concrete structures, notably highway and reservoir structures. However, there is limited experience of their use to evaluate typical nuclear safety related structures having thick sections, steel liners or access to one side only. There is a general lack of confidence in the techniques because there is very little independent advice on their applicability, capability, accuracy and reliability. The information obtained by techniques such as RADAR, ultrasonics, stress wave and radiography appears qualitative rather than quantitative and there is concern that NDE procedures lack the necessary qualification to permit their use on safety critical structures. There is no authoritative international guidance or standard for NDE of concrete structures. NDE of concrete structures is often based upon equipment developed for other materials and technologies, eg. examination of steel, evaluation of ground conditions. Other industries are developing equipment specifically for civil engineering applications and at the recent OECD workshop a number of relevant national and European programmes were identified. The nuclear industry maintain its awareness of developments and should seek to influence the development of equipment. The quantification of the capabilities of NDE techniques is seen as a priority area for development. The provision of authoritative documentation in the form of reports and Standards is desirable

  8. The development and implementation of a public information programme at the Kozloduy nuclear power plant, Bulgaria

    International Nuclear Information System (INIS)

    Parker, Keith

    1998-01-01

    This paper proposes to examine the aims, approach and evaluation of a Public Information Programme currently being undertaken by the British Nuclear Industry Forum. The Kozloduy Nuclear Power Plant has awarded a contract to the British Nuclear Industry Forum to carry out a Public Information Programme with the objective of informing the people of Bulgaria about the content of the Nuclear Safety Account Grant Agreement, the role of the NSA, the role of EBRD and the donor countries in providing the necessary funding, the safety improvement programme being undertaken and the progress being made. The overall aim of the Programme is to build a consensus among energy consumers of the principal issues to be faced in the development and implementation of a Bulgarian energy strategy. There are a number of key elements of the Programme which include: 1) To carry out a communications audit in order to assess current structures and methods of information dissemination. This is to establish the Programme's requirements and provide a detailed operational plan. 2) To assess public opinion and general levels of awareness of the general public, workforce, press, government and industry regarding Kozloduy in order to establish priority messages and create and produce the necessary information material. 3) To review the plant's existing information centre and to provide training and support to facilitate the handling of public and press enquiries and also presentation training for the centre's personnel. 4) To create bespoke outreach programmes which will inform local and national Government, the workforce and local communities, women and schools of safety improvement processes. 5) To implement a media programme which will enable staff at the plant to deal effectively with inquiries and to enable them to anticipate media interest in a range of issues including the safety upgrade. 6) To organise a workshop on issue management so that participants can create and communicate an issue

  9. Loviisa nuclear power plant analyzer

    International Nuclear Information System (INIS)

    Porkholm, K.; Nurmilaukas, P.; Tiihonen, O.; Haenninen, M.; Puska, E.

    1992-12-01

    The APROS Simulation Environment has been developed since 1986 by Imatran Voima Oy (IVO) and the Technical Research Centre of Finland (VTT). It provides tools, solution algorithms and process components for use in different simulation systems for design, analysis and training purposes. One of its main nuclear applications is the Loviisa Nuclear Power Plant Analyzer (LPA). The Loviisa Plant Analyzer includes all the important plant components both in the primary and in the secondary circuits. In addition, all the main control systems, the protection system and the high voltage electrical systems are included. (orig.)

  10. Organizing nuclear power plant operation

    International Nuclear Information System (INIS)

    Adams, H.W.; Rekittke, K.

    1987-01-01

    With the preliminary culmination in the convoy plants of the high standard of engineered safeguards in German nuclear power plants developed over the past twenty years, the interest of operators has now increasingly turned to problems which had not been in the focus of attention before. One of these problems is the organization of nuclear power plant operation. In order to enlarge the basis of knowledge, which is documented also in the rules published by the Kerntechnischer Ausschuss (Nuclear Technology Committee), the German Federal Minister of the Interior has commissioned a study of the organizational structures of nuclear power plants. The findings of that study are covered in the article. Two representative nuclear power plants in the Federal Republic of Germany were selected for the study, one of them a single-unit plant run by an independent operating company in the form of a private company under German law (GmbH), the other a dual-unit plant operated as a dependent unit of a utility. The two enterprises have different structures of organization. (orig.) [de

  11. Development of the methodology for application of revised source term to operating nuclear power plants in Korea

    International Nuclear Information System (INIS)

    Kang, M.S.; Kang, P.; Kang, C.S.; Moon, J.H.

    2004-01-01

    Considering the current trend in applying the revised source term proposed by NUREG-1465 to the nuclear power plants in the U.S., it is expected that the revised source term will be applied to the Korean operating nuclear power plants in the near future, even though the exact time can not be estimated. To meet the future technical demands, it is necessary to prepare the technical system including the related regulatory requirements in advance. In this research, therefore, it is intended to develop the methodology to apply the revised source term to operating nuclear power plants in Korea. Several principles were established to develop the application methodologies. First, it is not necessary to modify the existing regulations about source term (i.e., any back-fitting to operating nuclear plants is not necessary). Second, if the pertinent margin of safety is guaranteed, the revised source term suggested by NUREG-1465 may be useful to full application. Finally, a part of revised source term could be selected to application based on the technical feasibility. As the results of this research, several methodologies to apply the revised source term to the Korean operating nuclear power plants have been developed, which include: 1) the selective (or limited) application to use only some of all the characteristics of the revised source term, such as release timing of fission products and chemical form of radio-iodine and 2) the full application to use all the characteristics of the revised source term. The developed methodologies are actually applied to Ulchin 9 and 4 units and their application feasibilities are reviewed. The results of this research are used as either a manual in establishing the plan and the procedure for applying the revised source term to the domestic nuclear plant from the utility's viewpoint; or a technical basis of revising the related regulations from the regulatory body's viewpoint. The application of revised source term to operating nuclear

  12. Development of halogen-free flame-retardant cable for nuclear power plant. 2

    International Nuclear Information System (INIS)

    Matsumoto, Tetsuo; Kimura, Hitoshi; Ishii, Nobuhisa

    1997-01-01

    Halogen-free flame-retardant cables were developed for PWR nuclear power stations. It was confirmed that the developed cables possess flame retardant property, corrosion resistance, low toxicity and low smoke generation, and withstand the normal operation in the environment in PWR containment vessels for 60 years and loss of coolant accident. In the advancement of LWR technology, it is important to improve the reliability of machinery and equipment, to extend the period of continuous operation, to optimize the operation cycle and to improve the maintenance of plants. By improving halogen-free flame-retardant material and applying it to the cables for nuclear power stations, it can contribute to the above purposes. The required characteristics of these cables are explained, and the targets of development are power cables, control cables, instrumentation cables and insulated wires which do not contain halogen. The basic material is polyolefin, in which flame retardant magnesium hydroxide and the agent for improving radiation resistance are mixed. The corrosive property and toxicity of gases, smoke generation and the prevention of spread of flame when the cables burn and the durability in environment were evaluated. (K.I.)

  13. A review for identification of initiating events in event tree development process on nuclear power plants

    International Nuclear Information System (INIS)

    Riyadi, Eko H.

    2014-01-01

    Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logic model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events

  14. A review for identification of initiating events in event tree development process on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Riyadi, Eko H., E-mail: e.riyadi@bapeten.go.id [Center for Regulatory Assessment of Nuclear Installation and Materials, Nuclear Energy Regulatory Agency (BAPETEN), Jl. Gajah Mada 8 Jakarta 10120 (Indonesia)

    2014-09-30

    Initiating event is defined as any event either internal or external to the nuclear power plants (NPPs) that perturbs the steady state operation of the plant, if operating, thereby initiating an abnormal event such as transient or loss of coolant accident (LOCA) within the NPPs. These initiating events trigger sequences of events that challenge plant control and safety systems whose failure could potentially lead to core damage or large early release. Selection for initiating events consists of two steps i.e. first step, definition of possible events, such as by evaluating a comprehensive engineering, and by constructing a top level logic model. Then the second step, grouping of identified initiating event's by the safety function to be performed or combinations of systems responses. Therefore, the purpose of this paper is to discuss initiating events identification in event tree development process and to reviews other probabilistic safety assessments (PSA). The identification of initiating events also involves the past operating experience, review of other PSA, failure mode and effect analysis (FMEA), feedback from system modeling, and master logic diagram (special type of fault tree). By using the method of study for the condition of the traditional US PSA categorization in detail, could be obtained the important initiating events that are categorized into LOCA, transients and external events.

  15. Development of resilience evaluation method for nuclear power plants. Part 3. Study of evaluation method and applicability of resilience index

    International Nuclear Information System (INIS)

    Suzuki, Masaaki; Demachi, Kazuyuki; Miyano, Hiroshi

    2017-01-01

    We have developed a new index, called the resilience index, that evaluates the dynamic stability of the system safety of a nuclear power plant during a severe accident by considering the ability to recover system safety functions that have become lost in the situation. In this paper, a detailed evaluation procedure for the resilience index is described. The system safety of a pressurized water reactor plant during a severe accident is then assessed according to the resilience index in order to discuss the applicability of the index. We find that the resilience index successfully represents management capability and, therefore, the resilience capability of a nuclear power plant. (author)

  16. Development of resilience evaluation method for nuclear power plants. Part 3. Study on evaluation method and applicability of resilience index

    International Nuclear Information System (INIS)

    Suzuki, Masaaki; Demachi, Kazuyuki; Murakami, Kenta

    2016-01-01

    We have developed a new index, called the resilience index, that evaluates the dynamic stability of the system safety of a nuclear power plant during a severe accident by considering the ability to recover system safety functions that have become lost in the situation. In this paper, a detailed evaluation procedure for the resilience index is described. The system safety of a PWR plant during a severe accident is then assessed according to the resilience index in order to discuss the applicability of the index. We find that the resilience index successfully represents management capability and, therefore, the resilience capability of a nuclear power plant. (author)

  17. Development of resilience evaluation method for nuclear power plant. Part 1. Proposal of resilience index for assessment of safety of nuclear power plant under severe accident

    International Nuclear Information System (INIS)

    Demachi, Kazuyuki; Suzuki, Masaaki; Itoi, Tatsuya

    2016-01-01

    In this research, a new index 'The Resilience Index' was proposed to evaluate the capability of nuclear power plant to recover from the situation of safety function lost. Three elements assumed to evaluate the resilience index are the achievement rate, necessary time, and probability of success of each accident management activity. The resilience index is expected to visualize the improvement of safety of each nuclear power plant against severe accidents. (author)

  18. Maintenance Support System Development for Maintenance Personnel in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Yim, Ho Bin; Kim, In; Lee, Seung Min; Seong, Poong Hyun

    2009-01-01

    As plants or factories get old, the number of maintenance activities increases, and so does the necessity of maintenance itself. Industries have adopted the latest technologies such as RFID, wireless sensor networking, and augmented reality to develop systems which increase human performances as well as productivity. Due to safety related concerns, nuclear power industries have been reluctant to use these fancy technologies. However, from the economics point of view, it is necessary to take advantages of novel technologies. This empirical study considers not only ergonomics view, but also the economics view when building maintenance support systems. This paper introduces brief features of five subsystems which were built to reduce human errors and advantages when each system successfully reduces those human errors

  19. Development of general-purpose software to analyze the static thermal characteristic of nuclear power plant

    International Nuclear Information System (INIS)

    Nakao, Yoshinobu; Koda, Eiichi; Takahashi, Toru

    2009-01-01

    We have developed the general-purpose software by which static thermal characteristic of the power generation system is analyzed easily. This software has the notable features as follows. It has the new algorithm to solve non-linear simultaneous equations to analyze the static thermal characteristics such as heat and mass balance, efficiencies, etc. of various power generation systems. It has the flexibility for setting calculation conditions. It is able to be executed on the personal computer easily and quickly. We ensured that it is able to construct heat and mass balance diagrams of main steam system of nuclear power plant and calculate the power output and efficiencies of the system. Furthermore, we evaluated various heat recovery measures of steam generator blowdown water and found that this software could be a useful operation aid for planning effective changes in support of power stretch. (author)

  20. Development of failure diagnosis method based on transient information of nuclear power plant

    International Nuclear Information System (INIS)

    Washio, Takashi; Kitamura, Masaharu; Sugiyama, Kazusuke

    1987-01-01

    This paper proposes a new method of failure diagnosis of nuclear power plant (NPP). Transient behavior of the NPP includes ample failure information even for a limited period of time from the failure onset. We tried to develop a diagnosis system with high capability of identifying the failure cause and of estimating failure severeness. The Walsh function transformation of transient time series data and the reduction of the Walsh coefficients into ternary valued amplitude indicators were utilized to extract the essential characteristics of failure. The correspondences of the transient characteristics and causes were summarized in a failure symptom database. A method of ternary tree search using an information measure as a heuristic strategy was adopted to conduct the efficient retrieval of failure causes in the database. Through numerical experiments using a simulation model of a NPP, the diagnostic capability of the system was proved to be satisfactory. (author)

  1. Development of mobile manipulator for maintenance work in containment vessels of nuclear power plants

    International Nuclear Information System (INIS)

    Omichi, Takeo; Nishihara, Masatoshi; Hosaka, Shigetaka; Nakayama, Junji; Sato, Masatoshi; Ishida, Michiyasu

    1985-01-01

    The teleoperation system with robot is described for in the containment vessels of nuclear power plants. We have developed a high level robot system as the practical use level. The robot is designed to execute the locomotions and manipulations required for closing and opening the valve, tightening the bolt and others. The robot consists of a locomotor with four legs and two driving wheels, an articulated manipulator with seven joints, and an ITV arm with stereo-camera. The size of the robot is small, that is about 500 mm in length, 500 mm in width, 1200 mm in height and 420 kg in weight. The robot can be operated in a hostile environment, which has a 10 6 R gamma ray dose, 70 deg C temperature, 100 % relative humidity. We have added an advanced control method in order to reduce the operator's load. Also, an interlock and a fail-safe control are installed in the robot system. (author)

  2. Maintenance Support System Development for Maintenance Personnel in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Yim, Ho Bin; Kim, In; Lee, Seung Min; Seong, Poong Hyun [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2009-05-15

    As plants or factories get old, the number of maintenance activities increases, and so does the necessity of maintenance itself. Industries have adopted the latest technologies such as RFID, wireless sensor networking, and augmented reality to develop systems which increase human performances as well as productivity. Due to safety related concerns, nuclear power industries have been reluctant to use these fancy technologies. However, from the economics point of view, it is necessary to take advantages of novel technologies. This empirical study considers not only ergonomics view, but also the economics view when building maintenance support systems. This paper introduces brief features of five subsystems which were built to reduce human errors and advantages when each system successfully reduces those human errors.

  3. Development of inspection technology for inner wall pipe of aging nuclear power plant

    International Nuclear Information System (INIS)

    Ito, Fuyumi; Nishimura, Akihiko

    2013-01-01

    Careful inspection should be paid on aging nuclear power plants. Due to the Fukushima BWR accident, more advanced inspection techniques are now requested in Japan. To find SCC along welded sections by Ultrasonic Testing or Eddy Current Testing is difficult due to the low S/N. Here we propose to apply Magnetic particle Testing (MT) on the inspection. MT uses magnetic particles uniting fluorescent pigment. It is a weak point of MT that uniting particles and pigment is breakable. To extend the lifetime, we developed unique capsule for the magnetic particle to coexist with fluorescent pigment. In addition, Laser-Induced Breakdown Spectroscopy (LIBS) used for the laser cleaning of materials, is reported in this paper as a preliminary experiment. The intensity of 621nm peak gradually decreases over time. This result will become a measure of the degree of oxide layer removal. (author)

  4. Transfer system development for a remote inspection robot in nuclear power plants

    International Nuclear Information System (INIS)

    Mizuno, M.; Ohnuma, M.; Hamada, K.; Mizutani, T.; Shimada, A.; Segawa, M.; Kubo, K.

    1984-01-01

    A remote operated robot system has been developed for inspection inside the primary containment vessel (PCV) of nuclear power plants. This system consists of an inspection vehicle, a monorail driving system, a signal transmission system, a power supply system and an operator console.. The system has two main features. First is that the operator can transfer the vehicle at any time from outside the PCV to inside or vice versa through a personnel airlock. The second feature is that the vehicle can be transported from one inspection route to another route at junction points. A prototype inspection robot system was fabricated on a trial basis. Running and inspection performances were confirmed utilizing actual size test apparatus

  5. Development of automated equipment for reduction of personnel radiation exposure in nuclear power plants

    International Nuclear Information System (INIS)

    Ogushi, Akira; Fujii, Masaaki; Mizuno, Katsuhiro.

    1976-01-01

    Described are a mobile remote inspection system and an automatic analyzer for radioactive nuclides in reactor coolant now being developed as a means of reducing personnel radiation exposure in nuclear power plants. In the mobile remote inspection system ''TELEPAT'', a self-propelled vehicle equipped with a thermometer, accelerometer, microphone, ionization chamber, etc. is remote operated from the main control room to inspect the equipment in the reactor building. The automatic analyzer for radioactive nuclides in reactor coolant automates the series of operations ranging from sampling of reactor coolant to measurement of radioactivity and analyses of measured data, with a view to saving labor in radioactivity analysis work while reducing exposure of personnel to radiation. (auth.)

  6. Development of a system to support welding design of nuclear power plants

    International Nuclear Information System (INIS)

    Kataoka, S.; Okamoto, A.

    1991-01-01

    This paper describes about an expert system that supports engineers to make welding design for nuclear power plants. In the welding design, engineers must consider a lot of things; the weldability of the material, the weld deformation, the residual stress predicted, the strength of the weld joint, the configuration of the structure and so on. For easy consultation, the computer system that can provide advice and information about the welding technology is desirable to the engineers. The system was developed on a personal computer of Macintosh utilizing a card type data base, HyperCard. It supports an engineer in the following tasks. a. Advice using the instance of troubles caused by the welding design. b. Tutorial instruction for the knowledge of welding. c. Recommendation for the welding design. d. Data display of the base metal and the weld. (author)

  7. Nuclear plant life extension

    International Nuclear Information System (INIS)

    Negin, C.A.

    1989-01-01

    The nuclear power industry's addressing of life extension is a natural trend in the maturation of this technology after 20 years of commercial operation. With increasing emphasis on how plants are operated, and less on how to build them, attention is turning on to maximizing the use of these substantial investments. The first studies of life extension were conducted in the period from 1978 and 1982. These were motivated by the initiation, by the Nuclear Regulatory Commission (NRC), of studies to support decommissioning rulemaking. The basic conclusions of those early studies that life extension is feasible and worth pursuing have not been changed by the much more extensive investigations that have since been conducted. From an engineering perspective, life extension for nuclear plants is fundamentally the same as for fossil plants

  8. Development of ultrasonic testing technique to inspect containment liners embedded in concrete on nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ishida, H.; Kurozumi, Y. [Inst. of Nuclear Safety System, Incorporated, Mihama, Fukui (Japan); Kaneshima, Y. [The Kansai Electric Power Company, Inc., Mihama, Fukui (Japan)

    2004-07-01

    The purpose of this study is development of ultrasonic testing technique to inspect containment liners embedded in concrete on nuclear power plants. Integrity of containment liners on nuclear power plants can be secured by suitable present operation and maintenance. Furthermore, non-destructive testing technique to inspect embedded liners will ensure the integrity of the containment further. In order to develop the non-destructive testing technique, ultrasonic transducers were made newly and ultrasonic testing data acquisition and evaluation were carried out by using a mock-up. We adopted the surface shear horizontal (SH) wave, low frequency (0.3-0.5MHz), to be able to detect an echo from a defect against attenuation of ultrasonic waves due to long propagation in the liners and dispersion into concrete. We made transducers with three large active elements (40mm x 40mm) in a line which were equivalent to a 120mm width active element. Artificial hollows, {phi}200mm - 19mm depth (1/2thickness) and {phi}200mm - 9.5mm depth (1/4thickness), were made on a surface of a mock-up: carbon steel plate, 38mm thickness, 2,000mm length, 1000mm width. The surfaces of the plate were covered with concrete in order to simulate liners embedded in concrete. As a result of the examinations, the surface SH transducers could detect clearly the echo from the hollows at a distance of 1500mm. We evaluate that the newly made surface SH transducers with three elements have ability of detection of defects such as corrosion on the liners embedded in concrete. (author)

  9. Development of seismic damage assessment system for nuclear power plant structures in Korea

    International Nuclear Information System (INIS)

    Hyun, Chang-Hun; Lee, Sung-Kyu; Choi, Kang-Ryoung; Koh, Hyun-Moo; Cho, HoHyun

    2003-01-01

    A seismic damage assessment system that analyses in real-time the actual seismic resistance capacity and the damage level of power plant structures has been developed. The system consists of three parts: a 3-D inelastic seismic analysis, a damage assessment using a damage index based on the previous 3-D analysis, and a 3-D graphic representation. PSC containment structures are modelled by finite shell elements using layered method and analysis is performed by means of time history inelastic seismic analysis method, which takes into account material nonlinearities. HHT-α, one kind of direct integration method, is adopted for the seismic analysis. Two damage indices at finite element and structural levels are applied for the seismic damage assessment. 3-D graphical representation of dynamic responses and damage index expedites procedure for evaluating the damage level. The developed system is now being installed at the Earthquake Monitoring Center of KINS (Korea Institute of Nuclear Safety) to support site inspections after an earthquake occurrence, and decisions about effective emergency measures, repair and operations of the plant. (author)

  10. Development of an advanced human-machine interface for next generation nuclear power plants

    International Nuclear Information System (INIS)

    Chang, Soon Heung; Choi, Seong Soo; Park, Jin Kyun; Heo, Gyunyoung; Kim, Han Gon

    1999-01-01

    An advanced human-machine interface (HMI) has been developed to enhance the safety and availability of a nuclear power plant (NPP) by improving operational reliability. The key elements of the proposed HMI are the large display panels which present synopsis of plant status and the compact, computer-based work stations for monitoring, control and protection functions. The work station consists of four consoles such as a dynamic alarm console (DAC), a system information console (SIC), a computerized operating-procedure console (COC), and a safety system information console (SSIC). The DAC provides clean alarm pictures, in which information overlapping is excluded and alarm impacts are discriminated, for quick situation awareness. The SIC supports a normal operation by offering all necessary system information and control functions over non-safety systems. In addition, it is closely linked to the other consoles in order to automatically display related system information according to situations of the DAC and the COC. The COC aids operators with proper operating procedures during normal plant startup and shutdown or after a plant trip, and it also reduces their physical/mental burden through soft automation. The SSIC continuously displays safety system status and enables operators to control safety systems. The proposed HMI has been evaluated using the checklists that are extracted from various human factors guidelines. From the evaluation results, it can be concluded that the HMI is so designed as to address the human factors issues reasonably. After sufficient validation, the concept and the design features of the proposed HMI will be reflected in the design of the main control room of the Korean Next Generation Reactor (KNGR)

  11. Development of Heavy-Duty and High-Precision Hydraulic Manipulator for Inspection, Maintenance and Decommission of Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sung Uk; Seo, Yong-chil; Jung, Kyung Min; Kim, Chang-hoi; Choi, Byung-seon; Moon, Jei-kwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Robotic manipulators have been used for inspection, maintenance and decommission of nuclear power plants because nuclear power plants have high radiation and human workers cannot easily access the plants. And also, to inspecting, maintaining and decommissioning nuclear power plants require various manipulators. Only one manipulator cannot response to many required tasks. The existing manipulators that was used at nuclear power plants can only operate only focused specific task and cannot be used at several tasks. The actuators used at manipulators are varied and many companies sell actuators depending on power, torque and speed. However, the commercial product is not standardized. Therefore, the development of manipulator is time consuming and expensive. The essential item of a manipulator is an actuator module. If actuator module is standardized, it’s easier to develop a manipulator and also maintain a manipulator. Recently, manipulator having high-radiation, high-duty and high-precision is necessary to inspection, maintain and decommissioning of nuclear power plants. Hydraulic actuator has been used to development high-duty manipulator. But control performance of a hydraulic actuator is not better than that of an electric actuator so that hydraulic manipulator cannot easily satisfy the required precision. In this paper, we developed high-duty and high-precision actuator modules and hydraulic manipulator using the developed actuator modules. The developed hydraulic manipulator have a payload of 250kg and a precision of ±1mm. Four modularized hydraulic actuator modules were developed for inspection, maintenance and decommission. Using the developed actuator modules, the manipulator for decommissioning is easily developed. And also, various manipulators having different kinematic structure for specific tasks will be easily developed by using hydraulic modules.

  12. Development of Heavy-Duty and High-Precision Hydraulic Manipulator for Inspection, Maintenance and Decommission of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Lee, Sung Uk; Seo, Yong-chil; Jung, Kyung Min; Kim, Chang-hoi; Choi, Byung-seon; Moon, Jei-kwon

    2016-01-01

    Robotic manipulators have been used for inspection, maintenance and decommission of nuclear power plants because nuclear power plants have high radiation and human workers cannot easily access the plants. And also, to inspecting, maintaining and decommissioning nuclear power plants require various manipulators. Only one manipulator cannot response to many required tasks. The existing manipulators that was used at nuclear power plants can only operate only focused specific task and cannot be used at several tasks. The actuators used at manipulators are varied and many companies sell actuators depending on power, torque and speed. However, the commercial product is not standardized. Therefore, the development of manipulator is time consuming and expensive. The essential item of a manipulator is an actuator module. If actuator module is standardized, it’s easier to develop a manipulator and also maintain a manipulator. Recently, manipulator having high-radiation, high-duty and high-precision is necessary to inspection, maintain and decommissioning of nuclear power plants. Hydraulic actuator has been used to development high-duty manipulator. But control performance of a hydraulic actuator is not better than that of an electric actuator so that hydraulic manipulator cannot easily satisfy the required precision. In this paper, we developed high-duty and high-precision actuator modules and hydraulic manipulator using the developed actuator modules. The developed hydraulic manipulator have a payload of 250kg and a precision of ±1mm. Four modularized hydraulic actuator modules were developed for inspection, maintenance and decommission. Using the developed actuator modules, the manipulator for decommissioning is easily developed. And also, various manipulators having different kinematic structure for specific tasks will be easily developed by using hydraulic modules

  13. Benchmarking Nuclear Power Plants

    International Nuclear Information System (INIS)

    Jakic, I.

    2016-01-01

    One of the main tasks an owner have is to keep its business competitive on the market while delivering its product. Being owner of nuclear power plant bear the same (or even more complex and stern) responsibility due to safety risks and costs. In the past, nuclear power plant managements could (partly) ignore profit or it was simply expected and to some degree assured through the various regulatory processes governing electricity rate design. It is obvious now that, with the deregulation, utility privatization and competitive electricity market, key measure of success used at nuclear power plants must include traditional metrics of successful business (return on investment, earnings and revenue generation) as well as those of plant performance, safety and reliability. In order to analyze business performance of (specific) nuclear power plant, benchmarking, as one of the well-established concept and usual method was used. Domain was conservatively designed, with well-adjusted framework, but results have still limited application due to many differences, gaps and uncertainties. (author).

  14. Nuclear power plant decommissioning

    International Nuclear Information System (INIS)

    Yaziz Yunus

    1986-01-01

    A number of issues have to be taken into account before the introduction of any nuclear power plant in any country. These issues include reactor safety (site and operational), waste disposal and, lastly, the decommissioning of the reactor inself. Because of the radioactive nature of the components, nuclear power plants require a different approach to decommission compared to other plants. Until recently, issues on reactor safety and waste disposal were the main topics discussed. As for reactor decommissioning, the debates have been academic until now. Although reactors have operated for 25 years, decommissioning of retired reactors has simply not been fully planned. But the Shippingport Atomic Power Plant in Pennysylvania, the first large scale power reactor to be retired, is now being decommissioned. The work has rekindled the debate in the light of reality. Outside the United States, decommissioning is also being confronted on a new plane. (author)

  15. The development of a nuclear chemical plant human reliability management approach: HRMS and JHEDI

    International Nuclear Information System (INIS)

    Kirwan, Barry

    1997-01-01

    In the late 1980's, amidst the qualitative and quantitative validation of certain Human Reliability Assessment (HRA) techniques, there was a desire for a new technique specifically for a nuclear reprocessing plant being designed. The technique was to have the following attributes: it should be data-based rather than involving pure expert judgement; it was to be flexible, so that it would allow both relatively rapid screening and more detailed assessment; and it was to have sensitivity analysis possibilities, so that Human Factors design-related parameters, albeit at a gross level, could be brought into the risk assessment equation. The techniques and literature were surveyed, and it was decided that no one technique fulfilled these requirements, and so a new approach was developed. Two techniques were devised, the Human Reliability Management System (HRMS), and the Justification of Human Error Data Information (JHEDI) technique, the latter being essentially a quicker screening version of the former. Both techniques carry out task analysis, error analysis, and Performance Shaping Factor-based quantification, but JHEDI involves less detailed assessment than HRMS. Additionally, HRMS can be utilised to determine error reduction mechanisms, based on the way the Performance Shaping Factors are contributing to the assessed error probabilities. Both techniques are fully computerised and assessments are highly documentable and auditable, which was seen as a useful feature both by the company developing the techniques, and by the regulatory authorities assessing the final output risk assessments into which these two techniques fed data. This paper focuses in particular on the quantification process used by these techniques. The quantification approach for both techniques was principally one of extrapolation from real data to the desired Human Error Probability (HEP), based on a comparison between Performance Shaping Factor (PSF) profiles for the real, and the to

  16. Fire risk analysis for nuclear power plants: Methodological developments and applications

    International Nuclear Information System (INIS)

    Kazarians, M.; Apostolakis, G.; Siv, N.O.

    1985-01-01

    A methodology to quantify the risk from fires in nuclear power plants is described. This methodology combines engineering judgment, statistical evidence, fire phenomenology, and plant system analysis. It can be divided into two major parts: (1) fire scenario identification and quantification, and (2) analysis of the impact on plant safety. This article primarily concentrates on the first part. Statistical analysis of fire occurrence data is used to establish the likelihood of ignition. The temporal behaviors of the two competing phenomena, fire propagation and fire detection and suppression, are studied and their characteristic times are compared. Severity measures are used to further specialize the frequency of the fire scenario. The methodology is applied to a switchgear room of a nuclear power plant

  17. Next Generation Nuclear Plant Steam Generator and Intermediate Heat Exchanger Materials Research and Development Plan

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright

    2010-09-01

    DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Today’s high-temperature alloys and associated ASME Codes for reactor applications are approved up to 760°C. However, some primary system components, such as the Intermediate Heat Exchanger (IHX) for the NGNP will require use of materials that can withstand higher temperatures. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge. Examples include materials for the core barrel and core internals, such as the control rod sleeves. The requirements of the materials for the IHX are among the most demanding. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. A number of solid solution strengthened nickel based alloys have been considered for

  18. Next Generation Nuclear Plant Intermediate Heat Exchanger Materials Research and Development Plan (PLN-2804)

    Energy Technology Data Exchange (ETDEWEB)

    J. K. Wright

    2008-04-01

    DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium-cooled, prismatic or pebble-bed reactor and use low-enriched uranium, Tri-Isotopic (TRISO)-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Today’s high-temperature alloys and associated ASME Codes for reactor applications are approved up to 760°C. However, some primary system components, such as the Intermediate Heat Exchanger (IHX) for the NGNP will require use of materials that can withstand higher temperatures. The thermal, environmental, and service life conditions of the NGNP will make selection and qualification of some high-temperature materials a significant challenge. Examples include materials for the core barrel and core internals, such as the control rod sleeves. The requirements of the materials for the IHX are among the most demanding. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. A number of solid solution strengthened nickel based alloys have been considered for

  19. Elecnuc. Nuclear power plants in the world

    International Nuclear Information System (INIS)

    2005-01-01

    This 2005 edition of the Elecnuc booklet summarizes in tables all numerical data relative to the nuclear power plants worldwide. These data come from the PRIS database managed by the IAEA. The following aspects are reviewed: 2004 highlights; main characteristics of reactor types; map of the French nuclear power plants on 2005/01/01; worldwide status of nuclear power plants at the end of 2004; units distributed by countries; nuclear power plants connected to the grid by reactor-type group; nuclear power plants under construction on 2004; evolution of nuclear power plant capacities connected to the grid; first electric generations supplied by a nuclear unit; electrical generation from nuclear power plants by country at the end 2004; performance indicator of PWR units in France; trend of the generation indicator worldwide; 2004 load factor by owners; units connected to the grid by countries at 12/31/2004; status of licence renewal applications in USA; nuclear power plants under construction at 12/31/2004; shutdown reactors; exported nuclear capacity in net MWe; exported and national nuclear capacity connected to the grid; exported nuclear power plants under construction or order; exported and national nuclear capacity under construction or order; recycling of plutonium in LWR; Mox licence plant projects; Appendix - historical development; acronyms, glossary

  20. Development of Real Time Operating System for Safety Grade PLC (POSAFE-Q) for Nuclear Power Plants

    International Nuclear Information System (INIS)

    Son, Han Seong; Hwang, Sung Jae; Lee, Young Joon; Kim, Chang Hwoi; Lee, Dong Young

    2006-01-01

    POSAFE-Q is a newly developed programmable logic controller (PLC) in order to apply to digital safety system of nuclear power plants (NPP) according to Nuclear Power Plant safety requirements. POSAFE-Q hardware and software development process, including design, review, verification and validation, and configuration control and quality assurance, satisfies the requirements imposed by 10CFR50, Appendix B. This article introduces a real time operating system pCOS, which is the core of POSAFE-Q. Section 2 describes the structure of pCOS. Section 3 describes a few important features of pCOS, which are necessary to the application for the digital safety system of NPP

  1. ALARA at nuclear power plants

    International Nuclear Information System (INIS)

    Baum, J.W.

    1991-01-01

    Implementation of the ALARA principle at nuclear power plants presents a continuing challenge for health physicists at utility corporate and plant levels, for plant designers, and for regulatory agencies. The relatively large collective doses at some plants are being addressed through a variety of dose reduction techniques. Initiatives by the ICRP, NCRP, NRC, INPO, EPRI, and BNL ALARA Center have all contributed to a heightened interest and emphasis on dose reduction. The NCRP has formed Scientific Committee 46-9 which is developing a report on ALARA at Nuclear Power Plants. It is planned that this report will include material on historical aspects, management, valuation of dose reduction ($/person-Sv), quantitative and qualitative aspects of optimization, design, operational considerations, and training. The status of this work is summarized in this report

  2. Current status and development of heat exchangers for boiling water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Uchiyama, Yoshio; Nishioka, Shuji; Ito, Shizuo

    1975-01-01

    More efficient and reliable operation is required for BWR heat exchangers because of nuclear plant safety and other reasons. Heat exchangers are classified into two categories of systems, the system for normal operation and the system for emergency operation. The present state and future improvement of heat exchangers are described in view of heat transfer performance, material selection, structural design, vibration, and so on. When noncondensing gas exists in vapour, heat transfer performance deteriorates, so that the heat transfer characteristics should be corrected by the adaption of venting the non condensing gas from the system. Heat transfer tubes should have high corrosion resistance to working fluid as well as high thermal conductivity, strength and economy. From that point of view, 30% cupro-nickel tubes will be replaced with 10% cupro-nickel tubes or titanium tubes though some technical development is necessary. These heat exchangers are now designed according to the MITI criteria for the technology concerning nuclear and thermal power generation, ASME Boiler and Pressure Vessel Code Sec. III and some other criteria. Most of heat transfer tube failures are caused from the tube vibration induced by working fluid flow, so that the vibration test and analysis were performed on U-tube elements. Some correction was obtained for design and fabrication techniques. (Iwase, T.)

  3. Developing a program to identify and track corrosion in nuclear plant raw water systems

    International Nuclear Information System (INIS)

    Spires PE, G.V.; Pickles PE, S.B.

    2001-01-01

    Findings derived from a comprehensive plant performance survey at Ontario Power Generation's (OPG) nuclear units convinced management that it would be prudent to expand the ongoing power piping Flow Accelerated Corrosion (FAC) induced wall thinning base-lining and tracking program to encompass the raw cooling water systems as well. Such systems are subject to a distinctly different class of pipe wall thinning (PWT) mechanisms than the FAC that degrades high-energy power piping. This paper describes the PWT corrosion assessment and tracking program that has been developed and is currently being implemented by OPG for the raw cooling water (i.e., Service Water) systems within it's nuclear generating stations. Interim databases are used prior to initial inspection rounds to catalogue the prospective locations. For each piping system being surveyed, these interim databases include physical coordinates for the candidate locations, the type and wall thickness of the components comprising each location, ranking indications and recommended NDE methodologies as a function of the anticipated corrosion mechanisms. Rationales for assessing corrosion susceptibility and ranking prospective inspection sites are expounded by way of notations built into the database. (authors)

  4. New developments for optimization of automatic inspections in nuclear power plants

    International Nuclear Information System (INIS)

    Bollini, G.J.; Gonzalez, B.

    1989-01-01

    One of the most important factors to be taken into acount with respect to remote control mechanized inspection tasks carried out during refueling outages in nuclear power plants, which include components and systems as varied as the reactor vessel, nozzles, piping, vessel head closure bolts, turbine discs, etc., is the unification of systems used to control the inspection systems themselves. Likewise, the implementation in these systems of specific functions simplifying operational sequences and confirming correct execution of work implies a set of criteria that must be taken into account if the reliability of the tasks carried out is to be increased. The system presented in this paper, SIROCCO (Sistema Robotizado de Control) has been designed taking into consideration the above mentioned criteria, with two clear obectives in mind: a reduction in inspection times and in the dose received by the participating technical personnel, and an increase in the reliability of the electronic equipment used. In February and March of this year, during shutdown of a BMR-type nuclear power plant, several SIROCCO control units were usids in mechanized inspection systems of various kinds, applicable to components such as those mentioned above, with forecasts regarding the reduction of inspection times being met. The main reasons for this reduction were the familiarity of the operators with this single control system, the unification of replacement parts for equipment maintenance and the availability of interchangeable units. This philosophy of unification of electronic systems also includes data acquisition and the transmission of information between different areas, which require the unification of control systems already commented and accomplished. This course of action is currently under development at Tecnatom. (author). 4 figs

  5. Cost-effective instrumentation and control upgrades for commercial nuclear power plants surety principles developed at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Rochau, G.E.; Dalton, L.J.

    1998-01-01

    Many nuclear power plants use instrument and control systems based on analog electronics. The state of the art in process control and instrumentation has advanced to use digital electronics and incorporate advanced technology. This technology includes: distributed microprocessors, fiber optics, intelligent systems (neutral networks), and advanced displays. The technology is used to optimize processes and enhance the man-machine interface while maintaining control and safety of the processes. Nuclear power plant operators have been hesitant to install this technology because of the cost and uncertainty in the regulatory process. This technology can be directly applied in an operating nuclear power plant provided a surety principle-based 'administrator' hardware system is included in parallel with the upgrade. Sandia National Laboratories has developed a rigorous approach to High Consequence System Surety (HCSS). This approach addresses the key issues of safety, security, and control while satisfying requirements for reliability and quality. We believe that HCSS principles can be applied to nuclear power plants in a manner that allows the off-the-shelf use of process control instrumentation while maintaining a high level of safety and enhancing the plant performance. We propose that an HCSS Administrator be constructed as a standardized approach to address regulatory issues. Such an administrator would allow a plant control system to be constructed with commercially available, state-to-the-art equipment and be customized to the needs of the individual plant operator. (author)

  6. Cost-effective instrumentation and control upgrades for commercial nuclear power plants using surety principles developed at Sandia National Laboratories

    International Nuclear Information System (INIS)

    Rochau, G.E.; Dalton, L.J.

    1997-01-01

    Many nuclear power plants use instrument and control systems based on analog electronics. The state of the art in process control and instrumentation has advanced to use digital electronics and incorporate advanced technology. This technology includes distributed microprocessors, fiber optics, intelligent systems (neural networks), and advanced displays. The technology is used to optimize processes and enhance the man-machine interface while maintaining control and safety of the processes. Nuclear power plant operators have been hesitant to install this technology because of the cost and uncertainty in the regulatory process. This technology can be directly applied in an operating nuclear power plant provided a surety principle-based open-quotes administratorclose quotes hardware system is included in parallel with the upgrade Sandia National Laboratories has developed a rigorous approach to High Consequence System Surety (HCSS). This approach addresses the key issues of safety, security, and control while satisfying requirements for reliability and quality. HCSS principles can be applied to nuclear power plants in a manner that allows the off-the-shelf use of process control instrumentation while maintaining a high level of safety and enhancing the plant performance. We propose that an HCSS administrator be constructed as a standardized approach to address regulatory issues. Such an administrator would allow a plant control system to be constructed with commercially available, state-of-the-art equipment and be customized to the needs of the individual plant operator

  7. Procedure Development and Qualification of the Phased Array Ultrasonic Testing for the Nuclear Power Plant Piping Weld

    International Nuclear Information System (INIS)

    Yoon, Byung Sik; Yang, Seung Han; Kim, Yong Sik; Lee, Hee Jong

    2010-01-01

    The manual ultrasonic examination for the nuclear power plant piping welds has been demonstrated by using KPD(Korean Performance Demonstration) generic procedure. For automated ultrasonic examination, there is no generic procedure and it should be qualified by using applicable automated equipment. Until now, most of qualified procedures used pulse-echo technique and there is no qualified procedure using phased array technique. In this study, data acquisition and analysis software were developed and phased-array transducer and wedge were designed to implement phased array technique for nuclear power plant in-service inspection. The developed procedure are qualified for performance demonstration for the flaw detection, length sizing and depth sizing. The qualified procedure will be applied for the field examination in the nuclear power plant piping weld inspection

  8. Development of data acquisition and processing system for In-service inspection of nuclear power plants

    International Nuclear Information System (INIS)

    Takama, Shinkichi; Kobayashi, Koji; Satoh, Yoshio; Koga, Yoshihiro; Shimizu, Takakazu

    1981-01-01

    In-service inspection (ISI) is required during the plant outage to assure the reliability of the components of a nuclear power plant. IHI has developed the advanced ISI system which consists of remote controlled and mechanized ultrasonic inspection devices for reactor pressure vessel examination, semi-automatic inspection instruments for piping examination with manual scan and automatic recording, data acquisition and processing system with microprocessor and mini-computer. By this system, ISI can be performed fully satisfying the requirement of ASME Code Sec. XI and minimizing operation in the high radioactive areas. All ultrasonic information is processed by the computer and the examination results such as size and location of ultrasonic indication are printed out in the form of sectional and plan view of the part examined, reproduced screen image and polar plot, etc. as well as the evaluation sheet. This system saves the elaborate work of inspection personnel and is expected to contribute to the improvement of inspection quality and to the reduction of radiation exposure of inspection personnel. (author)

  9. Development of high performance condensers for thermal and nuclear power plants

    International Nuclear Information System (INIS)

    Okouchi, Isao; Takahashi, Sankichi; Tomita, Akira.

    1980-01-01

    As the trend toward the large capacity of thermal and nuclear power plants advances, condensers also become large, and from the viewpoint of energy saving in whole plants, the maintenance of high performance and reliability is strongly desired. Hitachi Ltd. responded to this demand, and repeated the basic investigation with a model condenser on the selection of condenser cooling tubes and their arrangement. As the result, balanced downflow type tube arrangement was developed, which enables smooth steam flow and the improvement of condenser performance by forming the intense flow of turbine exhaust from the upper part of tube nest downward and making the steam flow in the lower part of tube nest into radial form toward air ejecting port. This tube arrangement has been applied to actual machines, and the excellent results have been obtained. In particular, the improvement of the degree of vacuum due to the reduction of steam flow loss is advantageous for increasing the power output of turbines. Thereupon, based on the basic experiment with various models of tube arrangement and the consideration on the operational results of actual machines using this tube arrangement, the features of this technique are reported. The basic construction of tube arrangement, the steam flow in tube nest, the vacuum in condensers, the supercooling of condensate, and actual balanced downflow type condensers are described. (Kako, I.)

  10. Next Generation Nuclear Plant Methods Research and Development Technical Program Plan -- PLN-2498

    Energy Technology Data Exchange (ETDEWEB)

    Richard R. Schultz; Abderrafi M. Ougouag; David W. Nigg; Hans D. Gougar; Richard W. Johnson; William K. Terry; Chang H. Oh; Donald W. McEligot; Gary W. Johnsen; Glenn E. McCreery; Woo Y. Yoon; James W. Sterbentz; J. Steve Herring; Temitope A. Taiwo; Thomas Y. C. Wei; William D. Pointer; Won S. Yang; Michael T. Farmer; Hussein S. Khalil; Madeline A. Feltus

    2008-09-01

    One of the great challenges of designing and licensing the Very High Temperature Reactor (VHTR) is to confirm that the intended VHTR analysis tools can be used confidently to make decisions and to assure all that the reactor systems are safe and meet the performance objectives of the Generation IV Program. The research and development (R&D) projects defined in the Next Generation Nuclear Plant (NGNP) Design Methods Development and Validation Program will ensure that the tools used to perform the required calculations and analyses can be trusted. The Methods R&D tasks are designed to ensure that the calculational envelope of the tools used to analyze the VHTR reactor systems encompasses, or is larger than, the operational and transient envelope of the VHTR itself. The Methods R&D focuses on the development of tools to assess the neutronic and thermal fluid behavior of the plant. The fuel behavior and fission product transport models are discussed in the Advanced Gas Reactor (AGR) program plan. Various stress analysis and mechanical design tools will also need to be developed and validated and will ultimately also be included in the Methods R&D Program Plan. The calculational envelope of the neutronics and thermal-fluids software tools intended to be used on the NGNP is defined by the scenarios and phenomena that these tools can calculate with confidence. The software tools can only be used confidently when the results they produce have been shown to be in reasonable agreement with first-principle results, thought-problems, and data that describe the “highly ranked” phenomena inherent in all operational conditions and important accident scenarios for the VHTR.

  11. Robotics for nuclear power plants

    International Nuclear Information System (INIS)

    Nakayama, Ryoichi; Kimura, Motohiko; Abe, Akira

    1993-01-01

    A continuing need exists for automatic or remote-controlled machines or robots which can perform inspection and maintenance tasks in nuclear power plants. Toshiba has developed several types of monofunctional and multi- functional robots for such purposes over the past 20 years, some of which have already been used in actual plants. This paper describes new multifunctional robots for inspection and maintenance. An inspection robot has been applied in an actual plant for two years for performance testing. Maintenance robots for grinding tasks have also been developed, which can be easily teleoperated by the operator using automatic control. These new robots are expected to be applied to actual inspection and maintenance work in nuclear power plants. (author)

  12. Development of a strategic plan for an international R and D project on innovative nuclear fuel cycles and power plants

    International Nuclear Information System (INIS)

    Kendall, J.; Choi, J.S.

    2002-01-01

    The long-term outlook for nuclear energy should be considered in a broader perspective of future energy needs, operational safety, proliferation and environmental impacts. An Advisory Group Meeting (AGM) on Development of a Strategic Plan for an International R and D Project on Innovative Nuclear Fuel Cycles and Power Plants was convened in Vienna in October 1999 to assess the criteria, the needs for international cooperation, and to formulate a strategic plan for project integration. (author)

  13. Nuclear plant license renewal

    International Nuclear Information System (INIS)

    Gazda, P.A.; Bhatt, P.C.

    1991-01-01

    During the next 10 years, nuclear plant license renewal is expected to become a significant issue. Recent Electric Power Research Institute (EPRI) studies have shown license renewal to be technically and economically feasible. Filing an application for license renewal with the Nuclear Regulatory Commission (NRC) entails verifying that the systems, structures, and components essential for safety will continue to perform their safety functions throughout the license renewal period. This paper discusses the current proposed requirements for this verification and the current industry knowledge regarding age-related degradation of structures. Elements of a license renewal program incorporating NRC requirements and industry knowledge including a schedule are presented. Degradation mechanisms for structural components, their significance to nuclear plant structures, and industry-suggested age-related degradation management options are also reviewed

  14. World nuclear power plant capacity

    International Nuclear Information System (INIS)

    1991-01-01

    This report provides the background information for statistics and analysis developed by NUKEM in its monthly Market Report on the Nuclear Fuel Cycle. The assessments in this Special Report are based on the continuous review of individual nuclear power plant projects. This Special Report begins with tables summarizing a variety of nuclear power generating capacity statistics for 1990. It continues with a brief review of the year's major events regarding each country's nuclear power program. The standard NUKEM Market Report tables on nuclear plant capacity are given on pages 24 and 25. Owing to space limitations, the first year shown is 1988. Please refer to previous Special Reports for data covering earlier years. Detailed tables for each country list all existing plants as well as those expected by NUKEM to be in commercial operation by the end of 2005. An Appendix containing a list of abbreviations can be found starting on page 56. Only nuclear power plants intended for civilian use are included in this Special Report. Reactor lifetimes are assumed to be 35 years for all light water reactors and 30 years for all other reactor types, unless other data or definite decommissioning dates have been published by the operators. (orig./UA) [de

  15. Development of Advanced Technologies to Reduce Design, Fabrication and Construction Costs for Future Nuclear Power Plants

    International Nuclear Information System (INIS)

    DiNunzio, Camillo A.; Gupta, Abhinav; Golay, Michael; Luk, Vincent; Turk, Rich; Morrow, Charles; Geum-Taek Jin

    2002-01-01

    OAK-B135 This report presents a summation of the third and final year of a three-year investigation into methods and technologies for substantially reducing the capital costs and total schedule for future nuclear plants. In addition, this is the final technical report for the three-year period of studies

  16. Development of Advanced Technologies to Reduce Design, Fabrication and Construction Costs for Future Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    DiNunzio, Camillo A. [Framatome ANP DE& S, Marlborough, MA (United States); Gupta, Abhinav [Univ. of North Carolina, Raleigh, NC (United States); Golay, Michael [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Luk, Vincent [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Turk, Rich [Westinghouse Electric Company Nuclear Systems, Windsor, CT (United States); Morrow, Charles [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Jin, Geum-Taek [Korea Power Engineering Company Inc., Yongin-si, Gyeonggi-do (Korea, Republic of)

    2002-11-30

    This report presents a summation of the third and final year of a three-year investigation into methods and technologies for substantially reducing the capital costs and total schedule for future nuclear plants. In addition, this is the final technical report for the three-year period of studies.

  17. Recent developments in the field of nuclear power plant control and instrumentation in Hungary

    International Nuclear Information System (INIS)

    Pellionisz, P.

    1992-01-01

    A considerable percentage (32.8% in 1989) of electric energy in Hungary is produced by nuclear power plant Paks. The paper presents an overview of activities on control and instrumentation in the following areas: Control and instrumentation upgrading; training simulators; diagnostic systems. (author). 1 tab

  18. Worldwide nuclear-plant performance

    International Nuclear Information System (INIS)

    Surrey, J.; Thomas, S.

    1980-01-01

    The authors compare the performance of different reactor systems to identify the determinants of plant performance, to examine the evidence of technological maturation, and to discover the principal causes of outage or unavailability. In the light of the findings, they discuss the implications for the UK regarding reactor choice and technology development. They make no judgements about the relative merits of nuclear and fossil-fuel plants, or about safety. (author)

  19. Recent developments of and future prospects for nuclear power plant instrumentation and control in Belgium

    International Nuclear Information System (INIS)

    Naisse, J.C.

    1990-01-01

    The production of electricity in Belgium is covered for almost 70% by seven nuclear power plants all of pressurized water reactor type. Four major topics are representative of the evolution of nuclear power production in Belgium: The suspentions of the studies for the new Belgian PWR unit (DOEL 5), the commissioning of the simulators in DOEL and TIHANGE, the revamping of the oldest nuclear units, the studies of new storage installations for radioactive waste. The delay imposed on the construction of DOEL 5 will allow to go further into the examination of the implications arising from the new technologies in I and C. The use modern I and C systems for modifications in existing plants will allow a better estimation of their advantages and disadvantages. (author)

  20. Development of nuclear power plant monitoring system with neutral network using on-line PWR plant simulator

    International Nuclear Information System (INIS)

    Nabeshima Kunihiko; Suzuki Katsuo; Nose, Shoichi; Kudo, Kazuhiko

    1996-01-01

    The purpose of this paper is to demonstrate a nuclear power plant monitoring system using artificial neural network (ANN). The major advantages of the monitoring system are that a multi-output process system can be modelled using measurement information without establishing any mathematical expressions. The dynamics model of reactor plant was constructed by using three layered auto-associative neural network with backpropagation learning algorithm. The basic idea of anomaly detection method is to monitor the deviation between process signals measured from actual plant and corresponding output signals from the ANN plant model. The simulator used is a self contained system designed for training. Four kinds of simulated malfunction caused by equipment failure during steady state operation were used to evaluate the capability of the neural network monitoring system. The results showed that this monitoring system detected the symptom of small anomaly earlier than the prevailing alarm system. (author). 7 refs, 7 figs, 2 tabs

  1. Development of nuclear power plant monitoring system with neutral network using on-line PWR plant simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kunihiko, Nabeshima; Katsuo, Suzuki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Nose, Shoichi; Kudo, Kazuhiko [Kyushu Univ., Fukuoka (Japan). Faculty of Engineering

    1997-12-31

    The purpose of this paper is to demonstrate a nuclear power plant monitoring system using artificial neural network (ANN). The major advantages of the monitoring system are that a multi-output process system can be modelled using measurement information without establishing any mathematical expressions. The dynamics model of reactor plant was constructed by using three layered auto-associative neural network with backpropagation learning algorithm. The basic idea of anomaly detection method is to monitor the deviation between process signals measured from actual plant and corresponding output signals from the ANN plant model. The simulator used is a self contained system designed for training. Four kinds of simulated malfunction caused by equipment failure during steady state operation were used to evaluate the capability of the neural network monitoring system. The results showed that this monitoring system detected the symptom of small anomaly earlier than the prevailing alarm system. (author). 7 refs, 7 figs, 2 tabs.

  2. Medium-size nuclear plants

    International Nuclear Information System (INIS)

    Vogelweith, L.; Lavergne, J.C.; Martinot, G.; Weiss, A.

    1977-01-01

    CEA (TECHNICATOME) has developed a range of pressurized water reactors of the type ''CAS compact'' which are adapted to civil ship propulsion, or to electric power production, combined possibly with heat production, up to outputs equivalent to 125 MWe. Nuclear plants equipped with these reactors are suitable to medium-size electric networks. Among the possible realizations, two types of plants are mentioned as examples: 1) Floating electron-nuclear plants; and 2) Combined electric power and desalting plants. The report describes the design characteristics of the different parts of a 125 MWe unit floating electro-nuclear plant: nuclear steam system CAS 3 G, power generating plant, floating platform for the whole plant. The report gives attention to the different possibilities according to site conditions (the plant can be kept floating, in a natural or artificial basin, it can be put aground, ...) and to safety and environment factors. Such unit can be used in places where there is a growing demand in electric power and fresh water. The report describes how the reactor, the power generating plant and multiflash distillation units of an electric power-desalting plant can be combined: choice of the ratio water output/electric power output, thermal cycle combination, choice of the gain ratio, according to economic considerations, and to desired goal of water output. The report analyses also some technical options, such as: choice of the extraction point of steam used as heat supply of the desalting station (bleeding a condensation turbine, or recovering steam at the exhaust of a backpressure turbine), design making the system safe. Lastly, economic considerations are dealt with: combining the production of fresh water and electric power provides usually a much better energy balance and a lower cost for both products. Examples are given of some types of installations which combine medium-size reactors with fresh water stations yielding from 10000 to 120000 m 3 per day

  3. Nuclear plant safety

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    The four-member New York Power Pool Panel concluded that, for a number of reasons, no nuclear power plant in New York State is prone to the type of accident that occurred at Three Mile Island (TMI). The Panel further concluded that changes in operating practices, both regulatory and voluntary, and heightened sensitivity to reactor-core-cooling requirements will substantially reduce the chances for another such accident anywhere. Panel members found that New York State utilities have taken a responsible attitude with regard to requirements set forth by the Nuclear Regulatory Commission (NRC) as a result of the TMI accident. In a cover letter that accompanied the report to Federal and New York state officials, New York Power Pool Executive Committee Chairman Francis E. Drake, Jr. expressed hope that the report will alleviate public fears of nuclear reactors and promote wider acceptance of nuclear energy as an economic and safe means of power production. 17 references

  4. Nuclear plants - military hostages

    International Nuclear Information System (INIS)

    Ramberg, B.

    1986-01-01

    Recent events suggest that nuclear reactors could make tempting military or terrorist targets. Despite the care with which most reactors are built, studies document their vulnerability to willful destruction through disruption of coolant mechanisms both inside and outside the containment building. In addition to reactors, such nuclear support facilities as fuel fabrication, reprocessing, and waste storage installations may be attractive military targets. A nuclear bomb which exploded in the vicinity of a reactor could increase its lethal effects by one-third. The implications of this is vulnerability for Middle East stability as well as to other volatile regions. The author suggests several avenues for controlling the dangers: international law, military and civil defense, facility siting, increasing plant safety, and the international management of nuclear energy. 21 references

  5. Development of Information Datasheets of Nuclear Power Plant (NPP) Equipment using cfiXLM schema

    International Nuclear Information System (INIS)

    Lee, Jaiho; Song, Eunhye

    2014-01-01

    In 2009, EPRI (Electrical Power Research Institute) published a new NPP information handover guide to provide NPP owners and operators with data handover templates in consistent format for effective delivery of information during all stages of the handover process. Another difficult concern for NPP data information management is to exchange the data information among many organizations such as NPP owners, operators, engineering companies, suppliers, and vendors. As a matter of fact, the improperly formatted handover of information sometimes occurs due to the discrepancy of data format (e. g., data description language type). This improper delivery can make negative effects on NPP integrity and safety. Thus, the lack of proper exchange for different data information systems of organizations should be resolved by using an international standard data format. The standard data format can reduce the cost and time for data exchange in each phase for design, procurement, delivery, installation, operation and maintenance of equipment. The AEX(automating equipment information exchange) pilot implementation project team under EPRI advanced nuclear technology (ANT) program has been conducted a research for the use of XML equipment schemas for electronic data exchange(EDE). They applied XML equipment schema for the design, selection, quotation, purchase and mock install of a safety injection centrifugal pump using EDE standard HI(hydraulic institute) 50.7. For data exchange, FIATECH, an industry consortium, has equally developed library of templates and reference data for ISO-15926, which is an international standard capable of reducing data-error and delivery time for exchanging data among different organizations. KHNP as an only owner/operator company has not experienced much difficulty in data interoperability with other organizations, but continued its unremitting exertions to develop a robust system capable of managing data information generated in all the stages of NPP

  6. Development of Information Datasheets of Nuclear Power Plant (NPP) Equipment using cfiXLM schema

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jaiho; Song, Eunhye [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In 2009, EPRI (Electrical Power Research Institute) published a new NPP information handover guide to provide NPP owners and operators with data handover templates in consistent format for effective delivery of information during all stages of the handover process. Another difficult concern for NPP data information management is to exchange the data information among many organizations such as NPP owners, operators, engineering companies, suppliers, and vendors. As a matter of fact, the improperly formatted handover of information sometimes occurs due to the discrepancy of data format (e. g., data description language type). This improper delivery can make negative effects on NPP integrity and safety. Thus, the lack of proper exchange for different data information systems of organizations should be resolved by using an international standard data format. The standard data format can reduce the cost and time for data exchange in each phase for design, procurement, delivery, installation, operation and maintenance of equipment. The AEX(automating equipment information exchange) pilot implementation project team under EPRI advanced nuclear technology (ANT) program has been conducted a research for the use of XML equipment schemas for electronic data exchange(EDE). They applied XML equipment schema for the design, selection, quotation, purchase and mock install of a safety injection centrifugal pump using EDE standard HI(hydraulic institute) 50.7. For data exchange, FIATECH, an industry consortium, has equally developed library of templates and reference data for ISO-15926, which is an international standard capable of reducing data-error and delivery time for exchanging data among different organizations. KHNP as an only owner/operator company has not experienced much difficulty in data interoperability with other organizations, but continued its unremitting exertions to develop a robust system capable of managing data information generated in all the stages of NPP

  7. Design and development of innovative passive valves for Nuclear Power Plant applications

    Energy Technology Data Exchange (ETDEWEB)

    Sapra, M.K., E-mail: sapramk@barc.gov.in; Kundu, S.; Pal, A.K.; Vijayan, P.K.; Vaze, K.K.; Sinha, R.K.

    2015-05-15

    Highlights: • Passive valves are self-acting valves requiring no external energy to function. • These valves have been developed for Advanced Heavy Water Reactor (AHWR) of India. • Passive valves are core components of passive safety systems of the reactor. • Accumulator Isolation Passive Valve (AIPV) has been developed and tested for ECSS. • AIPV provided passive isolation and flow regulation in ECCS of Integral Test Loop. - Abstract: The recent Fukushima accident has resulted in an increased need for passive safety systems in upcoming advanced reactors. In order to enhance the global contribution and acceptability of nuclear energy, proven evidence is required to show that it is not only green but also safe, in case of extreme natural events. To achieve and establish this fact, we need to design, demonstrate and incorporate reliable ‘passive safety systems’ in our advanced reactor designs. In Nuclear Power Plants (NPPs), the use of passive safety systems such as accumulators, condensing and evaporative heat exchangers and gravity driven cooling systems provide enhanced safety and reliability. In addition, they eliminate the huge costs associated with the installation, maintenance and operation of active safety systems that require multiple pumps with independent and redundant electric power supplies. As a result, passive safety systems are preferred for numerous advanced reactor concepts. In current NPPs, passive safety systems which are not participating in day to day operation, are kept isolated, and require a signal and external energy source to open the valve. It is proposed to replace these valves by passive components and devices such as self-acting valves, rupture disks, etc. Some of these innovative passive valves, which do not require external power, have been recently designed, developed and tested at rated conditions. These valves are proposed to be used for various passive safety systems of an upcoming Nuclear Power Plant being designed

  8. Nuclear power development

    International Nuclear Information System (INIS)

    Povolny, M.

    1980-01-01

    The development and uses of nuclear power in Czechoslovakia and other countries are briefly outlined. In the first stage, the Czechoslovak nuclear programme was oriented to the WWER 440 type reactor while the second stage of the nuclear power plant construction is oriented to the WWER 10O0 type reactor. It is envisaged that 12 WWER 440 type reactors and four to five WWER 1000 type reactors will be commissioned till 1990. (J.P.)

  9. Pulsed nuclear power plant

    International Nuclear Information System (INIS)

    David, C.V.

    1986-01-01

    This patent describes a nuclear power plant. This power plant consists of: 1.) a cavity; 2.) a detonatable nuclear device in a central region of the cavity; 3.) a working fluid inside of the cavity; 4.) a method to denote a nuclear device inside of the cavity; 5.) a mechanical projection from an interior wall of the cavity for recoiling to absorb a shock wave produced by the detonation of the nuclear device and thereby protecting the cavity from damage. A plurality of segments defines a shell within the cavity and a plurality of shock absorbers, each connecting a corresponding segment to a corresponding location on the wall of the cavity. Each of these shock absorbers regulate the recoil action of the segments; and 6.) means for permitting controlled extraction of a quantity of hot gases from the cavity produced by the vaporization of the working fluid upon detonation of the nuclear device. A method of generating power is also described. This method consists of: 1.) introducing a quantity of water in an underground cavity; 2.) heating the water in the cavity to form saturated steam; 3.) detonating a nuclear device at a central location inside the cavity; 4.) recoiling plate-like elements inside the cavity away from the central location in a mechanically regulated and controlled manner to absorb a shock wave produced by the nuclear device detonation and thereby protect the underground cavity against damage; 5.) extracting a quantity of superheated steam produced by the detonation of the nuclear device; and 6.) Converting the energy in the extracted superheated steam into electrical power

  10. Development of decontamination technology for the decommissioned Bohunice A-1 nuclear power plant

    International Nuclear Information System (INIS)

    Krejci, F.; Majersky, D.; Solcanyi, M.; Sekely, S.; Kucharik, D.

    1991-01-01

    The main results of investigation into the decontamination technology for the equipment and buildings of the decommissioned A-1 nuclear power plant, achieved by the Nuclear Power Plants Research Institute in Trnava over the 1988-1990 period, are summarized. Mobile decontamination and recirculation equipment has been developed for pre-disassembling decontamination. A solution containing formic acid (19 g/l), EDTA-Na 4 (6 g/l) and thiourea (0.5 g/l) was used for decontamination of low-alloy steels; for materials from the steam generators and turbo-compressors, the decontamination factor (DF) of this solution was 30 to 150 per decontamination cycle. For high-alloy steels, a two-stage process comprising the use of an oxidation solution and a reduction solution appeared suitable. The oxidation solution contained potassium permanganate (0.6 g/l) and nitric acid (0.4 g/l), whereas the reduction solution, viz. Citrox 21, contained citric acid (0.5 g/l), oxalic acid (1.0 g/l) and EDTA-NA 4 (2.5 g/l). The DF is 10 to 50 in one oxidation-reduction cycle and 50-100 in two cycles. For the post-disassembling chemical decontamination, the contaminated material was cut into pieces 70 to 80 cm long, freed from grease and decontaminated chemically by submerging in the solution while applying treatment by ultrasound. A technology of electrochemical decontamination has also been developed. It appeared particularly suitable for structural materials of the primary coolant circuit comprising austenitic stainless steels and low-alloy steels after pre-disassembling chemical decontamination with remainders of the corrosion layer, and for structural materials of the secondary coolant circuit after chemical post-disassembling decontamination. Research in the field of decontamination of the building parts and of the outer surfaces of the structural materials concentrated mainly on the use of decontamination foams. Foaming solutions have been developed for the decontamination of PESL floors and

  11. Decommissioning of nuclear power plants

    International Nuclear Information System (INIS)

    Friske, A.; Thiele, D.

    1988-01-01

    The IAEA classification of decommissioning stages is outlined. The international development hitherto observed in decommissioning of nuclear reactors and nuclear power stations is presented. The dismantling, cutting and decontamination methods used in the decommissioning process are mentioned. The radioactive wastes from decommissioning are characterized, the state of the art of their treatment and disposal is given. The radiation burdens and the decommissioning cost in a decommissioning process are estimated. Finally, some evaluation of the trends in the decommissioning process of nuclear power plants is given. 54 refs. (author)

  12. SMART-P MMIS Software Development by Considering the Software License for Nuclear Power Plants and the Development Cost

    International Nuclear Information System (INIS)

    Suh, Yong Suk; Park, Jae Hong; Park, Heui Youn; Son, Ki Sung; Lee, Ki Hyun; Kim, Hyeon Soo

    2005-01-01

    The acceptance criteria of software for safety system functions in NPPs (Nuclear Power Plants) are as follows: 1) acceptable plans should be prepared to control the software development activities, 2) the plans should be followed in an acceptable software life cycle, and 3) the process should produce acceptable design outputs. The KINS (Korea Institute of Nuclear Safety) recommended that the software life cycle should be established based on the IEEE Std 1074 with a supplementary requirement of a software safety analysis. The KINS emphasized that the software should be developed to show its high qualities. This paper identifies the major requirements to achieve the software license from the KINS and presents the major facts reflected in the SMART-P (System-integrated Modular Advanced ReacTor-Pilot) MMIS (Man-Machine Interface Systems) which is being developed by KAERI and targeted to start operation in 2010. This paper also addresses major concerns on the development of a safety critical software and the facts reflected in the SMART-P MMIS

  13. Development of the job concept of a nuclear power plant operator

    International Nuclear Information System (INIS)

    Alpeev, A.S.; Bukrinskii, A.M.

    1994-01-01

    An important conclusion drawn in the aftermath of the 1979 accident at Three Mile Island 2 in the USA, in particular, in discussions, held in Stockholm in 1980 at the International Conference on Safety Problems, of the causes of this accident, was that the plant-operator concept employed at the time in western countries had to be changed. The concept presupposed that the operator's actions must be strictly limited by instructions which supposedly encompass all possible situations associated with the accidents. The principal problem facing the operator in this concept is to identify the situation in which the appropriate instruction can be selected for immediate execution. An error in identifying the situation could influence decisively the evolution of further events in the power block of the nuclear plant, and this is precisely what happened at Three Mile Island 2. For this reason, operator training was based mainly on identifying the initial events, which corresponded to a definite sequence of the further developments owing to the technical scheme and laws of thermophysical processes. Such operator training is based on an event tree which gives a graphical representation of the probable consequences of the development of events in an accident with a given initial event. It is believed that approximately 20 initial events (tree roots) form the basic collection of events, each of which has three to seven branches, i.e., an accident can evolve in several possible directions. Every branch point on each tree is identified by a definite collection of technological parameters, according to which the operator must be able to identify the situation. Under the present concept, when the situation is identified correctly, the search for the required control instruction and its execution is a secondary process

  14. Computer simulation of nuclear pollutant diffusion from Shimane nuclear power plants and development of an evacuation strategy in the event of a nuclear incident

    International Nuclear Information System (INIS)

    Yakura, Haruna; Kurimasa, Akihiro

    2012-01-01

    Beginning from Tohoku (northeastern) Earthquake on March 11, an accident at the Fukushima Dai-ichi Nuclear Power Plant resulted in a substantial release of radioactivity to the environment. The accident forced a large number of residents to evacuate from surrounding areas. Moreover, the nuclear incident was life-threatening for the elderly and for people with serious illnesses who were confined in hospital or nursing homes. Strikingly, the causes of death were not directly attributed to radiation exposure but to problems encountered during evacuation. Using nuclear diffusion data from the Fukushima incident, we simulated nuclear pollutant dispersion using computer software A2C in an area of Tottori and Shimane Prefectures surrounding the Shimane Nuclear Power Plant. We generated a model for the spread of nuclear pollutants around the Emergency Planning Zone, id est (i.e.) EPZ. From these findings, we proposed evacuation strategies for residents near the power plant to ensure a safe and reliable escape from nuclear pollutants. Our recommendations include: immediate evacuation from PAZ area (within 5 km), securing indoor shelter in the area of the UPZ (from 5 km to 30 km) and preparations to evacuate further outwards from the nuclear plant site, daytime evacuation within a 30 km area after considering wind direction and velocity, and preparation of a planned evacuation strategy that identifies secure facilities for weaker people in the event of a disaster. (author)

  15. Development of RCM methodology and tools for EDF nuclear power plants

    International Nuclear Information System (INIS)

    Jacquot, J.P.; Bouchet, J.L.; Despujols, A.; Dewailly, J.; Martin-Mattei, C.

    1995-01-01

    In 1990, EDF launched a Reliability-Centered Maintenance project for its nuclear plants. This 'OMF' project aims at developing methods and tools for analysis and in the first phase, applying these to one initial system (the pilot study). The results of the pilot study have confirmed the advantages of the 'OMF' analytical method: the prospects for the approach on an industrial scale are extremely promising. It should be noted that the precision of our 'OMF' analysis is not doubt superior to that common in other industrial domains (MSG/RCM analysis). The particular approach implies analysis of systems and components and, most importantly, integration of operation feedback, with a view to developing a rigorous maintenance program which can constantly be updated. In addition to the defining and implementing the method, the review of designing software aids has begun. The pilot study clearly pointed up the need for such aids in handling the necessary volume of information and assisting experts in their analysis. The EDF 'OMF' workstation (and its environment) will be used not only in preparing the 'initial' maintenance program but also in updating it during the 'living' program phase. (author)

  16. Development of integral type forgings for steam generator of nuclear power plant

    International Nuclear Information System (INIS)

    Suzuki, Komei; Sato, Ikuo; Murai, Etsuo

    1992-01-01

    The use of integral type steel forgings for the construction of pressure vessel enhances the structural integrity of components and makes the fabrication of components and the execution of in-service inspection (ISI) easier than those fabrication from plate and casting materials. Such steel forgings have been realized for steam generator (SG) for nuclear power plant as follows : (1) Forged shell ring : change from welding fabrication of formed plates to forging ; (2) Forged conical shell ring : ditto ; (3) Forged head integral with nozzles (s) : (i) Primary head : change from casting to forging ; (ii) Secondary head : change from welding fabrication of formed plates to forging. These steel forgings have been realized by recent development in manufacturing technologies, such as steel making, forging processes and heat treatment which are vital to the quality of steel forgings. Some examples of recent typical high quality steel forgings developed by the Japan Steel Works, Ltd. (JSW) are introduced, and the main points of the manufacturing technique and the quality attained are also described. (author)

  17. Integrated CAE system for nuclear power plants. Development of piping design check system

    International Nuclear Information System (INIS)

    Narikawa, Noboru; Sato, Teruaki

    1994-01-01

    Toshiba Corporation has developed and operated the integrated CAE system for nuclear power plants, the core of which is the engineering data base to manage accurately and efficiently enormous amount of data on machinery, equipment and piping. As the first step of putting knowledge base system to practical use, piping design check system has been developed. By automatically checking up piping design, this system aims at the prevention of overlooking mistakes, efficient design works and the overall quality improvement of design. This system is based on the thought that it supports designers, and final decision is made by designers. This system is composed of the integrated data base, a two-dimensional CAD system and three-dimensional CAD system. The piping design check system is one of the application systems of the integrated CAE system. Object-oriented programming is the base of the piping design check system, and design knowledge and CAD data are necessary. As to the method of realizing the check system, the flow of piping design, the checkup functions, the checkup of interference and attribute base, and the integration of the system are explained. (K.I)

  18. Gas cloud explosions and their effect on nuclear power plant, basic development of explosion codes

    International Nuclear Information System (INIS)

    Hall, S.F.; Martin, D.; MacKenzie, J.

    1985-01-01

    The study of factors influencing the pressure and velocity fields produced by the burning of flammable substances has been in progress at SRD for some years. This paper describes an extension of these studies by using existing codes for a parametric survey, and modifying codes to produce more realistic representations of explosions and developing a two dimensional combustion code, FLARE. The one dimensional combustion code, GASEX1, has been used to determine the pressure from a burning gas cloud for a number of different fuels, concentrations and burning velocities. The code was modified so that gas concentrations could be modelled. Results for concentration gradients showed the pressure depended on local conditions and the burning velocity. The two dimensional code, GASEX2, was modified to model the interaction of pressure waves with structures. It was used, with results from GASEX1, to model the interaction of a pressure wave from the combustion of a gas cloud with a rigid structure representing a nuclear power plant. The two dimensional code FLARE has been developed to model the interaction of flames and pressure waves with structures. The code incorporates a simple turbulence model with a turbulence dependent reaction rate. Validation calculations have been carried out for the code. (author)

  19. Development of RCM methodology and tools for EDF nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Jacquot, J.P.; Bouchet, J.L.; Despujols, A.; Dewailly, J.; Martin-Mattei, C. [Electricite de France, 78 - Chatou (France)

    1995-12-31

    In 1990, EDF launched a Reliability-Centered Maintenance project for its nuclear plants. This `OMF` project aims at developing methods and tools for analysis and in the first phase, applying these to one initial system (the pilot study). The results of the pilot study have confirmed the advantages of the `OMF` analytical method: the prospects for the approach on an industrial scale are extremely promising. It should be noted that the precision of our `OMF` analysis is not doubt superior to that common in other industrial domains (MSG/RCM analysis). The particular approach implies analysis of systems and components and, most importantly, integration of operation feedback, with a view to developing a rigorous maintenance program which can constantly be updated. In addition to the defining and implementing the method, the review of designing software aids has begun. The pilot study clearly pointed up the need for such aids in handling the necessary volume of information and assisting experts in their analysis. The EDF `OMF` workstation (and its environment) will be used not only in preparing the `initial` maintenance program but also in updating it during the `living` program phase. (author) 4 refs.

  20. Development of automated operating procedure system using fuzzy colored petri nets for nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Seong, Poong Hyun

    2004-01-01

    In this work, AuTomated Operating Procedure System (ATOPS) is developed. ATOPS is an automation system for emergency operation of a nuclear power plant (NPP) and it can monitor signals, diagnose statuses, and generate control actions according to corresponding operating procedures without any human operator's help. Main functions of ATOPS are an anomaly detection function and a procedure execution function but only the procedure execution function is implemented in this work because this work is just the first step. In the procedure execution function, operating procedures of NPPs are analyzed and modeled using Fuzzy Colored Petri Nets (FCPN) and executed depending on decision making of the inference engine. In this work, ATOPS prototype is developed to demonstrate its feasibility and it is also validated using the FISA-2/WS simulator. The validation is performed for the cases of a loss of coolant accident (LOCA) and a steam generator tube rupture (SGTR). The simulation results show that ATOPS works correctly in the emergency situations

  1. Nuclear power plant control and instrumentation in the Netherlands: Status and developments

    International Nuclear Information System (INIS)

    Plas, Y. van der

    1992-01-01

    This article has been prepared for the regular IAEA/IWG-NPPCI-meeting in May 1991. It provides an outline of the status of and prospects for nuclear energy in the Netherlands and a brief description of topics given specific attention by the authorities. It also gives an overview of recent developments and aims related to I and C in nuclear power stations. (author)

  2. Ardennes nuclear power plant

    International Nuclear Information System (INIS)

    1974-12-01

    The SENA nuclear power plant continued to operate, as before, at authorized rated power, namely 905MWth during the first half year and 950MWth during the second half year. Net energy production:2028GWh; hours phased to the line: 7534H; availability factor: 84%; utilization factor: 84%; total shutdowns:19; number of scrams:10; cost per KWh: 4,35 French centimes. Overall, the plant is performing very satisfactory. Over the last three years net production has been 5900GWh, corresponding to in average utilization factor of 83%

  3. Nuclear power plant

    International Nuclear Information System (INIS)

    Orlov, V.V.; Rineisky, A.A.

    1975-01-01

    The invention is aimed at designing a nuclear power plant with a heat transfer system which permits an accelerated fuel regeneration maintaining relatively high initial steam values and efficiency of the steam power circuit. In case of a plant with three circuits the secondary cooling circuit includes a steam generator with preheater, evaporator, steam superheater and intermediate steam superheater. At the heat supply side the latter is connected with its inlet to the outlet of the evaporator and with its outlet to the low-temperature side of the secondary circuit

  4. Current developments in mechanized non-destructive testing in nuclear power plants; Aktuelle Entwicklungen bei mechanisierten, zerstoerungsfreien Pruefungen in Kernkraftwerken

    Energy Technology Data Exchange (ETDEWEB)

    Zeilinger, R. [intelligeNDT System und Services GmbH und Co. KG, Erlangen (Germany)

    2008-01-15

    Nuclear power plants require frequent in-service activities to be carried out conscientiously in areas potentially hazardous to human operators (because of the associated radiation exposure), such as non-destructive testing of pressurized components of the steam system. Locations to be inspected in this way include the reactor pressure vessel, core internals, steam generators, pressurizers, and pipes. The codes to be used as a basis of these inspections demand high absolute positioning and repeating accuracy. These requirements can be met by mechanized test procedures. Accordingly, many new applications of, mostly mobile, robots have been developed over the past few years. The innovative control and sensor systems for stationary and mobile robots now on the market offer a potential for economic application in a large number of new areas in inspection, maintenance and service in nuclear power plants. More progress in this area is expected for the near future. Areva NP founded the new NDT Center, NETEC (Non-destructive Examination Technical Center), as a global technical center for non-destructive materials testing. NETEC is to advance research and development of all basic NDT technologies, robotics included. For many years, intelligeNDT has offered solutions and products for a variety of inspection and testing purposes and locations in nuclear power plants and is involved in continuous further development of the experience collected in nuclear power plants on the spot. (orig.)

  5. Basic Study of Establishment of Quality Assurance Processes to Develop an Integrated Quality Assurance System for Nuclear Power Plant Construction

    International Nuclear Information System (INIS)

    Lim, Byungki; Moon, Byeongsuk; Lee, Jae Kyoung

    2014-01-01

    An integrated quality assurance system has necessitated carrying out quality assurance programs in a systematic manner because the opportunities to expand business in overseas markets have increased since the export of a nuclear power plant to UAE in 2009. In this study, we use PDCA method to systematically analyze the quality assurance procedures that were used in previous projects for constructing nuclear power plants. We reached a classification system of quality assurance processes at each phase of nuclear power plant construction by integrating similar work related to quality such as planning, design, equipment manufacturing, construction and start-up. We also established a hierarchy of quality assurance processes to develop an integrated quality assurance system as a technology goal to be developed later. To obtain most updated quality assurance activities, a quality assurance process is structured by integrating similar works analyzed from quality assurance procedures through PDCA cycle method. At the implementation phase of Hierarchy of quality processes and sequence of processes for constructing nuclear power plant are established in this study. Integrated quality assurance system is to be developed by connecting organizations as well as stakeholders such as owners, Architect engineering, suppliers, contractors, and sub-contractors to carry out assigned work efficiently

  6. Basic Study of Establishment of Quality Assurance Processes to Develop an Integrated Quality Assurance System for Nuclear Power Plant Construction

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Byungki; Moon, Byeongsuk; Lee, Jae Kyoung [Korea Hydro and Nuclear Power Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    An integrated quality assurance system has necessitated carrying out quality assurance programs in a systematic manner because the opportunities to expand business in overseas markets have increased since the export of a nuclear power plant to UAE in 2009. In this study, we use PDCA method to systematically analyze the quality assurance procedures that were used in previous projects for constructing nuclear power plants. We reached a classification system of quality assurance processes at each phase of nuclear power plant construction by integrating similar work related to quality such as planning, design, equipment manufacturing, construction and start-up. We also established a hierarchy of quality assurance processes to develop an integrated quality assurance system as a technology goal to be developed later. To obtain most updated quality assurance activities, a quality assurance process is structured by integrating similar works analyzed from quality assurance procedures through PDCA cycle method. At the implementation phase of Hierarchy of quality processes and sequence of processes for constructing nuclear power plant are established in this study. Integrated quality assurance system is to be developed by connecting organizations as well as stakeholders such as owners, Architect engineering, suppliers, contractors, and sub-contractors to carry out assigned work efficiently.

  7. Nuclear power plant

    International Nuclear Information System (INIS)

    Aisaka, Tatsuyoshi; Kamahara, Hisato; Yanagisawa, Ko.

    1982-01-01

    Purpose: To prevent corrosion stress cracks in structural materials in a BWR type nuclear power plant by decreasing the oxygen concentration in the reactor coolants. Constitution: A hydrogen injector is connected between the condensator and a condensate clean up system of a nuclear power plant. The injector is incorporated with hydrogenated compounds formed from metal hydrides, for example, of alloys such as lanthanum-nickel alloy, iron titanium alloy, vanadium, palladium, magnesium-copper alloy, magnesium-nickel alloy and the like. Even if the pressure of hydrogen obtained from a hydrogen bomb or by way of water electrolysis is changed, the hydrogen can always be injected into a reactor coolant at a pressure equal to the equilibrium dissociation pressure for metal hydride by introducing the hydrogen into the hydrogen injector. (Seki, T.)

  8. Third generation nuclear plants

    Science.gov (United States)

    Barré, Bertrand

    2012-05-01

    After the Chernobyl accident, a new generation of Light Water Reactors has been designed and is being built. Third generation nuclear plants are equipped with dedicated systems to insure that if the worst accident were to occur, i.e. total core meltdown, no matter how low the probability of such occurrence, radioactive releases in the environment would be minimal. This article describes the EPR, representative of this "Generation III" and a few of its competitors on the world market.

  9. Nuclear power plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1976-01-01

    The invention concerns a quick-acting valve on the main-steam pipe of a nuclear power plant. The engineering design of the valve is to be improved. To the main valve disc, a piston-operated auxiliary valve disc is to be assigned closing a section of the area of the main valve disc. This way it is avoided that the drive of the main valve disc has to carry out different movements. 15 sub-claims. (UWI) [de

  10. Development of an EEG-based workload measurement method in nuclear power plants

    International Nuclear Information System (INIS)

    Choi, Moon Kyoung; Lee, Seung Min; Ha, Jun Su; Seong, Poong Hyun

    2018-01-01

    Highlights: •A human operator’s workload in nuclear power plants(NPPs) usually has been evaluated by using subjective ratings. •Subjective rating techniques have several weaknesses such as dependence on the operator’s memory as well as bias. •We suggested an electroencephalogram (EEG)-based workload index for measuring the workload of human operators. •The suggested index was applied to evaluate the effects of operating support systems. -- Abstract: The environment of main control rooms of large scale process control systems such as nuclear power plants (NPPs) has been changed from the conventional analog type to the digital type. In digitalized advanced main control rooms, human operators conduct highly cognitive work rather than physical work compared to the case of the original control rooms in NPPs. Various operating support systems (OSSs) have been developed to reduce an operator’s workload. Most representative techniques to evaluate the workload are based on subjective ratings. However, there are some limitations including the possibility of skewed results due to self-assessment of the workload and the impossibility of continuously measuring the workload due to freezing simulation for workload assessment. As opposed to subjective ratings techniques, physiological techniques can be used for objective and continuous measurements of a human operator’s mental status by sensing the physiological changes of the autonomic or central nervous system. In this study, electroencephalogram (EEG) was used to measure the operator’s mental workload because it had been proven to be sensitive to variations of mental workload in other studies, and it allows various types of analysis. Based on various research reviews on the characteristics of brainwaves, EEG-based Workload Index (EWI) was suggested and validated through experiments. As a result, EWI is concluded to be valid for measuring an operator’s mental workload and preferable to subjective techniques

  11. A Translator Verification Technique for FPGA Software Development in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Yeob; Kim, Eui Sub; Yoo, Jun Beom [Konkuk University, Seoul (Korea, Republic of)

    2014-10-15

    Although the FPGAs give a high performance than PLC (Programmable Logic Controller), the platform change from PLC to FPGA impose all PLC software engineers give up their experience, knowledge and practices accumulated over decades, and start a new FPGA-based hardware development from scratch. We have researched to fine the solution to this problem reducing the risk and preserving the experience and knowledge. One solution is to use the FBDtoVerilog translator, which translates the FBD programs into behavior-preserving Verilog programs. In general, the PLCs are usually designed with an FBD, while the FPGAs are described with a HDL (Hardware Description Language) such as Verilog or VHDL. Once PLC designer designed the FBD programs, the FBDtoVerilog translates the FBD into Verilog, mechanically. The designers, therefore, need not consider the rest of FPGA development process (e.g., Synthesis and Place and Routing) and can preserve the accumulated experience and knowledge. Even if we assure that the translation from FBD to Verilog is correct, it must be verified rigorously and thoroughly since it is used in nuclear power plants, which is one of the most safety critical systems. While the designer develops the FPGA software with the FBD program translated by the translator, there are other translation tools such as synthesis tool and place and routing tool. This paper also focuses to verify them rigorously and thoroughly. There are several verification techniques for correctness of translator, but they are hard to apply because of the outrageous cost and performance time. Instead, this paper tries to use an indirect verification technique for demonstrating the correctness of translator using the co-simulation technique. We intend to prove only against specific inputs which are under development for a target I and C system, not against all possible input cases.

  12. A Translator Verification Technique for FPGA Software Development in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Jae Yeob; Kim, Eui Sub; Yoo, Jun Beom

    2014-01-01

    Although the FPGAs give a high performance than PLC (Programmable Logic Controller), the platform change from PLC to FPGA impose all PLC software engineers give up their experience, knowledge and practices accumulated over decades, and start a new FPGA-based hardware development from scratch. We have researched to fine the solution to this problem reducing the risk and preserving the experience and knowledge. One solution is to use the FBDtoVerilog translator, which translates the FBD programs into behavior-preserving Verilog programs. In general, the PLCs are usually designed with an FBD, while the FPGAs are described with a HDL (Hardware Description Language) such as Verilog or VHDL. Once PLC designer designed the FBD programs, the FBDtoVerilog translates the FBD into Verilog, mechanically. The designers, therefore, need not consider the rest of FPGA development process (e.g., Synthesis and Place and Routing) and can preserve the accumulated experience and knowledge. Even if we assure that the translation from FBD to Verilog is correct, it must be verified rigorously and thoroughly since it is used in nuclear power plants, which is one of the most safety critical systems. While the designer develops the FPGA software with the FBD program translated by the translator, there are other translation tools such as synthesis tool and place and routing tool. This paper also focuses to verify them rigorously and thoroughly. There are several verification techniques for correctness of translator, but they are hard to apply because of the outrageous cost and performance time. Instead, this paper tries to use an indirect verification technique for demonstrating the correctness of translator using the co-simulation technique. We intend to prove only against specific inputs which are under development for a target I and C system, not against all possible input cases

  13. Development of reliability program for emergency diesel generators in domestic nuclear power plants

    International Nuclear Information System (INIS)

    Kim, Young Ho; Jung, Hyun Jong; Choi, Kwang Hee; Hong, Seoung Yeul

    2001-01-01

    Surveillance tests of Emergency Diesel Generators (EDGs) in Nuclear Power Plants (NPPs) have been conducted periodically to verify the reliability and integrity of the EDGs, however, it was found that these surveillance methods were so conservative and severe as to accelerate the degradation of the EDGs. Hence, new regulatory guideline, Reg. Guide 1.9 Rev. 3, was established by the U.S. NRC to resolve these problems. But it requires the additional implementation of reliability program of the EDGs to improve the actual reliability of them. In Korea, the EDGs of Yonggwang nuclear units 3 and 4 were the first plant applying new Reg. guide 1.9 rev.3 and implementing EDG reliability program. Furthermore it is expected that new guideline for the EDGs will be applied to other EDGs of Korean NPPs. In this paper, this reliability program is described, and it can be used as a reference for other EDGs in Korean NPPs

  14. The development of a task analysis method applicable to the tasks of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Wan Chul; Park, Ji Soo; Baek, Dong Hyeon; Ham, Dong Han; Kim, Huhn [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    While task analysis is one of the essential processes for human factors studies, traditional methods reveal weaknesses in dealing with the cognitive aspects, which become more critical in modern complex system. This report proposes a cognitive task analysis (CTA) method for identifying cognitive requirements of operators' tasks in nuclear power plants. The proposed CTA method is characterized by the information-oriented concept and procedure-based approach. The task prescription identifies the information requirements and trace the information flow to reveal the cognitive organization of task procedure with emphasis to the relations among the information requirements. The cognitive requirements are then analyzed in terms of cognitive span of task information, cognitive envelope and working memory relief point of t procedures, and working memory load. The proposed method is relatively simple and, possibly being incorporated in a full task analysis scheme, directly applicable to the design/evaluation of human-machine interfaces and operating procedures. A prototype of a computerized support system is developed for supporting the practicality of the proposed method. (Author) 104 refs., 8 tabs., 7 figs.

  15. Development of a water separator for steam drying in nuclear power plants

    International Nuclear Information System (INIS)

    Kall, H.

    1979-01-01

    In the wet steam of nuclear power plants with light water cooling the content of liquid phase before entering the high- and low-pressure turbine or at the inlet of the reheater connected in between is limited to a few parts per thousand. For the mechanical drying necessary to maintain this moisture limit there is suited the socalled lamellar separator, in which the vertical laminae, arranged one beside the other, confine a great number of separating ducts, bended in periodically changing sequence, through which wet steam is flowing. In the development of such a lamellar separator with internal drain channels described in this paper the droplet motion in the carrier gas stream and the discharge of the separated liquid is treated independently from each other. The mathematical and experimental study of the first-mentioned partial process leads to its complete description in a dimensionless separation diagram. During the subsequent discharge of the primarily separated droplet fluid there may be torn off droplets by the steam flow and carried away from the separator.The limit for the draining capacity reached with the occurrence of this phenomenon was determined in preliminary experiments with air and water and in experiments with wet steam carried out close to operating conditions. (orig.) 891 GL/orig. 892 RRD [de

  16. The development of monitoring techniques for thermal stratification in nuclear plant piping

    International Nuclear Information System (INIS)

    Sim, Cheul Muu; Joo, Young Sang; Yoon, Kwang Sik; Park, Chi Seung; Choi, Ha Lim; Moon, Jae Wha; Bae, Sang Ho.

    1996-12-01

    The conventional nondestructive testing has been performed in those area which are susceptible to thermal stress in according to NRC 88-08,11. In addition to that, it is necessary to set up a monitoring system to prevent severe thermal stress to pipes in early stages and to develop the non-intrusive techniques to diagnose the check valve because the thermal stratification has been caused by the malfunction of the check valve in ECCS pipe. Thermal stratification monitoring system has been designed and installed at ECCS line permanently and surge line temporally in YG nuclear power plant. The data is acceptable in according to TASCS guide line. Also, the data originated from ISMS is useful for the arrangement of a special UT program and stress analysis. Applying a togetherness of acoustics and magnetics signal, it is possible to determine the parameters of the function of the check valve internals without disassembling it. This series of tests show that the accelerometers can be use d to measure and to differentiate the three types of impacts; metal to metal impacts mechanical rubs, and worn internal parts. The magnet sensors can be used to detect the opening/closing of stainless check and fluttering of disk. (author). 50 refs., 5 tabs., 28 figs

  17. The development of basic glass formulations for solidifying HLW from nuclear fuel reprocessing plant

    International Nuclear Information System (INIS)

    Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

    1995-01-01

    Basic glass formulations 90U/19, 90U/20, 90Nd/7 and 90Nd/10 applied in electric melting process are developed by using the mathematical model of the viscosity and electric resistance of waste glass. The yellow phase does not occur for basic glass formulations 90U/19 and 90U/20 solidifying HLW from nuclear fuel reprocessing plant when the waste loading is 20%. Under the waste loading is 16%, the process and product properties of glass 90U/19 and 90U/20 come up to or surpass the properties of the same kind of foreign waste glasses, and other properties are about the same to them of foreign waste glasses. The process and product properties of basic glass formulations 90Nd/7 and 90Nd/10 used for the solidification of 'U replaced by Nd' liquid waste are almost similar to them of 90U/19 and 90U/20. These properties fairly meet the requirements of 'joint test' (performed at KfK-INE, Germany). Among these formulations, 90Nd/7 is applied in cold engineering scale electric melting test performed at KfK-INE in Germany. The main process properties of cold test is similar to laboratory results

  18. Development of corrosion models to ensure reliable performance of nuclear power plants

    International Nuclear Information System (INIS)

    Kritzky, V.G.; Stjazhkin, P.S.

    1993-01-01

    The safety and reliability of the coolant circuits in nuclear power plants depend much on corrosion and corrosion products transfer processes. Various empirical models have been developed which are applicable to particular sets of operational conditions. In our laboratory a corrosion model has been worked out, which is based on the thermodynamic properties of the compounds, participating in corrosion process and on the assumption, that the corrosion process is controlled by the solubilities of the corrosion products forming the surface oxide layer. The validity of the model has been verified by use of retrospective experimental data, which have been obtained for a series of structural materials such as e.g., carbon and stainless steels, Cu-, Al-, and Zr alloys. With regard for hydriding the model satisfactorily describes stress corrosion cracking process in water-salt environments. This report describes a model based on the thermodynamic properties of the compounds participating in the corrosion process, and on the assumption that the corrosion process is controlled by the solubilities of the corrosion products forming the surface oxide layer

  19. The development of monitoring techniques for thermal stratification in nuclear plant piping

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Cheul Muu; Joo, Young Sang; Yoon, Kwang Sik; Park, Chi Seung; Choi, Ha Lim; Moon, Jae Wha; Bae, Sang Ho

    1996-12-01

    The conventional nondestructive testing has been performed in those area which are susceptible to thermal stress in according to NRC 88-08,11. In addition to that, it is necessary to set up a monitoring system to prevent severe thermal stress to pipes in early stages and to develop the non-intrusive techniques to diagnose the check valve because the thermal stratification has been caused by the malfunction of the check valve in ECCS pipe. Thermal stratification monitoring system has been designed and installed at ECCS line permanently and surge line temporally in YG nuclear power plant. The data is acceptable in according to TASCS guide line. Also, the data originated from ISMS is useful for the arrangement of a special UT program and stress analysis. Applying a togetherness of acoustics and magnetics signal, it is possible to determine the parameters of the function of the check valve internals without disassembling it. This series of tests show that the accelerometers can be use d to measure and to differentiate the three types of impacts; metal to metal impacts mechanical rubs, and worn internal parts. The magnet sensors can be used to detect the opening/closing of stainless check and fluttering of disk. (author). 50 refs., 5 tabs., 28 figs.

  20. The development of a task analysis method applicable to the tasks of nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Wan Chul; Park, Ji Soo; Baek, Dong Hyeon; Ham, Dong Han; Kim, Huhn [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1996-07-01

    While task analysis is one of the essential processes for human factors studies, traditional methods reveal weaknesses in dealing with the cognitive aspects, which become more critical in modern complex system. This report proposes a cognitive task analysis (CTA) method for identifying cognitive requirements of operators' tasks in nuclear power plants. The proposed CTA method is characterized by the information-oriented concept and procedure-based approach. The task prescription identifies the information requirements and trace the information flow to reveal the cognitive organization of task procedure with emphasis to the relations among the information requirements. The cognitive requirements are then analyzed in terms of cognitive span of task information, cognitive envelope and working memory relief point of t procedures, and working memory load. The proposed method is relatively simple and, possibly being incorporated in a full task analysis scheme, directly applicable to the design/evaluation of human-machine interfaces and operating procedures. A prototype of a computerized support system is developed for supporting the practicality of the proposed method. (Author) 104 refs., 8 tabs., 7 figs.

  1. Development of a dynamic model of a BWR nuclear power plant

    International Nuclear Information System (INIS)

    Nonboel, E.

    1975-12-01

    A description is given of a one-dimensional steady-state model of a high-pressure steam turbine, a low-pressure steam turbine, a moisture separator, a reheater, a condenser, feedwater heaters and feedwater pump for a nuclear power plant. The model is contained in the program ''TURBPLANT''. The dynamic part of this model is presented in part II of this report. (author)

  2. Prospects of nuclear power plants for sustainable energy development in Islamic Republic of Iran

    International Nuclear Information System (INIS)

    Ghorashi, Amir Hossien

    2007-01-01

    This paper presents the feasible contributive share of electricity generation from each energy resources. This includes the economical feasibilities and all demographic projections involved in forecasting methodology, which explicitly reflect on overall national power demand projection in the Energy prospects of Islamic Republic of Iran till 2033. The Energy demand and reliability are presented with a view to elaborate on significant role and required capacity of Nuclear Power Plants (NPP) towards fulfillment of an energy mix policy in the country

  3. A belief network approach for development of a nuclear power plant diagnosis system

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, I K; Kim, J T; Lee, D Y; Jung, C H; Kim, J Y; Lee, J S; Ham, C S [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    Belief network (or Bayesian network) based on Bayes` rule in probabilistic theory can be applied to the reasoning of diagnostic system. This paper describes the basic theory of concept and feasibility of using the network for diagnosis of nuclear power plants. An example shows that the probabilities of root causes of a failure are calculated from the measured or believed evidences. 6 refs., 3 figs. (Author)

  4. A belief network approach for development of a nuclear power plant diagnosis system

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, I. K.; Kim, J. T.; Lee, D. Y.; Jung, C. H.; Kim, J. Y.; Lee, J. S.; Ham, C. S. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    Belief network (or Bayesian network) based on Bayes` rule in probabilistic theory can be applied to the reasoning of diagnostic system. This paper describes the basic theory of concept and feasibility of using the network for diagnosis of nuclear power plants. An example shows that the probabilities of root causes of a failure are calculated from the measured or believed evidences. 6 refs., 3 figs. (Author)

  5. Development of Inspection Technique for Socket Weld of Small Bore Piping in Nuclear Power Plant

    International Nuclear Information System (INIS)

    Yoon, Byungsik; Kim, Yongsik; Lee, Jeongseok

    2013-01-01

    The losses incurred by unplanned shutdowns are significant; consequently, early crack initiation and crack detection, including the detection of fillet weld manufacturing defects, is of the utmost importance. Current inspection techniques are not capable of reliably inspecting socket welds; therefore, new approaches are needed. The new technique must be sensitive to socket weld cracking, which usually initiates from the root, in order to detect the cracking during the early failure phase. In 2008, Kori unit 3 experienced leakage from the drain line socket weld of a steam generator. From this experience, KHNP enforced a management program to focus on enhancing the reliability of small bore socket weld piping inspections. Currently, conventional manual ultrasonic inspection techniques are used to detect service induced fatigue cracks. But there was uncertainty on manual ultrasonic inspection because of limited access to the welds and difficulties with contact between the ultrasonic probe and the OD surface of small bore piping. In this study, phased array ultrasonic inspection techniques are applied to increase inspection speed and reliability. Additionally a manually encoded scanner has been developed to enhance contact conditions and maintain constant signal quality. A phased array UT technique and system was developed to inspect small bore socket welds. The experimental results show all artificial flaws in the specimen were detected and measured. These experimental results show, that the newly developed inspection system, has improved the reliability and speed of small bore socket weld inspection. Based on these results, future works shall focus on additional experiments, with more realistic flaw responses. By applying this technique to the field, we expect that it can improve the integrity of small bore piping in nuclear power plants

  6. Development of a web-based monitoring system using operation parameters for the main component in nuclear power plants

    International Nuclear Information System (INIS)

    Son, Dong Chan; An, Kung Il; Hong, Suk Young; Lee, Jeong Soo; Jung, Duk Jin; Shin, Sun Hee; Son, So Hee

    2004-02-01

    The frequency of the damage is increasing, which is caused by the fatigue, according to the increase of running of nuclear power plants. So we need to acquire the reliance of design data to estimate the fatigue and damage of major machinery that might happen as time-dependent crack growth characterization. The research is focused on keeping operating record of nuclear power plants about major machinery which consists of a nuclear reactor pressure boarder on each excessive operating condition including normal operating and extraordinary operating by estimating fracture mechanical movements on real time and fatigue about major nuclear power plants machinery, which are acquired the pressure and temperature data. For further details about the scope and contents of R and D are following. Development of H/W that is necessary to acquire operating real time data of heating and hydraulic power. Selection of a safety variable about major system by each type (the four NPP, all unit). Communication protocol development for connecting between CARE system data base server and fatigue monitoring system data base server. Development of connecting database for controlling and storing of heating and hydraulic power operating data. Real time monitoring system development based on Web using JAVA

  7. Development of a web-based monitoring system using operation parameters for the main component in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Son, Dong Chan; An Kang Il; Hong, Suk Young; Lee, Jeong Soo; Lee, Kwang Yeol; Shin, Sun Hee; Lee, Chun Wha; Son, So Hee [Daesang Information Technology Co., Ltd., Seoul (Korea, Republic of)

    2003-03-15

    The frequency of the damage is increasing, which is caused by the fatigue, according to the increase of running of nuclear power plants. So we need to acquire the reliance of design data to estimate the fatigue and damage of major machinery that might happen as time-dependent crack growth characterization. The research is focused on keeping operating record of nuclear power plants about major machinery which consists of a nuclear reactor pressure boarder on each excessive operating condition including normal operating and extraordinary operating by estimating fracture mechanical movements on real time and fatigue about major nuclear power plants machinery, which are acquired the pressure and temperature data. For further details about the scope and contents of R and D are following: development of H/W that is necessary to acquire operating real time data of heating and hydraulic power, selection of a safety variable about major system by each type (the fourth unit), communication protocol development for connecting between CARE system data base server and fatigue monitoring system data base server, development of connecting database for controlling and storing of heating and hydraulic power operating data, real time monitoring system development based on web using JAVA.

  8. Development of a web-based monitoring system using operation parameters for the main component in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Son, Dong Chan; An, Kung Il; Hong, Suk Young; Lee, Jeong Soo; Jung, Duk Jin; Shin, Sun Hee; Son, So Hee [Daesang Information Technology Co., Ltd., Seoul (Korea, Republic of)

    2004-02-15

    The frequency of the damage is increasing, which is caused by the fatigue, according to the increase of running of nuclear power plants. So we need to acquire the reliance of design data to estimate the fatigue and damage of major machinery that might happen as time-dependent crack growth characterization. The research is focused on keeping operating record of nuclear power plants about major machinery which consists of a nuclear reactor pressure boarder on each excessive operating condition including normal operating and extraordinary operating by estimating fracture mechanical movements on real time and fatigue about major nuclear power plants machinery, which are acquired the pressure and temperature data. For further details about the scope and contents of R and D are following. Development of H/W that is necessary to acquire operating real time data of heating and hydraulic power. Selection of a safety variable about major system by each type (the four NPP, all unit). Communication protocol development for connecting between CARE system data base server and fatigue monitoring system data base server. Development of connecting database for controlling and storing of heating and hydraulic power operating data. Real time monitoring system development based on Web using JAVA.

  9. Development of nuclear power plant online monitoring system using statistical quality control

    International Nuclear Information System (INIS)

    An, Sang Ha

    2006-02-01

    Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCP) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And we made Control Chart Analyzer (CCA) to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability

  10. 78 FR 47012 - Developing Software Life Cycle Processes Used in Safety Systems of Nuclear Power Plants

    Science.gov (United States)

    2013-08-02

    ... for quality assurance programs in Appendix B to 10 CFR Part 50 as they apply to software development... is one of six RG revisions addressing computer software development and use in safety related systems... NUCLEAR REGULATORY COMMISSION [NRC-2012-0195] Developing Software Life Cycle Processes Used in...

  11. The development and evaluation of programmatic performance indicators associated with maintenance at nuclear power plants

    International Nuclear Information System (INIS)

    Wreathall, J.; Fragola, J.; Appignani, P.; Burlile, G.; Shen, Y.

    1990-05-01

    This report summarizes the development and evaluation of programmatic performance indicators of maintenance. These indicators were selected by: (1) creating a formal framework of plant processes; (2) identifying features of plant behavior considered important to safety; (3) evaluating existing indicators against these features; and (4) performing statistical analyses for the selected indicators. The report recommends additional testing. 32 refs., 29 figs., 11 tabs

  12. Expert environment for the development of nuclear power plants failure diagnosis systems

    International Nuclear Information System (INIS)

    Guido, P.N.; Oggianu, S.; Etchepareborda, A.; Fernandez, O.

    1996-01-01

    The present work explores some of the developing stages of an Expert Environment for plant failures Diagnosis Systems starting from Knowledge Based Systems. We present a prototype that carries out an inspection of anomalous symptoms and a diagnosis process based on a Plant Abnormality Model of a PHWR secondary system

  13. Nuclear power for developing countries

    International Nuclear Information System (INIS)

    Hirschmann, H.; Vennemann, J.

    1980-01-01

    The paper describes the energy policy quandary of developing countries and explains why nuclear power plants of a suitable size - the KKW 200 MW BWR nuclear power plant for electric power and/or process steam generation is briefly presented here - have an economic advantage over fossil-fuelled power plants. (HP) [de

  14. R/D and implement of temper bead welding as newly developed maintenance technique in nuclear power plant

    International Nuclear Information System (INIS)

    Hirano, Shinro; Sera, Takehiko; Chigusa, Naoki; Okimura, Koji; Nishimoto, Kazutoshi

    2011-01-01

    Japanese government has recently addressed a policy to increase capacity factor of existing nuclear PPs to achieve the goal to decrease the emission of CO 2 . Numerous preventive measures have taken in nuclear power plants to minimize the risk of unexpected long shutdown. Newly developed mitigation measures or repair methods need to be qualified to satisfy regulatory standards, before it is implemented to nuclear power plants. The qualification process needs to comply regulatory standards though it may consume time to go through each of the required steps. This paper describes such cases namely ambient temper-bead welding and clarifies the issues that need to be resolved regarding qualification process. The qualification process for new methods that has not been prescribed in regulatory standards temporarily completed by go through confirm testing by JAPEIC, RNP and issuance of no action letter in rush. Currently, the qualification process can only be applied on limited area so generalized qualification process needs to be established. (author)

  15. Development of advanced electrodes for corrosion monitoring in nuclear power plants

    International Nuclear Information System (INIS)

    Ku, Hee Kwon; Lim, Dong Seok; Cho, Jae Seon

    2014-01-01

    Much of corrosion-related due to the piping material and coolant are generated for Nuclear Power Plants (NPPs) operation time. During normal-operation, operators manage main factors such as pH, ECP and impurities density by using optimized water-purity operation technique to maintain integrity of piping and structural materials. Various correlations related to the corrosion are developed between metal corrosion and ECP, pH and the effect on the piping and structural, which are the water chemistry factors. In this study, Ag/AgCl and Cu/Cu 2 O(YSZ) pH-sensing electrode for measuring hydrochemistry factors such as ECP, pH operating in NPPs have been developed. The developed sensors are conducted performance tests to prove their validity under the NPPs conditions. In this study, the external Ag/AgCl electrode is developed to resolve the potential drift phenomenon, which is a general phenomenon of conventional Ag/AgCl electrodes, through using the ceramic tube as a potential sensing part and high temperature part, and a related equation is established to calculate the TLJP (Thermal Liquid Junction Potential). The Cu/Cu 2 O electrode as a working electrode is developed using 8%-YSZ membrane. Its relational expression for converting the pH value is established with the Ag/AgCl reference electrode. The developed electrodes are tested to evaluate their response characteristics and stability in low/high temperature conditions. A titration test of the developed electrode is performed using 0.1m-NaCl and 0.01m-HCl solution under the test conditions of 300degC and 2700psi. The test results show that the response characteristics, stability and reproducibility of the manufactured electrodes. Base on the test results, the corrosion environment of carbon-steel (SA106Gr.C) is evaluated by using electrodes completed performance evaluation, and Fepourbaix-diagram is calculated for performance evaluation referred to EPRI report. The conditions of performance evaluation are 1000 ppm

  16. On nuclear power plant uprating

    International Nuclear Information System (INIS)

    Ho, S. Allen; Bailey, James V.; Maginnis, Stephen T.

    2004-01-01

    Power uprating for commercial nuclear power plants has become increasingly attractive because of pragmatic reasons. It provides quick return on investment and competitive financial benefits, while involving low risks regarding plant safety and public objection. This paper briefly discussed nuclear plant uprating guidelines, scope for design basis analysis and engineering evaluation, and presented the Salem nuclear power plant uprating study for illustration purposes. A cost and benefit evaluation of the Salem power uprating was also included. (author)

  17. Emergency control centers for nuclear power plants

    International Nuclear Information System (INIS)

    Anon.

    1979-01-01

    Guidance is provided for the development and implementation of emergency control centers for nuclear power plants, including nuclear plant control room, nuclear plant company headquarters, emergency control center, and nuclear plant alternate emergency control center. Requirements and recommendations are presented for the mission, communications, instrumentation and equipment associated with each type of control center. Decisional aids, manning requirements and resources are also given; the decision aids cover both the accident assessment and protective action areas. Both normal and alternate means of communications are considered. Off-site emergency control centers, although not covered in the strict sense by this standard, are considered in an appendix

  18. Nuclear Power Plant 1996

    International Nuclear Information System (INIS)

    1997-01-01

    Again this year, our magazine presents the details of the conference on Spanish nuclear power plant operation held in February and that was devoted to 1996 operating results. The Protocol for Establishment of a New Electrical Sector Regulation that was signed last December will undoubtedly represent a new challenge for the nuclear industry. By clearing stating that current standards of quality and safety should be maintained or even increased if possible, the Protocol will force the Sector to improve its productivity, which is already high as demonstrated by the results of the last few years described during this conference and by recent sectorial economic studies. Generation of a nuclear kWh that can compete with other types of power plants is the new challenge for the Sector's professionals, who do not fear the new liberalization policies and approaching competition. Lower inflation and the resulting lower interest rates, apart from being representative indices of our economy's marked improvement, will be very helpful in facing this challenge. (Author)

  19. Nuclear power development

    International Nuclear Information System (INIS)

    Nealey, S.

    1990-01-01

    The objective of this study is to examine factors and prospects for a resumption in growth of nuclear power in the United States over the next decade. The focus of analysis on the likelihood that current efforts in the United States to develop improved and safer nuclear power reactors will provide a sound technical basis for improved acceptance of nuclear power, and contribute to a social/political climate more conducive to a resumption of nuclear power growth. The acceptability of nuclear power and advanced reactors to five social/political sectors in the U.S. is examined. Three sectors highly relevant to the prospects for a restart of nuclear power plant construction are the financial sector involved in financing nuclear power plant construction, the federal nuclear regulatory sector, and the national political sector. For this analysis, the general public are divided into two groups: those who are knowledgeable about and involved in nuclear power issues, the involved public, and the much larger body of the general public that is relatively uninvolved in the controversy over nuclear power

  20. Recent developments in the IAEA safety standards: design and operation of nuclear power plants

    International Nuclear Information System (INIS)

    Saito, Takehiko

    2004-01-01

    The IAEA has been publishing a wide variety of safety standards for nuclear and radiation related facilities and activities since 1978. In 1996, a more rigorously structured approach for the preparation and review of its safety standards was introduced. Currently, based on the approach, revision of most of the standards is in completion or near completion. The latest versions of the Safety Requirements for ''Design'' and ''Operation'' of nuclear power plants were respectively published in 2000. Currently, along with this revision of the Safety Requirements, many Safety Guides have been revised. In order to clarify the complicated revision procedure, an example of the entire revision process for a Safety Guide is provided. Through actual example of the revision process, enormous amount of work involved in the revision work is clearly indicated. The current status of all of the Safety Standards for Design and that for Operation of nuclear power plants are summarized. Summary of other IAEA safety standards currently revised and available related IAEA publications, together with information on the IAEA Web Site from where these documents can be downloaded, is also provided. The standards are reviewed to determine whether revision (or new issue) is necessary in five years following publication. The IAEA safety standards will continue to be updated through comprehensive and structured approach, collaboration of many experts of the world, and reflecting good practices of the world. The IAEA safety standards will serve to provide high level of safety assurance. (author)

  1. Development of integrated D/B system for the safety-related structures in nuclear power plant

    International Nuclear Information System (INIS)

    Cho, M. S.; Song, Y. C.; Lee, J. S.; Choi, W. S.

    2002-01-01

    The integrated D/B system is developed for digitalizing the history of the safety-related structures of nuclear power plant. It have 5 database which are consist of Generals, Structural and Design, Materials, Construction, Aging and repair information D/B. For efficient operation of the system, we are to set up the outline of the system, find out data field for target structures, and develop utilities. Utilities will be the aging and repair data management program, the close examination management program, the data search engine with various options which help users to find the information quickly, and the data management program restoring, updating and exchanging input data. Development of the integrated D/B system of the safety-related structures will contribute to management of the structures of nuclear power plant with advanced technology

  2. Engineering development of a digital replacement protection system at an operating US PWR nuclear power plant: Installation and operational experiences

    Energy Technology Data Exchange (ETDEWEB)

    Miller, M.H. [Duke Power Co., Seneca, SC (United States)

    1995-04-01

    The existing Reactor Protection Systems (RPSs) at most US PWRs are systems which reflect 25 to 30 year-old designs, components and manufacturing techniques. Technological improvements, especially in relation to modern digital systems, offer improvements in functionality, performance, and reliability, as well as reductions in maintenance and operational burden. The Nuclear power industry and the US nuclear regulators are poised to move forward with the issues that have slowed the transition to modern digital replacements for nuclear power plant safety systems. The electric utility industry is now more than ever being driven by cost versus benefit decisions. Properly designed, engineered, and installed digital systems can provide adequate cost-benefit and allow continued nuclear generated electricity. This paper describes various issues and areas related to an ongoing RPS replacement demonstration project which are pertinant for a typical US nuclear plant to consider cost-effective replacement of an aging analog RPS with a modern digital RPS. The following subject areas relative to the Oconee Nuclear Station ISAT{trademark} Demonstrator project are discussed: Operator Interface Development; Equipment Qualification; Validation and Verification of Software; Factory Testing; Field Changes and Verification Testing; Utility Operational, Engineering and Maintenance; Experiences with Demonstration System; and Ability to operate in parallel with the existing Analog RPS.

  3. Engineering development of a digital replacement protection system at an operating US PWR nuclear power plant: Installation and operational experiences

    International Nuclear Information System (INIS)

    Miller, M.H.

    1995-01-01

    The existing Reactor Protection Systems (RPSs) at most US PWRs are systems which reflect 25 to 30 year-old designs, components and manufacturing techniques. Technological improvements, especially in relation to modern digital systems, offer improvements in functionality, performance, and reliability, as well as reductions in maintenance and operational burden. The Nuclear power industry and the US nuclear regulators are poised to move forward with the issues that have slowed the transition to modern digital replacements for nuclear power plant safety systems. The electric utility industry is now more than ever being driven by cost versus benefit decisions. Properly designed, engineered, and installed digital systems can provide adequate cost-benefit and allow continued nuclear generated electricity. This paper describes various issues and areas related to an ongoing RPS replacement demonstration project which are pertinant for a typical US nuclear plant to consider cost-effective replacement of an aging analog RPS with a modern digital RPS. The following subject areas relative to the Oconee Nuclear Station ISAT trademark Demonstrator project are discussed: Operator Interface Development; Equipment Qualification; Validation and Verification of Software; Factory Testing; Field Changes and Verification Testing; Utility Operational, Engineering and Maintenance; Experiences with Demonstration System; and Ability to operate in parallel with the existing Analog RPS

  4. Siting nuclear power plants

    International Nuclear Information System (INIS)

    Yellin, J.; Joskow, P.L.

    1980-01-01

    The first edition of this journal is devoted to the policies and problems of siting nuclear power plants and the question of how far commercial reactors should be placed from urban areas. The article is divided into four major siting issues: policies, risk evaluation, accident consequences, and economic and physical constraints. One concern is how to treat currently operating reactors and those under construction that were established under less-stringent criteria if siting is to be used as a way to limit the consequences of accidents. Mehanical cost-benefit analyses are not as appropriate as the systematic use of empirical observations in assessing the values involved. Stricter siting rules are justified because (1) opposition because of safety is growing: (2) remote siting will make the industry more stable; (3) the conflict is eliminated between regulatory policies and the probability basis for nuclear insurance; and (4) joint ownership of utilities and power-pooling are increasing. 227 references, 7 tables

  5. Nuclear Power Plant Lifetime Management Study (I)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

  6. Nuclear Power Plant Lifetime Management Study (I)

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Sung Yull; Jeong, Ill Seok; Jang, Chang Heui; Song, Taek Ho; Song, Woo Young [Korea Electric Power Research Institute, Taejon (Korea, Republic of); Jin, Tae Eun [Korea Power Engineering Company Consulting and Architecture Engineers, (Korea, Republic of); Kim, Woo Chul [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-12-31

    As the operation-year of nuclear power plant increases and finding sites for new nuclear power plant becomes harder, a comprehensive and systematic nuclear plant lifetime management(PLIM) program including life extension has to be established for stable and safe supply of electricity. A feasibility study was conducted to systematically evaluate technical, economic and regulatory aspect of plant lifetime managements and plant life extension for Kori-1 nuclear power plant. For technical evaluation of nuclear power plant, 13 major components were selected for lifetime evaluation by screening system. structure, and components(SSCs) of the plant. It was found that except reactor pressure vessel, which needs detailed integrity analysis, and low pressure turbine, which is scheduled to be replaced, 11 out of 13 major components have sufficient service life, for more than 40 years. Because domestic rules and regulations related to license renewal has not yet been written, review on the regulatory aspect of life extensions was conducted using US NRC rules and regulations. A cooperative effort with nuclear regulatory body is needed for early completion of license renewal rules and regulations. For economic evaluation of plant lifetime extension, a computer program was developed and used. It was found that 10 to 20 year of extension operation of Kori-1 nuclear power plant was proved. Based on the results, next phase of plant lifetime management program for detailed lifetime evaluation and presenting detailed implementation schedule for plant refurbishment for lifetime extension should be followed. (author). 74 refs., figs.

  7. Development of Testing Platform for Digital I and C System in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Park, G. Y.; Kim, Y. M.; Jeong, C. H.

    2013-01-01

    According to digitalization of the NPP (Nuclear Power Plant) I and C (Instrumentation and Control) system, cyber threats against I and C system are increased. Moreover, the complexity of I and C system are increased due to adopt the up-to-date technologies (i. e., smart sensor, wireless network, and Field Programmable Gate Array / Complex Programmable Logic Device) into NPP's I and C system. For example, new issues such as cyber threat are introduced from digitalized I and C systems and components to replace obsolete analog equipment in existing NPPs. Furthermore, use of wireless communication, FPGA/CPLD, and smart sensor could introduce new considerations such as Defense-in-Depth and Diversity. Therefore, the proof testing for digital I and C system is required to verify the adverse effect from use of up-to-date digital technologies and identify the criteria to resolve and mitigate (or prevent) the (possibility of) effects. The objective of this study is developing the Testing Platform for the proof testing. The digital I and C System Test Platform is implemented using test platform hardware, component software, and architectural design. The digital I and C testing platform includes the safety-related PLC and relevant ladder logics, Windows-based C++ codes for host PC. For software, there are seven spike models to confirm the each module's functionality and generate/monitor the signals to/from PLCs. For future work, digital I and C System Test Platform architecture will be implemented using spike models. And a set of acceptance test against cyber security, smart sensor, wireless network, and FPGA/CPLD will be conducted using digital I and C System Test Platform

  8. Development of Testing Platform for Digital I and C System in Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Park, G. Y.; Kim, Y. M.; Jeong, C. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    According to digitalization of the NPP (Nuclear Power Plant) I and C (Instrumentation and Control) system, cyber threats against I and C system are increased. Moreover, the complexity of I and C system are increased due to adopt the up-to-date technologies (i. e., smart sensor, wireless network, and Field Programmable Gate Array / Complex Programmable Logic Device) into NPP's I and C system. For example, new issues such as cyber threat are introduced from digitalized I and C systems and components to replace obsolete analog equipment in existing NPPs. Furthermore, use of wireless communication, FPGA/CPLD, and smart sensor could introduce new considerations such as Defense-in-Depth and Diversity. Therefore, the proof testing for digital I and C system is required to verify the adverse effect from use of up-to-date digital technologies and identify the criteria to resolve and mitigate (or prevent) the (possibility of) effects. The objective of this study is developing the Testing Platform for the proof testing. The digital I and C System Test Platform is implemented using test platform hardware, component software, and architectural design. The digital I and C testing platform includes the safety-related PLC and relevant ladder logics, Windows-based C++ codes for host PC. For software, there are seven spike models to confirm the each module's functionality and generate/monitor the signals to/from PLCs. For future work, digital I and C System Test Platform architecture will be implemented using spike models. And a set of acceptance test against cyber security, smart sensor, wireless network, and FPGA/CPLD will be conducted using digital I and C System Test Platform.

  9. Development of nuclear power plant operator simulator for man-machine interface evaluation

    International Nuclear Information System (INIS)

    Nakagawa, Takashi; Nakatani, Yoshio; Sasaki, Kazunori; Yoshikawa, Hidekazu; Takahashi, Makoto; Furuta, Tomihiko; Hasegawa, Akira.

    1997-01-01

    The operational safety in nuclear power plants depends strongly on man machine interfaces (MMI), such as assignment of equipment on control boards and operation procedures in emergency situations. Therefore, the evaluation and analysis methods for the MMI are important. In order for the methods to be practical, the methods should be executed in each step of design and be easy for designers to use. We aim to develop SEAMAID system: a computer supported system for evaluating and analyzing the MMI by simulating the interaction between the operator and the machine. In this paper, we discuss problems of the conventional methods and the required functions of the operator simulator for the SEAMAID. The operator simulator executes not human errors but correct behavior which follows the operational procedure. The SEAMAID evaluates the MMI by finding potential human errors which could occur in the simulated interactions and points out the problematic interaction parts which could induce human errors. We construct the operator simulator by combining the human model which was proposed by Prof. Reason, and the knowledge base model based on the Petri net model. This simulator can treat frequency parameter which represents the degree of frequency of using a certain knowledge. We conduct two sample simulations in different frequency parameters in the same scenario. These simulation results show that even if the operator behaves correctly following the procedure, there are alternative task sequences. We verify that the simulated interactions are in good agreement with the actual one. Also, we propose the method to point out the problematic parts in the interactions based on the working memory consumption. (author)

  10. Development of nuclear power plant operator simulator for man-machine interface evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Nakagawa, Takashi; Nakatani, Yoshio; Sasaki, Kazunori [Mitsubishi Electric Corp., Amagasaki, Hyogo (Japan). Industrial Electronics and Systems Development Lab.; Yoshikawa, Hidekazu; Takahashi, Makoto; Furuta, Tomihiko; Hasegawa, Akira

    1997-08-01

    The operational safety in nuclear power plants depends strongly on man machine interfaces (MMI), such as assignment of equipment on control boards and operation procedures in emergency situations. Therefore, the evaluation and analysis methods for the MMI are important. In order for the methods to be practical, the methods should be executed in each step of design and be easy for designers to use. We aim to develop SEAMAID system: a computer supported system for evaluating and analyzing the MMI by simulating the interaction between the operator and the machine. In this paper, we discuss problems of the conventional methods and the required functions of the operator simulator for the SEAMAID. The operator simulator executes not human errors but correct behavior which follows the operational procedure. The SEAMAID evaluates the MMI by finding potential human errors which could occur in the simulated interactions and points out the problematic interaction parts which could induce human errors. We construct the operator simulator by combining the human model which was proposed by Prof. Reason, and the knowledge base model based on the Petri net model. This simulator can treat frequency parameter which represents the degree of frequency of using a certain knowledge. We conduct two sample simulations in different frequency parameters in the same scenario. These simulation results show that even if the operator behaves correctly following the procedure, there are alternative task sequences. We verify that the simulated interactions are in good agreement with the actual one. Also, we propose the method to point out the problematic parts in the interactions based on the working memory consumption. (author)

  11. Nuclear reactor plant

    International Nuclear Information System (INIS)

    Schabert, H.P.; Laurer, E.

    1977-01-01

    The invention is concerned with a quick-closing valve on the main-steam pipe of a nuclear reactor plant. The quick-closing valve serves as isolating valve and as safety valve permitting depressurization in case of an accident. For normal operation a tube-shaped gate valve is provided as valve disc, enclosing an auxiliary valve disc to be used in case of accidents and which is opened at increased pressure to provide a smaller flow cross-section. The design features are described in detail. (RW) [de

  12. Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Min; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2014-08-15

    This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness.

  13. Development of a New Safety Culture Assessment Method for Nuclear Power Plants (NPPs) (A study to suggest a new safety culture assessment method in nuclear power plants)

    International Nuclear Information System (INIS)

    Han, Sang Min; Seong, Poong Hyun

    2014-01-01

    This study is conducted to suggest a new safety culture assessment method in nuclear power plants. Criteria with various existing safety culture analysis methods are united, and reliability analysis methods are applied. The concept of the most representative methods, Fault Tree Analysis (FTA) and Failure Mode and Effect Analysis (FMEA), are adopted to assess safety culture. Through this application, it is expected that the suggested method will bring results with convenience and objectiveness

  14. EPRI nuclear power plant decommissioning technology program

    International Nuclear Information System (INIS)

    Kim, Karen S.; Bushart, Sean P.; Naughton, Michael; McGrath, Richard

    2011-01-01

    The Electric Power Research Institute (EPRI) is a non-profit research organization that supports the energy industry. The Nuclear Power Plant Decommissioning Technology Program conducts research and develops technology for the safe and efficient decommissioning of nuclear power plants. (author)

  15. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Byrd, J.S.; Fisher, J.J.; DeVries, K.R.; Martin, T.P.

    1987-01-01

    Expert robots enhance a safety and operations in nuclear plants. E.I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligence, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  16. Expert robots in nuclear plants

    International Nuclear Information System (INIS)

    Byrd, J.S.; Fisher, J.J.; DeVries, K.R.; Martin, T.P.

    1987-01-01

    Expert robots will enhance safety and operations in nuclear plants. E. I. du Pont de Nemours and Company, Savannah River Laboratory, is developing expert mobile robots for deployment in nuclear applications at the Savannah River Plant. Knowledge-based expert systems are being evaluated to simplify operator control, to assist in navigation and manipulation functions, and to analyze sensory information. Development work using two research vehicles is underway to demonstrate semiautonomous, intelligent, expert robot system operation in process areas. A description of the mechanical equipment, control systems, and operating modes is presented, including the integration of onboard sensors. A control hierarchy that uses modest computational methods is being used to allow mobile robots to autonomously navigate and perform tasks in known environments without the need for large computer systems

  17. Garigliano nuclear power plant

    International Nuclear Information System (INIS)

    1976-03-01

    During the period under review, the Garigliano power station produced 1,028,77 million kWh with a utilization factor of 73,41% and an availability factor of 85,64%. The disparity between the utilization and availability factors was mainly due to a shutdown of about one and half months owing to lack of staff at the plant. The reasons for nonavailability (14.36%) break down as follows: nuclear reasons 11,49%; conventional reasons 2,81%; other reasons 0,06%. During the period under review, no fuel replacements took place. The plant functioned throughout with a single reactor reticulation pump and resulting maximum available capacity of 150 MWe gross. After the month of August, the plant was operated at levels slightly below the maximum available capacity in order to lengthen the fuel cycle. The total number of outages during the period under review was 11. Since the plant was brought into commercial operation, it has produced 9.226 million kWh

  18. Passive Nuclear Plants Program (UPDATE)

    International Nuclear Information System (INIS)

    Chimeno, M. A.

    1998-01-01

    The light water passive plants program (PCNP), today Advanced Nuclear Power Plants Program (PCNA), was constituted in order to reach the goals of the Spanish Electrical Sector in the field of advanced nuclear power plants, optimize the efforts of all Spanish initiatives, and increase joint presence in international projects. The last update of this program, featured in revision 5th of the Program Report, reflects the consolidation of the Spanish sector's presence in International programs of the advanced power plants on the basis of the practically concluded American ALWR program. Since the beginning of the program , the PCNP relies on financing from the Electrical sector, Ocide, SEPI-Endesa, Westinghouse, General Electric, as well as from the industrial cooperators, Initec, UTE (Initec- Empresarios Agrupados), Ciemat, Enusa, Ensa and Tecnatom. The program is made up of the following projects, already concluded: - EPRI's Advanced Light Water Plants Certification Project - Westinghouse's AP600 Project - General Electric's SBWR Project (presently paralyzed) and ABWR project Currently, the following project are under development, at different degrees of advance: - EPP project (European Passive Plant) - EBWR project (European Advanced Boiling Water Reactor)

  19. Development of Iridium Solid-state Reference Electrode for the Water Chemistry Status Measurement in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Ku, Heekwon; Lim, Dongseok; Cho, Jaeseon

    2013-01-01

    The result of ECP measurement of piping material in nuclear power plant at low temperature using the developed iridium (SSRE) reference electrode is approximately -0.370V. Based on the various results of this study, the developed iridium (SSRE) reference electrode can be applied to the water chemistry environments of nuclear power plant. Various metallic materials used in a nuclear power plant have been exposed to a variety of water chemistry environments and the corrosion of metallic materials occurs due to the reactions between metal structures and water chemistry environments. Therefore, the management of the water chemistry factors is needed to prevent corrosion. The chemical factors affecting the corrosion are pH and Electrochemical Corrosion Potential (ECP). The world-wide studies suggest that ECP and pH are effective indicators for preventing the material damage from water chemistry condition. ECP and pH should be measured as the reference electrodes, and should show stable potential characteristics with fast responses. In this study, the iridium reference electrodes using a solid-state metal oxide electrode has been developed to measure effective indicators such as ECP and pH. The iridium (SSRE) reference electrode for the ECP measurement in water chemistry environment of nuclear power plants has been developed. A calibration for water chemistry measurement was performed by potential measurement of iridium (SSRE) reference electrode with Ag/AgCl (SSRE) reference electrode. The result exhibited a stable potential for 117 hours and a super-Nernst ian response with 63.12mV/p H. In this study, the iridium (SSRE) reference electrode shows super-Nernst ian characteristic and it may be caused by the property of electrolytically coated iridium oxide. Considering the long-term stability of the developed electrode, it is possible to apply as a reference electrode through calibration procedure

  20. Operational experience, evolution and developments in water chemistry in Indian Nuclear Power Plants - an overview

    International Nuclear Information System (INIS)

    Prasad, Y.S.R.

    2000-01-01

    Lessons learnt from the experiences at nuclear power plants have enriched the understanding of corrosion behaviour in water systems. The need for proper water chemistry control not only during operation but also during fabrication and preoperational tests is clearly seen. It should not be construed that maintenance of proper water chemistry is a panacea for all corrosion and other associated problems. Unless adequate care is taken in selection of material and sound design and fabrication practices are followed, no regime of water chemistry can help in eliminating failure due to corrosion

  1. Development of procedural requirements for life extension of nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Sang Hoon; Son, Moon Kyu; Jeong, Ji Hwan; Chang, Keun Sun; Ham, Chul Hoon; Chang, Soon Hong

    2002-03-01

    Technical issues relevant to life extension of NPP were investigated. The GALL report, domestic PSR and periodic inspection rules were reviewed. Technical issues appearing in the safety evaluation reports related to license renewal of Calvert Ciffs 1 and 2 and Qconee 1,2 and 3 NPPs were reviewed. Preliminary study on PSA usage in NPP life extension assessment was performed and further works were suggested. The environment of rules and regulations was analyzed from the viewpoint of plant life extension. Two alternatives are suggested to revise the current domestic nuclear acts

  2. Development of procedural requirements for life extension of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Hoon; Son, Moon Kyu [Korea Association for Nuclear Technology, Taejon (Korea, Republic of); Jeong, Ji Hwan [Baekseok College, Cheonan (Korea, Republic of); Chang, Keun Sun [Sunmoon Univ., Asan (Korea, Republic of); Ham, Chul Hoon [The Catholic University of Korea, Seoul (Korea, Republic of); Chang, Soon Hong [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    2002-03-15

    Technical issues relevant to life extension of NPP were investigated. The GALL report, domestic PSR and periodic inspection rules were reviewed. Technical issues appearing in the safety evaluation reports related to license renewal of Calvert Ciffs 1 and 2 and Qconee 1,2 and 3 NPPs were reviewed. Preliminary study on PSA usage in NPP life extension assessment was performed and further works were suggested. The environment of rules and regulations was analyzed from the viewpoint of plant life extension. Two alternatives are suggested to revise the current domestic nuclear acts.

  3. Trends in development of probability assessment of nuclear power plant safety

    International Nuclear Information System (INIS)

    Dach, K.

    1989-01-01

    A complete study of probability safety assessment (PSA) of nuclear power plants is a multidisciplinary endeavor, requiring a qualified decision-making team composed of experienced professionals in individual disciplines and requiring good coordination of effort. The main concerns for the execution of a PSA study and related tasks are schematically presented. Also shown is a summary of the main steps for a PSA study at all three levels, with the incorporation of analysis of external events and the reliability of humans, including the necessary uncertainty analyses. 1 ref., 2 figs., 3 tabs

  4. Analytic advanced development in the German Risk Study on Nuclear Power Plants

    International Nuclear Information System (INIS)

    1984-01-01

    The study carried out by the Ecological Institute (Oeko-Institut) deals with the difficulties encountered with the decision-making models and statistical models and methods of the German Risk Study on Nuclear Power Plants for the determination of care meltdown frequency and for the estimation of the behaviour of components in the event of a failure. Related deficiencies of the German Risk Study and the absence of completeness and lading representiveness of the course and causes of incidents considered are pointed out. The accident consequences model and the climatological-meteorological data base are considered non-representative. (RF) [de

  5. Guiding rules for development of intelligent monitoring system of nuclear power plants

    International Nuclear Information System (INIS)

    Kitamura, M.; Furukawa, H.; Kozma, R.; Washio, T.

    1996-01-01

    General frameworks and major component techniques for intelligent monitoring of nuclear power plants are presented. The key concept, diversity-based design, is to provide advisory information through consensus of multiple agents, each performing operational decision-making by focusing on mutually different information obtained from the plant. The multi-agent design scheme allows to attain high credibility and tolerance against sensor failure in fault detection and causal reasoning. The advantage of the proposed scheme realized by multiple neural networks was clearly demonstrated through numerical experiments with anomalies in a pressurized water reactor. Relevant techniques are also introduced for diagnostic information evaluation in specified symptoms, and for remedial procedure synthesis. A new architecture for future implementation of the proposed scheme, worm-type multi-agent system, is also proposed as a promising candidate. (author)

  6. Development of Export Control Comprehensive Management Model for Nuclear Power Plants and Others Projects

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chansuh; Seo, Hana; Choi, Sundo [Korea Institute of Nuclear Nonproliferation And Control, Daejeon (Korea, Republic of)

    2014-05-15

    It is required that there are lots of managements of care and concern if the project contains strategic items such as NPPs. The Korean nuclear industry and its related companies, such as the Korea Hydro and Nuclear Power (KHNP), are promoting greater exports of NPPs. It is likely that Korea will export more this technology to newcomer states in the future. As a result, the ROK has been improving its export control management system for NPPs. In keeping with this national effort, Korea Institute of Nuclear Nonproliferation And Control (KINAC) developed comprehensive export control management model for NPPs and other projects, in preparation for this projected growth in the industry. This model also applies to the nuclear export case of the UAE, aims to manage the project from bidding to the end of the contract. The recent Export Licensing of Nuclear Facility Technology was reflected in the Notice on Export and Import of Strategic Items in January 2014. Through this license, the large-scale project legislation framework was established. It can also minimize nonproliferation concerns of the international community through strict management. It is expected that the Korea will be able to enhance transparency and secure the nuclear use, while meeting nonproliferation purpose.

  7. Development of Export Control Comprehensive Management Model for Nuclear Power Plants and Others Projects

    International Nuclear Information System (INIS)

    Lee, Chansuh; Seo, Hana; Choi, Sundo

    2014-01-01

    It is required that there are lots of managements of care and concern if the project contains strategic items such as NPPs. The Korean nuclear industry and its related companies, such as the Korea Hydro and Nuclear Power (KHNP), are promoting greater exports of NPPs. It is likely that Korea will export more this technology to newcomer states in the future. As a result, the ROK has been improving its export control management system for NPPs. In keeping with this national effort, Korea Institute of Nuclear Nonproliferation And Control (KINAC) developed comprehensive export control management model for NPPs and other projects, in preparation for this projected growth in the industry. This model also applies to the nuclear export case of the UAE, aims to manage the project from bidding to the end of the contract. The recent Export Licensing of Nuclear Facility Technology was reflected in the Notice on Export and Import of Strategic Items in January 2014. Through this license, the large-scale project legislation framework was established. It can also minimize nonproliferation concerns of the international community through strict management. It is expected that the Korea will be able to enhance transparency and secure the nuclear use, while meeting nonproliferation purpose

  8. Development of the simulation system {open_quotes}IMPACT{close_quotes} for analysis of nuclear power plant severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi [Nuclear Power Corp. (Japan)] [and others

    1997-07-01

    The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system {open_quotes}IMPACT{close_quotes} for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT`s distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data.

  9. Development of Real Time Operating System for Safety Grade PLC (POSAFE-Q) for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Son, Han Seong [ENESYS, Taejon (Korea, Republic of); Hwang, Sung Jae [POSCON, Seoul (Korea, Republic of); Lee, Young Joon; Kim, Chang Hwoi; Lee, Dong Young [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    POSAFE-Q is a newly developed programmable logic controller (PLC) in order to apply to digital safety system of nuclear power plants (NPP) according to Nuclear Power Plant safety requirements. POSAFE-Q hardware and software development process, including design, review, verification and validation, and configuration control and quality assurance, satisfies the requirements imposed by 10CFR50, Appendix B. This article introduces a real time operating system pCOS, which is the core of POSAFE-Q. Section 2 describes the structure of pCOS. Section 3 describes a few important features of pCOS, which are necessary to the application for the digital safety system of NPP.0.

  10. Development of Behavioral Indicators of Competences for Safety Culture of Nuclear Power Plants: A Preliminary Study

    International Nuclear Information System (INIS)

    Moon, Kwangsu; Kim, Sa Kil; Oh, Yeon Ju; Shin, Youmin; Lee, Yong-Hee; Jang, Tong Il

    2015-01-01

    The term of safety competency in nuclear field was presented in the OECD/NEA workshop held in 1999. A model of the safety culture competencies in nuclear power plants was developed by KAERI (Korea Atomic Energy Research Institute). In general, a competency (competence) is defined as 'cluster of employee's attribute, knowledge, skill, ability or other characteristic that contributes to successful job performance'. We also defined safety culture competency as 'cluster of various internal characteristics (e.g., knowledge, skill, ability, motive, attitude and etc.) of employee that contribute to perform job safely and shape a healthy and strong safety culture.' By this definition, the safety culture competency is the broader construct including job competency. An employee having high level of safety culture competency shows extra discretionary effort to improve safety of peer, team and organization in addition to the individual's successful and safe job accomplishment. The behavioral indicators for each of the competencies are focal points of conversations on progress and are monitored continuously by self-assessment and managers or supervisors' intervention. Deficiencies in any of these indicators can point to coaching, training or other learning opportunities that employees may be required in order to improve. The purpose of this study was to derive a model of safety competencies for improving safety culture of NPPs and develop a set of behavioral indicators of each competency. In addition, the method of measuring behavioral indicators was suggested. For the application of developed safety culture competences and behavioral indicators, the most suitable measuring method for behavioral indicators must be developed. In the case of behavioral observations, behavioral dimensions (frequency, persistence and latency), observation possibility, occurrence basis of behavior (daily job performance, situational dependent) are considered to

  11. Development of Behavioral Indicators of Competences for Safety Culture of Nuclear Power Plants: A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Moon, Kwangsu; Kim, Sa Kil; Oh, Yeon Ju; Shin, Youmin; Lee, Yong-Hee; Jang, Tong Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The term of safety competency in nuclear field was presented in the OECD/NEA workshop held in 1999. A model of the safety culture competencies in nuclear power plants was developed by KAERI (Korea Atomic Energy Research Institute). In general, a competency (competence) is defined as 'cluster of employee's attribute, knowledge, skill, ability or other characteristic that contributes to successful job performance'. We also defined safety culture competency as 'cluster of various internal characteristics (e.g., knowledge, skill, ability, motive, attitude and etc.) of employee that contribute to perform job safely and shape a healthy and strong safety culture.' By this definition, the safety culture competency is the broader construct including job competency. An employee having high level of safety culture competency shows extra discretionary effort to improve safety of peer, team and organization in addition to the individual's successful and safe job accomplishment. The behavioral indicators for each of the competencies are focal points of conversations on progress and are monitored continuously by self-assessment and managers or supervisors' intervention. Deficiencies in any of these indicators can point to coaching, training or other learning opportunities that employees may be required in order to improve. The purpose of this study was to derive a model of safety competencies for improving safety culture of NPPs and develop a set of behavioral indicators of each competency. In addition, the method of measuring behavioral indicators was suggested. For the application of developed safety culture competences and behavioral indicators, the most suitable measuring method for behavioral indicators must be developed. In the case of behavioral observations, behavioral dimensions (frequency, persistence and latency), observation possibility, occurrence basis of behavior (daily job performance, situational dependent) are considered to

  12. Development of an integrated condition monitoring and diagnostic system for motor-operated valves used in nuclear power plant

    International Nuclear Information System (INIS)

    Carneiro, Alvaro Luiz Guimaraes

    2003-01-01

    The reliability question of the components, specifically of motor operated valves, became one of the most important issues to be investigated in nuclear power plants, considering security and life plant extension. Therefore, the necessity of improvements in monitoring and diagnosis methods started to be of extreme relevance in the maintenance predictive field, establishing as main goal the reliability and readiness of the system components. Specially in nuclear power plants, the predictive maintenance contributes in the security factor in order to diagnosis in advance the occurrence of a possible failure, preventing severe situations. It also presents a contribution on the economic side by establishing a better maintenance programming, and reducing unexpected shutdown. The development of non intrusive monitoring and diagnostic method makes it possible to identify malfunctions in plant components during normal plant operation. This dissertation presents the development of an integrated condition monitoring system for motor-operated valves used in nuclear power plants. The methodology used in this project is based on the electric motor power signatures analysis, during the closing and opening stroke time of the valve. Once the measurements baseline diagnostic of the motor-operated valve is taken, it is possible to detect long-term deviations during valve lifetime, detecting in advance valve failures. The system implements two parallel techniques for detection and categorization of anomalies: expert system using fuzzy logic based on rules and knowledge base, providing a systematic approach for decision making, and the Wavelet Transform Technique, where the main goal is to obtain more detailed information contained in the measured data, identifying and characterizing the transients phenomena in the time and frequency domains, correlating them to failures situations in the incipient stage. The conditioning monitoring and diagnostic system was designed and implemented at

  13. Technical and economical prerequisites of special district-heating nuclear power plant development

    International Nuclear Information System (INIS)

    Baturov, B.B.; Boldyrev, V.M.; Losev, V.L.; Sigal, M.V.

    1983-01-01

    Results are presented of technical and economical analysis of advisability of constructing combined Nuclear Power and Heating Plants (NPHP) assuring the possibility of their location near the areas of heat power consumption in case of observing a given degree of radiation safety for population and personnel. Specific features determining the choice of turbine-driven units for such plants are analyzed. Conditions of competiveness of a specialized NPHP with alternative power units, NPHP based on the WWER-1000 reactor and district heating plants (NDHP), are determined. Analysis of design specifications of NPHP with two VK-500 reactor units and structures of capital investments in such a plant reveal that an increase in the total capital investments in the NPHP would not exceed 2% with account of the difference in costs for grid heaters and a corresponding change in dimensions of operation rooms as well as changes in costs of heat removal system (within one site) at the TK turbine replacement by the T turbine

  14. Developing a virtual reality application for training Nuclear Power Plant operators: Setting up a database containing dose rates in the refuelling plant

    International Nuclear Information System (INIS)

    Rodenas, J.; Zarza, I.; Burgos, M. C.; Felipe, A.; Sanchez-Mayoral, M. L.

    2004-01-01

    Operators in Nuclear Power Plants can receive high doses during refuelling operations. A training programme for simulating refuelling operations will be useful in reducing the doses received by workers as well as minimising operation time. With this goal in mind, a virtual reality application is developed within the framework of the CIPRES project. The application requires doses, both instantaneous and accumulated, to be displayed at all times during operator training. Therefore, it is necessary to set up a database containing dose rates at every point in the refuelling plant. This database is based on radiological protection surveillance data measured in the plant during refuelling operations. Some interpolation routines have been used to estimate doses through the refuelling plant. Different assumptions have been adopted in order to perform the interpolation and obtain consistent data. In this paper, the procedures developed to set up the dose database for the virtual reality application are presented and analysed. (authors)

  15. Development of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

    International Nuclear Information System (INIS)

    Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun

    2004-02-01

    The objective of this research is to develop on efficient integrity evaluation technology and to investigate the applicability of the newly-developed technology such as internet-based cyber platform etc. to Nuclear Power Plant(NPP) components. The development of an efficient structural integrity evaluation system is necessary for safe operation of NPP as the increase of operating periods. Moreover, material test data as well as emerging structural integrity assessment technology are also needed for the evaluation of aged components. The following five topics are covered in this project: development of the wall-thinning evaluation program for nuclear piping; development of structural integrity evaluation criteria for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for major components of NPP; ingegration of internet-based cyber platform and integrity evaluation program for primary components of NPP; effects of aging on strength of dissimilar welds

  16. Development of integrity evaluation technology for pressurized components in nuclear power plant and IT based integrity evaluation system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Jin; Choi, Jae Boong; Shim, Do Jun [Sungkyunkwan Univ., Seoul (Korea, Republic of)] (and others)

    2004-02-15

    The objective of this research is to develop on efficient integrity evaluation technology and to investigate the applicability of the newly-developed technology such as internet-based cyber platform etc. to Nuclear Power Plant(NPP) components. The development of an efficient structural integrity evaluation system is necessary for safe operation of NPP as the increase of operating periods. Moreover, material test data as well as emerging structural integrity assessment technology are also needed for the evaluation of aged components. The following five topics are covered in this project: development of the wall-thinning evaluation program for nuclear piping; development of structural integrity evaluation criteria for steam generator tubes with cracks of various shape; development of fatigue life evaluation system for major components of NPP; ingegration of internet-based cyber platform and integrity evaluation program for primary components of NPP; effects of aging on strength of dissimilar welds.

  17. Nuclear power plant emergency preparedness

    International Nuclear Information System (INIS)

    2005-01-01

    The guide sets forth detailed requirements on how the licensee of a nuclear power plant shall plan, implement and maintain emergency response arrangements. The guide is also applied to nuclear material and nuclear waste transport in situations referred to in guide YVL 6.5. Requirements on physical protection are presented in a separate guide of Finnish Radiation and Nuclear Safety Authority (STUK)

  18. Nuclear plant analyzer desktop workstation

    International Nuclear Information System (INIS)

    Beelman, R.J.

    1990-01-01

    In 1983 the U.S. Nuclear Regulatory Commission (USNRC) commissioned the Idaho National Engineering Laboratory (INEL) to develop a Nuclear Plant Analyzer (NPA). The NPA was envisioned as a graphical aid to assist reactor safety analysts in comprehending the results of thermal-hydraulic code calculations. The development was to proceed in three distinct phases culminating in a desktop reactor safety workstation. The desktop NPA is now complete. The desktop NPA is a microcomputer based reactor transient simulation, visualization and analysis tool developed at INEL to assist an analyst in evaluating the transient behavior of nuclear power plants by means of graphic displays. The NPA desktop workstation integrates advanced reactor simulation codes with online computer graphics allowing reactor plant transient simulation and graphical presentation of results. The graphics software, written exclusively in ANSI standard C and FORTRAN 77 and implemented over the UNIX/X-windows operating environment, is modular and is designed to interface to the NRC's suite of advanced thermal-hydraulic codes to the extent allowed by that code. Currently, full, interactive, desktop NPA capabilities are realized only with RELAP5

  19. The development of robotic system for the nuclear power plants -The development of advanced robotics for the nuclear industry-

    International Nuclear Information System (INIS)

    Kim, Seung Ho; Kim, Chang Hoi; Kim, Byung Soo; Lee, Yong Bum; Hwang, Suk Yeoung; Kim, Woong Ki; Park, Soon Yong; Lee, Young Kwang; Lee, Jae Gyeong; Seo, Yong Chil

    1994-07-01

    The omni-directional planetary wheel type mobile robot (KAEROT/ml) has been developed on the concepts of the modularity, reliability, and stability. Manipulator system is designed in order to be able to install on mobile system and to upgrade operating capability. Control system consists of 68030 processor board, servo motor controller and I/O board. The 6 DOFs hydraulic manipulator is designed for nozzle dam installation and removal. The reachable length of manipulator is 1.7 m with the wrist configuration of pitch-yaw-roll. For the easy installation of manipulator from outside steam generator, specially designed guider is considered. Also master manipulator is designed for force feedback control. RCP (Remote Control Part) is constructed with Sparc processor boards and servo control boards. Graphic simulation has done for the evaluation of manipulator performance of nozzle dam works. (Author)

  20. The Evaluation and Application Plan Report for the Development of Nuclear Power Plant DCS Using CASE Tools

    Energy Technology Data Exchange (ETDEWEB)

    Lee, B.Y.; Moon, H.J.; Yoon, M.H.; Lee, Y.K. [Korea Electric Power Research Institute, Taejon (Korea)

    2000-06-01

    This report contains the evaluation and application plan report for the development of nuclear power plant DCS using CASE tools. In this report, the necessity of using CASE tools is considered and a available CASE environment is suggested. And, also according to the IEEE Std 1209 Recommended Practice for Evaluation and Selection of CASE Tools, their functional and economical evaluation about available commercial CASE tools is performed and described. (author). 6 figs., 3 tabs.

  1. Real-time variables dictionary (RTVD), and expert system for development of real-time applications in nuclear power plants

    International Nuclear Information System (INIS)

    Senra Martinez, A.; Schirru, R.; Dutra Thome Filho, Z.

    1990-01-01

    It is presented in this paper a computerized methodology based on a data dictionary managed by an expert system called Real-Time Variables Dictionary (RTVD). This system is very usefull for development of real-time applications in nuclear power plants. It is described in details the RTVD functions and its implantation in a VAX 8600 computer. It is also pointed out the concepts of artificial intelligence used in teh RTVD

  2. Development of RCM methodology and tools for EDF nuclear power plants

    International Nuclear Information System (INIS)

    Jacquot, J.P.; Bouchet, J.L.; Despujols, A.; Dewailly, J.; Martin-Mattei, C.

    1992-06-01

    This paper outlines the development of Reliability-Centered Maintenance procedures in the nuclear power industry. It presents the pilot study undertaken by EDF in the overall framework of its 'OMF' (RCM) project, as well as the potential for further improving and enriching analytical methods. Lastly, it gives the prospects for the future design of an 'OMF' workstation

  3. Development of Intelligent Database Program for PSI/ISI Data Management of Nuclear Power Plant (Part II)

    International Nuclear Information System (INIS)

    Park, Un Su; Park, Ik Keun; Um, Byong Guk; Lee, Jong Po; Han, Chi Hyun

    2000-01-01

    In a previous paper, we have discussed the intelligent Windows 95-based data management program(IDPIN) which was developed for effective and efficient management of large amounts of pre-/in-service inspection(PSI/ISI) data of Kori nuclear power plants. The IDPIN program enables the prompt extraction of previously conducted PSI/ISI conditions and results so that the time-consuming data management, painstaking data processing and analysis of the past are avoided. In this study, the intelligent Windows based data management program(WS-IDPIN) has been developed as an effective data management of PSI/ISI data for the Wolsong nuclear power plants. The WS-IDPIN program includes the modules of comprehensive management and analysis of PSI/ISI results, statistical reliability assessment program of PSI/ISI results(depth and length sizing performance etc), standardization of UT report form and computerization of UT results. In addition, the program can be further developed as a unique PSI/ISI data management expert system which can be part of the PSI/ISI total support system for Korean nuclear power plants

  4. Development of a mobile manipulator for nuclear plant disaster, HELIOS X. Mechanical design and basic experiments

    International Nuclear Information System (INIS)

    Noda, Satsuya; Hirose, Shigeo; Ueda, Koji; Nakano, Hisami; Horigome, Atsushi; Endo, Gen

    2016-01-01

    In places such as nuclear power plant disaster area, which it is difficult for human workers to enter, robots are required to scout those places instead of human workers. In this paper, we present a mobile manipulator HELIOS X for a nuclear plant decommissioning task. Firstly, we address demands and specifications for the robot, considering the mission of reconnaissance. Then we outline the system of the robot, mainly focusing on the following mechanism: 'Crank Wheel', 'Main Arm', 'Sphere Link Wrist', 'Camera Arm', 'Control System' and 'System architecture'. Especially, we installed 3 degree of freedom 'Camera Arm' on the 'Main Arm', in order to improve functionality of remote control system. This enables the operator to monitor both the gripper and its overall view of the robot. 'Camera Arm' helps the operator to recognize the distance from an object to the gripper, because the operator can interactively move the viewpoint of the camera, and monitor from another camera angle without changing the gripper's position. We confirmed the basic functionality of mobile base, 'Main Arm' and 'Camera Arm' through hardware experiments. We also demonstrated that HELIOS X could pass through the pull-to-open door with a substantial closing force when the operator watched camera view only. (author)

  5. Support analysis for safety analysis development for CANDU nuclear power plant

    International Nuclear Information System (INIS)

    Bedreaga, L.; Florescu, Gh.; Apostol, M.; Nitoi, M.

    2004-01-01

    Probabilistic Safety Assessment analysis (PSA) is a technique used to assess the safety of a nuclear power plant. Assessments of the nuclear plant systems/components from safety point of view consist in accomplishment of a lot of support analyses that are the base for the main analysis, in order to evaluate the impact of occurrences of abnormal states for these systems. Evaluation of initiating events frequency and components failure rate is based on underlying probabilistic theory and mathematic statistics. Some of these analyses are detailed analyses and are known very well in PSA. There are also some analyses, named support analyses for PSA, which are very important but less applicable because they involve a huge human effort and hardware facilities to accomplish. The usual methods applicable in PSA such as input data extracted from the specific documentation (operation procedures, testing procedures, maintenance procedures and so on) or conservative evaluation provide a high level of uncertainty for both input and output data. The paper describes support analysis required to improve the certainty level in evaluation of reliability parameters and also in the final results (either risk, reliability or safety assessment). (author)

  6. Development of a Relap based Nuclear Plant Analyser with 3-D graphics using OpenGL and Object Relap

    International Nuclear Information System (INIS)

    Lee, Young Jin

    2010-01-01

    A 3-D Graphic Nuclear Plant Analyzer (NPA) program was developed using GLScene and the TRelap. GLScene is an OpenGL based 3D graphics library for the Delphi object-oriented program language, and it implements the OpenGL functions in forms suitable for programming with Delphi. TRelap is an object wrapper developed by the author to easily implement the Relap5 thermal hydraulic code under object oriented programming environment. The 3-D Graphic NPA was developed to demonstrate the superiority of the object oriented programming approach in developing complex programs

  7. Development of a Desktop Simulator for APR1400 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Lee, J. B.

    2016-01-01

    It is essential for utilities to possess a full-scope simulator for operator training and operation test for operators. But it is very expensive and sometimes lack of fidelity if processes of developing the simulator and designing the plant are in parallel. It is due to the situation that simulator development stage sometimes precedes the plant design stage and modifications may occur to the design of the plant in construction stage. In an attempt to build a low cost and efficient simulator, a desktop simulator has been developed. This model is described herein. Using desktop simulators for training operators is an efficient method for familiarizing operators with their plant’s operation. A low cost and efficient desktop simulator for APR1400 has been developed, and brief features are introduced here. It is configured to mimic a full-scale simulator, and can be used for operators to be familiarized to their plant’s operation. Since the size of the simulator is small enough to be fit in a desk, it can be used in a classroom or in an office at any time. It can also be used to evaluate design changes or modifications of the plant before implementing them to the plant

  8. Development of a Desktop Simulator for APR1400 Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J. B. [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    It is essential for utilities to possess a full-scope simulator for operator training and operation test for operators. But it is very expensive and sometimes lack of fidelity if processes of developing the simulator and designing the plant are in parallel. It is due to the situation that simulator development stage sometimes precedes the plant design stage and modifications may occur to the design of the plant in construction stage. In an attempt to build a low cost and efficient simulator, a desktop simulator has been developed. This model is described herein. Using desktop simulators for training operators is an efficient method for familiarizing operators with their plant’s operation. A low cost and efficient desktop simulator for APR1400 has been developed, and brief features are introduced here. It is configured to mimic a full-scale simulator, and can be used for operators to be familiarized to their plant’s operation. Since the size of the simulator is small enough to be fit in a desk, it can be used in a classroom or in an office at any time. It can also be used to evaluate design changes or modifications of the plant before implementing them to the plant.

  9. Development of a dose database in the refuelling scenario of a nuclear power plant for a virtual reality application

    International Nuclear Information System (INIS)

    Rodenas, J.; Zarza, I.; Pascual, A.; Felipe, A.; Sanchez-Mayoral, M.L.

    2002-01-01

    Operators in Nuclear Power Plants can receive high doses during refuelling operation. A training program simulating refuelling operations will be useful to reduce doses received by workers as well as to minimise operation time. With this goal in mind a Virtual Reality application is developed in the frame of CIPRES Project (Calculos Interactivos de Proteccion Radiologica en un Entorno de Simulacion - Interactive Calculations of Radiological Protection in a Simulation Environment), a RD project sponsored by IBERINCO and developed jointly by IBERINCO and the Nuclear Engineering Department of the Polytechnic University of Valencia. The Virtual Reality application requires the possibility of displaying doses, both instantaneous and accumulated, at all times during the operator training. Therefore, it is necessary to elaborate a database containing dose rates at every point of the refuelling plant. This database is elaborated from Radiological Protection Surveillance data measured throughout the plant during refuelling operation. To estimate doses throughout the refuelling plant some interpolation routines have been used. Different assumptions have been adopted in order to perform the interpolation and obtain consistent data. In this paper, procedures developed to elaborate the dose database for the Virtual Reality application are presented and analysed

  10. Basic Study on Data-Centric design information integration system framework development for adapting Nuclear Power Plant construction in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Byung Ki [KHNP, Gyeongju (Korea, Republic of)

    2016-05-15

    This study established the concept of data-centric design, which is the latest design technique, by analyzing the existing literature so that the data-centric design would be applied to the nuclear power plant projects in Korea and analyzed the status of data-centric design application by the advanced companies and the domestic design companies participating in the nuclear power plant projects. By analyzing the function of the 3D CAD commercial system and all design drawings used in the nuclear power plant projects in Korea, a data-centric design integrated system model has been developed. This study established the concept of data-centric design technology, analyzed the functions of the plant architect engineering (A/E) software being globally used in the plant field and the design process status of nuclear power plant projects in Korea. A design information integration system building model, which is capable of data-centric design, in the place of the existing document-centric system design such as P and ID and SLD, has been suggested through the investigation on the data-centric design cases of the advanced companies. The major functions of the suggested model required for the application to the domestic industry were drawn. The suggested framework builds the field design, which was performed in the 3D system of the constructor, as an owner's field design system, which can manage all design drawings generated from the field design and the related information in integrated way. An as-built full model integrated of plant architect engineering, supplier design and field design is built. It is handed over to the operation team at the O and M stage and utilized in the maintenance and repair. As a power plant full model of future construction project has been enabled, an improved design process has been suggested, in which only the design change information during the plant architect engineering (A/E) and the design change information during the field design

  11. Development and application of objective uncertainty measures for nuclear power plant transient analysis[Dissertation 3897

    Energy Technology Data Exchange (ETDEWEB)

    Vinai, P

    2007-10-15

    For the development, design and licensing of a nuclear power