WorldWideScience

Sample records for safety improvement plant

  1. Safety improvement of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Vamos, G.

    1999-01-01

    Safety upgrading completed in the early nineties at the Paks NPP include: replacement of steam generator safety valves and control valves; reliability improvement of the electrical supply system; modification of protection logic; enhancement of the fire protection; construction of full scope Training Simulator. Design safety upgrading measures achieved in recent years were concerned with: relocation of steam generator emergency feed-water supply; emergency gas removal from the primary coolant system; hydrogen management in the containment; protection against sumps; preventing of emergency core cooling system tanks from refilling. Increasing seismic resistance, containment assessment, refurbishment of reactor protection system, improving reliability of emergency electrical supply, analysis of internal hazards are now being implemented. Safety upgrading measures which are being prepared include: bleed and feed procedures; reactor over-pressurisation protection in cold state; treatment of steam generator primary to secondary leak accidents. Operational safety improvements are dealing with safety culture, training measures and facilities; symptom based emergency operating procedures; in-service inspection; fire protection. The significance of international cooperation is emphasised in view of achieving nuclear safety standards recognised in EU

  2. Improving the safety of LWR power plants. Final report

    International Nuclear Information System (INIS)

    1980-04-01

    This report documents the results of the Study to identify current, potential research issues and efforts for improving the safety of Light Water Reactor (LWR) power plants. This final report describes the work accomplished, the results obtained, the problem areas, and the recommended solutions. Specifically, for each of the issues identified in this report for improving the safety of LWR power plants, a description is provided in detail of the safety significance, the current status (including information sources, status of technical knowledge, problem solution and current activities), and the suggestions for further research and development. Further, the issues are ranked for action into high, medium, and low priority with respect to primarily (a) improved safety (e.g. potential reduction in public risk and occupational exposure), and secondly (b) reduction in safety-related costs

  3. South Ukraine NPP: Safety improvements through Plant Computer upgrade

    International Nuclear Information System (INIS)

    Brenman, O.; Chernyshov, M. A.; Denning, R. S.; Kolesov, S. A.; Balakan, H. H.; Bilyk, B. I.; Kuznetsov, V. I.; Trosman, G.

    2006-01-01

    This paper summarizes some results of the Plant Computer upgrade at the Units 2 and 3 of South Ukraine Nuclear Power Plant (NPP). A Plant Computer, which is also called the Computer Information System (CIS), is one of the key safety-related systems at VVER-1000 nuclear plants. The main function of the CIS is information support for the plant operators during normal and emergency operational modes. Before this upgrade, South Ukraine NPP operated out-of-date and obsolete systems. This upgrade project wax founded by the U.S. DOE in the framework of the International Nuclear Safety Program (INSP). The most efficient way to improve the quality and reliability of information provided to the plant operator is to upgrade the Human-System Interface (HSI), which is the Upper Level (UL) CIS. The upgrade of the CIS data-acquisition system (DAS), which is the Lower Level (LL) CIS, would have less effect on the unit safety. Generally speaking, the lifetime of the LL CIS is much higher than one of the UL CIS. Unlike Plant Computers at the Western-designed plants, the functionality of the WER-1000 CISs includes a control function (Centralized Protection Testing) and a number of the plant equipment monitoring functions, for example, Protection and Interlock Monitoring and Turbo-Generator Temperature Monitoring. The new system is consistent with a historical migration of the format by which information is presented to the operator away from the traditional graphic displays, for example, Piping and Instrument Diagrams (P and ID's), toward Integral Data displays. The cognitive approach to information presentation is currently limited by some licensing issues, but is adapted to a greater degree with each new system. The paper provides some lessons learned on the management of the international team. (authors)

  4. Improving plant state information for better operational safety

    International Nuclear Information System (INIS)

    Girard, C.; Olivier, E.; Grimaldi, X.

    1994-01-01

    Nuclear Power Plant (NPP) safety is strongly dependent on components' reliability and particularly on plant state information reliability. This information, used by the plant operators in order to produce appropriate actions, have to be of a high degree of confidence, especially in accidental conditions where safety is threatened. In this perspective, FRAMATOME, EDF and CEA have started a joint research program to prospect different solutions aiming at a better reliability for critical information needed to safety operate the plant. This paper gives the main results of this program and describes the developments that have been made in order to assess reliability of different information systems used in a Nuclear Power Plant. (Author)

  5. To improve nuclear plant safety by learning from accident's experience

    International Nuclear Information System (INIS)

    Matsumoto, Hidezo; Kida, Masanori; Kato, Hiroyuki; Hara, Shin-ichi

    1994-01-01

    The ultimate goal of this study is to produce an expert system that enables the experience (records and information) gained from accidents to be put to use towards improving nuclear plant safety. A number of examples have been investigated, both domestic and overseas, in which experience gained from accidents was utilized by utilities in managing and operating their nuclear power stations to improve safety. The result of investigation has been used to create a general 'basic flow' to make the best use of experience. The ultimate goal is achieved by carrying out this 'basic flow' with artificial intelligence (AI). To do this, it is necessary (1) to apply language analysis to process the source information (primary data base; domestic and overseas accident's reports) into the secondary data base, and (2) to establish an expert system for selecting (screening) significant events from the secondary data base. In the processing described in item (1), a multi-lingual thesaurus for nuclear-related terms become necessary because the source information (primary data bases) itself is multi-lingual. In the work described in item (2), the utilization of probabilistic safety assessment (PSA), for example, is a candidate method for judging the significance of events. Achieving the goal thus requires developing various new techniques. As the first step of the above long-term study project, this report proposes the 'basic flow' and presents the concept of how the nuclear-related AI can be used to carry out this 'basic flow'. (author)

  6. Improving the safety of future nuclear fission power plants

    International Nuclear Information System (INIS)

    Frisch, W.; Gros, G.

    2001-01-01

    The main objectives and principles in nuclear fission reactor safety are presented, e.g. the defence in depth strategy and technical principles such as redundancy, diversity and physical separation. After a brief historical review of the continuous development of safety improvement, the most recent international discussion is presented. This includes mainly the international activities within IAEA and its International Nuclear Safety Advisory Group (INSAG). The safety improvement, presented in recommendations of IAEA and INSAG is expressed as an improvement of all elements and all levels of the defence in depth concept. Special emphasis is put on improvement of the highest level, which requires the implementation of means to mitigate consequences of accidents with severe core damage. The different future concepts are briefly characterised. Some examples from the French-German safety approach are taken to demonstrate how requirements for safety improvement by means of an enhancement of the defence in depth principle are developed

  7. Improving nuclear power plant safety through operator aids

    International Nuclear Information System (INIS)

    1987-12-01

    In October 1986, the IAEA convened a one-week Technical Committee Meeting on Improving Nuclear Power Plant Safety Through Operator Aids. The term ''operator aid'' or more formally ''operator support system'' refers to a class of devices designed to be added to a nuclear power plant control station to assist an operator in performing his job and thereby decrease the probability of operator error. The addition of a carefully planned and designed operator aid should result in an increase in nuclear power plant safety and reliability. Operator aids encompass a wide range of devices from the very simple, such as color coding a display to distinguish it out of a group of similar displays, to the very complex, such as a computer-generated video display which concentrates a number of scattered indicator readings located around a control room into a concise display in front of the operator. This report provides guidelines and information to help make a decision as to whether an operator aid is needed, what kinds of operator aids are available and whether it should be purchased or developed by the utility. In addition, a discussion is presented on advanced operator aids to provide information on what may become available in the future. The broad scope of these guidelines makes it most suitable for use by a multi-disciplinary team. The document consists of two parts. The recommendations and results of the meeting discussions are given in the first part. The second part is the annex where the papers presented at the Technical Committee Meeting are printed. A separate abstract was prepared for each of the 10 papers. Refs, figs and tabs

  8. Improving Chemical Plant Safety Training Using Virtual Reality

    OpenAIRE

    Nasios, Konstantinos

    2002-01-01

    The chemical engineering industry often requires people to work in hazardous environments and to operate complicated equipment which often limits the type of training that be carried out on site. The daily job of chemical plant operators is becoming more demanding due to the increasing plant complexity together with increasing requirements on plant safety, production capacity, product quality and cost effectiveness. The importance of designing systems and environments that are as safe as poss...

  9. Improved technical specifications and related improvements to safety in commercial Nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Hoffman, D.R.; Demitrack, T.; Schiele, R.; Jones, J.C. [EXCEL Services Corporation, 11921 Rockville Pike, Suite 100, Rockville, MD 20852 (United States)]. e-mail: donaldh@excelservices.com

    2004-07-01

    Many of the commercial nuclear power plants in the United States (US) have been converting a portion of the plant operating license known as the Technical Specifications (TS) in accordance with a document published by the US Nuclear Regulatory Commission (NRC). The TS prescribe commercial nuclear power plant operating requirements. There are several types of nuclear power plants in the US, based on the technology of different vendors, and there is an NRC document that supports each of the five different vendor designs. The NRC documents are known as the Improved Standard Technical Specifications (ISTS) and are contained in a separate document (NUREG series) for each one of the designs. EXCEL Services Corporation (hereinafter EXCEL) has played a major role in the development of the ISTS and in the development, licensing, and implementation of the plant specific Improved Technical Specifications (ITS) (which is based on the ISTS) for the commercial nuclear power plants in the US that have elected to make this conversion. There are currently 103 operating commercial nuclear power plants in the US and 68 of them have successfully completed the conversion to the ITS and are now operating in accordance with their plant specific ITS. The ISTS is focused mainly on safety by ensuring the commercial nuclear reactors can safely shut down and mitigate the consequences of any postulated transient and accident. It accomplishes this function by including requirements directly associated with safety in a document structured systematically and taking into account some key human factors and technical initiatives. This paper discusses the ISTS including its format, content, and detail, the history of the ISTS, the ITS development, licensing, and implementation process, the safety improvements resulting from a plant conversion to ITS, and the importance of the ITS Project to the industry. (Author)

  10. Improved technical specifications and related improvements to safety in commercial Nuclear power plants

    International Nuclear Information System (INIS)

    Hoffman, D.R.; Demitrack, T.; Schiele, R.; Jones, J.C.

    2004-01-01

    Many of the commercial nuclear power plants in the United States (US) have been converting a portion of the plant operating license known as the Technical Specifications (TS) in accordance with a document published by the US Nuclear Regulatory Commission (NRC). The TS prescribe commercial nuclear power plant operating requirements. There are several types of nuclear power plants in the US, based on the technology of different vendors, and there is an NRC document that supports each of the five different vendor designs. The NRC documents are known as the Improved Standard Technical Specifications (ISTS) and are contained in a separate document (NUREG series) for each one of the designs. EXCEL Services Corporation (hereinafter EXCEL) has played a major role in the development of the ISTS and in the development, licensing, and implementation of the plant specific Improved Technical Specifications (ITS) (which is based on the ISTS) for the commercial nuclear power plants in the US that have elected to make this conversion. There are currently 103 operating commercial nuclear power plants in the US and 68 of them have successfully completed the conversion to the ITS and are now operating in accordance with their plant specific ITS. The ISTS is focused mainly on safety by ensuring the commercial nuclear reactors can safely shut down and mitigate the consequences of any postulated transient and accident. It accomplishes this function by including requirements directly associated with safety in a document structured systematically and taking into account some key human factors and technical initiatives. This paper discusses the ISTS including its format, content, and detail, the history of the ISTS, the ITS development, licensing, and implementation process, the safety improvements resulting from a plant conversion to ITS, and the importance of the ITS Project to the industry. (Author)

  11. Safety improvement plant modifications at Forsmark 3, 1986-1995

    Energy Technology Data Exchange (ETDEWEB)

    Kjellander, M. [Kaernkraftsaekerhet och utbildning, Nykoeping (Sweden)

    1998-10-01

    All important plant modifications implemented in safety-related equipment or software at Forsmark 3 are compiled in this report. The report covers the period from the start of commercial operation in 1985 up to and including 1995. The plant modifications, which were carried out by different suppliers during the guarantee period, are not included in the report since they have not been administered by the Forsmark organisation. The report contains references to relevant modification notices and to files and file divider numbers. These data refer to the Safety Department central archives. The report is based on Forsmark 3 Technical Specifications (STF) which means that Chapter 3 is divided into the same sections as in the STF. Modifications, which cannot be directly attributed to any specific STF chapter, and major modifications are described separately

  12. Alternative off-site power supply improves nuclear power plant safety

    International Nuclear Information System (INIS)

    Gjorgiev, Blaže; Volkanovski, Andrija; Kančev, Duško; Čepin, Marko

    2014-01-01

    Highlights: • Additional power supply for mitigation of the station blackout event in NPP is used. • A hydro power plant is considered as an off-site alternative power supply. • An upgrade of the probabilistic safety assessment from its traditional use is made. • The obtained results show improvement of nuclear power plant safety. - Abstract: A reliable power system is important for safe operation of the nuclear power plants. The station blackout event is of great importance for nuclear power plant safety. This event is caused by the loss of all alternating current power supply to the safety and non-safety buses of the nuclear power plant. In this study an independent electrical connection between a pumped-storage hydro power plant and a nuclear power plant is assumed as a standpoint for safety and reliability analysis. The pumped-storage hydro power plant is considered as an alternative power supply. The connection with conventional accumulation type of hydro power plant is analysed in addition. The objective of this paper is to investigate the improvement of nuclear power plant safety resulting from the consideration of the alternative power supplies. The safety of the nuclear power plant is analysed through the core damage frequency, a risk measure assess by the probabilistic safety assessment. The presented method upgrades the probabilistic safety assessment from its common traditional use in sense that it considers non-plant sited systems. The obtained results show significant decrease of the core damage frequency, indicating improvement of nuclear safety if hydro power plant is introduced as an alternative off-site power source

  13. Safety Performance Improvement for Nuclear Power Plants Utilizing THOMAS

    International Nuclear Information System (INIS)

    Kim, Won June; Ryu, Jung Uk; Suh, Kune Y.

    2005-01-01

    THOMAS (Thermal Hydraulics Online Monitoring Advisory System) is equipped with a couple of salient features compared with existing monitoring systems. The first has to do with the three-dimensional (3D) visualization technique to support the nuclear power plant (NPP) operators and personnel using the virtual reality (VR) technology. VR depicts an environment simulated by a computer. Most of the VR environments primarily include visual experiences, displayed either on a monitor or though special stereoscopic goggles. Users can often interactively manipulate a VR environment, either through standard input devices like a keyboard, or through specially designed devices like a cybergloves. Additional devices were not applied the in THOMAS. The visualized model file is brought to the VR space from the computer-aided design (CAD) tool. In the VR space, using mapping, the component color is changed with linked value of the safety variables. Operators thus can easily recognize the plant condition. This is related with the human factor engineering. The second is the function of decision making using the influence diagram logic. The influence diagram logic is based on the total probability and Bayesian theory. The accident modeling is rooted in the emergency operating procedure (EOP). The final goal of this system is, in the accident situation, to present a success path to the operator for the recovery of the NPP system. At the current developing level, the database signals THOMAS. In other words, a spectrum of system analysis codes provides the safety parameter values to the database, which are subsequently supplied to THOMAS through the network

  14. Safety assessment of proposed improvements to RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1993-03-01

    The purpose of this report is to summarize the findings and recommendations of a Consultants Meeting convened by the IAEA in Vienna (27 October - 5 November 1992) to review new design features and modifications proposed or already implemented for RBMK reactors. This information was provided in four technical areas, namely: Core Monitoring and Control, Pressure Boundary Integrity, Accident Mitigation and Electric Power Supply. The report also presents the status of the modifications at the plants as given by the RBMK specialists. The limited information available and the time constraints did not allow the review to be conducted at the level of a peer review, and the findings and recommendations made reflect the limited scope of the review. More detailed reviews and analysis focusing on selected safety issues are required and should be conducted on a generic and plant specific basis as appropriate. In Chapters 2-5 of the report the main findings and recommendations for the four topical areas reviewed are summarized. Appendices I-IV reflect the results of the discussions held at the meeting and provide more detailed information on the review. 17 refs, 27 figs, 17 tabs

  15. Transportable nuclear power plant T3C-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Ogloblin, B.G.; Gromov, B.F.

    1993-01-01

    Development and cultivating of districts in Siberia, North, Far East, Kamchatka and other remote or almost inaccessible district of the country depends to a large degree on their providing with power. The specific character of these districts imposes in turn a wide variety of special requirements upon the power sources. In particular, it is essential to provide the following; maximum manufacture availability of the whole equipment at the minimum volume of construction and installation work on operation site, high safety, longterm service life, ecologically, minimum scope of work on equipment in-service maintenance and inspection, etc. Taking into account the well-known difficulties connected with the delivery of conventional energy carriers to the above-mentioned districts and the situation with the alternative power sources, the application of the low-power nuclear plants (NPP) for these purposes looks definitely promising. Among the probable trends in creating the NPPs of this type as very promising is considered the possibility to apply the two-circuit reactor plant of the vessel type with the liquid lead as a primary coolant and free air as a secondary coolant and working medium in the open gas-turbine cycle. The nuclear plant T3C-M of improved safety with two of this type reactor plants with total electric power of 8 MW is developed by CDB of Machine Building with participation of several enterprises of St. Petersburg under the scientific leadership and is intended for generation of electric power and up to 4 Gcal/h of heat for populated areas and installations placed at long distance from the main electric power supply sources where it is difficult or non-efficient economically to deliver the conventional kinds of fuel. The main principles being laid as a basis when developing the proposed NPP will allow one to create mobile power sources which possess a high degree of safety and inherent self-protection

  16. Improving the safety of workers in the vicinity of mobile plant

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    This report investigates possible approaches and technologies to improve the safety of pedestrians around mobile plant. The study has considered mobile plant in underground mining, surface mineral extraction and warehousing and materials handling. It investigates vehicle ergonomics and visual task analysis, aids to improve reversing and sight lines, direct body detection techniques, conventional transponder (RFID) and VLF magnetic dipole detection technologies. 43 refs., 5 apps.

  17. Technical concepts of further improvement of nuclear power plant safety

    International Nuclear Information System (INIS)

    Sochor, R.

    1983-01-01

    The following technical concepts are described which secure the integrity of the containment in case of an accident whose scale exceeds the so-called design basis accident: siting nuclear power plants underground which raises construction costs by 20 - 25%; completing the containment with equipment preventing the outflow of molten corium; completing the containment with emergency pressure space for discharging overpressure - this emergency space is filled with gravel which will trap approximately 50% of fission waste. (Ha)

  18. Approach to improving safety for TEPCO's nuclear power plants

    International Nuclear Information System (INIS)

    Watanabe, Kunimichi

    2008-01-01

    TEPCO (Tokyo Electric Power Company) has still continued implementing countermeasures to guard against a reoccurrence of incidents in 2002, such as promoting disclosure of information and ensuring transparency of nuclear operations, creating a work environment where proper operations can be carried out, strengthening internal surveillance and reforming the corporate culture and promoting observance of corporate ethics. It is essential to give top priority to safety in nuclear operations and actualize the above properly. Discrepancy resolution management and human errors preventive measures were discussed. Lacks of communications, risk prediction and management will result in human errors related issues to be solved. (T. Tanaka)

  19. Kozloduy nuclear power plant. Units 1-4. Status of safety improvements. Rev. 2

    International Nuclear Information System (INIS)

    1999-01-01

    This paper presents the results of the safety improvements activities carried out by the Kozloduy Nuclear Power Plant (KNPP) within the period 1990-1998. The steam supply system of this units is based of the reactor WWER-440/ B-230, which is a PWR of russian design developed according to the safety standards in force in USSR in late sixties. Up to now 10 reactor units of this type are in operation in four NPPs. Despite of efforts of the different plants to implement safety improvements measures during first 10-15 years of operation of this type of reactor its major safety problems were not eliminated and were a subject of international concern. The systematic evaluation of the deficiencies of the original design of this type of reactors have been initiated by IAEA in the beginning of 1990 and brought to developing a comprehensive list of safety problems which required urgent implementation of safety measures in all plants. To solve this problems in 1991 KNPP initiated implementation of so called 'short term' safety improvement program, developed with the help of WANO under agreement with Bulgarian Nuclear Safety Authority (BNSA) and consortium RISKAUDIT. The program was based on a stage approach and was foreseen to be implemented by tree stages in very tight time schedule in order to achieve significant and rapid improvements of the level of safety in operation of the units. The Short term program was implemented between from 1991 to 1997 owing to strong safety commitment of NEC and KNPP staff as well as broad international cooperation and financial support. Important part of resources were supplied under PHARE program of CEC, EBRD grant agreement and EDF support. In parallel a special assessment process started in 1995 in order to evaluate the level of safety, achieved by Short Term Program, according to current safety standards and to define the measures, which should be implemented by the Utility to complete the process of improving the safety in future

  20. Safety improvements at Canadian nuclear power plants in the aftermath of Fukushima accident

    International Nuclear Information System (INIS)

    Rzentkowski, G.; Khouaja, H.

    2014-01-01

    This paper describes the safety review of operating nuclear power plants undertaken by the Canadian Nuclear Safety Commission in light of the March 11, 2011 accident at the Fukushima Daiichi Nuclear Power Plants (NPPs). The review confirmed that the Canadian NPPs are robust and have a strong design relying on multiple layers of defence to protect the public from credible external events. Nevertheless, in the spirit of continuous safety improvements, the review identified a number of recommendations to further strengthen reactor defence-in-depth in preventing and mitigating the consequences of beyond design basis accidents, enhance onsite and offsite emergency response, and improve the CNSC regulatory framework. Progress achieved to date, in implementing these measures, is described in this paper along with a summary of safety benefits for each level of the reactor defence-in-depth. (author)

  1. Safety improvements at Canadian nuclear power plants in the aftermath of Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

    Rzentkowski, G.; Khouaja, H. [Canadian Nuclear Safety Commission, Ottawa, ON (Canada)

    2014-07-01

    This paper describes the safety review of operating nuclear power plants undertaken by the Canadian Nuclear Safety Commission in light of the March 11, 2011 accident at the Fukushima Daiichi Nuclear Power Plants (NPPs). The review confirmed that the Canadian NPPs are robust and have a strong design relying on multiple layers of defence to protect the public from credible external events. Nevertheless, in the spirit of continuous safety improvements, the review identified a number of recommendations to further strengthen reactor defence-in-depth in preventing and mitigating the consequences of beyond design basis accidents, enhance onsite and offsite emergency response, and improve the CNSC regulatory framework. Progress achieved to date, in implementing these measures, is described in this paper along with a summary of safety benefits for each level of the reactor defence-in-depth. (author)

  2. Plan for research to improve the safety of light-water nuclear power plants

    International Nuclear Information System (INIS)

    1978-03-01

    This is the U.S. Nuclear Regulatory Commission's first annual report to Congress on recommendations for research on improving the safety of light-water nuclear power plants. Suggestions for reactor safety research were identified in, or received from, various sources, including the Advisory Committee on Reactor Safeguards, the NRC regulatory staff, and the consultants to the Research Review Group. After an initial screening to eliminate those not related to improved reactor safety, all the suggestions were consolidated into research topics. It is recommended that the following research projects be carried out: alternate containment concepts, especially vented containments; alternate decay heat removal concepts, especially add-on bunkered systems; alternate emergency core cooling concepts; improved in-plant accident response; and advanced seismic designs

  3. Transportable nuclear power plant TEC-M with two reactor plants of improved safety

    International Nuclear Information System (INIS)

    Ogloblin, B.G.; Sazonov, A.G.; Svishchev, A.M.; Gromov, B.F.; Zelensky, V.N.; Komkova, O.I.; Sidorov, V.I.; Tolstopyatov, V.P.; Toshinsky, G.I.

    1993-01-01

    Liquid metals are the best to meet the requirements of inherently safety nuclear power plants among the coolants used. A great experience has been gained in lead coolant power plant development and operation as applied to transportable power set-ups. Low chemical activity of this coolant with respect to air-water interaction is a determining factor for this coolant. The transportable nuclear power plant is described. It is intended to generate electric power for populated areas placed a long distance from the main electric power supply sources where it is difficult or not economical to deliver the conventional types of fuel. There are several remote areas in Siberia, Kamchatka in need of this type of power plant

  4. Improved safety of the system 80+TM standard plants design through increased diversity and redundancy of safety systems

    International Nuclear Information System (INIS)

    Matzie, Regis A.; Carpentino, Frederick L.; Robertson, James E.

    1996-01-01

    Safely systems in the System 80+ TM Standard Plant are designed with more redundancy, diversity and simplicity than earlier nuclear power plant designs. These gains were accomplished by an evolutionary process that preserved the desirable and proven features in currently operating nuclear plants, while improving reliability and defense-in-depth. The System 80+ safety systems are the primary contributors to a core damage frequency that is more than 100 times lower than 1980's vintage U. S. designs, including the predecessor System 80 R standard nuclear steam supply system (NSSS) design. The System 80+ design includes significant improvements to the safety injection system, emergency feedwater system, shutdown cooling system, containment spray system, reactor coolant gas vent system, and to their vital support systems. These improvements enhance performance for traditional design basis events and significantly reduce the probability of a severe accident. The System 80+ design also incorporates safety systems to mitigate a severe accident. The added systems include the rapid depressurization system, the in-containment refueling water storage tank, the cavity flooding system. These systems fully address the U. S. Nuclear Regulatory Commission's (US NRC) severe accident policy. The System 80+ safety systems are integrated with the System 80+ Nuclear Island (NI) design. The NI general arrangement provides quadrant separation of the safety systems for protection from fire and flooding, and large equipment pull spaces and lay down areas for maintenance. This paper will describe the System 80+ safety systems advanced design features, the improved accident prevention and mitigation capabilities, and startup, operating and maintenance benefits

  5. Efforts to improve safety and reliability of nuclear power plants in Kyushu Electric Power

    International Nuclear Information System (INIS)

    Yamamoto, Satoshi

    2014-01-01

    After the Fukushima accident, Kyushu Electric Power Co. took emergency safety measures requested by government to ensure power supply, coolant supply pumps and cooling water so as to keep cooling fuels in the reactor and spent fuel storage pool in case of losses of ordinary cooling capability caused by earthquake and tsunami. In order to improve safety and reliability of nuclear power plants, further efforts based on lessons learned from the Fukushima accident had been made to diversify corresponding equipment of safety measures in terms of prevention of core damage, prevention of containment failure, mitigation of radioactive materials release, cooling of spent fuel pit and ensurance of power supply, and to enhance emergency response capability so as to make operational management more complete. Additional safety measures applicable to new regulatory requirements against severe accidents were in progress. This article introduced details of such activities. (T. Tanaka)

  6. Using game technologies to improve the safety of construction plant operations.

    Science.gov (United States)

    Guo, Hongling; Li, Heng; Chan, Greg; Skitmore, Martin

    2012-09-01

    Many accidents occur world-wide in the use of construction plant and equipment, and safety training is considered by many to be one of the best approaches to their prevention. However, current safety training methods/tools are unable to provide trainees with the hands-on practice needed. Game technology-based safety training platforms have the potential to overcome this problem in a virtual environment. One such platform is described in this paper - its characteristics are analysed and its possible contribution to safety training identified. This is developed and tested by means of a case study involving three major pieces of construction plant, which successfully demonstrates that the platform can improve the process and performance of the safety training involved in their operation. This research not only presents a new and useful solution to the safety training of construction operations, but illustrates the potential use of advanced technologies in solving construction industry problems in general. Copyright © 2011 Elsevier Ltd. All rights reserved.

  7. Application of Advanced Technology to Improve Plant Performance. Safety and Performance in Current NPPs

    International Nuclear Information System (INIS)

    Hashemian, H.M.

    2011-01-01

    Advances in computer technologies, signal processing, analytical modeling, and the advent of wireless sensors have provided the nuclear industry with ample means to automate and optimize maintenance activities and improve safety, efficiency, and availability, while reducing costs and radiation exposure to maintenance personnel. This paper provides a review of these developments and presents examples of their use in the nuclear power industry and the financial and safety benefits that they have produced. As the current generation of nuclear power plants have passed their mid-life, increased monitoring of their health is critical to their safe operation. This is especially true now that license renewal of nuclear power plants has accelerated, allowing some plants to operate up to 60 years or more. Furthermore, many utilities are maximizing their power output through uprating projects and retrofits. This puts additional demand and more stress on the plant equipment such as the instrumentation and control (I and C) systems and the reactor internal components making them more vulnerable to the effects of aging, degradation, and failure. In the meantime, the nuclear power industry is working to reduce generation costs by adopting condition-based maintenance strategies and automation of testing activities. These developments have stimulated great interest in on-line monitoring (OLM) technologies and new diagnostic and prognostic methods to anticipate, identify, and resolve equipment and process problems and ensure plant safety, efficiency, and immunity to accidents. The foundation for much of the required technologies has already been established through 40 years of research and development (R and D) efforts performed by numerous organizations, scientists, and engineers around the world including the author. This paper provides examples of these technologies and demonstrates how the gap between some of the more important R and D efforts and end users have been filled

  8. 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)

  9. Improved people detection in nuclear plants by video processing for safety purpose

    Energy Technology Data Exchange (ETDEWEB)

    Jorge, Carlos Alexandre F.; Mol, Antonio Carlos A.; Carvalho, Paulo Victor R., E-mail: calexandre@ien.gov.br, E-mail: mol@ien.gov.br, E-mail: paulov@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Seixas, Jose M.; Silva, Eduardo Antonio B., E-mail: seixas@lps.ufrj.br, E-mail: eduardo@smt.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), RJ (Brazil). Programa de Engenharia Eletrica; Waintraub, Fabio, E-mail: fabiowaintraub@hotmail.com [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil). Escola Politecnica. Departamento de Engenharia Eletronica e de Computacao

    2013-07-01

    This work describes improvements in a surveillance system for safety purposes in nuclear plants. The objective is to track people online in video, in order to estimate the dose received by personnel, during working tasks executed in nuclear plants. The estimation will be based on their tracked positions and on dose rate mapping in a nuclear research reactor, Argonauta. Cameras have been installed within Argonauta room, supplying the data needed. Video processing methods were combined for detecting and tracking people in video. More specifically, segmentation, performed by background subtraction, was combined with a tracking method based on color distribution. The use of both methods improved the overall results. An alternative approach was also evaluated, by means of blind source signal separation. Results are commented, along with perspectives. (author)

  10. Improved people detection in nuclear plants by video processing for safety purpose

    International Nuclear Information System (INIS)

    Jorge, Carlos Alexandre F.; Mol, Antonio Carlos A.; Carvalho, Paulo Victor R.; Seixas, Jose M.; Silva, Eduardo Antonio B.; Waintraub, Fabio

    2013-01-01

    This work describes improvements in a surveillance system for safety purposes in nuclear plants. The objective is to track people online in video, in order to estimate the dose received by personnel, during working tasks executed in nuclear plants. The estimation will be based on their tracked positions and on dose rate mapping in a nuclear research reactor, Argonauta. Cameras have been installed within Argonauta room, supplying the data needed. Video processing methods were combined for detecting and tracking people in video. More specifically, segmentation, performed by background subtraction, was combined with a tracking method based on color distribution. The use of both methods improved the overall results. An alternative approach was also evaluated, by means of blind source signal separation. Results are commented, along with perspectives. (author)

  11. Modernisation for maintaining and improving safety at Nordic nuclear power plants

    International Nuclear Information System (INIS)

    Hammer, L.; Wahlstroem, B.; Simola, K.

    1998-02-01

    The safety practices in Finland and Sweden are described and compared in regard of effecting modernisation for safety of the nuclear plants in the two countries, considering new technology and advancing safety requirements as proposed for new reactors. Particular attention is given to strategies for applying new safety requirements to reactors built to earlier standards, and to the interplay between the nuclear utilities and the safety authorities. Overviews are given of past and current modernisation of the nuclear power plants in Finland and Sweden. The management procedures in controlling the implementation of modifications to the nuclear power plants are described and discussed in regard of prevailing differences between Finnish and Swedish practices. A formal modelling technique (SADT) was applied for capture of the essential contents of the relevant documented procedures. Two examples of recent plant modifications in the Finnish nuclear plants in Olkiluoto and Loviisa are described and discussed in greater detail. Recommendations are given. (au)

  12. A computer based I and C to improve nuclear power plant operation and safety

    International Nuclear Information System (INIS)

    Appell, Bernard; Beltranda, Guy

    1995-01-01

    Chooz B1 is the head of 1400 MW N4 series (up to now, a programme of 4 identical units have been ordered) will be put on the French national grid this year (1995). Hot tests were performed last autumn. Loading fuel is scheduled in June 95. Chooz B2 will follow 6 months later. This nuclear power plant series is fully French designed. It comprises, compared to the former 1300 MW series, evolution concerning mainly the turbine, the steam generators, the primary pumps, the I and C and the man machine interface. A major improvement for the plant operation and safety is the computerised I and C which comprises the protection system (specific controllers), the process automation system (off the shelf equipment), the computer based and conventional man machine interfaces associated. These systems were put in service last year. This paper is focused on them. (author)

  13. Improved nuclear power plant operations through performance-based safety regulation

    International Nuclear Information System (INIS)

    Golay, M.W.

    1998-01-01

    The US Nuclear Regulatory Commission (NRC) has recently instituted use of Risk-Informed, Performance-Based Regulation (RIPBR) for protecting public safety in the use of nuclear power. This was done most importantly during June 1997 in issuance of revised Regulatory Guides and Standard Review Plan (SRP) guidance to licensees and the NRC staff. The propose of RIPBR is to replace the previously-used system of prescriptive regulation, which focuses upon what licensees must do, to a system which focuses upon what they must achieve. RIPBR is goals-oriented and the previous system is means-oriented. This regulatory change is potentially revolutionary, and offers many opportunities for improving the efficiency of improving both nuclear power operations and safety. However, it must be nurtured carefully if is to be successful. The work reported in this paper is concerned with showing how RIPBR can be implemented successfully, with benefits in both areas being attained. It is also concerned with how several of the practical barriers to establishing a workable new regulatory system can be overcome. This work, sponsored by the US Dept. of Energy, is being performed in collaboration with Northeast Utilities Services Crop. and the Idaho National Engineering Laboratory. In our work we have examined a practical safety-related example at the Millstone 3 nuclear power station for implementation of RIPBR. In this examination we have formulated a set of modifications to the plant's technical specifications, and are in the process of investigating their bases and refining the modifications. (author)

  14. Continuous Improvement and the Safety Case for the Waste Isolation Pilot Plant Geologic Repository - 13467

    Energy Technology Data Exchange (ETDEWEB)

    Van Luik, Abraham; Patterson, Russell; Nelson, Roger [US Department of Energy, Carlsbad Field Office, 4021 S. National parks Highway, Carlsbad, NM 88220 (United States); Leigh, Christi [Sandia National Laboratories Carlsbad Operations, 4100 S. National parks Highway, Carlsbad, NM 88220 (United States)

    2013-07-01

    The Waste Isolation Pilot Plant (WIPP) is a geologic repository 2150 feet (650 m) below the surface of the Chihuahuan desert near Carlsbad, New Mexico. WIPP permanently disposes of transuranic waste from national defense programs. Every five years, the U.S. Department of Energy (DOE) submits an application to the U.S. Environmental Protection Agency (EPA) to request regulatory-compliance re-certification of the facility for another five years. Every ten years, DOE submits an application to the New Mexico Environment Department (NMED) for the renewal of its hazardous waste disposal permit. The content of the applications made by DOE to the EPA for re-certification, and to the NMED for permit-renewal, reflect any optimization changes made to the facility, with regulatory concurrence if warranted by the nature of the change. DOE points to such changes as evidence for its having taken seriously its 'continuous improvement' operations and management philosophy. Another opportunity for continuous improvement is to look at any delta that may exist between the re-certification and re-permitting cases for system safety and the consensus advice on the nature and content of a safety case as being developed and published by the Nuclear Energy Agency's Integration Group for the Safety Case (IGSC) expert group. DOE at WIPP, with the aid of its Science Advisor and teammate, Sandia National Laboratories, is in the process of discerning what can be done, in a reasonably paced and cost-conscious manner, to continually improve the case for repository safety that is being made to the two primary regulators on a recurring basis. This paper will discuss some aspects of that delta and potential paths forward to addressing them. (authors)

  15. Improved Management of Part Safety Classification System for Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Young; Park, Youn Won; Park, Heung Gyu; Park, Hyo Chan [BEES Inc., Daejeon (Korea, Republic of)

    2016-10-15

    As, in recent years, many quality assurance (QA) related incidents, such as falsely-certified parts and forged documentation, etc., were reported in association with the supply of structures, systems, components and parts to nuclear power plants, a need for a better management of safety classification system was addressed so that it would be based more on the level of parts . Presently, the Korean nuclear power plants do not develop and apply relevant procedures for safety classifications, but rather the safety classes of parts are determined solely based on the experience of equipment designers. So proposed in this paper is a better management plan for safety equipment classification system with an aim to strengthen the quality management for parts. The plan was developed through the analysis of newly introduced technical criteria to be applied to parts of nuclear power plant.

  16. Reprocessing plants safety

    International Nuclear Information System (INIS)

    Davies, A.G.; Leighton, C.; Millington, D.

    1989-01-01

    The reprocessing of irradiated nuclear fuel at British Nuclear Fuels (BNFL) Sellafield site consists of a number of relatively self-contained activities carried out in separate plants across the site. The physical conditions and time scales applied in reprocessing and storage make it relatively benign. The potential for minor releases of radioactivity under fault conditioning is minimised by plant design definition of control procedures, training and supervision. The risks to both the general public and workforce are shown to be low with all the safety criteria being met. Normal operating conditions also have the potential for some occupational radiation exposure and the plant and workers are monitored continuously. Exposure levels have been reduced steadily and will continue to fall with plant improvements. (U.K.)

  17. 3D representation of radioisotopic dose rates within nuclear plants for improved radioprotection and plant safety

    International Nuclear Information System (INIS)

    Vabo, Rune; Rindahl, Grete; Piotrowski, Leon

    2010-01-01

    A better awareness of the origin and nature of gamma doses in nuclear environments is demonstrated by visualizing the dose maps created by individual radionucleides that are present in radioactive contaminations. This isotopic representation of doses is much more informative than showing a map of the measured total doses. Two practical examples are given : (a) the placing of protections, and (b) using radiation decay to help plan dismantling operations. The necessary radionucleide information can be easily obtained by the new EDF CZT gamma spectrometer that is now used by all its NPPs. Defining radioactive sources based on such information enables the reconstruction of the radiation situation in a virtual 3-D environment. In such a virtual environment, dose rates can be calculated in any position in space and information about how much each radionucleide contributes can be extracted. Such 3D visualisations increase the awareness and knowledge of the distribution of radiation in a nuclear facility and can be considered as an educational tool for training and improved ALARA procedures. (author)

  18. Recommended research program for improving seismic safety of light-water nuclear power plants

    International Nuclear Information System (INIS)

    1979-04-01

    Recommendations are presented for research areas concerned with seismic safety. These recommendations are based on an analysis of the answers to a questionnaire which was sent to over 80 persons working in the area of seismic safety of nuclear power plants. In addition to the answers of the 55 questionnaires which were received, the recommendations are based on ideas expressed at a meeting of an ad hoc group of professionals formed by Sandia, review of literature, current research programs, and engineering judgement

  19. Creativity Support to Improve Health-and-Safety in Manu-facturing Plants: Demonstrating Everyday Creativity

    OpenAIRE

    Zachos, K.; Maiden, N.; Levis, S.

    2015-01-01

    This paper reports the development and deployment of digi-tal support for human creativity in a domain outside of the creative industries -- health-and-safety management in man-ufacturing plants. It reports applied research to extend a risk detection and resolution process at a world-class manufac-turing plant that produces tractors with creativity techniques and new digital support for the plant employees to use these techniques effectively as part of the risk detection and reso-lution proce...

  20. Reactor limitation system improves the safety and availability of the Angra 2 nuclear power plant

    International Nuclear Information System (INIS)

    Souza Mendes, J.E. de

    1987-01-01

    Beyond the classic Reactor Protection System and Reactor Control System, nuclear plant Angra 2 has a third system called Reactor Limitation System which combines the intelligence features of the control systems with the high reliability of the protection systems. In determined events, which are not controlled by the control system (e.g.: load rejection, failure of one main reactor coolant pump), the Reactor Limitation System actuates automatically in order to lead the plant to a safe operating condition and so it avoids the actuation of the Reactor Protection System and consequently the reactor trip. This increases safety and availability of the plant and reduces component stresses. After the safe operating condition is reached, the process guidance automatically returns to the control systems. (Author) [pt

  1. INPO's role in the improvement of nuclear plant safety and reliability

    International Nuclear Information System (INIS)

    Pate, Z.T.

    1985-01-01

    The Institute of Nuclear Power Operations (INPO) is a nonprofit, nongovernment corporation dedicated to promoting safe and reliable operation and quality construction of nuclear power plants. The institute is supported by the US nuclear utility industry. All US utilities building or operating nuclear power plants are INPO members. INPO's many activities fall within four major program areas: (1) Evaluations: INPO conducts detailed evaluations of all US operating plants, utility corporate support of nuclear programs, and nuclear projects under active construction. Events analysis and information exchange: INPO reviews plant reports of abnormal or unusual events and reports the lessons learned to the industry. INPO also fosters improved industry communications through a worldwide telecommunications network. (3) Training and accreditation: INPO assists utilities, including international participants, in developing quality training programs; and (4) Assistance: the Institute provides various forms of assistance to members and participants. INPO places emphasis on visits to facilities to aid members and participants in various program or operational areas

  2. Nuclear power plant safety and reliability improvements derived from operational experience analysis

    International Nuclear Information System (INIS)

    Zebroski, E.L.; Rosen, S.L.

    1983-01-01

    The frequency and distribution of significant events at operating nuclear power plants can be used as one of several indices to obtain a preliminary indication of unit and/or utility performance in the development and implementation of improvements aimed at reducing the probability or consequences of troublesome events. Initial data of this type are presented along with qualifications on the validity of such indicators. Planned further steps to improve this as a performance indicator are noted

  3. Methods for the computerized control of nuclear power plants for improved safety, quality, and productivity

    International Nuclear Information System (INIS)

    Heising, C.D.

    1988-12-01

    The purpose of this research project was to translate the Total Quality Control (TQC) management system to the operation of nuclear power plants. This work began on September 30, 1986 and continued through December 30, 1988 at an original funding level of $251,000. Four graduate research assistants were employed during the two year project period, with all four completing master's level degree theses or projects. In addition, several papers were presented at technical conferences regarding this work. Currently, one paper has been accepted for publication in the journal of Operations Research, and two more papers are currently being written for submission in the near future. The project also received favorable publicity at the university and elsewhere. The major findings of this work was that, indeed, the Japanese management system of Total Quality Control (TQC) may be applied to nuclear power plant operation to achieve enhanced safety and increased efficiency. Two utilities are already heavily engaged in implementing TQC at their nuclear power plants: the Kansai Electric Company of Osaka, Japan and the Florida Power and Light Company of Miami, Florida. We have documented in detail the progress of these utilities in their attempts to implement TQC, as well as highlighting ''success stories'' of TQC implementation in other industries

  4. Efficient improvement of nuclear power plant safety by reorganization of risk-informed safety importance evaluation methods for piping welded portions

    Energy Technology Data Exchange (ETDEWEB)

    Irie, Takashi; Hanafusa, Hidemitsu; Suyama, Takeshi [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Morota, Hidetsugu; Kojima, Sigeo; Mizuno, Yoshinobu [Computer Software Development Co., Ltd., Tokyo (Japan)

    2002-09-01

    In this work, risk information was used to evaluate the safety importance of piping welded portions which were important for plant operation and maintenance of nuclear power plants. There are two types of risk-informed safety importance evaluation methods, namely the ASME method and the EPRI method. Since both methods have advantages and disadvantages, elements of each method were combined and reorganized. Considerations included whether the degradation mechanisms would be objectively evaluated and whether plant safety would be efficiently improved. The most objective and efficient method was as follows. Piping failure potential is quantitatively and objectively evaluated for failure with probabilistic fracture mechanics (PFM) and for other degradation mechanisms with empirical failure rates, and conditional core damage probability (CCDP) is calculated with PSA. This method reduces the inspected segment numbers to 1/4 of the deterministic method and increases the ratio of risk, which is covered by the inspected segments, to total risk from 80% of the deterministic method to 95%. Piping inspection numbers decreased for safety injection systems that were required the inspections by the deterministic method. Piping inspections were required for part of main feed water and main steam systems that were not required the inspections by the deterministic method. (author)

  5. 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

  6. Review of nuclear power plant safety cable aging studies with recommendations for improved approaches and for future work

    International Nuclear Information System (INIS)

    Gillen, Kenneth Todd; Bernstein, Robert

    2010-01-01

    Many U. S. nuclear power plants are approaching 40 years of age and there is a desire to extend their life for up to 100 total years. Safety-related cables were originally qualified for nuclear power plant applications based on IEEE Standards that were published in 1974. The qualifications involved procedures to simulate 40 years of life under ambient power plant aging conditions followed by simulated loss of coolant accident (LOCA). Over the past 35 years or so, substantial efforts were devoted to determining whether the aging assumptions allowed by the original IEEE Standards could be improved upon. These studies led to better accelerated aging methods so that more confident 40-year lifetime predictions became available. Since there is now a desire to potentially extend the life of nuclear power plants way beyond the original 40 year life, there is an interest in reviewing and critiquing the current state-of-the-art in simulating cable aging. These are two of the goals of this report where the discussion is concentrated on the progress made over the past 15 years or so and highlights the most thorough and careful published studies. An additional goal of the report is to suggest work that might prove helpful in answering some of the questions and dealing with some of the issues that still remain with respect to simulating the aging and predicting the lifetimes of safety-related cable materials.

  7. IAEA meeting report of the consultants' meeting on safety improvements of WWER-440 model 230 nuclear power plants

    International Nuclear Information System (INIS)

    1998-05-01

    In 1992 the identification and ranking of safety issues for WWER-440/230 NPPs were completed. Some 100 safety issues have been identified and ranked according to their safety impact on the plant's defence in depth. The IAEA work was based on a conceptual design review and a series of safety review missions to each one of the four sites (Bohunice, Kozloduy, Novovoronezh and Kola) where WWER-440/230 NPPs were in operation. Results were published by the IAEA in February and May 1992. Since the initial publications the IAEA has continued its activities and conducted a number of follow-up missions and technical visits to the NPP sites to review the status of implementation of safety upgradings. In September of 1994 a Consultants Meeting (CM) was held in Vienna to consolidate all the information available to the AEA. The results of that meeting were included in the report WWER-SC-107. During 1995-1997 further information became available from technical visits conducted by the IAEA at the four sites. Therefore, in February 1998 the IAEA convened a CM in Vienna to prepare an up-to-date report on the status of safety improvements at WWER-440/230 NPPs. This report contains the results of that meeting. It reviews the status of remedial actions implemented at each plant in response to the safety issues originally identified in the IAEA reports and the remaining safety concerns. For convenience the issues are presented under broad titles related to: integrity of barriers (Section 2); performance of basic functions (Section 3); performance of supporting functions (Section 4); internal hazards (Section 5); external hazards (Section 6); accident analysis (Section 7); operational functions (Section 8); Section 9 provides an overview of the results of PSA studies performed and under way. Section 10 provides specific information on the status of Armenia NPP. It is presented separately because the IAEA assistance to this plant was provided in the framework of the national technical co

  8. Water chemistry data acquisition, processing, evaluation and diagnostic systems in Light Water Reactors: Future improvement of plant reliability and safety

    International Nuclear Information System (INIS)

    Uchida, S.; Takiguchi, H.; Ishigure, K.

    2006-01-01

    Data acquisition, processing and evaluation systems have been applied in major Japanese PWRs and BWRs to provide (1) reliable and quick data acquisition with manpower savings in plant chemical laboratories and (2) smooth and reliable information transfer among chemists, plant operators, and supervisors. Data acquisition systems in plants consist of automatic and semi-automatic instruments for chemical analyses, e. g., X-ray fluorescence analysis and ion chromatography, while data processing systems consist of PC base-sub-systems, e.g., data storage, reliability evaluation, clear display, and document preparation for understanding the plant own water chemistry trends. Precise and reliable evaluations of water chemistry data are required in order to improve plant reliability and safety. For this, quality assurance of the water chemistry data acquisition system is needed. At the same time, theoretical models are being applied to bridge the gaps between measured water chemistry data and the information desired to understand the interaction of materials and cooling water in plants. Major models which have already been applied for plant evaluation are: (1) water radiolysis models for BWRs and PWRs; (2) crevice radiolysis model for SCC in BWRs; and (3) crevice pH model for SG tubing in PWRs. High temperature water chemistry sensors and automatic plant diagnostic systems have been applied in only restricted areas. ECP sensors are gaining popularity as tools to determine the effects of hydrogen injection in BWR systems. Automatic plant diagnostic systems based on artificial intelligence will be more popular after having sufficient experience with off line diagnostic systems. (author)

  9. Improving operating room safety

    Directory of Open Access Journals (Sweden)

    Garrett Jill

    2009-11-01

    Full Text Available Abstract Despite the introduction of the Universal Protocol, patient safety in surgery remains a daily challenge in the operating room. This present study describes one community health system's efforts to improve operating room safety through human factors training and ultimately the development of a surgical checklist. Using a combination of formal training, local studies documenting operating room safety issues and peer to peer mentoring we were able to substantially change the culture of our operating room. Our efforts have prepared us for successfully implementing a standardized checklist to improve operating room safety throughout our entire system. Based on these findings we recommend a multimodal approach to improving operating room safety.

  10. Safety improvements of Temelin NPP

    International Nuclear Information System (INIS)

    Vita, J.

    2000-01-01

    A detailed overview is given of the efforts made to enhance the safety level of the plant considering recommendations of a number of assessment missions. A list is presented of 10 international missions of the IAEA at the Temelin plant, covering the period 1990 to 1998. For each mission the date and objective is given, the focus of the assessment is characterized, the international participation of experts is specified, and the main conclusions of the experts is reproduced. A commented list of 60 main design changes and safety improvements is also included, as they were implemented in the wake of various safety assessments. An overview of the Temelin safety improvement programme is attached, comprising brief descriptions of 30 planned improvement items together with the time schedules. (A.K.)

  11. Application of space and aviation technology to improve the safety and reliability of nuclear power plant operations. Final report

    International Nuclear Information System (INIS)

    1980-04-01

    This report investigates various technologies that have been developed and utilized by the aerospace community, particularly the National Aeronautics and Space Administration (NASA) and the aviation industry, that would appear to have some potential for contributing to improved operational safety and reliability at commercial nuclear power plants of the type being built and operated in the United States today. The main initiator for this study, as well as many others, was the accident at the Three Mile Island (TMI) nuclear power plant in March 1979. Transfer and application of technology developed by NASA, as well as other public and private institutions, may well help to decrease the likelihood of similar incidents in the future

  12. Operating plant safety analysis needs

    International Nuclear Information System (INIS)

    Young, M.Y.; Love, D.S.

    1992-01-01

    The primary objective for nuclear power station owners is to operate and manage their plants safely. However, there is also a need to provide economical electric power, which requires that the unit be operated as efficiently as possible, consistent with the safety requirements. The objectives cited above can be achieved through the identification and use of available margins inherent in the plant design. As a result of conservative licensing and analytical approaches taken in the past, many of these margins may be found in the safety analysis limits within which plants currently operate. Improvements in the accuracy of the safety analysis, and a more realistic treatment of plant initial and boundary conditions, can make this margin available for a variety of uses which enhance plant performance, help to reduce O and M costs, and may help to extend licensed operation. Opportunities for improvement exist in several areas in the accident analysis normally performed for Chapter 15 of the FSAR. For example, recent modifications to the ECCS rule, 10CFR50.46 and Appendix K, allow use of margins previously unavailable in the analysis of the Loss of Coolant Accident (LOCA). To take advantage of this regulatory change, new methods are being developed to analyze both the large and small break loss of coolant accident (LOCA). As this margin is used, enhancements in the analysis of other transients will become necessary. The paper discusses accident analysis methods, future development needs, and analysis margin utilization in specific accident scenarios

  13. Improving versus maintaining nuclear safety

    International Nuclear Information System (INIS)

    2002-01-01

    The concept of improving nuclear safety versus maintaining it has been discussed at a number of nuclear regulators meetings in recent years. National reports have indicated that there are philosophical differences between NEA member countries about whether their regulatory approaches require licensees to continuously improve nuclear safety or to continuously maintain it. It has been concluded that, while the actual level of safety achieved in all member countries is probably much the same, this is difficult to prove in a quantitative way. In practice, all regulatory approaches require improvements to be made to correct deficiencies and when otherwise warranted. Based on contributions from members of the NEA Committee on Nuclear Regulatory Activities (CNRA), this publication provides an overview of current nuclear regulatory philosophies and approaches, as well as insights into a selection of public perception issues. This publication's intended audience is primarily nuclear safety regulators, but government authorities, nuclear power plant operators and the general public may also be interested. (author)

  14. Probabilistic safety criteria for improvement of Nuclear Power Plant design and operation

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Nam Jin; Chung, Woo Sick; Park, Moon Kyu [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1991-12-15

    The procedure of this study is to : research on the status of IAEA(International Atomic Energy Agency) member states about the policy of safety goals, study figures of merit and demerit that inherently exist in the existing methodology for reliability allocation, develop an efficient methodology for allocating reliability from top-level safety goals to intermediate and low-level PSC, write a computer code on the basis of the methodology proposed in the study, and apply the methodology to Surry Unit 1 that is the type of PWR.

  15. An external domino effects investment approach to improve cross-plant safety within chemical clusters

    International Nuclear Information System (INIS)

    Reniers, Genserik

    2010-01-01

    Every company situated within a chemical cluster faces the risk of being struck by an escalating accident at one of its neighbouring plants (the so-called external domino effect risks). These cross-plant risks can be reduced or eliminated if neighbouring companies are willing to invest in systems and measures to prevent them. However, since reducing such multi-plant risks does not lead to direct economic benefits, enterprises tend to be reluctant to invest more than needed for meeting minimal legal requirements and they tend to invest without collaborating. The suggested approach in this article indicates what information is required to evaluate the available investment options in external domino effects prevention. To this end, game theory is used as a promising scientific technique to investigate the decision-making process on investments in prevention measures simultaneously involving several plants. The game between two neighbouring chemical plants and their strategic investment behaviour regarding the prevention of external domino effects is described and an illustrative example is provided. Recommendations are formulated to advance cross-plant prevention investments in a two-company cluster.

  16. An external domino effects investment approach to improve cross-plant safety within chemical clusters.

    Science.gov (United States)

    Reniers, Genserik

    2010-05-15

    Every company situated within a chemical cluster faces the risk of being struck by an escalating accident at one of its neighbouring plants (the so-called external domino effect risks). These cross-plant risks can be reduced or eliminated if neighbouring companies are willing to invest in systems and measures to prevent them. However, since reducing such multi-plant risks does not lead to direct economic benefits, enterprises tend to be reluctant to invest more than needed for meeting minimal legal requirements and they tend to invest without collaborating. The suggested approach in this article indicates what information is required to evaluate the available investment options in external domino effects prevention. To this end, game theory is used as a promising scientific technique to investigate the decision-making process on investments in prevention measures simultaneously involving several plants. The game between two neighbouring chemical plants and their strategic investment behaviour regarding the prevention of external domino effects is described and an illustrative example is provided. Recommendations are formulated to advance cross-plant prevention investments in a two-company cluster. Copyright (c) 2009 Elsevier B.V. All rights reserved.

  17. Improving Patient Safety: Improving Communication.

    Science.gov (United States)

    Bittner-Fagan, Heather; Davis, Joshua; Savoy, Margot

    2017-12-01

    Communication among physicians, staff, and patients is a critical element in patient safety. Effective communication skills can be taught and improved through training and awareness. The practice of family medicine allows for long-term relationships with patients, which affords opportunities for ongoing, high-quality communication. There are many barriers to effective communication, including patient factors, clinician factors, and system factors, but tools and strategies exist to address these barriers, improve communication, and engage patients in their care. Use of universal precautions for health literacy, appropriate medical interpreters, and shared decision-making are evidence-based tools that improve communication and increase patient safety. Written permission from the American Academy of Family Physicians is required for reproduction of this material in whole or in part in any form or medium.

  18. Nuclear safety. Improvement programme

    International Nuclear Information System (INIS)

    2000-01-01

    In this brochure the improvement programme of nuclear safety of the Mochovce NPP is presented in detail. In 1996, a 'Mochovce NPP Nuclear Safety Improvement Programme' was developed in the frame of unit 1 and 2 completion project. The programme has been compiled as a continuous one, with the aim to reach the highest possible safety level at the time of commissioning and to establish good preconditions for permanent safety improvement in future. Such an approach is in compliance with the world's trends of safety improvement, life-time extension, modernisation and nuclear station power increase. The basic document for development of the 'Programme' is the one titled 'Safety Issues and their Ranking for WWER 440/213 NPP' developed by a group of IAEA experts. The following organisations were selected for solution of the safety measures: EUCOM (Consortium of FRAMATOME, France, and SIEMENS, Germany); SKODA Prague, a.s.; ENERGOPROJEKT Prague, a.s. (EGP); Russian organisations associated in ATOMENERGOEXPORT; VUJE Trnava, a.s

  19. Power plants and safety 1982

    International Nuclear Information System (INIS)

    1982-01-01

    The papers of this volume deal with the whole range of safety issues from planning and construction to the operation of power plants, and discuss also issues like availability and safety of power plants, protective clothes and their incommodating effect, alternatives for rendering hot-water generators safe and the safety philosophy in steam turbine engineering. (HAG) [de

  20. Industry use of operating experience to achieve improved nuclear plant safety

    International Nuclear Information System (INIS)

    Zebroski, E.L.

    1981-01-01

    A principal lesson drawn from the accident at Three Mile Island was the need for a comprehensive and rigorous system for analysis and feedback of operating experience to reactor operators. Chief executives of US utilities directed in mid-1979 that an intensive and rigorous system of analysis and feedback of operating experience be established. This system is commonly referred to as the ''Significant Events Program''. Since April 1980, the Nuclear Safety Analysis Center (NSAC) has been joined by the Institute for Nuclear Power Operations (INPO) in the field investigation of significant operating events. NSAC has responsibility for analysis of the design and physical events aspects, while INPO has primary responsibility for the operators' aspects, including procedures and training. The process of screening, analysis and feedback of operating experience is now functioning as a seven-step process. A variety of data sources is used, including License Event Reports and outage and major maintenance reports. These are compiled and indexed in convenient form. However, such data bases are used only as incidental tools for the basic investigations and analytical efforts. Rapid dissemination of results is provided by a computer-aided conferencing system, which links 70 operating LWR reactors in the USA, and which has now been extended to four utilities outside the USA, representing several dozen more reactors. Major safety and economic incentives are evident for the rigorous use of such operating experience and for participation in a comprehensive system. Traditional habits of secrecy are recognized as obstacles to timely communication. A principal responsibility of top management of reactor-operating organizations is to overcome such habits where they are counter to the public interest, as well as to the health and survival interest of the utility itself

  1. IMPROVING PATIENT SAFETY:

    DEFF Research Database (Denmark)

    Bagger, Bettan; Taylor Kelly, Hélène; Hørdam, Britta

    Improving patient safety is both a national and international priority as millions of patients Worldwide suffer injury or death every year due to unsafe care. University College Zealand employs innovative pedagogical approaches in educational design. Regional challenges related to geographic......, social and cultural factors have resulted in a greater emphasis upon digital technology. Attempts to improve patient safety by optimizing students’ competencies in relation to the reporting of clinical errors, has resulted in the development of an interdisciplinary e-learning concept. The program makes...

  2. Industrial safety in power plants

    International Nuclear Information System (INIS)

    1987-01-01

    The proceedings of the VGB conference 'Industrial safety in power plants' held in the Gruga-Halle, Essen on January 21 and 22, 1987, contain the papers reporting on: Management responsibility for and legal consequences of industrial safety; VBG 2.0 Industrial Accident Prevention Regulation and the power plant operator; Operational experience gained with wet-type flue gas desulphurization systems; Flue gas desulphurization systems: Industrial-safety-related requirements to be met in planning and operation; the effects of the Hazardous Substances Ordinance on power plant operation; Occupational health aspects of heat-exposed jobs in power plants; Regulations of the Industrial Accident Insurance Associations concerning heat-exposed jobs and industrial medical practice; The new VBG 30 Accident Prevention Regulation 'Nuclear power plants'; Industrial safety in nuclear power plants; safe working on and within containers and confined spaces; Application of respiratory protection equipment in power plants. (HAG) [de

  3. Safety improvements made at the Loviisa nuclear power plant to reduce fire risks originating from the turbine generators

    International Nuclear Information System (INIS)

    Virolainen, T.; Marttila, J.; Aulamo, H.

    1998-01-01

    Comprehensive upgrading measures have been completed for the Loviisa Nuclear Power Plant (modified VVER440/V213). These were carried out from the start of the design phase and during operation to ensure safe plant shutdown in the event of a large turbine generator oil fire. These modifications were made mainly on a deterministic basis according to specific risk studies and fire analyses. As part of the probabilistic safety assessment, a fire risk analysis was made that confirmed the importance of these upgrading measures. In fact, they should be considered as design basis modifications for all VVER440 plants. (author)

  4. Modifications to nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA's programme for safety standards for nuclear power plants. It supplements Section 7 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation, which establishes the safety requirements for the modification of nuclear power plants. Reasons for carrying out modifications to nuclear power plants may include: (1) maintaining or strengthening existing safety provisions and thus maintaining consistency with or improving on the current design. (2) recovering from plant faults. (3) improving the thermal performance or increasing the power rating of the plant. (4) increasing the maintainability of the plant, reducing the radiation exposure of personnel or reducing the costs of plant maintenance. And (5) extending the design life of the plant. Most modifications, made on the basis of operating experience, are intended to improve on the design or to improve operational performance and flexibility. Some are rendered necessary by new regulatory requirements, ageing of the plant or obsolescence of equipment. However, the benefits of regularly updating the plant design can be jeopardized if modifications are not kept under rigorous control throughout the lifetime of the plant. The need to reduce costs and improve efficiency, in combination with changes to the structure of the electricity generation sector of the economy in many countries, has led many companies to make changes in the structure of the operating organization for nuclear power plants. Whatever the reason for such organizational changes, consideration should be given to the effects of those changes with the aim of ensuring that they would have no impacts that would compromise the safety of the plant. The objective of this Safety Guide is to provide guidance and recommendations on controlling activities relating to modifications at nuclear power plants in order to reduce risk and to ensure that the configuration of the plant is at all times under

  5. Modifications to nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2007-01-01

    This Safety Guide was prepared under the IAEA's programme for safety standards for nuclear power plants. It supplements Section 7 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation, which establishes the safety requirements for the modification of nuclear power plants. Reasons for carrying out modifications to nuclear power plants may include: (1) maintaining or strengthening existing safety provisions and thus maintaining consistency with or improving on the current design. (2) recovering from plant faults. (3) improving the thermal performance or increasing the power rating of the plant. (4) increasing the maintainability of the plant, reducing the radiation exposure of personnel or reducing the costs of plant maintenance. And (5) extending the design life of the plant. Most modifications, made on the basis of operating experience, are intended to improve on the design or to improve operational performance and flexibility. Some are rendered necessary by new regulatory requirements, ageing of the plant or obsolescence of equipment. However, the benefits of regularly updating the plant design can be jeopardized if modifications are not kept under rigorous control throughout the lifetime of the plant. The need to reduce costs and improve efficiency, in combination with changes to the structure of the electricity generation sector of the economy in many countries, has led many companies to make changes in the structure of the operating organization for nuclear power plants. Whatever the reason for such organizational changes, consideration should be given to the effects of those changes with the aim of ensuring that they would have no impacts that would compromise the safety of the plant. The objective of this Safety Guide is to provide guidance and recommendations on controlling activities relating to modifications at nuclear power plants in order to reduce risk and to ensure that the configuration of the plant is at all times under

  6. Project of SVBR-75/100 reactor plant with improved safety for nuclear sources of small and medium power

    International Nuclear Information System (INIS)

    Dragunov, Yu. G.; Stepanov, V. S.; Klimov, N. N.; Dedul, A. V.; Bolvanchikov, S. N.; Zrodnikov, A. V.; Tolhinsky, G. I.; Komlev, O. G.

    2004-01-01

    As a result of the joint work performed recently by FSUE OKB Gidropress, SNC RF-IPPE and other organizations the technical feasibility is shown for creation and usage in nuclear power engineering of the unified reactor plant (RP) SVBR-75/100 with fast neutron reactor core and lead-bismuth coolant (LBC) in the primary circuit. Technical design of SVBR-75/100 reactor plant is based on the following: 50-year operation experience in development and operation of RP with LBC for nuclear submarines; experience in development and operation of fast reactor with sodium coolant; experience in optimization of LBC technology at nuclear submarines and ground-based test benches; conceptual design of SVBR-75 reactor plant (for renovation of Units 2, 3 and 4 of Novovoronezh NPP). Technical solutions laid down in the basis of SVBR-75/100 reactor plant design are oriented towards the industrial basis, structural materials existing in Russia, as well as the unique LBC technology with experimental and practical support. The concept of SVBR-75/100 reactor plant safety assurance is based on the following provisions: maximum usage of inherent safety supported by physical features of fast neutron reactor, chemically inert LBC in the primary circuit, integral layout and special design solutions; maximum possible combination of normal operation and safety functions in RP systems. Small power of SVBR-75/100 RP makes it possible to manufacture the complete set of RP main equipment at the factory and delivery it to NPP site as-finished practically using any transport including railway. Possible fields of application of SVBR-75/100 reactor plants: modular NPPs of different power; renovation of NPPs with light water reactors exhausted their service life; independent nuclear power sources for different applications (ground-based nuclear water-desalinating plants, etc. )(author)

  7. Technical specification improvement through safety margin considerations

    International Nuclear Information System (INIS)

    Howard, R.C.; Jansen, R.L.

    1986-01-01

    Westinghouse has developed an approach for utilizing safety analysis margin considerations to improve plant operability through technical specification revision. This approach relies on the identification and use of parameter interrelations and sensitivities to identify acceptable operating envelopes. This paper summarizes technical specification activities to date and presents the use of safety margin considerations as another viable method to obtain technical specification improvement

  8. Dukovany nuclear power plant safety

    International Nuclear Information System (INIS)

    1999-01-01

    Presentation covers recommended safety issues for the Dukovany NPP which have been solved with satisfactory conclusions. Safety issues concerned include: radiation safety; nuclear safety; security; emergency preparedness; health protection at work; fire protection; environmental protection; chemical safety; technical safety. Quality assurance programs at all stages on NPP life time is described. Report includes description of NPP staff training provision, training simulator, emergency operating procedures, emergency preparedness, Year 2000 problem, inspections and life time management. Description of Dukovany Plant Safety Analysis Projects including integrity of the equipment, modernisation, equipment innovation and safety upgrading program show that this approach corresponds to the actual practice applied in EU countries, and fulfilment of current IAEA requirements for safety enhancement of the WWER 440/213 units in the course of MORAWA Equipment Upgrading program

  9. More safety by improving the safety culture

    International Nuclear Information System (INIS)

    Laaksonen, J.

    1993-01-01

    In its meeting in 1986, after Chernobyl accident, the INSAG group concluded, that the most important reason for the accident was lack of safety culture. Later the group realized that the safety culture, if it is well enough, can be used as a powerful tool to assess and develop practices affecting safety in any country. A comprehensive view on the various aspects of safety culture was presented in the INSAG-4 report published in 1991. Finland was among the first nations include the concept of safety culture in its regulations. This article describes the roles of government and the regulatory body in creating a national safety culture. How safety culture is seen in the operation of a nuclear power plant is also discussed. (orig.)

  10. A Study on SE Methodology for Design of Big Data Pilot Platform to Improve Nuclear Power Plant Safety

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Junguk; Cha, Jae-Min; Kim, Jun-Young; Park, Sung-Ho; Yeom, Choong-Sub [Institute for Advanced Engineering (IAE), Yongin (Korea, Republic of)

    2016-10-15

    A big data concept is expected to have a large impact on the safety of the nuclear power plant (NPP) from the beginning of the big data era. Though there are high interests on the NPP safety with the big data, almost no studies on the logical and physical structures and the systematic design methods of the big data platform for the NPP safety have been conducted. For the current study, a new big data pilot platform for the NPP safety is designed with the main focus on the health monitoring and early warning systems, and a tailored design process based on the systems engineering approaches is proposed to manage inherent high complexity of the platform design. The big data concept is expected to have a large impact on the safety of the NPP. So, in this study, the big data pilot platform for the health monitoring and early warning of the NPP is designed. For this, the development process based on the SE approach for the pilot platform is proposed and the design results along with the proposed process are also presented. Implementation of the individual modules and integrations of those are in currently progress.

  11. A Study on SE Methodology for Design of Big Data Pilot Platform to Improve Nuclear Power Plant Safety

    International Nuclear Information System (INIS)

    Shin, Junguk; Cha, Jae-Min; Kim, Jun-Young; Park, Sung-Ho; Yeom, Choong-Sub

    2016-01-01

    A big data concept is expected to have a large impact on the safety of the nuclear power plant (NPP) from the beginning of the big data era. Though there are high interests on the NPP safety with the big data, almost no studies on the logical and physical structures and the systematic design methods of the big data platform for the NPP safety have been conducted. For the current study, a new big data pilot platform for the NPP safety is designed with the main focus on the health monitoring and early warning systems, and a tailored design process based on the systems engineering approaches is proposed to manage inherent high complexity of the platform design. The big data concept is expected to have a large impact on the safety of the NPP. So, in this study, the big data pilot platform for the health monitoring and early warning of the NPP is designed. For this, the development process based on the SE approach for the pilot platform is proposed and the design results along with the proposed process are also presented. Implementation of the individual modules and integrations of those are in currently progress

  12. Tools for plant safety engineer

    International Nuclear Information System (INIS)

    Fabic, S.

    1996-01-01

    This paper contains: - review of tools for monitoring plant safety equipment reliability and readiness, before and accident (performance indicators for monitoring the risk and reliability performance and for determining when degraded performance alert levels are achieved) - brief reviews of tools for use during an accident: Emergency Operating Procedures (EOPs), Emergency Response Data System (ERDS), Reactor Safety Assessment System (RSAS), Computerized Accident Management Support

  13. Safety of industrial irradiation plants

    International Nuclear Information System (INIS)

    1992-01-01

    Radiation is nowadays used in many applications in industry and medicine; accidental exposure, however, can have grave consequences as large doses of radiation occur in the 600 accelerator or gamma source plants in use around the world. This film explains the operation of irradiation plants and the safety procedures that must be followed to prevent accidents and to ensure safe use

  14. The development of design technology on the safety parameter display system for the operability improvement of the nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Young Joon; Choi, Hae Yoon; Ahn, Jang Sun; Lee, Tae Woo; Lee, Ki Won; Kim, Kil Kon; Baek, Seong Min; Sul, Young Sil [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-12-31

    The purpose of this study is, basically, threefold. Firstly, through detailed comparison, the difference between the safety parameters used in the EPG and CFMS is analyzed. Furthermore, to anticipate and extrapolate the problems that might be encountered when developing the CFMS system that utilizes safety parameters consistent with that of the EPG. Secondly, the setpoint analysis of the CFMS alarm algorithm was done for there is a possibility of causing spurious alarms since the alarm setpoint of the YGN 3,4 CFMS is not reflective of the plant operating conditions nor accident progression. Lastly, the analysis of the success path for each accident was done to help operator in mitigating the accident by using the pictorial path of the success path during an accident condition. Moreover, in this analysis, the contents of the concerns that KINS raised regarding the YGN 3,4 SPDS has been addressed from the designer`s perspective. 33 figs., 16 refs. (Author) .new.

  15. Process and plant safety

    CERN Document Server

    Hauptmanns, Ulrich

    2015-01-01

    Accidents in technical installations are random events. Hence they cannot be totally avoided. Only the probability of their occurrence may be reduced and their consequences be mitigated. The book proceeds from hazards caused by materials and process conditions to indicating technical and organizational measures for achieving the objectives of reduction and mitigation. Qualitative methods for identifying weaknesses of design and increasing safety as well as models for assessing accident consequences are presented. The quantitative assessment of the effectiveness of safety measures is explained. The treatment of uncertainties plays a role there. They stem from the random character of the accident and from lacks of knowledge on some of the phenomena to be addressed. The reader is acquainted with the simulation of accidents, safety and risk analyses and learns how to judge the potential and limitations of mathematical modelling. Risk analysis is applied amongst others to “functional safety” and the determinat...

  16. Report of the Fukushima nuclear accident by the National Academy of Science. Lessons learned from the Fukushima nuclear accident for improving safety of U.S. nuclear plants

    International Nuclear Information System (INIS)

    Nariai, Hideki

    2014-01-01

    U.S. National Academy of Science investigated the accident at the Fukushima Daiichi nuclear plant initiated by the Great East Japan Earthquake for two years and published a draft report in July 24, 2014. Investigation results were summarized in nine new findings and made ten recommendations in a wide horizon; (1) hardware countermeasures against severe accidents and training of operators, (2) upgrade of risk assessment capability for beyond design basis accident, (3) incorporation of new information about hazards in safety regulations, (4) needed improvement of off-site emergency preparedness, and (5) improvements of nuclear safety culture. New information about hazards related with tsunami assessment, new risk assessment for beyond design basis accident, advice of foreigner resident evacuations, regulatory capture, and safety culture and regulator's specialty were discussed as Japanese issues. (T. Tanaka)

  17. Safety of Nuclear Power Plants: Design. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  18. Evolutionary CANDU 9 plant improvements

    International Nuclear Information System (INIS)

    Yu, S.K.W.

    1999-01-01

    The CANDU 9 is a 935 MW(e) nuclear power plant (NPP) based on the multi-unit Darlington and Bruce B designs with additional enhancements from our ongoing engineering and research programs. Added to the advantages of using proven systems and components, CANDU 9 offers improvement features with enhanced safety, improved operability and maintenance including a control centre with advanced man-machine interface, and improved project delivery in both engineering and construction. The CANDU 9 NPP design incorporated safety enhancements through careful attention to emerging licensing and safety issues. The designers assessed, revised and evolved such systems as the moderator, end shield, containment and emergency core cooling (ECC) systems while providing an integrated final design that is more passive and severe-accident-immune. AECL uses a feedback process to incorporate lessons learned from operating plants, from current projects experiences and from the implementation or construction phase of previous projects. Most of the requirements for design improvements are based on a systematic review of current operating CANDU stations in the areas of design and reliability, operability, and maintainability. The CANDU 9 Control Centre provides plant staff with improved operability and maintainability capabilities due to the combination of systematic design with human factors engineering and enhanced operating and diagnostics features. The use of advanced engineering tools and modem construction methods will reduce project implementation risk on project costs and schedules. (author)

  19. Optimization of safety equipment outages improves safety

    International Nuclear Information System (INIS)

    Cepin, Marko

    2002-01-01

    Testing and maintenance activities of safety equipment in nuclear power plants are an important potential for risk and cost reduction. An optimization method is presented based on the simulated annealing algorithm. The method determines the optimal schedule of safety equipment outages due to testing and maintenance based on minimization of selected risk measure. The mean value of the selected time dependent risk measure represents the objective function of the optimization. The time dependent function of the selected risk measure is obtained from probabilistic safety assessment, i.e. the fault tree analysis at the system level and the fault tree/event tree analysis at the plant level, both extended with inclusion of time requirements. Results of several examples showed that it is possible to reduce risk by application of the proposed method. Because of large uncertainties in the probabilistic safety assessment, the most important result of the method may not be a selection of the most suitable schedule of safety equipment outages among those, which results in similarly low risk. But, it may be a prevention of such schedules of safety equipment outages, which result in high risk. Such finding increases the importance of evaluation speed versus the requirement of getting always the global optimum no matter if it is only slightly better that certain local one

  20. Availability Improvement of German Nuclear Power Plants

    International Nuclear Information System (INIS)

    Wilhelm, Oliver

    2008-01-01

    High availability is important for the safety and economical performance of Nuclear Power Plants (NPP). The strategy for availability improvement in a typical German PWR shall be discussed here. Key parameters for strategy development are plant design, availability of safety systems, component reliability, preventive maintenance and outage organization. Plant design, availability of safety systems and component reliability are to a greater extent given parameters that can hardly be influenced after the construction of the plant. But they set the frame for maintenance and outage organisation which have shown to have a large influence on the availability of the plant. (author)

  1. Planned activities to improve safety

    International Nuclear Information System (INIS)

    1998-01-01

    This document presents the fulfilling of the Brazilian obligations under the Convention on Nuclear Safety. The Chapter 6 of the document contains some details about the planed activities to safety improvements

  2. Robots and plant safety

    International Nuclear Information System (INIS)

    Christensen, P.

    1996-02-01

    The application of robots in the harsh environments in which TELEMAN equipment will have to operate has large benefits, but also some drawbacks. The main benefit is the ability gained to perform tasks where people cannot go, while there is a possibility of inflicting damage to the equipment handled by the robot, and the plant when mobile robots are involved. The paper describes the types of possible damage and the precautions to be taken in order to reduce the frequency of the damaging events. A literature study for the topic only gave some insight into examples, but no means for a systematic treatment of the topic. (au) 16 refs

  3. Safety in nuclear power plants

    International Nuclear Information System (INIS)

    Koeberlein, K.

    1987-01-01

    In nuclear power plants large amounts of radioactive fission products ensue from the fission of uranium. In order to protect the environment, the radioactive material is confined in multiple 'activity barriers' (crystal matrix of the fuel, fuel cladding, coolant boundary, safety containment, reactor building). These barriers are protected by applying a defense-in-depth concept (high quality requirements, protection systems which recognize and terminate operational incidents, safety systems to cope with accidents). In spite of a favorable safety record of German nuclear power plants it is obvious - and became most evident by the Chernobyl accident - that absolute safety is not achievable. At Chernobyl, however, design disadvantages of that reactor type (like positive reactivity feedback of coolant voiding, missing safety containment) played an important role in accident initiation and progression. Such features of the Russian 'graphite-moderated pressure tube boiling water reactor' are different from those of light water reactors operating in western countries. The essential steps of the waste management of the nuclear fuel cycle ('Entsorgung') are the interim storage, the shipment, and the reprocessing of the spent fuel and the final repository of radioactive waste. Reprocessing means the separation of fossil material (uranium, plutonium) from radioactive waste. Legal requirements for radiological protection of the environment, which are identical for nuclear power plants and reprocessing plant, are complied with by means of comprehensive filter systems. Safety problems of a reprocessing plant are eased considerably by the fact that system pressures, process temperatures and energy densities are low. In order to confine the radioactive waste from the biosphere for a very long period of time, it is to be discarded after appropriate treatment into the deep geological underground of salt domes. (orig./HP) [de

  4. Monitoring and diagnosis systems to improve nuclear power plant reliability and safety. Proceedings of the specialists` meeting

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-31

    The 50 participants, representing 21 Member States (Argentina, Austria, Belgium, Canada, Czech Republic, France, Germany, Hungary, Japan, Netherlands, Norway, Russian Federation, Slovak Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, Ukraine, UK and USA), reviewed the current approaches in Member States in the area of monitoring and diagnosis systems. The Meeting attempted to identify advanced techniques in the field of diagnostics of electrical and mechanical components for safety and operation improvements, reviewed actual practices and experiences related to the application of those systems with special emphasis on real occurrences, exchanged current experiences with diagnostics as a means for predictive maintenance. Refs, figs, tabs.

  5. Monitoring and diagnosis systems to improve nuclear power plant reliability and safety. Proceedings of the specialists' meeting

    International Nuclear Information System (INIS)

    1996-01-01

    The 50 participants, representing 21 Member States (Argentina, Austria, Belgium, Canada, Czech Republic, France, Germany, Hungary, Japan, Netherlands, Norway, Russian Federation, Slovak Republic, Slovenia, Spain, Sweden, Switzerland, Turkey, Ukraine, UK and USA), reviewed the current approaches in Member States in the area of monitoring and diagnosis systems. The Meeting attempted to identify advanced techniques in the field of diagnostics of electrical and mechanical components for safety and operation improvements, reviewed actual practices and experiences related to the application of those systems with special emphasis on real occurrences, exchanged current experiences with diagnostics as a means for predictive maintenance. Refs, figs, tabs

  6. 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

  7. Improved safety at CERN

    CERN Multimedia

    2006-01-01

    As announced in Weekly Bulletin No. 43/2006, a new approach to the implementation of Safety at CERN has been decided, which required taking some managerial decisions. The guidelines of the new approach are described in the document 'New approach to Safety implementation at CERN', which also summarizes the main managerial decisions I have taken to strengthen compliance with the CERN Safety policy and Rules. To this end I have also reviewed the mandates of the Safety Commission and the Safety Policy Committee (SAPOCO). Some details of the document 'Safety Policy at CERN' (also known as SAPOCO42) have been modified accordingly; its essential principles, unchanged, remain the basis for the safety policy of the Organisation. I would also like to inform you that I have appointed Dr M. Bona as the new Head of the Safety Commission until 31.12.2008, and that I will proceed soon to the appointment of the members of the new Safety Policy Committee. All members of the personnel are deemed to have taken note of the d...

  8. Plant air systems safety study: Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    1982-05-01

    The Portsmouth Gaseous Diffusion Plant Air System facilities and operations are reviewed for potential safety problems not covered by standard industrial safety procedures. Information is presented under the following section headings: facility and process description (general); air plant equipment; air distribution system; safety systems; accident analysis; plant air system safety overview; and conclusion

  9. Improvement of the safety regulations in the management of radioactive waste accumulated in the liquid radioactive waste water basins of the PO 'Majak' (Ozersk), the Siberian Chemical Plant (Seversk) and the Mining-Chemical Plant (Zheleznogorsk)

    International Nuclear Information System (INIS)

    Vishnevski, Y.G.; Kislov, A.I.; Irushkin, V.M.

    2002-01-01

    One of the most important problems of radiation safety in Russia is the decommissioning of the liquid radioactive waste water basins of the PO 'Majak' (Ozersk), Siberian Chemical Plant (Seversk) and Mining-Chemical Plant (Zheleznogorsk). The liquid radioactive waste water basins were constructed in 1950-1960 for the collection and storage of liquid waste from the radiochemical plants. The potential hazards of the liquid in the radioactive waste water basins are: migration of radionuclides into the soil of the liquid radioactive waste water basin floors; wind-induced carry-over of radionuclides from the liquid radioactive waste water basins; hazards (radiation included) to the environment and population arising in case physical barriers and hydraulic structures are damaged; and criticality hazards. The classification of the liquid radioactive waste water basins were developed based on the collection and analyzes of the information on liquid radioactive waste water basin characteristics and the method of multicriterion expert assessment of potential hazards. Three main directions for the improvement of safety regulation in the management of radioactive waste accumulated in the liquid radioactive waste water basins were defined: 1. Common directions for the improvement of safety regulation in the area of rehabilitation of the territories contaminated with radioactive substances. 2. Common directions for the improvement of safety regulation in the area of rehabilitation of the territories, such as the liquid radioactive waste water basins. 3. Special directions for the regulatory activities in the area of operation and decommissioning of the liquid radioactive waste water basins of the PO 'Majak' (Ozersk), Siberian Chemical Plant (Seversk) and Mining-Chemical Plant (Zheleznogorsk). As a result, concrete recommendations on safety regulation for the management of radioactive waste accumulated in the water basins were developed. (author)

  10. Research on the improvement of nuclear safety

    International Nuclear Information System (INIS)

    Yoo, Keon Joong; Kim, Dong Soo; Kim, Hui Dong; Park, Chang Kyu

    1993-06-01

    To improve the nuclear safety, this project is divided into three areas which are the development of safety analysis technology, the development of severe accident analysis technology and the development of integrated safety assessment technology. 1. The development of safety analysis technology. The present research aims at the development of necessary technologies for nuclear safety analysis in Korea. Establishment of the safety analysis technologies enables to reduce the expenditure both by eliminating excessive conservatisms incorporated in nuclear reactor design and by increasing safety margins in operation. It also contributes to improving plant safety through realistic analyses of the Emergency Operating Procedures (EOP). 2. The development of severe accident analysis technology. By the computer codes (MELCOR and CONTAIN), the in-vessel and the ex-vessel severe accident phenomena are simulated. 3. The development of integrated safety assessment technology. In the development of integrated safety assessment techniques, the included research areas are the improvement of PSA computer codes, the basic study on the methodology for human reliability analysis (HRA) and common cause failure (CCF). For the development of the level 2 PSA computer code, the basic research for the interface between level 1 and 2 PSA, the methodology for the treatment of containment event tree are performed. Also the new technologies such as artificial intelligence, object-oriented programming techniques are used for the improvement of computer code and the assessment techniques

  11. Twenty years of improvements in LWR safety

    International Nuclear Information System (INIS)

    Franks, S. III; Mulkey, J.P.; Moonka, A.

    1996-01-01

    Substantial improvements have been made in the safety of light-water reactors in the US during the past two decades, making currently operating reactors safer than ever before. Safety improvements have resulted both from regulatory and operational changes and from new knowledge and technology. The US Nuclear Regulatory Commission, the US Department of Energy, and the American nuclear power industry have worked together and with the international community to enhance the safety of existing plants and to incorporate lessons learned from prior operation into designs for a new generation of advanced, inherently safer reactors

  12. Discussion of important safety requirements for new nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Lin; Jia Xiang; Yan Tianwen; Li Wenhong; Li Chun

    2014-01-01

    This paper presents the analysis of several important safety requirements and improvement direction. Technical view of security goals on site safety evaluation, internal and external events fortification, serious accident prevention and mitigation, as well as the core, containment system and instrument control system design and engineering optimization, and etc are indicated. It will be useful for new plant design, construction and safety improvement. (authors)

  13. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This publication supports the Safety Requirements on the Safety of Nuclear Power Plants: Design. This Safety Guide was prepared on the basis of a systematic review of all the relevant publications including the Safety Fundamentals, Safety of Nuclear Power Plants: Design, current and ongoing revisions of other Safety Guides, INSAG reports and other publications that have addressed the safety of nuclear power plants. This Safety Guide also provides guidance for Contracting Parties to the Convention on Nuclear Safety in meeting their obligations under Article 14 on Assessment and Verification of Safety. The Safety Requirements publication entitled Safety of Nuclear Power Plants: Design states that a comprehensive safety assessment and an independent verification of the safety assessment shall be carried out before the design is submitted to the regulatory body. This publication provides guidance on how this requirement should be met. This Safety Guide provides recommendations to designers for carrying out a safety assessment during the initial design process and design modifications, as well as to the operating organization in carrying out independent verification of the safety assessment of new nuclear power plants with a new or already existing design. The recommendations for performing a safety assessment are suitable also as guidance for the safety review of an existing plant. The objective of reviewing existing plants against current standards and practices is to determine whether there are any deviations which would have an impact on plant safety. The methods and the recommendations of this Safety Guide can also be used by regulatory bodies for the conduct of the regulatory review and assessment. Although most recommendations of this Safety Guide are general and applicable to all types of nuclear reactors, some specific recommendations and examples apply mostly to water cooled reactors. Terms such as 'safety assessment', 'safety analysis' and 'independent

  14. The safety of future nuclear power plants in France

    International Nuclear Information System (INIS)

    Queniart, D.

    1988-10-01

    The present paper concerns certain personal thoughts on the safety of future French power plants, which will come into operation at the beginning of the next century. These reflections, which are made on the author's own behalf and, under no circumstances, implicate at this stage the official views of the French safety authorities, are aimed at defining some directions for the improvement of safety in these future plants as compared with that of plants presently in operation or under construction

  15. Improving safety in mining

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-08-15

    AcuMine is a spin-out company from CRC Mining Australia and the University of Sydney's Australian Centre for Field Robotics (ACFR). Its focus is to provide safety and fatigue management in mining environments. The AcuLine Haul Check system was its first development. Of greater benefit to safety in mines will be the AcuMine Proximity System (APPS) developed to reliably detect and warn drivers when in proximity to other trucks and utility vehicles and to detect personnel near to those heavy vehicles. 6 figs.

  16. Safety Assessment - Swedish Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B. [Luleaa Univ. of Technology (Sweden)

    1996-12-31

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs.

  17. Safety Assessment - Swedish Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kjellstroem, B.

    1996-01-01

    After the reactor accident at Three Mile Island, the Swedish nuclear power plants were equipped with filtered venting of the containment. Several types of accidents can be identified where the filtered venting has no effect on the radioactive release. The probability for such accidents is hopefully very small. It is not possible however to estimate the probability accurately. Experiences gained in the last years, which have been documented in official reports from the Nuclear Power Inspectorate indicate that the probability for core melt accidents in Swedish reactors can be significantly larger than estimated earlier. A probability up to one in a thousand operating years can not be excluded. There are so far no indications that aging of the plants has contributed to an increased accident risk. Maintaining the safety level with aging nuclear power plants can however be expected to be increasingly difficult. It is concluded that the 12 Swedish plants remain a major threat for severe radioactive pollution of the Swedish environment despite measures taken since 1980 to improve their safety. Closing of the nuclear power plants is the only possibility to eliminate this threat. It is recommended that until this is done, quantitative safety goals, same for all Swedish plants, shall be defined and strictly enforced. It is also recommended that utilities distributing misleading information about nuclear power risks shall have their operating license withdrawn. 37 refs

  18. Methods for safety culture improvement

    International Nuclear Information System (INIS)

    Sivintsev, Yu.V.

    1998-01-01

    New IAEA publication concerning the problems of safety assurance covering different aspects beginning from terminology applied and up to concrete examples of well and poor safety culture development at nuclear facilities is discussed. The safety culture is defined as such set of characteristics and specific activities of institutions and individual persons which states that safety problems of a nuclear facility are given the attention determined by their importance as being of highest priority. The statements of the new document have recommended, not mandatory character. It is emphasized that the process of safety culture improvement at nuclear facilities should be integral component of management procedure, not a bolt on extra

  19. Elevating standards, improving safety.

    Science.gov (United States)

    Clarke, Richard

    2014-08-01

    In our latest 'technical guidance' article, Richard Clarke, sales and marketing director at one of the UK's leading lift and escalator specialists, Schindler, examines some of the key issues surrounding the specification, maintenance, and operation of lifts in hospitals to help ensure the highest standards of safety and reliability.

  20. Safety assessment, safety performance indicators at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Baji, C.; Vamos, G.; Toth, J.

    2001-01-01

    The Paks Nuclear Power Plant has been using different methods of safety assessment (event analysis, self-assessment, probabilistic safety analysis), including performance indicators characterizing both operational and safety performance since the early years of operation of the plant. Regarding the safety performance, the indicators include safety system performance, number of scrams, release of radioactive materials, number of safety significant events, industrial safety indicator, etc. The Paks NPP also reports a set of ten indicators to WANO Performance Indicator Programme which, among others, include safety related indicators as well. However, a more systematic approach to structuring and trending safety indicators is needed so that they can contribute to the enhancement of the operational safety. A more comprehensive set of indicators and a systematic evaluation process was introduced in 1996. The performance indicators framework proposed by the IAEA was adapted to Paks in this year to further improve the process. Safety culture assessment and characterizing safety culture is part of the assessment process. (author)

  1. Bohunice Nuclear Power Plant Safety Upgrading Program

    International Nuclear Information System (INIS)

    Toth, A.; Fagula, L.

    1996-01-01

    Bohunice nuclear Power Plant generation represents almost 50% of the Slovak republic electric power production. Due to such high level of commitment to nuclear power in the power generation system, a special attention is given to safe and reliable operation of NPPs. Safety upgrading and operational reliability improvement of Bohunice V-1 NPP was carried out by the Bohunice staff continuously since the plant commissioning. In the 1990 - 1993 period extensive projects were realised. As a result of 'Small Reconstruction of the Bohunice V-1 NPP', the standards of both the nuclear safety and operational reliability have been significantly improved. The implementation of another modifications that will take place gradually during extended refuelling outages and overhauls in the course of 1996 through 1999, is referred to as the Gradual Reconstruction of the Bohunice V-1 Plant. The general goal of the V-1 NPP safety upgrading is the achievement of internationally acceptable level of nuclear safety. Extensive and financially demanding modification process of Bohunice V-2 NPP is likely to be implemented after a completion of the Gradual Reconstruction of the Bohunice V-1 NPP, since the year 1999. With this in mind, a first draft of the strategy of the Bohunice V-2 NPP upgrading program based on Probabilistic Safety assessment consideration was developed. A number of actions with a general effect on Bohunice site safety is evident. All these activities are aimed at reaching the essential objective of Bohunice NPP Management - to ensure a safe, reliable and effective electric energy and heat generation at the Bohunice site. (author)

  2. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  3. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2003-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations. To be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources. And to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  4. Safety of nuclear power plants: Operation. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    The safety of a nuclear power plant is ensured by means of its proper siting, design, construction and commissioning, followed by the proper management and operation of the plant. In a later phase, proper decommissioning is required. This Safety Requirements publication supersedes the Code on the Safety of Nuclear Power Plants: Operation, which was issued in 1988 as Safety Series No. 50-C-O (Rev. 1). The purpose of this revision was: to restructure Safety Series No. 50-C-O (Rev. 1) in the light of the basic objectives, concepts and principles in the Safety Fundamentals publication The Safety of Nuclear Installations; to be consistent with the requirements of the International Basic Safety Standards for Protection against Ionizing Radiation and for the Safety of Radiation Sources; and to reflect current practice and new concepts and technical developments. Guidance on fulfillment of these Safety Requirements may be found in the appropriate Safety Guides relating to plant operation. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the basic objectives, concepts and principles that are presented in the Safety Fundamentals publication The Safety of Nuclear Installations. This publication deals with matters specific to the safe operation of land based stationary thermal neutron nuclear power plants, and also covers their commissioning and subsequent decommissioning

  5. Basic safety principles for nuclear power plant

    International Nuclear Information System (INIS)

    Zhang Shiguan

    1989-01-01

    To ensure the safety operation of nuclear power plant, one should strictly adhere to the implelmentation of safety codes and the establishment of nuclear safety code system, as well as the applicable basic safety principles of nuclear power plants. This article briefly introduce the importance of nuclear codes and its economic benefits and the implementation of basic safety principles to be accumulated in practice for many years by various countries

  6. Technical basis for instrumentation and control design improvements in WWER-440/230 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-01-01

    Instrumentation and control (I and C) has been recognized as an area which requires substantial improvements in WWER NNPs, particularly for model 230 plants. Under contract with the IAEA the Spanish company Empresarios Agrupados (EA) developed a basic document proposing a technical basis for improvements related to the following most significant aspects of I and C: criteria for safety classification; remote shutdown panel; I and C support to operation and control room design; instrumentation set point margins; accident monitoring instrumentation. This publication is derived from the original report of EA which was circulated by the IAEA for review by staff members and experts from various Member States. It was finally agreed upon at a Consultants' meeting convened by the IAEA in Vienna in May 1994 with the participation of experts from France, Germany and Spain. The guidance expressed in this report is based on the IAEA/NUSS standards, safety guides and practices, and on regulations in use in various Member States. It is proposed as a way of carrying out the necessary studies to improve safety by upgrading the vital part of instrumentation an control in WWER-440 model 230 nuclear power plants. 28 refs, 3 figs

  7. Safety design of Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ouyang Yu; Zhang Lian; Du Shenghua; Zhao Jiayu

    1984-01-01

    Safety issues have been greatly emphasized through the design of the Qinshan Nuclear Power Plant. Reasonable safety margine has been taken into account in the plant design parameters, the design incorporated various safeguard systems, such as engineering safety feature systems, safety protection systems and the features to resist natural catastrophes, e. g. earthquake, hurricanes, tide and so on. Preliminary safety analysis and environmental effect assessment have been done and anti-accident provisions and emergency policy were carefully considered. Qinshan Nuclear Power Plant safety related systems are designed in accordance with the common international standards established in the late 70's, as well as the existing engineering standard of China

  8. Deterministic Safety Analysis for Nuclear Power Plants. Specific Safety Guide (Russian Edition)

    International Nuclear Information System (INIS)

    2014-01-01

    The objective of this Safety Guide is to provide harmonized guidance to designers, operators, regulators and providers of technical support on deterministic safety analysis for nuclear power plants. It provides information on the utilization of the results of such analysis for safety and reliability improvements. The Safety Guide addresses conservative, best estimate and uncertainty evaluation approaches to deterministic safety analysis and is applicable to current and future designs. Contents: 1. Introduction; 2. Grouping of initiating events and associated transients relating to plant states; 3. Deterministic safety analysis and acceptance criteria; 4. Conservative deterministic safety analysis; 5. Best estimate plus uncertainty analysis; 6. Verification and validation of computer codes; 7. Relation of deterministic safety analysis to engineering aspects of safety and probabilistic safety analysis; 8. Application of deterministic safety analysis; 9. Source term evaluation for operational states and accident conditions; References

  9. Periodic safety reviews of nuclear power plants

    International Nuclear Information System (INIS)

    Toth, Csilla

    2009-01-01

    Operational nuclear power plants (NPPs) are generally subject to routine reviews of plant operation and special safety reviews following operational events. In addition, many Member States of the International Atomic Energy Agency (IAEA) have initiated systematic safety reassessment, termed periodic safety review (PSR), to assess the cumulative effects of plant ageing and plant modifications, operating experience, technical developments, site specific, organizational and human aspects. These reviews include assessments of plant design and operation against current safety standards and practices. PSRs are considered an effective way of obtaining an overall view of actual plant safety, to determine reasonable and practical modifications that should be made in order to maintain a high level of safety throughout the plant's operating lifetime. PSRs can be used as a means to identify time limiting features of the plant. The trend is to use PSR as a condition for deciding whether to continue operation of the plant beyond the originally established design lifetime and for assessing the status of the plant for long term operation. To assist Member States in the implementation of PSR, the IAEA develops safety standards, technical documents and provides different services: training courses, workshops, technical meetings and safety review missions for the independent assessment of the PSR at NPPs, including the requirements for PSR, the review process and the PSR final reports. This paper describes the PSR's objectives, scopes, methods and the relationship of PSR with other plant safety related activities and recent experiences of Member States in implementation of PSRs at NPPs. (author)

  10. Design of plant safety model in plant enterprise engineering environment

    International Nuclear Information System (INIS)

    Gabbar, Hossam A.; Suzuki, Kazuhiko; Shimada, Yukiyasu

    2001-01-01

    Plant enterprise engineering environment (PEEE) is an approach aiming to manage the plant through its lifecycle. In such environment, safety is considered as the common objective for all activities throughout the plant lifecycle. One approach to achieve plant safety is to embed safety aspects within each function and activity within such environment. One ideal way to enable safety aspects within each automated function is through modeling. This paper proposes a theoretical approach to design plant safety model as integrated with the plant lifecycle model within such environment. Object-oriented modeling approach is used to construct the plant safety model using OO CASE tool on the basis of unified modeling language (UML). Multiple views are defined for plant objects to express static, dynamic, and functional semantics of these objects. Process safety aspects are mapped to each model element and inherited from design to operation stage, as it is naturally embedded within plant's objects. By developing and realizing the plant safety model, safer plant operation can be achieved and plant safety can be assured

  11. Safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    Vuorinen, A.

    1993-01-01

    The role and purpose of safety principles for nuclear power plants are discussed. A brief information is presented on safety objectives as given in the INSAG documents. The possible linkage is discussed between the two mentioned elements of nuclear safety and safety culture. Safety culture is a rather new concept and there is more than one interpretation of the definition given by INSAG. The defence in depth is defined by INSAG as a fundamental principle of safety technology of nuclear power. Discussed is the overall strategy for safety measures, and features of nuclear power plants provided by the defence-in-depth concept. (Z.S.) 7 refs

  12. Safety aspects and operating experience of LWR plants in Japan

    International Nuclear Information System (INIS)

    Aoki, S.; Yoshioka, T.; Toyota, M.; Hinoki, M.

    1977-01-01

    To develop nuclear power generation for the future, it is necessary to put further emphasis on safety assurance and to endeavour to devise measures to improve plant availability, based on the careful analysis of causes that reduce plant availability. The paper discusses the results of studies on the following items from such viewpoints: (1) Safety and operating experience of LWR nuclear power plants in Japan: operating experience with LWRs; improvements in LWR design during the past ten years; analysis of the factors affecting plant availability; (2) Assurance of safety and measures to increase availability: measures for safety and environmental protection; measures to reduce radiation exposure of employees; appropriateness of maintenance and inspection work; measures to increase plant availability; measures to improve reliability of equipment and components; (3) Future technical problems. (author)

  13. Safety goals and safety culture opening plenary. 1. WANO's Role in Maintaining and Improving Safety Culture

    International Nuclear Information System (INIS)

    Tsutsumi, Ryosuke

    2001-01-01

    Over the past several years, operators of the world's nuclear plants have compiled an increasingly impressive record of operational performance. Among the many factors that have led to this improvement are the unprecedented cooperation and information exchange among the world's nuclear operators. This paper presents the World Association of Nuclear Operators (WANO) operating experience program and WANO peer review program as examples of the kinds of interaction that are occurring around the globe to maintain and improve the nuclear safety culture. In addition, some unique features of WANO are discussed. WANO has established four programs to help its members communicate effectively with each other. These include the exchange of operating experiences, voluntary peer reviews, professional and technical development, and technical support and exchange. The operating experience program alerts members to events that have occurred at other NPPs and enables members to take appropriate actions to prevent event recurrence. When an event occurs at a plant, management at that plant analyses the event and completes an event report, which is then sent to the WANO regional center to which the plant belongs. After a regional center review and necessary iteration, the report is posted onto the WANO Web site to make it available to all WANO members. By the end of 2000, more than 1500 event reports had been posted. The WANO Peer Review Program is a unique opportunity for members to learn and share the best worldwide insights into safe and reliable nuclear operations. The peer review program has become one of WANO's most important activities containing all essential elements of WANO's mission. A WANO peer review team consists of 15 to 16 people with NPP experience; most team members are from countries outside the one that they are visiting. These teams of peers from plants around the world visit host plants upon request to identify strengths and areas for improvement, with a strong

  14. A BWR Safety and Operability Improvements

    International Nuclear Information System (INIS)

    Sawyer, Craig D.

    1993-01-01

    The A BWR is the culmination of 30 years of design, development and operating experience of BWRs around the world. It represents across the board improvements is safety, operation and maintenance practices (O and M), economics, radiation exposure and rad waste generation. More than ten years and $20m5 went into the design and development of its new features, and it is now under construction in Japan. This paper concentrates on the safety and operability improvements. In the safety area, more than a decade improvement in core damage frequency (CDFR) has been assessed by formal PIRA techniques, with CDFR less than 10 -6 /year. Severe accident mitigation has also been formally addressed in the design. Plant operations were simplified by incorporation of better materials, optimum use of redundancy in mechanical and electrical equipment so that on-line maintenance can be performed, by better arrangements which account for required maintenance practices, and by an advanced control room

  15. Safety culture improvement. An adaptive management framework

    International Nuclear Information System (INIS)

    Obadia, Isaac Jose

    2005-01-01

    After the Chernobyl nuclear accident in 1986, the International Atomic Energy Agency (IAEA) established the safety culture concept as a proactive mean to contribute to safety improvement, starting a worldwide safety culture enhancement program within nuclear organizations mainly focused on nuclear power plants. More recently, the safety culture concept has been extended to non-power applications such as nuclear research reactors and nuclear technological research and development organizations. In 1999, the Nuclear Engineering Institute (IEN), a research and technological development unit of the Brazilian Nuclear Energy Commission (CNEN), started a management change program aiming at improving its performance level of excellence. This change program has been developed assuming the occurrence of complex causal inter-relationships between the organizational culture and the implementation of the management process. A systematic and adaptive management framework comprised of a safety culture improvement practice integrated to a management process based on the Criteria for Excellence of the Brazilian Quality Award Model, has been developed and implemented at IEN. The case study has demonstrated that the developed framework makes possible an effective safety culture improvement and simultaneously facilitates an effective implementation of the management process, thus providing some governance to the change program. (author)

  16. Safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Selvatici, E.

    1981-01-01

    A study about the safety analysis of nuclear power plant, giving emphasis to how and why to do is presented. The utilization of the safety analysis aiming to perform the licensing requirements is discussed, and an example of the Angra 2 and 3 safety analysis is shown. Some presented tendency of the safety analysis are presented and examples are shown.(E.G.) [pt

  17. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2000-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  18. Safety of nuclear power plants: Design. Safety requirements

    International Nuclear Information System (INIS)

    2004-01-01

    The present publication supersedes the Code on the Safety of Nuclear Power Plants: Design (Safety Series No. 50-C-D (Rev. 1), issued in 1988). It takes account of developments relating to the safety of nuclear power plants since the Code on Design was last revised. These developments include the issuing of the Safety Fundamentals publication, The Safety of Nuclear Installations, and the present revision of various safety standards and other publications relating to safety. Requirements for nuclear safety are intended to ensure adequate protection of site personnel, the public and the environment from the effects of ionizing radiation arising from nuclear power plants. It is recognized that technology and scientific knowledge advance, and nuclear safety and what is considered adequate protection are not static entities. Safety requirements change with these developments and this publication reflects the present consensus. This Safety Requirements publication takes account of the developments in safety requirements by, for example, including the consideration of severe accidents in the design process. Other topics that have been given more detailed attention include management of safety, design management, plant ageing and wearing out effects, computer based safety systems, external and internal hazards, human factors, feedback of operational experience, and safety assessment and verification. This publication establishes safety requirements that define the elements necessary to ensure nuclear safety. These requirements are applicable to safety functions and the associated structures, systems and components, as well as to procedures important to safety in nuclear power plants. It is expected that this publication will be used primarily for land based stationary nuclear power plants with water cooled reactors designed for electricity generation or for other heat production applications (such as district heating or desalination). It is recognized that in the case of

  19. EDITORIAL: Safety aspects of fusion power plants

    Science.gov (United States)

    Kolbasov, B. N.

    2007-07-01

    This special issue of Nuclear Fusion contains 13 informative papers that were initially presented at the 8th IAEA Technical Meeting on Fusion Power Plant Safety held in Vienna, Austria, 10-13 July 2006. Following recommendation from the International Fusion Research Council, the IAEA organizes Technical Meetings on Fusion Safety with the aim to bring together experts to discuss the ongoing work, share new ideas and outline general guidance and recommendations on different issues related to safety and environmental (S&E) aspects of fusion research and power facilities. Previous meetings in this series were held in Vienna, Austria (1980), Ispra, Italy (1983), Culham, UK (1986), Jackson Hole, USA (1989), Toronto, Canada (1993), Naka, Japan (1996) and Cannes, France (2000). The recognized progress in fusion research and technology over the last quarter of a century has boosted the awareness of the potential of fusion to be a practically inexhaustible and clean source of energy. The decision to construct the International Thermonuclear Experimental Reactor (ITER) represents a landmark in the path to fusion power engineering. Ongoing activities to license ITER in France look for an adequate balance between technological and scientific deliverables and complying with safety requirements. Actually, this is the first instance of licensing a representative fusion machine, and it will very likely shape the way in which a more common basis for establishing safety standards and policies for licensing future fusion power plants will be developed. Now that ITER licensing activities are underway, it is becoming clear that the international fusion community should strengthen its efforts in the area of designing the next generations of fusion power plants—demonstrational and commercial. Therefore, the 8th IAEA Technical Meeting on Fusion Safety focused on the safety aspects of power facilities. Some ITER-related safety issues were reported and discussed owing to their potential

  20. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6`` cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author).

  1. Research on the improvement of nuclear safety -The development of LOCA analysis codes for nuclear power plant-

    International Nuclear Information System (INIS)

    Jang, Won Pyo; Jung, Yung Jong; Kim, Kyung Doo; Jung, Jae Joon; Kim, Won Suk; Han, Doh Heui; Hah, Kooi Suk; Jung, Bub Dong; Lee, Yung Jin; Hwang, Tae Suk; Lee, Sang Yong; Park, Chan Uk; Choi, Han Rim; Lee, Sang Jong; Choi, Jong Hoh; Ban, Chang Hwan; Bae, Kyoo Hwan

    1995-07-01

    The present research aims at development of both a best estimate methodology on LOCA analysis and, as an application, performance analyses of safety systems. SBLOCA analyses have been continued to examine the capacity reduction effect of ECCS since the second project year. As a results, core uncovery, which is requirement of URD has not been occurred in 6'' cold leg break. Although core uncovery has been predicted when DVI line has been broken for DVI+4-Train HPIS, the calculated PCT has lied well within the criterion. The effect of safety injection position and SIT characteristics are also analyzed for LBLOCA. The results show that cold leg injection is the most effective way and the adaption of advanced SIT could lead to elimination of LPSI pump from the safety system. On the other hand, the quantified uncertainties obtained from THTF and FLECHT/SEASET which represents blowdown and reflood phenomena, respectively, have been confirmed using IET(LOFT test). The application uncertainty for Kori unit 3 has been analyzed. Finally, application of the best estimate methodology using the uncertainties concerned with the code, the bais, and the application, leads to overall uncertainty of about 200K for Kori unit 3. 244 figs, 22 tabs, 92 refs. (Author)

  2. Safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Weihe, G. von; Pamme, H.

    2003-01-01

    Experience shows that German nuclear power plants have always been operated reliably and safely. Over the years, the safety level in these plants has been raised considerably so that they can stand any comparison with other countries. This is confirmed by the two reports published by the Federal Ministry for the Environment on the nuclear safety convention. Behind this, there must obviously stand countless appropriate 'good practices' and a safety management system in nuclear power plants. (orig.) [de

  3. The German nuclear power plant safety study

    International Nuclear Information System (INIS)

    1979-01-01

    With this study a new approach has been chosen, taking nuclear power plants as an example to assess and to describe the risks arising from the use of modern technology, including those hazards emanating from the rather hypothetical possibility of occurrence of very serious accidents. Following the definition of basic concepts and methods to be applied in risk assessment studied, as well as a brief account of the design and operating mode of nuclear power plants with PWRs', accidents and failures to be considered in a safety study are described. Using the course-of-event and fault tree analysis, the probability of fission product release as a consequence of failures in safety systems or of core meltdown is evaluated. Subsequently, the theoretical model for assessment of reactor accident consequences is presented, discussing such aspects as the dispersion of radioactivity in the atmosphere, the radiation dose model, safety and countermeasures, the model for the evaluation of health hazards as well as methods and calculations for estimating the reliability of risk assessments together with the remaining uncertainties. In an appendix to this study, the analyses presented in the study are discussed in the light of the TMI-2 event. This safety study showing the possibilities of detecting, keeping in check and minimizing harmful effects, can be regarded as a contribution to a better understanding of our modern, highly industrialised society, and eventually to an improvement of the quality of life. (GL) 891 GL/GL 892 MB [de

  4. Improved safety in ski jumping.

    Science.gov (United States)

    Wester, K

    1988-01-01

    Among approximately 2,600 licensed Norwegian ski jumpers, only three injuries that caused a permanent medical disability of at least 10% were incurred during the 5 year period from 1982 through 1986. When compared to the previous 5 year period (1977 to 1981), a dramatic improvement in safety is seen, as both number and severity of such injuries were markedly reduced. There are several probable reasons for this improved safety record: better preparation of the jumps, the return to using only one standard heel block, and the fact that coaches are being more responsible, especially with younger jumpers.

  5. Organizational factors influencing improvements in safety

    International Nuclear Information System (INIS)

    Marcus, A.; Nichols, M.L.; Olson, J.; Osborn, R.; Thurber, J.

    1992-01-01

    Research reported here seeks to identify the key organizational factors that influence safety-related performance indicators in nuclear power plants over time. It builds upon organizational factors identified in NUREG/CR-5437, and begins to develop a theory of safety-related performance and performance improvement based on economic and behavioral theories of the firm. Central to the theory are concepts of past performance, problem recognition, resource availability, resource allocation, and business strategies that focus attention. Variables which reflect those concepts are combined in statistical models and tested for their ability to explain scrams, safety system actuations, significant events, safety system failures, radiation exposure, and critical hours. Results show the performance indicators differ with respect to the sets of variables which serve as the best predictors of future performance, and past performance is the most consistent predictor of future performance

  6. NPP Temelin. Status of safety improvements

    International Nuclear Information System (INIS)

    1999-01-01

    The WWER-1000 Temelin NPP under construction has been subjected as other NPPs of the same type to numerous project reviews resulting in quite a number of recommendations for design changes. Results of the IAEA mission to review the resolution of WWER-1000 safety issues at Temelin NPP are cited in this paper. The main conclusions emphasize that a combination of eastern and western technology and practices led to safety improvements in comparison with the international practices. Plant managers are clearly committed to implementation of operational programs which are consistent with effective western operational safety practices. Considerable effort remains to bring planned programs to successful implementation, in particular in meeting the need to foster strong safety culture among all personnel

  7. Lessons drawn from the operation of EDF 900MW nuclear power plants to evaluate and improve safety

    International Nuclear Information System (INIS)

    Brisbois, J.; Dupuis, M.C.; Marcille, R.

    1986-06-01

    The lessons drawn from operation are used to detect the weak points not taken sufficiently into account by the design studies and to check that system availability remains at least equal to the level defined at the design stage. In both cases, they aim at deciding what remedial actions are to be taken. The latter may result in additional obligations (modification or addition of equipment or operating rules) and, in certain cases, may slacken obligations. This paper examines in turn: how are the events recorded and then analyzed, what lessons are drawn using concrete examples, what contribution is made by probabilistic risk studies using experience feedback in evaluating overall safety [fr

  8. Domestic Regulation for Periodic Safety Review of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Kim, Daesik; Ahn, Seunghoon; Auh, Geunsun; Lee, Jonghyeok

    2015-01-01

    The so-called Periodic Safety Review (PSR) has been carried out such safety assessment throughout its life, on a periodic basis. In January 2001, the Atomic Energy Act and related regulations were amended to adopt the PSR institutional scheme from IAEA Nuclear Safety Guide 50-SG-O12. At that time the safety assessment was made to review the plant safety on 10 safety factors, such as aging management and emergency planning, where the safety factor indicates the important aspects of safety of an operating NPP to be addressed in the PSR. According to this legislation, the domestic utility, the KHNP has conducted the PSR for the operating NPP of 10 years coming up from operating license date, starting since May 2000. Some revisions in the PSR rule were made to include the additional safety factors last year. This paper introduces the current status of the PSR review and regulation, in particular new safety factors and updated technical regulation. Comprehensive safety assessment for Korea Nuclear Power Plants have performed a reflecting design and procedure changes and considering the latest technology every 10 years. This paper also examined the PSR system changes in Korea. As of July 2015, reviews for PSR of 18 units have been completed, with 229 nuclear safety improvement items. And implementation have been completed for 165 of them. PSR system has been confirmed that it has contributed to improvement of plant safety. In addition, this paper examined the PSR system change in Korea

  9. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  10. Safety assessment and verification for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  11. Summary of the contractor information exchange meeting for improving the safety of Soviet-Designed Nuclear Power Plants, February 19, 1997

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-04-01

    This report summarizes a meeting held on February 19, 1997, in Washington, D.C. The meeting was held primarily to exchange information among the contractors involved in the U.S. Department of Energy`s efforts to improve the safety of Soviet-designed nuclear power plants. Previous meetings have been held on December 5-6, 1995, and May 22, 1996. The meetings are sponsored by the U.S. Department of Energy and coordinated by the Pacific Northwest National Laboratory. The U.S. Department of Energy works with countries to increase the level of safety at 63 Soviet-designed nuclear reactors operating in Armenia, Bulgaria, the Czech Republic, Hungary, Lithuania, Russia, Slovakia, and Ukraine. The work is implemented largely by commercial companies and individuals who provide technologies and services to the countries with Soviet-designed nuclear power plants. Attending the meeting were 71 representatives of commercial contractors, the U.S. Department of Energy, the U.S. Department of State, national laboratories, and other federal agencies. The presentations and discussions that occurred during the exchange are summarized in this report. While this report captures the general presentation and discussion points covered at the meeting, it is not a verbatim, inclusive record. To make the report useful, information presented at the meeting has been expanded to clarify issues, respond to attendees` requests, or place discussion points in a broader programmatic context. Appendixes A through F contain the meeting agenda, list of attendees, copies of presentation visuals and handouts, the Strategy Document discussed at the meeting, and a summary of attendees` post-meeting evaluation comments. As with past information exchanges, the participants found this meeting valuable and useful. In response to the participant`s requests, a fourth information exchange will be held later in 1997.

  12. Summary of the contractor information exchange meeting for improving the safety of Soviet-Designed Nuclear Power Plants, February 19, 1997

    International Nuclear Information System (INIS)

    1997-04-01

    This report summarizes a meeting held on February 19, 1997, in Washington, D.C. The meeting was held primarily to exchange information among the contractors involved in the U.S. Department of Energy's efforts to improve the safety of Soviet-designed nuclear power plants. Previous meetings have been held on December 5-6, 1995, and May 22, 1996. The meetings are sponsored by the U.S. Department of Energy and coordinated by the Pacific Northwest National Laboratory. The U.S. Department of Energy works with countries to increase the level of safety at 63 Soviet-designed nuclear reactors operating in Armenia, Bulgaria, the Czech Republic, Hungary, Lithuania, Russia, Slovakia, and Ukraine. The work is implemented largely by commercial companies and individuals who provide technologies and services to the countries with Soviet-designed nuclear power plants. Attending the meeting were 71 representatives of commercial contractors, the U.S. Department of Energy, the U.S. Department of State, national laboratories, and other federal agencies. The presentations and discussions that occurred during the exchange are summarized in this report. While this report captures the general presentation and discussion points covered at the meeting, it is not a verbatim, inclusive record. To make the report useful, information presented at the meeting has been expanded to clarify issues, respond to attendees' requests, or place discussion points in a broader programmatic context. Appendixes A through F contain the meeting agenda, list of attendees, copies of presentation visuals and handouts, the Strategy Document discussed at the meeting, and a summary of attendees' post-meeting evaluation comments. As with past information exchanges, the participants found this meeting valuable and useful. In response to the participant's requests, a fourth information exchange will be held later in 1997

  13. Safety of Nuclear Power Plants: Commissioning and Operation

    International Nuclear Information System (INIS)

    2011-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  14. Safety of Nuclear Power Plants: Commissioning and Operation (French Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  15. Safety of Nuclear Power Plants: Commissioning and Operation. Arabic Edition

    International Nuclear Information System (INIS)

    2011-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  16. Safety of Nuclear Power Plants: Commissioning and Operation (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe operation of nuclear power plants. Over recent years there have been developments in areas such as long term operation, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. It became necessary to revise the IAEA's safety requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the Fundamental Safety Principles. Contents: 1. Introduction; 2. Safety objectives and principles; 3. The management and organizational structure of the operating organization; 4. Management of operational safety; 5. Operational safety programmes; 6. Plant commissioning; 7. Plant operations; 8. Maintenance, testing, surveillance and inspection; 9. Preparation for decommissioning.

  17. Improving safety on rural local and tribal roads safety toolkit.

    Science.gov (United States)

    2014-08-01

    Rural roadway safety is an important issue for communities throughout the country and presents a challenge for state, local, and Tribal agencies. The Improving Safety on Rural Local and Tribal Roads Safety Toolkit was created to help rural local ...

  18. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Guinn, W.M.; Porter, N.J.

    1981-01-01

    The operator's role in nuclear safety is outlined and the concept of ''safety functions'' introduced. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. The plant safety evaluation uses four inputs in predicting the results of an event: the event initiator, the plant design, the initial plant conditions and setup, and the operator actions. If any of these inputs are not as assumed in the evaluation, confidence that the consequences will be as predicted is reduced. Based on the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results: Maintain plant setup in readiness to properly respond. Operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events. Monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of safety functions introduces this systematic approach and presents a hierarchy of protection. If the operator has difficulty identifying an event for any reason, the systematic safety function approach allows accomplishing the overall path of mitigating consequences. Ten functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions are identified

  19. EC6 safety design improvements

    Energy Technology Data Exchange (ETDEWEB)

    Yu, S.; Lee, A.G.; Soulard, M. [Candu Energy Inc., Mississauga, ON (Canada)

    2014-07-01

    The Enhanced CANDU 6 (EC6) builds on the proven high performance design such as the Qinshan CANDU 6 reactor, and has made improvements to safety, operational performance, and has incorporated extensive operational feedback. Completion of all three phases of the pre-licensing design review by the Canadian Regulator - the Canadian Nuclear Safety Commission has provided a higher level of assurance that the EC6 reference design has taken modern regulatory requirements and expectations into account and further confirmed that there are no fundamental barriers to licensing the EC6 design in Canada. The EC6 design is based on the defence-in-depth principles in INSAG-10 and provides further safety features that address the lessons learned from Fukushima. With these safety features, the EC6 design has strengthened accident prevention as the first priority in the defence-in-depth strategy, as outlined in INSAG-10. As well, the EC6 design has incorporated further mitigation measures to provide additional protection of the public and the environment if the preventive measures fail. The EC6 design has an appropriate combination of inherent, passive safety characteristics, engineered features and administrative safety measures to effectively prevent and mitigate severe accident progressions. A strong contributor to the robustness and redundancy of CANDU design is the two-group separation philosophy. This ensures a high degree of independence between safety systems as well as physical separation and functional independence in how fundamental safety functions are provided. This paper will describe the following safety features based on the application of defence-in-depth and design approach to prevent beyond design basis events progressing to severe accidents and to mitigate the consequences if it occurs: Improved steam generator heat sink via a more reliable emergency heat removal system; Increased time before manual field actions are required via enhanced capacity of

  20. Safety provisions of nuclear power plants

    International Nuclear Information System (INIS)

    Niehaus, F.

    1994-01-01

    Safety of nuclear power plants is determined by a deterministic approach complemented by probabilistic considerations. Much use has been made of the wealth of information from more than 6000 years of reactor operation. Design, construction and operation is governed by national and international safety standards and practices. The IAEA has prepared a set of Nuclear Safety Standards as recommendations to its Member States, covering the areas of siting, design, operations, quality assurance, and governmental organisations. In 1988 the IAEA published a report by the International Nuclear Safety Advisory Group on Basic Safety Principles for Nuclear Power Plants, summarizing the underlying objectives and principles of excellence in nuclear safety and the way in which its aspects are interrelated. The paper will summarize some of the key safety principles and provisions, and results and uses of Probabilistic Safety Assessments. Some comments will be made on the safety of WWER 440/230 and WWER-1000 reactors which are operated on Bulgaria. 8 figs

  1. Good safety culture maintenance at Leningrad nuclear power plant

    International Nuclear Information System (INIS)

    Ardanov, A.

    1996-01-01

    The evidence in favour of the Leningrad NPP commitment to safety tasks, as the case is in the international practice, is The Safety Policy Statement document where safety is declared to be more significant than the power generation related issues, with the entire responsibility for the safety provision taken over by the operating utility. To avoid the situation when the stated safety tasks and policy remain only a declaration, the organizational structure of the operating utility was expanded to include The Safety Control Department and The Quality Control Department whose tasks encompass the control of the achieved safety level, development of recommendations, measures and actions aimed at the safety culture improvement, assessment and revision of the criteria and requirements to the personnel and management. Each individual at LNPP whose activity affects the plant safety has been familiarized with The Safety Policy Statement document

  2. Good safety culture maintenance at Leningrad nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Ardanov, A [Safety Control Dept., Leningrad Nuclear Power Plant, Leningrad (Russian Federation)

    1997-12-31

    The evidence in favour of the Leningrad NPP commitment to safety tasks, as the case is in the international practice, is The Safety Policy Statement document where safety is declared to be more significant than the power generation related issues, with the entire responsibility for the safety provision taken over by the operating utility. To avoid the situation when the stated safety tasks and policy remain only a declaration, the organizational structure of the operating utility was expanded to include The Safety Control Department and The Quality Control Department whose tasks encompass the control of the achieved safety level, development of recommendations, measures and actions aimed at the safety culture improvement, assessment and revision of the criteria and requirements to the personnel and management. Each individual at LNPP whose activity affects the plant safety has been familiarized with The Safety Policy Statement document.

  3. Safety assessment principles for nuclear plants

    International Nuclear Information System (INIS)

    1992-01-01

    The present Safety Assessment Principles result from the revision of those which were drawn up following a recommendation arising from the Sizewell-B enquiry. The principles presented here relate only to nuclear safety; there is a section on risks from normal operation and accident conditions and the standards against which those risks are assessed. A major part of the document deals with the principles that cover the design of nuclear plants. The revised Safety assessment principles are aimed primarily at the safety assessment of new nuclear plants but they will also be used in assessing existing plants. (UK)

  4. Safety performance indicators used by the Russian Safety Regulatory Authority in its practical activities on nuclear power plant safety regulation

    International Nuclear Information System (INIS)

    Khazanov, A.L.

    2005-01-01

    The Sixth Department of the Nuclear, Industrial and Environmental Regulatory Authority of Russia, Scientific and Engineering Centre for Nuclear and Radiation Safety process, analyse and use the information on nuclear power plants (NPPs) operational experience or NPPs safety improvement. Safety performance indicators (SPIs), derived from processing of information on operational violations and analysis of annual NPP Safety Reports, are used as tools to determination of trends towards changing of characteristics of operational safety, to assess the effectiveness of corrective measures, to monitor and evaluate the current operational safety level of NPPs, to regulate NPP safety. This report includes a list of the basic SPIs, those used by the Russian safety regulatory authority in regulatory activity. Some of them are absent in list of IAEA-TECDOC-1141 ('Operational safety performance indicators for nuclear power plants'). (author)

  5. Effort on Nuclear Power Plants safety

    International Nuclear Information System (INIS)

    Prayoto.

    1979-01-01

    Prospects of nuclear power plant on designing, building and operation covering natural safety, technical safety, and emergency safety are discussed. Several problems and their solutions and nuclear energy operation in developing countries especially control and permission are also discussed. (author tr.)

  6. Criticality safety evaluation in Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Shirai, Nobutoshi; Nakajima, Masayoshi; Takaya, Akikazu; Ohnuma, Hideyuki; Shirouzu, Hidetomo; Hayashi, Shinichiro; Yoshikawa, Koji; Suto, Toshiyuki

    2000-04-01

    Criticality limits for equipments in Tokai Reprocessing Plant which handle fissile material solution and are under shape and dimension control were reevaluated based on the guideline No.10 'Criticality safety of single unit' in the regulatory guide for reprocessing plant safety. This report presents criticality safety evaluation of each equipment as single unit. Criticality safety of multiple units in a cell or a room was also evaluated. The evaluated equipments were ones in dissolution, separation, purification, denitration, Pu product storage, and Pu conversion processes. As a result, it was reconfirmed that the equipments were safe enough from a view point of criticality safety of single unit and multiple units. (author)

  7. Organizational factors influencing improvements in safety

    International Nuclear Information System (INIS)

    Marcus, A.; Nichols, M.L.; Olson, J.; Osborn, R.; Thurber, J.

    1991-01-01

    Results of conceptual and empirical research conducted by this research team, and published in NUREG-CR 5437, suggested that processes of organizational problem solving and learning provide a promising area for understanding improvement in safety-related performance in nuclear power plants. In this paper the authors describe the way in which they have built upon that work and gone much further in empirically examining a range of potentially important organizational factors related to safety. The paper describes (1) overall trends in plant performance over time on the Nuclear Regulatory Commission performance indicators, (2) the major elements in the conceptual framework guiding the current work, which seeks among other things to explain those trends, (3) the specific variables used as measures of the central concepts, (4) the results to date of the quantitative empirical work and qualitative work in progress, and (5) conclusions from the research

  8. Perspectives on plant vulnerabilities ampersand other plant and containment improvements

    International Nuclear Information System (INIS)

    LaChance, J.; Kolaczkowski, A.; Kahn, J.

    1996-01-01

    The primary goal of the Individual Plant Examination (IPE) Program was for licensees to identify plant-unique vulnerabilities and actions to address these vulnerabilities. A review of these vulnerabilities and plant improvements that were identified in the IPEs was performed as part of the IPE Insights Program sponsored by the U.S. Nuclear Regulatory Commission (NRC). The purpose of this effort was to characterize the identified vulnerabilities and the impact of suggested plant improvements. No specific definition for open-quotes vulnerabilityclose quotes was provided in NRC Generic Letter 88-20 or in the subsequent NRC IPE submittal guidance documented in NUREG-1335. Thus licensees were left to use their own definitions. Only 20% of the plants explicitly stated that they had vulnerabilities. However, most licensees identified other plant improvements to address issues not explicitly classified as vulnerabilities, but pertaining to areas in which overall plant safety could potentially be increased. The various definitions of open-quotes vulnerabilityclose quotes used by the licensees, explicitly identified vulnerabilities, proposed plant improvements to address these vulnerabilities, and other plant improvements are summarized and discussed

  9. Advanced power reactors with improved safety characteristics

    International Nuclear Information System (INIS)

    Birkhofer, A.

    1994-01-01

    The primary objective of nuclear safety is the protection of individuals, society and environment against radiological hazards from accidental releases of radioactive materials contained in nuclear reactors. Hereto, these materials are enclosed by several successive barriers and the barriers protected against mishaps and accidents by a multi-level system of safety precautions. The evolution of reactor technology continuously improves this concept and its implementation. At a world-wide scale, several advanced reactor concepts are currently being considered, some of them already at a design stage. Essential safety objectives include both further strengthening the prevention of accidents and improving the containment of fission products should an accident occur. The proposed solutions differ considerably with regard to technical principles, plant size and time scales considered for industrial application. Two typical approaches can be distinguished: The first approach basically aims at an evolution of power reactors currently in use, taking into account the findings from safety research and from operation of current plants. This approach makes maximum use of proven technology and operating experience but may nevertheless include new safety features. The corresponding designs are often termed 'large evolutionary'. The second approach consists in more fundamental changes compared to present designs, often with strong emphasis on specific passive features protecting the fuel and fuel cladding barriers. Owing to the nature and capability of those passive features such 'innovative designs' are mostly smaller in power output. The paper describes the basic objectives of such developments and illustrates important technical concepts focusing on next generation plants, i.e. designs to be available for industrial application until the end of this decade. 1 tab. (author)

  10. USSR orders computers to improve nuclear safety

    International Nuclear Information System (INIS)

    Anon.

    1990-01-01

    Control Data Corp (CDC) has received an order valued at $32-million from the Soviet Union for six Cyber 962 mainframe computer systems to be used to increase the safety of civilian nuclear powerplants. The firm is now waiting for approval of the contract by the US government and Western Allies. The computers, ordered by the Soviet Research and Development Institute of Power Engineering (RDIPE), will analyze safety factors in the operation of nuclear reactors over a wide range of conditions. The Soviet Union's civilian nuclear program is one of the largest in the world, with over 50 plants in operation. Types of safety analyses the computers perform include: neutron-physics calculations, radiation-protection studies, stress analysis, reliability analysis of equipment and systems, ecological-impact calculations, transient analysis, and support activities for emergency response. They also include a simulator with realistic mathematical models of Soviet nuclear powerplants to improve operator training

  11. Safety standards and safety record of nuclear power plants

    International Nuclear Information System (INIS)

    Davis, A.B.

    1984-01-01

    This paper focuses on the use of standards and the measurement and enforcement of these standards to achieve safe operation of nuclear power plants. Since a discussion of the safety standards that the Nuclear Regulatory Commission (NRC) uses to regulate the nuclear power industry can be a rather tedious subject, this discussion will provide you with not only a description of what safety standards are, but some examples of their application, and various indicators that provide an overall perspective on safety. These remarks are confined to the safety standards adopted by the NRC. There are other agencies such as the Environmental Protection Agency, the Occupational Safety and Health Administration, and the state regulatory agencies which impact on a nuclear power plant. The NRC has regulatory authority for the commercial use of the nuclear materials and facilities which are defined in the Atomic Energy Act of 1954 to assure that the public health and safety and national security are protected

  12. Nuclear power plant's safety and risk

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1975-01-01

    Starting with a comprehensive safety strategy as evolved over the past years and the present legal provisions for the construction and operation of nuclear power plants, the risk of the intended operation, of accidents and unforeseen events is discussed. Owing to the excellent safety record of nuclear power plants, main emphasis in discussing accidents is given to the precautionary analysis within the framework of the licensing procedure. In this context, hypothetical accidents are mentioned only as having been utilized for general risk comparisons. The development of a comprehensive risk concept for a completely objective safety assessment of nuclear power plants remains as a final goal. (orig.) [de

  13. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEA’s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

  14. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2017-01-01

    This publication is a revision of IAEA Safety Standards Series No. NS-R-2, Safety of Nuclear Power Plants: Operation, and has been extended to cover the commissioning stage. It describes the requirements to be met to ensure the safe commissioning, operation, and transition from operation to decommissioning of nuclear power plants. Over recent years there have been developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis review and risk informed decision making processes. It became necessary to revise the IAEA’s Safety Requirements in these areas and to correct and/or improve the publication on the basis of feedback from its application by both the IAEA and its Member States. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications, initiated in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan, revealed no significant areas of weakness but resulted in a small set of amendments to strengthen the requirements and facilitate their implementation. These are contained in the present publication.

  15. Can we improve patient safety?

    Science.gov (United States)

    Corbally, Martin Thomas

    2014-01-01

    Despite greater awareness of patient safety issues especially in the operating room and the widespread implementation of surgical time out World Health Organization (WHO), errors, especially wrong site surgery, continue. Most such errors are due to lapses in communication where decision makers fail to consult or confirm operative findings but worryingly where parental concerns over the planned procedure are ignored or not followed through. The WHO Surgical Pause/Time Out aims to capture these errors and prevent them, but the combination of human error and complex hospital environments can overwhelm even robust safety structures and simple common sense. Parents are the ultimate repository of information on their child's condition and planned surgery but are traditionally excluded from the process of Surgical Pause and Time Out, perhaps to avoid additional stress. In addition, surgeons, like pilots, are subject to the phenomenon of "plan-continue-fail" with potentially disastrous outcomes. If we wish to improve patient safety during surgery and avoid wrong site errors then we must include parents in the Surgical Pause/Time Out. A recent pilot study has shown that neither staff nor parents found it added to their stress, but, moreover, 100% of parents considered that it should be a mandatory component of the Surgical Pause nor does it add to the stress of surgery. Surgeons should be required to confirm that the planned procedure is in keeping with the operative findings especially in extirpative surgery and this "step back" should be incorporated into the standard Surgical Pause. It is clear that we must improve patient safety further and these simple measures should add to that potential.

  16. Safety criteria of uranium enrichment plants

    International Nuclear Information System (INIS)

    Nardocci, A.C.; Oliveira Neto, J.M. de

    1994-01-01

    The applicability of nuclear reactor safety criteria applied to uranium enrichment plants is discussed, and a new criterion based on the soluble uranium compounds and hexafluoride chemical toxicities is presented. (L.C.J.A.). 21 refs, 4 tabs

  17. Safety in waste management plants: An Indian perspective

    International Nuclear Information System (INIS)

    Shekhar, P.; Ozarde, P.D.; Gandhi, P.M.

    2000-01-01

    Assurance of safety of public and plant workers and protection of the environment are prime objectives in the design and construction of Waste Management Plants. In India, waste management principles and strategies have been evolved in accordance with national and international regulations and standards for radiation protection. The regulations governing radiation protection have a far-reaching impact on the management of the radioactive waste. The wastes arise at each stages of the fuel cycle with varying chemical nature, generation rate and specific activity levels depending upon the type of the facility. Segregation of waste based on its chemical nature and specific activity levels is an essential feature, as its aids in selection of treatment and conditioning process. Selection of the process, equipment and materials in the plant, are governed by safety consideration alongside factors like efficiency and simplicity. The plant design considerations like physical separation, general arrangement, ventilation zoning, access control, remote handling, process piping routing, decontamination etc. have major role in realizing waste safety. Stringent quality control measures during all stages of construction have helped in achieving the design intended safety. These aspects together with operating experience gained form basis for the improved safety features in the design and construction of waste management plants. The comprehensive safety is derived from adoption of waste management strategies and appropriate plant design considerations. The paper briefly brings safety in waste management programme in India, in its current perspective. (author)

  18. Safety culture of nuclear power plant

    International Nuclear Information System (INIS)

    Zheng Beixin

    2008-01-01

    This paper is a summary on the basis of DNMC safety culture training material for managerial personnel. It intends to explain the basic contents of safety, design, management, enterprise culture, safety culture of nuclear power plant and the relationship among them. It explains especially the constituent elements of safety culture system, the basic requirements for the three levels of commitments: policy level, management level and employee level. It also makes some analyses and judgments for some typical safety culture cases, for example, transparent culture and habitual violation of procedure. (authors)

  19. Rolls-Royce I and C long term support a dedicated solution to manage plant condition, optimize operating life and maximize plant value while improving safety

    International Nuclear Information System (INIS)

    Baillonmartos, F.

    2010-01-01

    Rolls-Royce supplied safety I and C equipment as Original Equipment Manufacturer (OEM) for 200 reactors in operation from 1970', including various technological steps (analog to digital hardware). The hardware scope covers various systems, either more than 500 hardware references, either more than 150000 parts in operation. Rolls-Royce contributes to life time extension ensuring system availability and reliability on long term basis by determining the optimum solution between different maintenance approaches: maintenance and repair with limited retrofit or major system retrofit. Based on Long Term Support (LTS) agreement with customers, Rolls-Royce commits to maintain the capability to manufacture, modify, repair, test, at board, rack and systems level over a long time period. This means finding solution to hardware ageing, technology evolution, skills and tools maintenance. This paper describes how Rolls-Royce has built a global process based on delivering a package of services and developing long term support engineering solutions and tools. Our main experience on EDF partnership will serve to illustrate Rolls-Royce I and C Long Term Support activity. (authors)

  20. IAEA Completes Safety Review at Czech Nuclear Power Plant

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: An international team of nuclear safety experts, led by the International Atomic Energy Agency (IAEA), today completed a review of safety practices at Temelin Nuclear Power Station in the Czech Republic. The team highlighted the Power Plant's good practices and also recommended improvements to some safety measures. At the request of the Government of the Czech Republic, the IAEA assembled a team of nuclear installation safety experts to send an Operational Safety Review Team (OSART) to the Power Plant, and the mission was conducted from 5 to 22 November 2012. The team was comprised of experts from Brazil, Hungary, Slovakia, South Africa, Sweden, Ukraine and the United Kingdom. An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the Plant's overall safety status. The team at Temelin conducted an in-depth review of the functions essential to the safe operation of the Power Plant, which are under the responsibility of the site's management. The review covered the areas of management, organization and administration; operations; maintenance; technical support; operating experience; radiation protection; chemistry; and severe accident management. The conclusions of the review are based on the IAEA's Safety Standards and proven good international practices. The OSART team has identified good plant practices, which will be shared with the rest of the nuclear industry for consideration of potential application elsewhere. Examples include the following: - The Power Plant has adopted effective computer software to improve the efficiency of the plant to prepare and isolate equipment for maintenance; - The Power Plant undertakes measures to control precisely the chemical parameters that limit corrosion in the reactor's coolant system, which in turn reduce radiation exposure to the workforce; and - The Temelin

  1. Safety aspects and operating experience of LWR plants in Japan

    International Nuclear Information System (INIS)

    Aoki, S.; Hinoki, M.

    1977-01-01

    From the outset of nuclear power development in Japan, major emphasis has been placed on the safety of the nuclear power plants. There are now twelve nuclear power plants in operation with a total output of 6600 MWe. Their operating records were generally satisfactory, but in the 1974 to 1975 period, they experienced somewhat declined availability due to the repair work under the specific circumstances. After investigation of causes of troubles and the countermeasures thereof were made to ensure safety, they are now keeping good performance. In Japan, nuclear power plants are strictly subject to sufficient and careful inspection in compliance with the safety regulation, and are placed under stringent radiation control of employees. Under the various circumstances, however, the period of annual inspection tends to be prolonged more than originally planned, and this consequently is considered to be one of the causes of reduced availability. In order to develop nuclear power generation for the future, it is necessary to put further emphasis on the assurance of safety and to endeavor to devise measures to improve availability of the plants, based on the careful analysis of causes which reduce plant availability. This paper discusses the results of studies made for the following items from such viewpoints: (1) Safety and Operating Experience of LWR Nuclear Power Plants in Japan; a) Operating experience with light water reactors b) Improvements in design of light water reactors during the past ten years c) Analysis of the factors which affect plant availability; 2) Assurance of Safety and Measures to Increase Availability a) Measures for safety and environmental protection b) Measures to reduce radiation exposure of employees c) Appropriateness of maintenance and inspection work d) Measures to increase plant availability e) Measures to improve reliability of equipments and components; and 3) Future Technical Problems

  2. Barsebaeck power plant - safety and emergency measures

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    A Swedish-Danish Committee on safety at the Swedish nuclear power plant Barsebaeck was established in 1979 in order to evaluate the nuclear safety at Barsebaeck with a view to the reactor accident at the Three-Mile-Island nuclear power plant March 28, 1979. According to the committees mandate the investigations of the Kemeny Commission, the Rogouin investigation, investigations of the American Nuclear Regulatory Commission, and the Swedish report ''Safe nuclear power'' have been taken into consideration by the Committee. Furthermore, it has formed the basis for the Committees work that the authority responsibility for the safety at Barsebaeck lies with the Swedish authorities, and that these authorities have evaluated the safety aspects before the permissions for operation of the Barsebaeck power plant were given and hereafter currently in connection with the inspection of the power plant. The report prepared by the Commission treats aspects as: a) Nuclear safety at the Barsebaeck power plant, b) reactor safety and emergency provisions, c) common elements in the emergency provision situation in Sweden and Denmark, d) ongoing investigations on course of events during accidents and release limiting safety systems. (BP)

  3. Seismic safety of nuclear power plants

    International Nuclear Information System (INIS)

    Guerpinar, A.; Godoy, A.

    2001-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in Eastern Europe and in particular WWER type nuclear power plants during the past seven years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on 'Benchmark study for the seismic analysis and testing of WWER type nuclear power plants'. These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  4. The safety of nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    1988-01-01

    Do nuclear power plants present an unjustifiable risk Can there be confidence in their safety The Uranium Institute invited a group of senior safety experts from eight different Western countries operating different types of reactors to provide an authoritative explanation for non-specialists of the basic principles of reactor safety, their application and their implications. The report presents the group's opinion on the level of safety achieved in the Western nuclear power plants with which the authors are directly familiar. Although many of the points made may well also be true for non-Western reactors, the report does not cover them except where specifically stated. It does describe and discuss the causes of the Chernobyl disaster. It does not compare nuclear power with other fuels, nor does it deal with its benefits, since however great the benefits from the peaceful use of nuclear power, and its own advantages over other fuels, they could not compensate for lack of safety. The conclusion reached is that the risk associated with electricity production at nuclear power plants can be kept very low. Proper use of the extensive knowledge available today can guarantee operation of nuclear power plants at very high safety levels, carrying very low risks, both to health and of contamination of the environment: risks that are continually lowered by upgrading existing plants and their operation, and by the design of future power plants. (author).

  5. Safety targets for nuclear power plants

    International Nuclear Information System (INIS)

    Herttrich, P.M.

    1985-01-01

    By taking as an example the safety targets of the American nuclear energy authority US-NRC, this paper explains what is meant by global, quantitative safety targets for nuclear power plants and what expectations are associated with the selecton of such safety targets. It is shown how probabilistic methods can be an appropriate completion of proven deterministic methods and what are the sectors where their application may become important in future. (orig./HP) [de

  6. Transportation Safety Excellence in Operations Through Improved Transportation Safety Document

    International Nuclear Information System (INIS)

    Dr. Michael A. Lehto; MAL

    2007-01-01

    A recent accomplishment of the Idaho National Laboratory (INL) Materials and Fuels Complex (MFC) Nuclear Safety analysis group was to obtain DOE-ID approval for the inter-facility transfer of greater-than-Hazard-Category-3 quantity radioactive/fissionable waste in Department of Transportation (DOT) Type A drums at MFC. This accomplishment supported excellence in operations through safety analysis by better integrating nuclear safety requirements with waste requirements in the Transportation Safety Document (TSD); reducing container and transport costs; and making facility operations more efficient. The MFC TSD governs and controls the inter-facility transfer of greater-than-Hazard-Category-3 radioactive and/or fissionable materials in non-DOT approved containers. Previously, the TSD did not include the capability to transfer payloads of greater-than-Hazard-Category-3 radioactive and/or fissionable materials using DOT Type A drums. Previous practice was to package the waste materials to less-than-Hazard-Category-3 quantities when loading DOT Type A drums for transfer out of facilities to reduce facility waste accumulations. This practice allowed operations to proceed, but resulted in drums being loaded to less than the Waste Isolation Pilot Plant (WIPP) waste acceptance criteria (WAC) waste limits, which was not cost effective or operations friendly. An improved and revised safety analysis was used to gain DOE-ID approval for adding this container configuration to the MFC TSD safety basis. In the process of obtaining approval of the revised safety basis, safety analysis practices were used effectively to directly support excellence in operations. Several factors contributed to the success of MFC's effort to obtain approval for the use of DOT Type A drums, including two practices that could help in future safety basis changes at other facilities. (1) The process of incorporating the DOT Type A drums into the TSD at MFC helped to better integrate nuclear safety

  7. Experience of safety and performance improvement for fuel handling equipment

    International Nuclear Information System (INIS)

    Gyoon Chang, Sang; Hee Lee, Dae

    2014-01-01

    The purpose of this study is to provide experience of safety and performance improvement of fuel handling equipment for nuclear power plants in Korea. The fuel handling equipment, which is used as an important part of critical processes during the refueling outage, has been improved to enhance safety and to optimize fuel handling procedures. Results of data measured during the fuel reloading are incorporated into design changes. The safety and performance improvement for fuel handling equipment could be achieved by simply modifying the components and improving the interlock system. The experience provided in this study can be useful lessons for further improvement of the fuel handling equipment. (authors)

  8. Evolution of the future plants operation for a better safety

    International Nuclear Information System (INIS)

    Papin, B.; Malvache, P.

    1994-01-01

    This paper describes a coordinated research project of the french CEA, addressing to the evolutions in plant operation apt to bring perceptible and assessable improvement in the operational safety. This program has been scheduled for the 1992-1996 period, with a global 40 men/year effort. The present status of the two main parts of the project is presented: ESCRIME (program aiming at defining the optimal share of tasks between humans and computers in plant operation), IMAGIN (research in the domain of plant information management, in order to ensure the global coherence of the image of the plant, used by the different actors in plant operation). (authors). 3 refs., 4 figs

  9. Construct ability Improvement for Nuclear Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dae Soo; Lee, Jong Rim; Kim, Jong Ku [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    The purpose of this study was to identify methods for improving the construct ability of nuclear power plants. This study reviewed several references of current construction practices of domestic and overseas nuclear plants in order to identify potential methods for improving construct ability. The identified methods for improving construct ability were then evaluated based on the applicability to domestic nuclear plant construction. The selected methods are expected to reduce the construction period, improve the quality of construction, cost, safety, and productivity. Selection of which methods should be implemented will require further evaluation of construction modifications, design changes, contract revisions. Among construction methods studied, platform construction methods can be applied through construction sequence modification without significant design changes, and Over the Top construction method of the NSSS, automatic welding of RCL pipes, CLP modularization, etc., are considered to be applied after design modification and adjustment of material lead time. (author). 49 refs., figs., tabs.

  10. Licensee responsibility for nuclear power plant safety

    International Nuclear Information System (INIS)

    Schneider, Horst

    2010-01-01

    Simple sentences easy to grasp are desirable in regulations and bans. However, in a legal system, their meaning must be unambiguous. Article 6, Paragraph 1 of the EURATOM Directive on a community framework for the nuclear safety of nuclear facilities of June 2009 states that 'responsibility for the nuclear safety of a nuclear facility is incumbent primarily on the licensee.' The draft 'Safety Criteria for Nuclear Power Plants, Revision D, April 2009' of the German Federal Ministry for the Environment, Nature Conservation, and Nuclear Safety (BMU) (A Module 1, 'Safety Criteria for Nuclear Power Plants: Basic Safety Criteria' / '0 Principles' Paragraph 2) reads: 'Responsibility for ensuring safety rests with the licensee. He shall give priority to compliance with the safety goal over the achievement of other operational objectives.' In addition, the existing rules and regulations, whose rank is equivalent to that of international regulations, assign priority to the safety goal to be pursued by the licensee over all other objectives of the company. The operator's responsibility for nuclear safety can be required and achieved only on the basis of permits granted, which must meet legal requirements. The operator's proximity to plant operation is the reason for his 'primary responsibility.' Consequently, verbatim incorporation of Article 6, Paragraph 1 of the EURATOM Directive would only be a superscript added to existing obligations of the operator - inclusive of a safety culture designed as an incentive to further 'the spirit of safety-related actions' - without any new legal contents and consequences. In the reasons of the regulation, this would have to be clarified in addition to the cryptic wording of 'responsibility.. primarily,' at the same time expressing that operators and authorities work together in a spirit of openness and trust. (orig.)

  11. Problems of nuclear power plant safety evaluation

    International Nuclear Information System (INIS)

    Suchomel, J.

    1977-01-01

    Nuclear power plant safety is discussed with regard to external effects on the containment and to the human factor. As for external effects, attention is focused on shock waves which may be due to explosions or accidents in flammable material transport and storage, to missiles, and to earthquake effects. The criteria for evaluating nuclear power plant safety in different countries are shown. Factors are discussed affecting the reliability of man with regard to his behaviour in a loss-of-coolant accident in the power plant. Different types of PWR containments and their functions are analyzed, mainly in case of accident. Views are discussed on the role of destructive accidents in the overall evaluation of fast reactor safety. Experiences are summed up gained with the operation of WWER reactors with respect to the environmental impact of the nuclear power plants. (Z.M.)

  12. Safety culture in nuclear power plants. Proceedings

    International Nuclear Information System (INIS)

    1994-12-01

    As a consequence of the INSAG-4 report on 'safety culture', published by the IAEA in 1991, the Federal Commission for the Safety of Nuclear Power Plants (KSA) decided to hold a one-day seminar as a first step in this field. The KSA is an advisory body of the Federal Government and the Federal Department of Transport and Energy (EVED). It comments on applications for licenses, observes the operation of nuclear power plants, assists with the preparation of regulations, monitors the progress of research in the field of nuclear safety, and makes proposals for research tasks. The objective of this seminar was to familiarise the participants with the principles of 'safety culture', with the experiences made in Switzerland and abroad with existing concepts, as well as to eliminate existing prejudices. The main points dealt with at this seminar were: - safety culture from the point of view of operators, - safety culture from the point of view of the authorities, - safety culture: collaboration between power plants, the authorities and research organisations, - trends and developments in the field of safety culture. Invitations to attend this seminar were extended to the management boards of companies operating Swiss nuclear power plants, and to representatives of the Swiss authorities responsible for the safety of nuclear power plants. All these organisations were represented by a large number of executive and specialist staff. We would like to express our sincerest thanks to the Head of the Federal Department of Transport and Energy for his kind patronage of this seminar. (author) figs., tabs., refs

  13. Safety of uranium enrichment plant

    International Nuclear Information System (INIS)

    Yonekawa, Shigeru; Morikami, Yoshio; Morita, Minoru; Takahashi, Tsukasa; Tokuyasu, Takashi.

    1991-01-01

    With respect to safety evaluation of the gas centrifuge enrichment facility, several characteristic problems are described as follows. Criticality safety in the cascade equipments can be obtained to maintain the enrichment of UF 6 below 5 %. External radiation dose equivalent rate of the 30B cylinder is low enough, the shield is not necessary. Penetration ratio of the two-stage HEPA filters for UF 6 aerosol is estimated at 10 -9 . From the experimental investigation, vacuum tightness is not damaged by destruction of gas centrifuge rotor. Carbon steel can be used for uranium enrichment equipments under the condition below 100degC. (author)

  14. Improving nuclear safety of VVER-440 units

    International Nuclear Information System (INIS)

    Nochev, T.; Sabinov, S.

    2001-01-01

    In this paper authors deals with improvement of nuclear safety of WWER-440 units in Kozloduy NPP. Main directions for improving nuclear safety of WWER-440 units were: - to expand number of the design accident; - to increase reliability of equipment important for the safety; - to decrease the probability of initiating events; - improvements the integrity of the primary circuit (application LBB concept, qualification of the pressure safety valves to avoid pressurized thermal shock); - improvement of the fire protection; - improvement of the operation including upgrading and improvement of operational documents, implementation of new system for training the operators and etc.; - reassessment of the seismic response of the plant. Main actions were made at NPP Kozloduy to increase nuclear safety of VVER-440 units. 1. Modernization of Emergency High Pressure Safety Injection System. The modernization includes dividing of independent channels with reservation of active elements. Pumps were exchanged with more effective and reliable ones. HPSIS was increased reliability in general through decrease number of active elements and exchanged with passive. 2. For the purpose of avoiding fast cooling at the primary circuit and obtaining thermal shock of reactor vessel, Main Safety Insulation Valves are installed at NPP Kozloduy. 3. Modernization of Emergency power supplies AC. Oil breakers VMP-10 are exchanged with gas FS-4. 4. Generator breakers are installed to decrease probability of loss power supply and blackout. They provide reliable power supply to the system important for the safety in case of failure on generator. 5. I and C system has been qualified and optimized. 6. Reassessments of Limiting Conditions of Operation and new scram signals have been introduced. 7. An operators-oriented Informational System has been developed. It includes ensuring and updating of equipment data, new informational support of operator and etc. 8. A new auxiliary independent system for

  15. Operational safety of nuclear power plants

    International Nuclear Information System (INIS)

    Tanguy, P.

    1987-01-01

    The operational safety of nuclear power plants has become an important safety issue since the Chernobyl accident. A description is given of the various aspects of operational safety, including the importance of human factors, responsibility, the role and training of the operator, the operator-machine interface, commissioning and operating procedures, experience feedback, and maintenance. The lessons to be learnt from Chernobyl are considered with respect to operator errors and the management of severe accidents. Training of personnel, operating experience feedback, actions to be taken in case of severe accidents, and international cooperation in the field of operational safety, are also discussed. (U.K.)

  16. An engineer-constructor's view of nuclear power plant safety

    International Nuclear Information System (INIS)

    Landis, J.W.; Jacobs, S.B.

    1984-01-01

    At SWEC we have been involved in the development of safety features of nuclear power plants ever since we served as the engineer-constructur for the first commerical nuclear power station at Shippingport, Pennsylvania, in the 1950s. Our personnel have pioneered a number of safety innovations and improvements. Among these innovations is the subatmospheric containment for pressurized water reactor (PWR) power plants. This type of containment is designed so that leakage will terminate within 1 to 2 hours of the worst postulated loss of coolant accident. Other notable contributions include first use of reinforced-concrete atmospheric containments for PWR power plants and of reinforced-concrete, vapor-suppression containments for boiling water reactor (BWR) power plants. Both concepts meet rigorous U.S. safety requirements. SWEC has performed a substantial amount of work on developing standardized plant designs and has developed standardized engineering and construction techniques and procedures. Standardization concepts are being developed in Canada, France, USSR, and Germany, as well as in the United States. The West German convoy concept, which involves developing a number of standardized plants in a common effort, has been quite successful. We believe standardization contributes to safety in a number of ways. Use of standardized designs, procedures, techniques, equipment, and methods increases efficiency and results in higher quality. Standardization also reduces the design variations with which plant operators, emergency teams, and regulatory personnel must be familiar, thus increasing operator capability, and permits specialized talents to be focused on important safety considerations. (orig./RW)

  17. The critical safety functions and plant operation

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Porter, N.J.; Cross, M.T.; Guinn, W.M.

    1981-01-01

    The paper outlines the operator's role in nuclear safety and introduces the concept of ''safety functions''. Safety functions are a group of actions that prevent core melt or minimize radiation releases to the general public. They can be used to provide a hierarchy of practical plant protection that an operator should use. ''An accident identical to that at Three Mile Island is not going to happen again'', said the Rogovin investigators. The concepts put forward in this paper are intended to help the operator avoid serious consequence from the next unexpected threat. On the basis of the safety evaluation, the operator has three roles in assuring that the consequences of an event will be no worse than the predicted acceptable results. These three operator roles are: first, maintain plant setup in readiness to properly respond; second, operate the plant in a manner such that fewer, milder events minimize the frequency and the severity of adverse events; third, the operator needs to monitor the plant to verify that the safety functions are accomplished. The operator needs a systematic approach to mitigating the consequences of an event. The concept of ''safety function'' introduces that systematic approach and prevents a hierarchy of protection. If the operator has difficulty in identifying an event for any reason, the systematic safety function approach allows ones to accomplish the overall path of mitigating consequences. There are ten identified functions designed to protect against core melt, preserve containment integrity, prevent indirect release of radioactivity, and maintain vital auxiliaries needed to support the other safety functions. The paper describes in detail the operator's role and the safety functions, and provides many examples of the use of alternative success paths to accomplish the safety function

  18. Using game theory to improve safety within chemical industrial parks

    CERN Document Server

    Reniers, Genserik

    2013-01-01

    Though the game-theoretic approach has been vastly studied and utilized in relation to economics of industrial organizations, it has hardly been used to tackle safety management in multi-plant chemical industrial settings. Using Game Theory for Improving Safety within Chemical Industrial Parks presents an in-depth discussion of game-theoretic modelling which may be applied to improve cross-company prevention and -safety management in a chemical industrial park.   By systematically analyzing game-theoretic models and approaches in relation to managing safety in chemical industrial parks, Using Game Theory for Improving Safety within Chemical Industrial Parks explores the ways game theory can predict the outcome of complex strategic investment decision making processes involving several adjacent chemical plants. A number of game-theoretic decision models are discussed to provide strategic tools for decision-making situations.   Offering clear and straightforward explanations of methodologies, Using Game Theor...

  19. The role of probabilistic safety assessment and probabilistic safety criteria in nuclear power plant safety

    International Nuclear Information System (INIS)

    1992-01-01

    The purpose of this Safety Report is to provide guidelines on the role of probabilistic safety assessment (PSA) and a range of associated reference points, collectively referred to as probabilistic safety criteria (PSC), in nuclear safety. The application of this Safety Report and the supporting Safety Practice publication should help to ensure that PSA methodology is used appropriately to assess and enhance the safety of nuclear power plants. The guidelines are intended for use by nuclear power plant designers, operators and regulators. While these guidelines have been prepared with nuclear power plants in mind, the principles involved have wide application to other nuclear and non-nuclear facilities. In Section 2 of this Safety Report guidelines are established on the role PSA can play as part of an overall safety assurance programme. Section 3 summarizes guidelines for the conduct of PSAs, and in Section 4 a PSC framework is recommended and guidance is provided for the establishment of PSC values

  20. Reviewing industrial safety in nuclear power plants

    International Nuclear Information System (INIS)

    1990-02-01

    This document contains guidance and reference materials for Operational Safety Review Team (OSART) experts, in addition to the OSART Guidelines (TECDOC-449), for use in the review of industrial safety activities at nuclear power plants. It sets out objectives for an excellent industrial safety programme, and suggests investigations which should be made in evaluating industrial safety programmes. The attributes of an excellent industrial safety programme are listed as examples for comparison. Practical hints for reviewing industrial safety are discussed, so that the necessary information can be obtained effectively through a review of documents and records, discussions with counterparts, and field observations. There are several annexes. These deal with major features of industrial safety programmes such as safety committees, reporting and investigation systems and first aid and medical facilities. They include some examples which are considered commendable. The document should be taken into account not only when reviewing management, organization and administration but also in the review of related areas, such as maintenance and operations, so that all aspects of industrial safety in an operating nuclear power plant are covered

  1. Programs to improve plant performance

    International Nuclear Information System (INIS)

    Felmus, N.L.

    1987-01-01

    Looking toward the 1990's, we see a period in which our industry will face the challenge of improving the performance of the nuclear plants which are built and operating. The skills and technology are at hand to make good plant performance a reality and we believe the time has come to use them to achieve that end. As reserve margins decline, utilities and their regulators will increasingly seek to tap the unexploited capacity tied up in plants operating below their optimum availability. This paper describes a number of the programs, plant improvements and operations improvements which can yield a significant increase in nuclear plant availability and capacity factor now and into the 1990's. (author)

  2. Safety goals for commercial nuclear power plants

    International Nuclear Information System (INIS)

    Roe, J.W.

    1988-01-01

    In its official policy statement on safety goals for the operation of nuclear power plants, the Nuclear Regulatory Commission (NRC) set two qualitative goals, supported by two quantitative objectives. These goals are that (1) individual members of the public should be provided a level of protection from the consequences of nuclear power plant operation such that individuals bear no significant additional risk to life and health; and (2) societal risks to life and health from nuclear power plant operation should be comparable to or less than the risks of generating electricity by viable competing technologies and should not be a significant addition to other societal risks. As an alternative, this study proposes four quantitative safety goals for nuclear power plants. It begins with an analysis of the NRC's safety-goal development process, a key portion of which was devoted to delineating criteria for evaluating goal-development methods. Based on this analysis, recommendations for revision of the NRC's basic benchmarks for goal development are proposed. Using the revised criteria, NRC safety goals are evaluated, and the alternative safety goals are proposed. To further support these recommendations, both the NRC's goals and the proposed goals are compared with the results of three major probabilistic risk assessment studies. Finally, the potential impact of these recommendations on nuclear safety is described

  3. Safety aspects of nuclear power plant ageing

    International Nuclear Information System (INIS)

    1990-01-01

    The nuclear community is facing new challenges as commercial nuclear power plants (NPPs) of the first generation get older. At present, some of the plants are approaching or have even exceeded the end of their nominal design life. Experience with fossil fired power plants and in other industries shows that reliability of NPP components, and consequently general plant safety and reliability, may decline in the middle and later years of plant life. Thus, the task of maintaining operational safety and reliability during the entire plant life and especially, in its later years, is of growing importance. Recognizing the potential impact of ageing on plant safety, the IAEA convened a Working Group in 1985 to draft a report to stimulate relevant activities in the Member States. This report provided the basis for the preparation of the present document, which included a review in 1986 by a Technical Committee and the incorporation of relevant results presented at the 1987 IAEA Symposium on the Safety Aspects of the Ageing and Maintenance of NPPs and in available literature. The purpose of the present document is to increase awareness and understanding of the potential impact of ageing on plant safety; of ageing processes; and of the approach and actions needed to manage the ageing of NPP components effectively. Despite the continuing growth in knowledge on the subject during the preparation of this report it nevertheless contains much that will be of interest to a wide technical and managerial audience. Furthermore, more specific technical publications on the evaluation and management of NPP ageing and service life are being developed under the Agency's programme, which is based on the recommendations of its 1988 Advisory Group on NPP ageing. Refs, figs and tabs

  4. Safety upgrading at PAKS Nuclear Power Plant

    International Nuclear Information System (INIS)

    Bajsz, J.; Elter, J.

    2000-01-01

    The operation of Paks NPP has reached its half time. Until this time the plant fulfilled expectations raised before its construction: the four units have produced safely and reliably more than 200 TWh electricity. The production of the plant has been at the stable level since its construction and has provided 43-38 % of electricity consumed in Hungary. The annual production is around 14 TWh, which means a load factor higher than 85 %. Safety upgrading activities [1] at Paks had started in the late eighties, when the commissioning work of units 3 and 4 were carried out. That time the main emphases were put to lessons learned of the TMI and Chernobyl accidents. The international reviews hosted by our plant widened our review's scope. To systematize our approach a complete safety review, the AGNES (Advanced General Safety and New Evaluation of Safety) project was started in 1991. The goal of the project was to evaluate to what extent Paks NPP satisfied the current international safety expectations and to help in determining the priorities for safety enhancement and upgrading measures. The project completed in 1994 ranked our safety upgrading measures by safety significance, which became a basis for technical design work and financial scheduling. The other important outcome of the AGNES project was the introduction the Periodical Safety Review regime by our nuclear authority. These periodical reviews held after 10 years of operation offer the possibility - and obligation for the licensee - to perform a comprehensive assessment of the safety of the plant, to evaluate the integral effects of changes of circumstances happened during the review period. The goal of these reviews is to deal with cumulative effects of NPP ageing, modifications, operating experience and technical developments aimed at ensuring a high level of safety throughout plant service life. The execution of our safety-upgrading program is well advancing. For the whole program from 1996 to 2002 250

  5. Leadership Actions to Improve Nuclear Safety Culture

    International Nuclear Information System (INIS)

    Clewett, L.K.

    2016-01-01

    The challenge many leaders face is how to effectively implement and then utilise the results of Safety Culture surveys. Bruce Power has recently successfully implemented changes to the Safety Culture survey process including how corrective actions were identified and implemented. The actions taken in response to the latest survey have proven effective with step change performance noted. Nuclear Safety is a core value for Bruce Power. Nuclear Safety at Bruce Power is based on the following four pillars: reactor safety, industrial safety, radiological safety and environmental safety. Processes and practices are in place to achieve a healthy Nuclear Safety Culture within Bruce Power such that nuclear safety is the overriding priority. This governance is based on industry leading practices which monitor, asses and take action to drive continual improvements in the Nuclear Safety Culture within Bruce Power.

  6. Researchers' Roles in Patient Safety Improvement.

    Science.gov (United States)

    Pietikäinen, Elina; Reiman, Teemu; Heikkilä, Jouko; Macchi, Luigi

    2016-03-01

    In this article, we explore how researchers can contribute to patient safety improvement. We aim to expand the instrumental role researchers have often occupied in relation to patient safety improvement. We reflect on our own improvement model and experiences as patient safety researchers in an ongoing Finnish multi-actor innovation project through self-reflective narration. Our own patient safety improvement model can be described as systemic. Based on the purpose of the innovation project, our improvement model, and the improvement models of the other actors in the project, we have carried out a wide range of activities. Our activities can be summarized in 8 overlapping patient safety improvement roles: modeler, influencer, supplier, producer, ideator, reflector, facilitator, and negotiator. When working side by side with "practice," researchers are offered and engage in several different activities. The way researchers contribute to patient safety improvement and balance between different roles depends on the purpose of the study, as well as on the underlying patient safety improvement models. Different patient safety research paradigms seem to emphasize different improvement roles, and thus, they also face different challenges. Open reflection on the underlying improvement models and roles can help researchers with different backgrounds-as well as other actors involved in patient safety improvement-in structuring their work and collaborating productively.

  7. Nuclear power plant safety in Brazil

    International Nuclear Information System (INIS)

    Lederman, L.

    1980-01-01

    The Code of Practice for the Safe Operation of Nuclear Power Plants states that: 'In discharging its responsibility for public health and safety, the government should ensure that the operational safety of a nuclear reactor is subject to surveillance by a regulatory body independent of the operating organization'. In Brazil this task is being carried out by the Comissao Nacional de Energia Nuclear in accordance with the best international practice. (orig./RW)

  8. Improved core monitoring for improved plant operations

    International Nuclear Information System (INIS)

    Mueller, N.P.

    1987-01-01

    Westinghouse has recently installed a core on-line surveillance, monitoring and operations systems (COSMOS), which uses only currently available core and plant data to accurately reconstruct the core average axial and radial power distributions. This information is provided to the operator in an immediately usable, human-engineered format and is accumulated for use in application programs that provide improved core performance predictive tools and a data base for improved fuel management. Dynamic on-line real-time axial and radial core monitoring supports a variety of plant operations to provide a favorable cost/benefit ratio for such a system. Benefits include: (1) relaxation or elimination of certain technical specifications to reduce surveillance and reporting requirements and allow higher availability factors, (2) improved information displays, predictive tools, and control strategies to support more efficient core control and reduce effluent production, and (3) expanded burnup data base for improved fuel management. Such systems can be backfit into operating plants without changing the existing instrumentation and control system and can frequently be implemented on existing plant computer capacity

  9. Safety prediction technique for nuclear power plants

    International Nuclear Information System (INIS)

    Henry, C.D. III; Anderson, R.T.

    1985-01-01

    This paper presents a safety prediction technique (SPT) developed by Reliability Technology Associates (RTA) for nuclear power plants. It is based on a technique applied by RTA to assess the flight safety of US Air Force aircraft. The purpose of SPT is to provide a computerized technique for objective measurement of the effect on nuclear plant safety of component failure or procedural, software, or human error. A quantification is determined, called criticality, which is proportional to the probability that a given component or procedural-human action will cause the plant to operate in a hazardous mode. A hazardous mode is characterized by the fact that there has been a failure/error and the plant, its operating crew, and the public are exposed to danger. Whether the event results in an accident, an incident, or merely the exposure to danger is dependent on the skill and reaction of the operating crew as well as external influences. There are three major uses of SPT: (a) to predict unsafe situations so that corrective action can be taken before accidents occur, (b) to quantify the impact of equipment malfunction or procedural, software, or human error on safety and thereby establish priorities for proposed modifications, and (c) to provide a means of evaluating proposed changes for their impact on safety prior to implementation and to provide a method of tracking implemented changes

  10. Human performance improvement for nuclear power plants

    International Nuclear Information System (INIS)

    2005-01-01

    The IAEA assists NPP operating organizations to improve plant performance through a focus on human performance improvement in areas like organizational and leadership development, senior management decision making, organization and management of HPI programmes including tools needed for effective HPI implementation, safety culture enhancement, knowledge management, personnel selection and staffing, career development, training and development, work design, scheduling and conditions, procedure and other job-aid development and use, effective communications, human performance monitoring, motivation. Many NPP operating organizations in Member States, are not yet achieving the full potential of their NPP technology/equipment regarding safety, operational or economic performance due to human performance weaknesses. The IAEA's HPI (Human Performance Improvement) services provide a means for these organizations to efficiently and effectively learn from international experts and the experiences of others in improving plant performance through human performance improvements. NPP operating organizations can benefit from these services in a number of ways, including requesting a national project, participating in a regional project, or requesting an assist visit. The types of activities provided through these services include assistance in benchmarking practices of successful organizations, providing information exchange and reviews of current practices through assist missions, conducting workshops on focused human performance topics, evaluating current human performance methods, including assistance in implementing self assessment programmes and providing support to safety culture enhancement programmes based on self-assessment

  11. Improving safety culture through the health and safety organization: a case study.

    Science.gov (United States)

    Nielsen, Kent J

    2014-02-01

    International research indicates that internal health and safety organizations (HSO) and health and safety committees (HSC) do not have the intended impact on companies' safety performance. The aim of this case study at an industrial plant was to test whether the HSO can improve company safety culture by creating more and better safety-related interactions both within the HSO and between HSO members and the shop-floor. A quasi-experimental single case study design based on action research with both quantitative and qualitative measures was used. Based on baseline mapping of safety culture and the efficiency of the HSO three developmental processes were started aimed at the HSC, the whole HSO, and the safety representatives, respectively. Results at follow-up indicated a marked improvement in HSO performance, interaction patterns concerning safety, safety culture indicators, and a changed trend in injury rates. These improvements are interpreted as cultural change because an organizational double-loop learning process leading to modification of the basic assumptions could be identified. The study provides evidence that the HSO can improve company safety culture by focusing on safety-related interactions. © 2013. Published by Elsevier Ltd and National Safety Council.

  12. Ways of improving safety for future PWRs in France

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Manesse, D.; Mattei, J.M.

    1994-06-01

    Results of thinkings and studies, conducted within the Institute for Nuclear Safety and Protection (IPSN) on various fields of nuclear power plant safety, on the definition of safety objectives and principles for future PWRs. The aim of the studies is to identify ways of improving the design of future plants in France and Germany, with the main following objectives: significant reduction of the global probability of core damage, significant reduction of radioactive releases, mainly for severe accident conditions, and reduction of individual and collective doses received by workers. (R.P.) 3 refs., 1 tab

  13. Augmented reality for improved safety

    CERN Multimedia

    Stefania Pandolfi

    2016-01-01

    Sometimes, CERN experts have to operate in low visibility conditions or in the presence of possible hazards. Minimising the duration of the operation and reducing the risk of errors is therefore crucial to ensuring the safety of personnel. The EDUSAFE project integrates different technologies to create a wearable personnel safety system based on augmented reality.    The EDUSAFE integrated safety system uses a camera mounted on the helmet to monitor the working area.  In its everyday operation of machines and facilities, CERN adopts a whole set of measures and safety equipment to ensure the safety of its personnel, including personal wearable safety devices and access control systems. However, sometimes, scheduled and emergency maintenance work needs to be done in zones with potential cryogenic hazards, in the presence of radioactive equipment or simply in demanding conditions where visibility is low and moving around is difficult. The EDUSAFE Marie Curie Innovative&...

  14. Nuclear power plant with a safety enclosure

    International Nuclear Information System (INIS)

    Keller, W.; Krueger, J.; Ropers, J.; Schabert, H.P.

    1976-01-01

    A nuclear power plant has a safety enclosure for a nuclear reactor. A fuel element storage basin is also located in this safety enclosure and a fuel element lock extends through the enclosure, with a cross-sectional size proportioned for the endwise passage of fuel elements, the lock including internal and external valves so that a fuel element may be locked endwise safely through the lock. The lock, including its valves, being of small size, does not materially affect the pressure resistance of the safety enclosure, and it is more easily operated than a lock large enough to pass people and fuel element transport vessels

  15. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  16. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  17. Software for computer based systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2000-01-01

    Computer based systems are of increasing importance to safety in nuclear power plants as their use in both new and older plants is rapidly increasing. They are used both in safety related applications, such as some functions of the process control and monitoring systems, as well as in safety critical applications, such as reactor protection or actuation of safety features. The dependability of computer based systems important to safety is therefore of prime interest and should be ensured. With current technology, it is possible in principle to develop computer based instrumentation and control systems for systems important to safety that have the potential for improving the level of safety and reliability with sufficient dependability. However, their dependability can be predicted and demonstrated only if a systematic, fully documented and reviewable engineering process is followed. Although a number of national and international standards dealing with quality assurance for computer based systems important to safety have been or are being prepared, internationally agreed criteria for demonstrating the safety of such systems are not generally available. It is recognized that there may be other ways of providing the necessary safety demonstration than those recommended here. The basic requirements for the design of safety systems for nuclear power plants are provided in the Requirements for Design issued in the IAEA Safety Standards Series.The IAEA has issued a Technical Report to assist Member States in ensuring that computer based systems important to safety in nuclear power plants are safe and properly licensed. The report provides information on current software engineering practices and, together with relevant standards, forms a technical basis for this Safety Guide. The objective of this Safety Guide is to provide guidance on the collection of evidence and preparation of documentation to be used in the safety demonstration for the software for computer based

  18. Safety related terms for advanced nuclear plants

    International Nuclear Information System (INIS)

    1995-12-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

  19. Safety related terms for advanced nuclear plants

    International Nuclear Information System (INIS)

    1991-09-01

    The terms considered in this document are in widespread current use without a universal consensus as to their meaning. Other safety related terms are already defined in national or international codes and standards as well as in IAEA's Nuclear Safety Standards Series. Most of the terms in those codes and standards have been defined and used for regulatory purposes, generally for application to present reactor designs. There is no intention to duplicate the description of such regulatory terms here, but only to clarify the terms used for advanced nuclear plants. The following terms are described in this paper: Inherent safety characteristics, passive component, active component, passive systems, active system, fail-safe, grace period, foolproof, fault-/error-tolerant, simplified safety system, transparent safety

  20. Plutonium Finishing Plant safety evaluation report

    International Nuclear Information System (INIS)

    1995-01-01

    The Plutonium Finishing Plant (PFP) previously known as the Plutonium Process and Storage Facility, or Z-Plant, was built and put into operation in 1949. Since 1949 PFP has been used for various processing missions, including plutonium purification, oxide production, metal production, parts fabrication, plutonium recovery, and the recovery of americium (Am-241). The PFP has also been used for receipt and large scale storage of plutonium scrap and product materials. The PFP Final Safety Analysis Report (FSAR) was prepared by WHC to document the hazards associated with the facility, present safety analyses of potential accident scenarios, and demonstrate the adequacy of safety class structures, systems, and components (SSCs) and operational safety requirements (OSRs) necessary to eliminate, control, or mitigate the identified hazards. Documented in this Safety Evaluation Report (SER) is DOE's independent review and evaluation of the PFP FSAR and the basis for approval of the PFP FSAR. The evaluation is presented in a format that parallels the format of the PFP FSAR. As an aid to the reactor, a list of acronyms has been included at the beginning of this report. The DOE review concluded that the risks associated with conducting plutonium handling, processing, and storage operations within PFP facilities, as described in the PFP FSAR, are acceptable, since the accident safety analyses associated with these activities meet the WHC risk acceptance guidelines and DOE safety goals in SEN-35-91

  1. Westinghouse Advances in Passive Plant Safety

    International Nuclear Information System (INIS)

    Bruschi, H. J.; Manager, General; Gerstenhaber, E.

    1993-01-01

    On June 26, 1992, Westinghouse submitted the Ap600 Standard Safety Analysis Report and comprehensive PIRA results to the U. S. NRC for review as part of the Ap600 design certification program. This major milestone was met on time on a schedule set more than 3 years before submittal and is the result of the cooperative efforts of the U. S. Department of Energy (DOE), the Electric Power Requirements Program, and the Westinghouse Ap600 design team. These efforts were initiated in 1985 to develop a 600 MW advanced light water reactor plant design based on specific technical requirements established to provide the safety, simplicity, reliability, and economics necessary for the next generation of nuclear power plants. The Ap600 design achieves the ALRR safety requirements through ample design margins, simplified safety systems based on natural driving forces, and on a human-engineered man-machine interface system. Extensive Probabilistic Risk evolution, have recently shown that even if none of the active defense-in-depth safety systems are available, the passive systems alone meet safety goals. Furthermore, many tests in an extensive test program have begun or have been completed. Early tests show that passive safety perform well and meet design expectations

  2. Design safety improvements of Kozloduy NPP

    International Nuclear Information System (INIS)

    Hinovski, I.

    1999-01-01

    Design safety improvements of Kozloduy NPP, discussed in detail, are concerned with: primary circuit integrity; reactor pressure vessel integrity; primary coolant piping integrity; primary coolant overpressure protection; leak before break status; design basis accidents and transients; severe accident analysis; improvements of safety and support systems; containment/confinement leak tightness and strength; seismic safety improvements; WWER-1000 control rod insertion; upgrading and modernization of Units 5 and 6; Year 2000 problem

  3. Fire safety study of Dodewaard and Borssele nuclear power plants

    International Nuclear Information System (INIS)

    1988-03-01

    From the nuclear and conventional fire safety audits of both Dutch nuclear power plants performed under supervision of the Nuclear Safety Inspectorate and the Inspectorate for the Fire Services it turns out that the fire safety is sufficient however amenable for improvement. Besides a detailed description of the method of examination, the study 'Fire Safety' contains the results of the audit and 14 respectively 15 recommendations for improvement of the fire safety in Dodewaard and Borssele. The suggested recommendations which are applicable to both power plants are the following: fire fighting training for operators and guards, a complete emergency ventilation system of the control room, increase in the number of detectors and alarms, an increase in the quantity of water available for high-pressure fire fighting, improvement of fire barriers between several redundancies of nuclear safety systems, an investigation and possible improvement of the heat and radiation protection offered by presently used fire fighting suits. For Dodewaard a closed emergency staircase in the reactor building (secondary containment) is deemed necessary for personnel emergency escape routes and continued fire fighting if required

  4. Upgrading the safety assessment of exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

    An examination of the safety aspects of exported nuclear power plants demonstrates that additional and somewhat special considerations exist for these plants, and thus that some new approaches may be required to insure their safety. In view of the generally small regulatory staffs of importing countries, suggestions are given for measures which should be taken by the various organizations involved in the export and import of nuclear power facilities to raise the level of the very essential safety assessment. These include the upgrading of the 'export edition' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which highlights the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical Operating Regulations (Technical Specifications for Operation). Consideration is also given to upgrading the regulatory effort and to the obligations of principal organizations involved with exported nuclear plants, including national and international, for insuring the importing countries' technical readiness and the adequacy of the regulatory effort. Special attention is directed towards the project contract as a means of implementing programmes to achieve these goals. (author)

  5. 15 years after Chernobyl, nuclear power plant safety improved world-wide, but regional strains on health, economy and environment remain

    International Nuclear Information System (INIS)

    2001-01-01

    Fifteen years after the Chernobyl accident, exhaustive studies by the IAEA and others provide a solid understanding of the causes and consequences of the accident, which stemmed from design deficiencies in the reactor compounded by violation of operating procedures. These deficiencies and the lack of an international notification mechanism led to the speedy adoption of early Notification and Assistance Conventions as well as later establishment of the landmark Convention on Nuclear Safety. Lessons learned from the accident were also a significant driving force behind a decade of IAEA assistance to the countries of Central and eastern Europe and the Former Soviet Union. Much of this work was focused on identifying the weaknesses in and improving the design safety of WWER and RBMK reactors

  6. Upgrading safety documentation for exported nuclear power plants

    International Nuclear Information System (INIS)

    Rosen, M.

    1978-01-01

    In view of the generally small regulatory staffs of importing countries, suggestions are given for upgrading the ''export edition'' of the traditionally supplied safety documentation by use of a Supplementary Information Report, written specifically for the needs of a smaller and/or less technically qualified staff, which would highlight the differences that exist between the facility to be constructed and the supposedly similar reference plant of the supplier country; by improvement of supporting safety documentation to allow for adequate understanding of significant safety parameters; and by attention to the needs of smaller countries in the critical operating regulations (Technical Specifications for Operation). (author)

  7. IAEA Leads Operational Safety Mission to Muehleberg Nuclear Power Plant

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: An international team of nuclear safety experts led by the International Atomic Energy Agency today concluded a review of the safety practices at the Muehleberg Nuclear Power Plant (NPP) near Bern in Switzerland. The team noted a series of good practices and made recommendations and suggestions to reinforce them. The IAEA assembled the Operational Safety Review Team at the request of the Swiss government. The team, led by the IAEA's Division of Nuclear Installation Safety, performed an in-depth operational safety review from 8 to 25 October 2012. The team comprised experts from Belgium, the Czech Republic, Finland, Germany, Hungary, Slovakia, Sweden, the United Kingdom and the United States as well as experts from the IAEA. The team conducted an in-depth review of the aspects essential to the safe operation of the Muehleberg NPP. The conclusions of the review are based on the IAEA's Safety Standards and proven good international practices. The review covered the areas of Management, Organization and Administration; Training; Operations; Maintenance; Technical Support; Operating Experience; Radiation Protection; Chemistry, Emergency Planning and Preparedness, Severe Accident Management and Long-Term Operation. The OSART team made 10 recommendations and 11 suggestions related to areas where operations of Muehleberg NPP could be further improved, for example: - Plant management could improve the operating experience program and methods throughout the plant to ensure corrective actions are taken in a timely manner; - In the area of Long-Term Operation, the ageing management review for some systems and components is not complete and the environmental qualification of originally installed safety cables has not yet been revalidated for long-term operation; and - The plant provisions for the protection of persons on the site during an emergency with radioactive release can be improved to minimize health risks to plant personnel. The team also identified 10 good

  8. Safety assessment of plant food supplements (PFS)

    NARCIS (Netherlands)

    Berg, van den S.J.P.L.; Serra-Majem, L.; Coppens, P.; Rietjens, I.

    2011-01-01

    Botanicals and botanical preparations, including plant food supplements (PFS), are widely used in Western diets. The growing use of PFS is accompanied by an increasing concern because the safety of these PFS is not generally assessed before they enter the market. Regulatory bodies have become more

  9. Safety aspects of a fuel reprocessing plant

    International Nuclear Information System (INIS)

    Donoghue, J.K.; Charlesworth, F.R.; Fairbairn, A.

    1977-01-01

    The establishment of the basic process must include the determination of the sensitivity of the process to operational errors or plant failures. The probability, and consequences of escapes of activity must be evaluated and emergency procedures set up to deal with accidents which might lead to such escapes. The administrative arrangements for safety should include a safety evaluation and advisory service independent of line management. A quality assurance strategy for the construction and commissioning stages is important. The design and construction of the plant must include: (i) Attention to plant reliability. Maintenance and inspection procedures to maintain reliability must be adopted and the design should include measures to facilitate in-service inspection of highly-active plant. (ii) Suitable and sufficient means of detection and prevention of malfunction, including criticality, bearing in mind both the timescale of development of the fault and its consequences. (iii) Measures for containment of activity. Penetrations from active into operating areas should be eliminated or minimised and maintenance should be separated from operational areas. Secondary containment beyond that provided for operations of a significant magnitude. A ventilation system with appropriate gas clean-up, monitoring and discharge facilities is required. (iv) Adequate shielding, with particular attention paid to multiple activities in a single operational area which might lead to an operator being exposed to radiation from operations which are beyond his control. (v) Means of accounting for active materials and for their recovery, transfer and disposal in the event of a forced shut down. (vi) Suitable methods for segregation and control of wastes within the plant and for their discharge. Solid or liquid wastes should be subject to delay and monitoring procedures before release. Facilities for storage of waste must be subject to the same safety principles as the plant itself. (vii) Final

  10. PROPOSAL OF VOIVODESHIP ROAD SAFETY IMPROVEMENT PROGRAMME

    OpenAIRE

    Tomasz SZCZURASZEK; Jan KEMPA

    2016-01-01

    The article presents a proposal of the ‘GAMBIT KUJAWSKO-POMORSKI’ Road Safety Improvement Programme. The main idea of the Programme is to establish and initiate systems that will be responsible for the most important areas of activity within road safety, including road safety control, supervision, and management systems in the whole Voivodeship. In total, the creation and start of nine such systems has been proposed, namely: the Road Safety Management, the Integrated Road Rescue Service, the ...

  11. Different aspects of safety in Nuclear Fuel Plant at Pitesti, Romania

    International Nuclear Information System (INIS)

    Ivana, T.; Epure, Gh.

    2009-01-01

    Nuclear Fuel Plant (FCN) is a facility that produces fuel bundles of CANDU-6 type for the CANDU nuclear power plant. Only natural and depleted uranium in bulk and itemized form are present as nuclear materials in this facility. Uranium and wastes from the plant are handled, processed, treated and stored throughout the entire facility. The nuclear materials with natural and depleted uranium are entirely under nuclear safeguards. The amount of uranium present in the plant in different forms and activities together with zircaloy, beryllium and other hazardous substances, wastes, explosive materials at high temperatures, etc. lead to special measures undertaken by Nuclear Safety Department (DNS) to ensure nuclear safety. Different aspects of safety are continuously monitored in the plant: operational safety, industrial safety, radiological safety, labour safety, informational safety. The emergency preparedness and response, physical protection and the security of the plant and of the transportation of radioactive materials are contributing to cover the multitude of safety aspects. The safety culture of workers built directly on the safety components completes this activity in the plant. In addition the aspects of safety, security and safeguards are in permanent synergy, parts of the three components being included in each other. In the future the policy of FCN will be focused so that any improvement of one of the safety components will be reflected in improving the other safety aspects. (authors)

  12. Safety strategy and safety analysis of nuclear power plants

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1976-01-01

    The safety strategy for nuclear power plants is characterized by the fact that the high level of safety was attained not as a result of experience, but on the basis of preventive accident analyses and the finding derived from such analyses. Although, in these accident analyses, the deterministic approach is predominant, it is supplemented by reliability analyses. The accidents analyzed in nuclear licensing procedures cover a wide spectrum from minor incidents to the design basis accidents which determine the design of the safety devices. The initial and boundary conditions, which are essentail for accident analyses, and the determination of the loads occurring in various states during regular operation and in accidents flow into the design of the individual systems and components. The inevitable residual risk and its origins are discussed. (orig.) [de

  13. Total safety management: An approach to improving safety culture

    International Nuclear Information System (INIS)

    Blush, S.M.

    1993-01-01

    A little over 4 yr ago, Admiral James D. Watkins became Secretary of Energy. President Bush, who had appointed him, informed Watkins that his principal task would be to clean up the nuclear weapons complex and put the US Department of Energy (DOE) back in the business of producing tritium for the nation's nuclear deterrent. Watkins recognized that in order to achieve these objectives, he would have to substantially improve the DOE's safety culture. Safety culture is a relatively new term. The International Atomic Energy Agency (IAEA) used it in a 1986 report on the root causes of the Chernobyl nuclear accident. In 1990, the IAEA's International Nuclear Safety Advisory Group issued a document focusing directly on safety culture. It provides guidelines to the international nuclear community for measuring the effectiveness of safety culture in nuclear organizations. Safety culture has two principal aspects: an organizational framework conducive to safety and the necessary organizational and individual attitudes that promote safety. These obviously go hand in hand. An organization must create the right framework to foster the right attitudes, but individuals must have the right attitudes to create the organizational framework that will support a good safety culture. The difficulty in developing such a synergistic relationship suggests that achieving and sustaining a strong safety culture is not easy, particularly in an organization whose safety culture is in serious disrepair

  14. New requirements on safety of nuclear power plants according to the IAEA safety standards

    International Nuclear Information System (INIS)

    Misak, J.

    2005-01-01

    In this presentation author presents new requirements on safety of nuclear power plants according to the IAEA safety standards. It is concluded that: - New set of IAEA Safety Standards is close to completion: around 40 standards for NPPs; - Different interpretation of IAEA Safety Standards at present: best world practices instead of previous 'minimum common denominator'; - A number of safety improvements required for NPPs; - Requirements related to BDBAs and severe accidents are the most demanding due to degradation of barriers: hardware modifications and accident management; - Large variety between countries in implementation of accident management programmes: from minimum to major hardware modifications; -Distinction between existing and new NPPs is essential from the point of view of the requirements; WWER 440 reactors have potential to reflect IAEA Safety Standards for existing NPPs; relatively low reactor power offers broader possibilities

  15. Research on the improvement of nuclear safety -Development of a nuclear power plant system analysis code TASS (Transient and setpoint simulation)

    Energy Technology Data Exchange (ETDEWEB)

    Shim, Suk Koo; Jang, Won Pyo; Kim, Heui Chul; Kim, Kyung Doo; Lee, Sung Jae; Hah, Kyooi Suk; Song, Soon Jah; Um, Kil Sub; Yoon, Han Yung; Kim, Doo Il; Yoo, Hyung Keun; Choi, Jae Don; Lee, Byung Il; Kim, Jung Jin [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-07-01

    During the third year of the project the development of TASS 1.0 code has been completed and validated its capability in applying for the licensing transient analyses of the Westinghouse and CE type operating reactors as well as the PWR reactors under construction in Korea. The validation of the TASS 1.0 code has been achieved through the comparison calculations of the YGN-3/4 FSAR transients, Kori-3 loss of AC power transient, plant data, Kori-4 load rejection and YGN-3 startup test data as well as the BETHSY loop steam generator tube rupture test data. TASS 1.0 calculation agrees well with the best estimate RELAP5/MOD 3.1 calculation for the YGN-3/4 FASR transients and shows its capability in simulating plant transient and startup data as well as the thermal hydraulic transient test data. Topical reports on TASS 1.0 code have been prepared and will be submitted to Korea Institute of Nuclear Safety for its licensing application to Westinghouse and CE type PWR transient analyses. The development of TASS 2.0 code has been head started in this year to timely utilize the TASS 2.0 code for the KNGR design certification. 65 figs, 30 tabs, 44 refs. (Author).

  16. Waste Isolation Pilot Plant Safety Analysis Report

    International Nuclear Information System (INIS)

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions'' (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.'' This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment

  17. Waste Isolation Pilot Plant Safety Analysis Report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-11-01

    The following provides a summary of the specific issues addressed in this FY-95 Annual Update as they relate to the CH TRU safety bases: Executive Summary; Site Characteristics; Principal Design and Safety Criteria; Facility Design and Operation; Hazards and Accident Analysis; Derivation of Technical Safety Requirements; Radiological and Hazardous Material Protection; Institutional Programs; Quality Assurance; and Decontamination and Decommissioning. The System Design Descriptions`` (SDDS) for the WIPP were reviewed and incorporated into Chapter 3, Principal Design and Safety Criteria and Chapter 4, Facility Design and Operation. This provides the most currently available final engineering design information on waste emplacement operations throughout the disposal phase up to the point of permanent closure. Also, the criteria which define the TRU waste to be accepted for disposal at the WIPP facility were summarized in Chapter 3 based on the WAC for the Waste Isolation Pilot Plant.`` This Safety Analysis Report (SAR) documents the safety analyses that develop and evaluate the adequacy of the Waste Isolation Pilot Plant Contact-Handled Transuranic Wastes (WIPP CH TRU) safety bases necessary to ensure the safety of workers, the public and the environment from the hazards posed by WIPP waste handling and emplacement operations during the disposal phase and hazards associated with the decommissioning and decontamination phase. The analyses of the hazards associated with the long-term (10,000 year) disposal of TRU and TRU mixed waste, and demonstration of compliance with the requirements of 40 CFR 191, Subpart B and 40 CFR 268.6 will be addressed in detail in the WIPP Final Certification Application scheduled for submittal in October 1996 (40 CFR 191) and the No-Migration Variance Petition (40 CFR 268.6) scheduled for submittal in June 1996. Section 5.4, Long-Term Waste Isolation Assessment summarizes the current status of the assessment.

  18. The safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Hoehn, J.; Niehaus, F.

    1997-01-01

    Nuclear power plant operators and nuclear organizations from the West and from the East cooperate at many levels. The G7 and G24 nations have taken it upon themselves to improve the safety of Eastern nuclear power plants. The European Union has launched support programs, i.e. Technical Assistance to the Commonwealth of Independent States (Tacis) and Pologne-Hangrie: Aide a la Reconstruction Economique (Phare), and founded the European Bank for Reconstruction and Development. The countries of Central and Eastern Europe operate nuclear power plants equipped with VVER-type pressurized water reactors and those equipped with RBMK-type reactors. The safety of these two types of plants is judged very differently. Among the VVER plants, a distinction is made between the older and the more recent 440 MWe lines and the 1000 MWe line. Especially the RBMK plants (Chernobyl-type plants) differ greatly as a function of location and year of construction. Even though they do not meet Western safety standards and at best can be backfitted up to a certain level, it must yet be assumed that they will remain in operation to the end of their projected service lives for economic reasons. (orig.) [de

  19. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Chinese Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  20. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  1. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Ed.)

    International Nuclear Information System (INIS)

    2012-01-01

    On the basis of the principles included in the Fundamental Safety Principles, IAEA Safety Standards Series No. SF-1, this Safety Requirements publication establishes requirements applicable to the design of nuclear power plants. It covers the design phase and provides input for the safe operation of the power plant. It elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  2. Complementary assessment of the safety of French nuclear power plants

    International Nuclear Information System (INIS)

    Camarcat, N.; Pouget-Abadie, X.

    2011-01-01

    As an immediate consequence of the Fukushima accident the French nuclear safety Authority (ASN) asked EDF to perform a complementary safety assessment for each nuclear power plant dealing with 3 points: 1) the consequences of exceptional natural disasters, 2) the consequences of total loss of electrical power, and 3) the management of emergency situations. The safety margin has to be assessed considering 3 main points: first a review of the conformity to the initial safety requirements, secondly the resistance to events overdoing what the facility was designed to stand for, and the feasibility of any modification susceptible to improve the safety of the facility. This article details the specifications of such assessment, the methodology followed by EDF, the task organization and the time schedule. (A.C.)

  3. Basic safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    1988-01-01

    Nuclear power plant safety requires a continuing quest for excellence. All individuals concerned should constantly be alert to opportunities to reduce risks to the lowest practicable level. The quest, however, is most likely to be fruitful if it is based on an understanding of the underlying objectives and principles of nuclear safety, and the way in which its aspects are interrelated. This report is an attempt to provide a logical framework for such an understanding. The proposed objectives and principles of nuclear safety are interconnected and must be taken as a whole; they do not constitute a menu from which selection can be made. The report takes account of current issues and developments. It includes the concept of safety objectives and the use of probabilistic safety assessment. Reliability targets for safety systems are discussed. The concept of a 'safety culture' is crucial. Attention has been paid to the need for planning for accident management. The report contains objectives and principles. The objectives state what is to be achieved; the principles state how to achieve it. In each case, the basic principle is stated as briefly as possible. The accompanying discussion comments on the reasons for the principle and its importance, as well as exceptions, the extent of coverage and any necessary clarification. The discussion is as important as the principle it augments. 4 figs

  4. An optimization model for improving highway safety

    Directory of Open Access Journals (Sweden)

    Promothes Saha

    2016-12-01

    Full Text Available This paper developed a traffic safety management system (TSMS for improving safety on county paved roads in Wyoming. TSMS is a strategic and systematic process to improve safety of roadway network. When funding is limited, it is important to identify the best combination of safety improvement projects to provide the most benefits to society in terms of crash reduction. The factors included in the proposed optimization model are annual safety budget, roadway inventory, roadway functional classification, historical crashes, safety improvement countermeasures, cost and crash reduction factors (CRFs associated with safety improvement countermeasures, and average daily traffics (ADTs. This paper demonstrated how the proposed model can identify the best combination of safety improvement projects to maximize the safety benefits in terms of reducing overall crash frequency. Although the proposed methodology was implemented on the county paved road network of Wyoming, it could be easily modified for potential implementation on the Wyoming state highway system. Other states can also benefit by implementing a similar program within their jurisdictions.

  5. Plant safety review from mass criticality accident

    International Nuclear Information System (INIS)

    Susanto, B.G.

    2000-01-01

    The review has been done to understand the resent status of the plant in facing postulated mass criticality accident. From the design concept of the plant all the components in the system including functional groups have been designed based on favorable mass/geometry safety principle. The criticality safety for each component is guaranteed because all the dimensions relevant to criticality of the components are smaller than dimensions of 'favorable mass/geometry'. The procedures covering all aspects affecting quality including the safety related are developed and adhered to at all times. Staff are indoctrinated periodically in short training session to warn the important of the safety in process of production. The plant is fully equipped with 6 (six) criticality detectors in strategic places to alert employees whenever the postulated mass criticality accident occur. In the event of Nuclear Emergency Preparedness, PT BATAN TEKNOLOGI has also proposed the organization structure how promptly to report the crisis to Nuclear Energy Control Board (BAPETEN) Indonesia. (author)

  6. Assessment of IAEA safety series no. 75-INSAG-3 - ''basic safety principles for nuclear power plants''

    International Nuclear Information System (INIS)

    1989-01-01

    The International Atomic Energy Agency Safety Series No. 75-INSAG--3, 'Basic Safety Principles for Nuclear Power Plants' is reviewed in the light of the Advisory Committee on Nuclear Safety reports ACNS--2, 'Safety Objectives for Nuclear Activities in Canada', and ACNS--4, 'Recommended General Safety Requirements for Nuclear Power Plants'. The INSAG safety objectives are consistent with the safety objectives stated in ACNS--2 but are less general, applying only to nuclear power plants. The INSAG safety principles are, in general, consistent with the requirements stated in ACNS--4 but put more emphasis on 'safety culture'. They give little attention to reactor plant effluents, waste management, or decommissioning. (fig., 5 refs.)

  7. Mathematical Safety Assessment Approaches for Thermal Power Plants

    Directory of Open Access Journals (Sweden)

    Zong-Xiao Yang

    2014-01-01

    Full Text Available How to use system analysis methods to identify the hazards in the industrialized process, working environment, and production management for complex industrial processes, such as thermal power plants, is one of the challenges in the systems engineering. A mathematical system safety assessment model is proposed for thermal power plants in this paper by integrating fuzzy analytical hierarchy process, set pair analysis, and system functionality analysis. In the basis of those, the key factors influencing the thermal power plant safety are analyzed. The influence factors are determined based on fuzzy analytical hierarchy process. The connection degree among the factors is obtained by set pair analysis. The system safety preponderant function is constructed through system functionality analysis for inherence properties and nonlinear influence. The decision analysis system is developed by using active server page technology, web resource integration, and cross-platform capabilities for applications to the industrialized process. The availability of proposed safety assessment approach is verified by using an actual thermal power plant, which has improved the enforceability and predictability in enterprise safety assessment.

  8. Nuclear safety in Slovak Republic. Status of safety improvements

    International Nuclear Information System (INIS)

    Toth, A.

    1999-01-01

    Status of the safety improvements at Bohunice V-1 units concerning WWER-440/V-230 design upgrading were as follows: supplementing of steam generator super-emergency feed water system; higher capacity of emergency core cooling system; supplementing of automatic links between primary and secondary circuit systems; higher level of secondary system automation. The goal of the modernization program for Bohunice V-1 units WWER-440/V-230 was to increase nuclear safety to the level of the proposals and IAEA recommendations and to reach probability goals of the reactor concerning active zone damage, leak of radioactive materials, failures of safety systems and damage shields. Upgrading program for Mochovce NPP - WWER-440/V-213 is concerned with improving the integrity of the reactor pressure vessel, steam generators 'leak before break' methods applied for the NPP, instrumentation and control of safety systems, diagnostic systems, replacement of in-core monitoring system, emergency analyses, pressurizers safety relief valves, hydrogen removal system, seismic evaluations, non-destructive testing, fire protection. Implementation of quality assurance has a special role in improvement of operational safety activities as well as safety management and safety culture, radiation protection, decommissioning and waste management and training. The Year 2000 problem is mentioned as well

  9. Current collectors for improved safety

    Science.gov (United States)

    Abdelmalak, Michael Naguib; Allu, Srikanth; Dudney, Nancy J.; Li, Jianlin; Simunovic, Srdjan; Wang, Hsin

    2017-12-19

    A battery electrode assembly includes a current collector with conduction barrier regions having a conductive state in which electrical conductivity through the conduction barrier region is permitted, and a safety state in which electrical conductivity through the conduction barrier regions is reduced. The conduction barrier regions change from the conductive state to the safety state when the current collector receives a short-threatening event. An electrode material can be connected to the current collector. The conduction barrier regions can define electrical isolation subregions. A battery is also disclosed, and methods for making the electrode assembly, methods for making a battery, and methods for operating a battery.

  10. Organizational processes and nuclear power plant safety

    International Nuclear Information System (INIS)

    Landy, F.J.; Jacobs, R.R.; Mathieu, J.

    1991-01-01

    The paper describes the effects organizational factors have on the risk associated with the operation of nuclear power plants. The described research project addresses three methods for identifying the organizational factors that impact safety. The first method consists of an elaborate theory-based protocol dealing with decision making procedures, interdepartmental coordination of activities, and communications. The second, known as goals/means/measures protocol, deals with identifying safey related goals. The third method is known as behaviorally anchored rating scale development. The paper discusses the importance of the convergence of these three methods to identify organizational factors essential to reactor safety

  11. Self-assessment of operational safety for nuclear power plants

    International Nuclear Information System (INIS)

    1999-12-01

    Self-assessment processes have been continuously developed by nuclear organizations, including nuclear power plants. Currently, the nuclear industry and governmental organizations are showing an increasing interest in the implementation of this process as an effective way for improving safety performance. Self-assessment involves the use of different types of tools and mechanisms to assist the organizations in assessing their own safety performance against given standards. This helps to enhance the understanding of the need for improvements, the feeling of ownership in achieving them and the safety culture as a whole. Although the primary beneficiaries of the self-assessment process are the plant and operating organization, the results of the self-assessments are also used, for example, to increase the confidence of the regulator in the safe operation of an installation, and could be used to assist in meeting obligations under the Convention on Nuclear Safety. Such considerations influence the form of assessment, as well as the type and detail of the results. The concepts developed in this report present the basic approach to self-assessment, taking into consideration experience gained during Operational Safety Review Team (OSART) missions, from organizations and utilities which have successfully implemented parts of a self-assessment programme and from meetings organized to discuss the subject. This report will be used in IAEA sponsored workshops and seminars on operational safety that include the topic of self-assessment

  12. Special safety requirements applied to Brazilian nuclear power plant

    International Nuclear Information System (INIS)

    Lepecki, W.P.S.; Hamel, H.J.E.; Koenig, N.; Vieira, P.C.R.; Fritzsche, J.C.

    1981-01-01

    Some safety aspects of the Angra 2 and 3 nuclear power plants are presented. An analysis of the civil and mechanical project of these nuclear power plant having in view a safety analysis is done. (E.G.) [pt

  13. Code on the safety of nuclear power plants: Siting

    International Nuclear Information System (INIS)

    1988-01-01

    This Code provides criteria and procedures that are recommended for safety in nuclear power plant siting. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  14. Mochovce NPP safety improvement and completion

    International Nuclear Information System (INIS)

    1997-01-01

    6th Nuclear society information meeting dealt with the completion of the Mochovce NPP with regard to implementation of safety measures. It was aimed to next problems: I. 'Survey' presentation on the situation of the nuclear power industry in partner countries; II. Basic technical presentations; III. Presentations of operators of the other VVER 440/213 NPPs on their activities in the field of safety improvement in relation to IAEA recommendations; IV. Technical solutions of safety improvements ranked with IAEA degree 3 (Report SC 108 VVER); V: Technical solutions of selected Safety Measures ranked with IAEA degree 2 and 1 (Report SC 108 VVER)

  15. Can we Improve Patient Safety?

    Directory of Open Access Journals (Sweden)

    Martin Thomas Corbally

    2014-09-01

    Full Text Available Despite greater awareness of patient safety issues especially in the operating room and the widespread implementation of surgical time out (WHO,errors, especially wrong site surgery, continue. Most such errors are due to lapses in communication where decision makers fail to consult or confirm operative findings but worryingly where parental concerns over the planned procedure are ignored or not followed through. The WHO surgical pause / Time Out aims to capture these errors and prevent them but the combination of human error and complex hospital environments can overwhelm even robust safety structures and simple common sense. Parents are the ultimate repository of information on their child's condition and planned surgery but are traditionally excluded from the process of Surgical pause and Time Out perhaps to avoid additional stress. In addition surgeons, like pilots, are subject to the phenomenon of plan continue fail with potentially disastrous outcomes.

  16. NPP Krsko core calculations to improve operational safety

    International Nuclear Information System (INIS)

    Ivekovic, I.; Grgic, D.; Nemec, T.

    2007-01-01

    Calculation tools and methodology used to perform independent calculations of cumulative influence of different changes related to fuel and core operation of NPP Krsko were described. Some examples of steady state and transient results are used to illustrate potential improvements to understanding and reviewing plant safety. (author)

  17. A PIP chart for nuclear plant safety

    International Nuclear Information System (INIS)

    Suzuki, Tatsujiro; Yamaoka, Taiji

    1992-01-01

    While it is known that social and political aspects of nuclear safety issues are important, little study has been done on identifying the breadth of stakeholders whose policies have important influences over nuclear plant safety in a comprehensive way. The objectives of this study are to develop a chart that visually identifies important stakeholders and their policies and illustrates these influences in a hierarchical representation so that the relationship between stakeholders and nuclear safety will be better understood. This study is based on a series of extensive interviews with major stakeholders, such as nuclear plant managers, corporate planning vice presidents, state regulators, news media, and public interest groups, and focuses on one US nuclear power plant. Based on the interview results, the authors developed a conceptual policy influence paths (PIP) chart. The PIP chart illustrates the hierarchy of influence among stakeholders. The PIP chart is also useful in identifying possible stakeholders who can be easily overlooked without the PIP chart. In addition, it shows that influence flow is circular rather than linear in one direction

  18. Safety and licensing of nuclear heating plants

    International Nuclear Information System (INIS)

    Snell, V.G.; Hilborn, J.W.; Lynch, G.F.; McAuley, S.J.

    1989-09-01

    World attention continues to focus on nuclear district heating, a low-cost energy from a non-polluting fuel. It offers long-term security for countries currently dependent on fossil fuels, and can reduce the burden of fossil fuel transportation on railways and roads. Current initiatives encompass large, centralized heating plants and small plants supplying individual institutions. The former are variants of their power reactor cousins but with enhanced safety features. The latter face the safety and licensing challenges of urban siting and remotely monitored operation, through use of intrinsic safety features such as passive decay heat removal, low stored energy and limited reactivity speed and depth in the control systems. Small heating reactor designs are compared, and the features of the SLOWPOKE Energy System, in the forefront of these designs, are summarized. The challenge of public perception must be met by clearly presenting the characteristics of small heating reactors in terms of scale and transparent safety in design and operation, and by explaining the local benefits

  19. IAEA Leads Operational Safety Mission to Armenian Nuclear Power Plant

    International Nuclear Information System (INIS)

    2011-01-01

    Full text: An international team of nuclear installation safety experts, led by the International Atomic Energy Agency (IAEA), has reviewed the Armenian Nuclear Power Plant (ANPP) near Metsamor for its safety practices and has noted a series of good practices, as well as recommendations to reinforce them. The IAEA assembled an international team of experts at the request of the Government of the Republic of Armenia to conduct an Operational Safety Review (OSART) of the NPP. Under the leadership of the IAEA's Division of Nuclear Installation Safety, the OSART team performed an in-depth operational safety review from 16 May to 2 June 2011. The team was made up of experts from Finland, France, Lithuania, Hungary, Netherlands, Slovakia, UK, USA, EC and the IAEA. An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the plant's overall safety status. Experts participating in the IAEA's June 2010 International Conference on Operational Safety of Nuclear Power Plants (NPP) reviewed the experience of the OSART programme and concluded: In OSART missions NPPs are assessed against IAEA safety standards which reflect the current international consensus on what constitutes a high level of safety; and OSART recommendations and suggestions are of utmost importance for operational safety improvement of NPPs. Armenia is commended for openness to the international nuclear community and for actively inviting IAEA safety review missions to submit their activities to international scrutiny. Examples of IAEA safety reviews include: Design Safety Review in 2003; Review of Probabilistic Safety Assessment in 2007; and Assessment of Seismic Safety Re-Evaluation in 2009. The team at ANPP conducted an in-depth review of the aspects essential to the safe operation of the plant, which is largely under the control of the site management

  20. 46. The goals of safety engineering department of the plant

    International Nuclear Information System (INIS)

    Ivanov, A.V.

    1993-01-01

    The goals of safety engineering department of the plant, including elaboration of instructions on safety engineering on all specialities, safety engineering training of all labours working on the plant and control for abidance by the instructions on safety engineering were discussed.

  1. Improved design on Qinshan 300 MWe nuclear power plant

    International Nuclear Information System (INIS)

    Shi Peihua; Cheng Wanli; Lu Rongliang

    1993-01-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented

  2. Improved design on Qinshan 300 MWe nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Peihua, Shi; Wanli, Cheng; Rongliang, Lu [Shanghai Nuclear Engineering Research and Design Inst. (China)

    1993-06-01

    The main aim, guiding ideology, general performance and parameters of improved design on Qinshan 300 MWe nuclear power plant are presented. Improved items are also introduced including the characteristic of layout in nuclear island building, decreasing unnecessary devices increasing necessary safety facilities and unifying code and standard. The progress of improved design is presented.

  3. ESRS guidelines for software safety reviews. Reference document for the organization and conduct of Engineering Safety Review Services (ESRS) on software important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    The IAEA provides safety review services to assist Member States in the application of safety standards and, in particular, to evaluate and facilitate improvements in nuclear power plant safety performance. Complementary to the Operational Safety Review Team (OSART) and the International Regulatory Review Team (IRRT) services are the Engineering Safety Review Services (ESRS), which include reviews of siting, external events and structural safety, design safety, fire safety, ageing management and software safety. Software is of increasing importance to safety in nuclear power plants as the use of computer based equipment and systems, controlled by software, is increasing in new and older plants. Computer based devices are used in both safety related applications (such as process control and monitoring) and safety critical applications (such as reactor protection). Their dependability can only be ensured if a systematic, fully documented and reviewable engineering process is used. The ESRS on software safety are designed to assist a nuclear power plant or a regulatory body of a Member State in the review of documentation relating to the development, application and safety assessment of software embedded in computer based systems important to safety in nuclear power plants. The software safety reviews can be tailored to the specific needs of the requesting organization. Examples of such reviews are: project planning reviews, reviews of specific issues and reviews prior final acceptance. This report gives information on the possible scope of ESRS software safety reviews and guidance on the organization and conduct of the reviews. It is aimed at Member States considering these reviews and IAEA staff and external experts performing the reviews. The ESRS software safety reviews evaluate the degree to which software documents show that the development process and the final product conform to international standards, guidelines and current practices. Recommendations are

  4. Safety criteria for nuclear chemical plants

    International Nuclear Information System (INIS)

    Ball, P.W.; Curtis, L.M.

    1983-01-01

    Safety measures have always been required to limit the hazards due to accidental release of radioactive substances from nuclear power plants and chemical plants. The risk associated with the discharge of radioactive substances during normal operation has also to be kept acceptably low. BNFL (British Nuclear Fuels Ltd.) are developing risk criteria as targets for safe plant design and operation. The numerical values derived are compared with these criteria to see if plants are 'acceptably safe'. However, the criteria are not mandatory and may be exceeded if this can be justified. The risk assessments are subject to independent review and audit. The Nuclear Installations Inspectorate also has to pass the plants as safe. The assessment principles it uses are stated. The development of risk criteria for a multiplant site (nuclear chemical plants tend to be sited with many others which are related functionally) is discussed. This covers individual members of the general public, societal risks, risks to the workforce and external hazards. (U.K.)

  5. Organization and safety in nuclear power plants

    International Nuclear Information System (INIS)

    Marcus, A.A.; Nichols, M.L.; Bromiley, P.; Olson, J.; Osborn, R.N.; Scott, W.; Pelto, P.; Thurber, J.

    1990-05-01

    Perspectives from industry, academe, and the NRC are brought together in this report and used to develop a logical framework that links management and organization factors and safety in nuclear power plant performance. The framework focuses on intermediate outcomes which can be predicted by organizational and management factors, and which are subsequently linked to safety. The intermediate outcomes are efficiency, compliance, quality, and innovation. The organization and management factors can be classified in terms of environment, context, organizational governance, organizational design, and emergent processes. Initial empirical analyses were conducted on a limited set of hypotheses derived from the framework. One set of hypotheses concerned the relationships between one of the intermediate outcome variables, efficiency, as measured by critical hours and outage rate, and safety, as measured by 5 NRC indicators. Results of the analysis suggest that critical hours and outage rates and safety, as measured in this study, are not related to each other. Hypotheses were tested concerning the effects on safety and efficiency of utility financial resources and the lagged recognition and correction of problems that accompanies the reporting of major violations and licensee event reports. The analytical technique employed was regression using polynomial distributed lags. Results suggest that both financial resources and organizational problem solving/learning have significant effects on the outcome variables when time is properly taken into account. Conclusions are drawn which point to this being a promising direction to proceed, though with some care, due to the current limitations of the study. 138 refs., 36 figs., 9 tabs

  6. Biotechnological improvement of ornamental plants

    OpenAIRE

    Flavia Soledad Darqui; Laura Mabel Radonic; Horacio Esteban Hopp; Marisa Lopez Bilbao

    2017-01-01

    The discovery of commercial transgenic varieties of orange petunias sold in Europe and the United States although they had never reached the approved status, and the consequent recommendation to destroy them, was the trigger to discuss about biotechnological improvement of ornamental plants. Inside the restricted world of 26 vegetal transgenic species, according to the ISAAA’s reports (http://www.isaaa.org), there are three ornamental species: carnation, rose and the Beijing University develo...

  7. Safety implications of computerized process control in nuclear power plants

    International Nuclear Information System (INIS)

    1991-02-01

    Modern nuclear power plants are making increasing use of computerized process control because of the number of potential benefits that accrue. This practice not only applies to new plants but also to those in operation. Here, the replacement of both conventional process control systems and outdated computerized systems is seen to be of benefit. Whilst this contribution is obviously of great importance to the viability of nuclear electricity generation, it must be recognized that there are major safety concerns in taking this route. However, there is the potential for enhancing the safety of nuclear power plants if the full power of microcomputers and the associated electronics is applied correctly through well designed, engineered, installed and maintained systems. It is essential that areas where safety can be improved be identified and that the pitfalls are clearly marked so that they can be avoided. The deliberations of this Technical Committee Meeting are a step on the road to this goal of improved safety through computerized process control. This report also contains the papers presented at the technical committee meeting by participants. A separate abstract was prepared for each of these 15 presentations. Refs, figs and tabs

  8. Integrated safety assessment report, Haddam Neck Plant (Docket No. 50-213): Integrated Safety Assessment Program: Draft report

    International Nuclear Information System (INIS)

    1987-07-01

    The integrated assessment is conducted on a plant-specific basis to evaluate all licensing actions, licensee initiated plant improvements and selected unresolved generic/safety issues to establish implementation schedules for each item. Procedures allow for a periodic updating of the schedules to account for licensing issues that arise in the future. The Haddam Neck Plant is one of two plants being reviewed under the pilot program. This report indicates how 82 topics selected for review were addressed, and presents the staff's recommendations regarding the corrective actions to resolve the 82 topics and other actions to enhance plant safety. 135 refs., 4 figs., 5 tabs

  9. Organizational factors and nuclear power plant safety

    International Nuclear Information System (INIS)

    Haber, S.B.

    1995-01-01

    There are many organizations in our society that depend on human performance to avoid incidents involving significant adverse consequences. As our culture and technology have become more sophisticated, the management of risk on a broad basis has become more and more critical. The safe operation of military facilities, chemical plants, airlines, and mass transit, to name a few, are substantially dependent on the performance of the organizations that operate those facilities. The nuclear power industry has, within the past 15 years, increased the attention given to the influence of human performance in the safe operation of nuclear power plants (NPP). While NPPs have been designed through engineering disciplines to intercept and mitigate events that could cause adverse consequences, it has been clear from various safety-related incidents that human performance also plays a dominant role in preventing accidents. Initial efforts following the 1979 Three Mile Island incident focused primarily on ergonomic factors (e.g., the best design of control rooms for maximum performance). Greater attention was subsequently directed towards cognitive processes involved in the use of NPP decision support systems and decision making in general, personnel functions such as selection systems, and the influence of work scheduling and planning on employees' performance. Although each of these approaches has contributed to increasing the safety of NPPS, during the last few years, there has been a growing awareness that particular attention must be paid to how organizational processes affect NPP personnel performance, and thus, plant safety. The direct importance of organizational factors on safety performance in the NPP has been well-documented in the reports on the Three Mile Island and Chernobyl accidents as well as numerous other events, especially as evaluated by the U.S. Nuclear Regulatory Commission (NRC)

  10. A proposed approach for enhancing design safety assurance of future plants

    International Nuclear Information System (INIS)

    Oh, Kyu Myeng; Ahn, Sang Kyu; Lee, Chang Ju; Kim, Inn Seock

    2010-01-01

    This paper provides various insights from a detailed review of deterministic approaches typically applied to ensure design safety of nuclear power plants (NPPs) and risk-informed approaches proposed to evaluate safety of advanced reactors such as Generation IV reactors. Also considered herein are the risk-informed safety analysis (RISA) methodology suggested by Westinghouse as a means to improve the conventional accident analysis, together with the Technology Neutral Framework recently suggested by the U.S. NRC for safety evaluation of future plants. These insights from the comparative review of deterministic and risk-informed approaches could be used in further enhancing the methodology for design safety assurance of future plants

  11. A new approach to preparing safety cases for existing nuclear plant (COSR)

    International Nuclear Information System (INIS)

    Rice, S.A.; Buchan, A.B.

    2000-01-01

    BNFL is committed to achieving world class safety performance, through a process of continuously reviewing and improving its safety practices. In the mid 1990s, as part of this process, the company began to develop a new type of safety case, for existing non-reactor nuclear plants, called the continued operation safety report (COSR). Following a significant amount of development work from experts within BNFL and important contributions from its regulators, the first approved COSR was recently completed and submitted to the Nuclear Installations Inspectorate. The COSR aims to provide a visibly integrated safety and engineering case for the adequacy of continued operation of a nuclear facility. It achieves this by identifying the main plant structures, systems and components that have a safety function and provides the appropriate supporting engineering substantiation. The COSR aims to explore plant safety and identify worthwhile improvements. The document also aims to be reader-friendly by focusing on the main safety issues. It is therefore a slim safety summary which provides operators, safety specialists and regulators with an overview and introduction into the broader, more detailed safety case. This paper provides an overview of the COSR and its production process, describing the safety case improvements that have been made by comparing it to its predecessor, the fully developed safety case. The paper also illustrates the benefits of the COSR by providing current examples of its application on existing BNFL plant. Finally, the paper describes ongoing development work aimed at further improving the COSR and its production process. (author)

  12. The minimum attention plant inherent safety through LWR simplification

    International Nuclear Information System (INIS)

    Turk, R.S.; Matzie, R.A.

    1987-01-01

    The Minimum Attention Plant (MAP) is a unique small LWR that achieves greater inherent safety, improved operability, and reduced costs through design simplification. The MAP is a self-pressurized, indirect-cycle light water reactor with full natural circulation primary coolant flow and multiple once-through steam generators located within the reactor vessel. A fundamental tenent of the MAP design is its complete reliance on existing LWR technology. This reliance on conventional technology provides an extensive experience base which gives confidence in judging the safety and performance aspects of the design

  13. IAEA Concludes Safety Review at Gravelines Nuclear Power Plant, France

    International Nuclear Information System (INIS)

    2012-01-01

    Full text: An IAEA-led international team of nuclear safety experts noted a series of good practices and made recommendations to reinforce some safety measures during a review of operational safety at France's Gravelines Nuclear Power Plant (NPP) that concluded today. The Operational Safety Review Team (OSART) was assembled at the French Government's request. The in-depth review, which began 12 November 2012, focused on aspects essential to the safe operation of the NPP. The team was composed of experts from Bulgaria, China, Germany, Hungary, Japan, Romania, Slovakia, South Africa, Spain, Ukraine and the IAEA. The review covered the areas of management, organization and administration; training and qualification; operations; maintenance; technical support; operating experience; radiation protection; chemistry; emergency planning and preparedness; and severe accident management. The conclusions of the review are based on the IAEA's Safety Standards. The OSART team has identified good plant practices, which will be shared with the rest of the nuclear industry for consideration of their possible use elsewhere. Examples include the following: - The Power Plant uses a staff-skills mapping process that significantly enhances knowledge of the facility's collective and individual skills and provides proactive management to address the loss of such skills; - As a measure to reduce the risk of workers' radiation exposure, the Power Plant uses a system to ensure that dose rate measurements are carried out at a precise distance from the source of radiation; and - Flood protection of the Power Plant is supported by special technical guidance documents and associated arrangements. The team identified a number of proposals for improvements to operational safety at Gravelines NPP. Examples include the following: - The Power Plant should reinforce its measures to prevent foreign objects from entering plant systems; - The Power Plant should ensure the 24-hour presence of an operator

  14. Sign up to Safety: developing a safety improvement plan.

    Science.gov (United States)

    Dight, Carol; Peters, Hayley

    2015-04-01

    The Sign up to Safety (SutS) programme was launched in June 2014 by health secretary Jeremy Hunt. It focuses on listening to patients, carers and staff, learning from what they say when things go wrong, and then taking action to improve patient safety. The programme aims to make the NHS the safest healthcare system in the world by creating a culture devoted to continuous learning and improvement (NHS England 2014). Musgrove Park Hospital, part of Taunton and Somerset NHS Foundation Trust, was one of 12 NHS organisations that signed up to the SutS programme, making public its commitment to the national pledges to be 'open and transparent' and to develop a safety improvement plan. This paper describes the development of the strategy.

  15. Safety assessment of UP3-A reprocessing plant

    International Nuclear Information System (INIS)

    Mercier, J.P.; Guezenec, J.Y.; Poirier, M.C.

    1992-02-01

    This presentation describes how the safety assessment was made of UP3-A plant of the La Hague establishment for the building permit and operating license within the context of French nuclear regulations and the national debate on the need for reprocessing. Other factors discussed are how the public was involved, how the regulations were improved in the process and what the different stages of commissioning consisted of. (author)

  16. Safety classification of items in Tianwan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sun Yongbin

    2005-01-01

    The principle of integrality, moderation and equilibrium should be considered in the safety classification of items in nuclear power plant. The basic ways for safety classification of items is to classify the safety function based on the effect of the outside enclosure damage of the items (parts) on the safety. Tianwan Nuclear Power Plant adopts Russian VVER-1000/428 type reactor, it safety classification mainly refers to Russian Guidelines and standards. The safety classification of the electric equipment refers to IEEE-308(80) standard, including 1E and Non 1E classification. The safety classification of the instrumentation and control equipment refers to GB/T 15474-1995 standard, including safety 1E, safety-related SR and NC non-safety classification. The safety classification of Tianwan Nuclear Power Plant has to be approved by NNSA and satisfy Chinese Nuclear Safety Guidelines. (authors)

  17. PROPOSAL OF VOIVODESHIP ROAD SAFETY IMPROVEMENT PROGRAMME

    Directory of Open Access Journals (Sweden)

    Tomasz SZCZURASZEK

    2016-07-01

    Full Text Available The article presents a proposal of the ‘GAMBIT KUJAWSKO-POMORSKI’ Road Safety Improvement Programme. The main idea of the Programme is to establish and initiate systems that will be responsible for the most important areas of activity within road safety, including road safety control, supervision, and management systems in the whole Voivodeship. In total, the creation and start of nine such systems has been proposed, namely: the Road Safety Management, the Integrated Road Rescue Service, the Personnel Continuing Education, the Hazardous Road Behaviour Monitoring, the Social Education for Safe Behaviour on Road, the Teaching Personnel Improvement, the Area Development and Planning Process Improvement, the Road Infrastructure Design Quality Improvement, and the Road and Traffic Management Process Efficiency Improvement. The basic aim of each system has been discussed as well as the most important tasks implemented as its part. The Road Safety Improvement Programme for the Kujawsko-Pomorskie Voivodeship presented in this article is a part of the National Road Safety Programme 2013-2020. Moreover, it is not only an original programme in Poland, but also a universal project that may be adapted for other voivodeships as well.

  18. A study of the international trend and comprehensive enhancement program on the Nuclear Power Plant safety

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Soon Hong; Cho, Nam Jin; Paek, Won Phil [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1990-12-15

    The objectives of this study are as follows : overview of the international trend related to the safety of Nuclear Power Plant(NPPs), study of the present status of NPP safety in Korea in aspects of design, construction and operation, suggestion of the comprehensive program to improve NPP safety in Korea. The results of this study can contribute to improve the safety of existing and future NPPs, and to establish the severe accident policy in Korea.

  19. A study of the international trend and comprehensive enhancement program on the Nuclear Power Plant safety

    International Nuclear Information System (INIS)

    Jang, Soon Hong; Cho, Nam Jin; Paek, Won Phil

    1990-12-01

    The objectives of this study are as follows : overview of the international trend related to the safety of Nuclear Power Plant(NPPs), study of the present status of NPP safety in Korea in aspects of design, construction and operation, suggestion of the comprehensive program to improve NPP safety in Korea. The results of this study can contribute to improve the safety of existing and future NPPs, and to establish the severe accident policy in Korea

  20. Safety goals for nuclear power plant operation

    International Nuclear Information System (INIS)

    1983-05-01

    This report presents and discusses the Nuclear Regulatory Commission's, Policy Statement on Safety Goals for the Operation of Nuclear Power Plants. The safety goals have been formulated in terms of qualitative goals and quantitative design objectives. The qualitative goals state that the risk to any individual member of the public from nuclear power plant operation should not be a significant contributor to that individual's risk of accidental death or injury and that the societal risks should be comparable to or less than those of viable competing technologies. The quantitative design objectives state that the average risks to individual and the societal risks of nuclear power plant operation should not exceed 0.1% of certain other risks to which members of the US population are exposed. A subsidiary quantitative design objective is established for the frequency of large-scale core melt. The significance of the goals and objectives, their bases and rationale, and the plan to evaluate the goals are provided. In addition, public comments on the 1982 proposed policy statement and responses to a series of questions that accompanied the 1982 statement are summarized

  1. The safety of RBMK nuclear power plants

    International Nuclear Information System (INIS)

    Holloway, N.J.

    1993-01-01

    The accident at Chernobyl coincided with the beginning of the era of ''perestroika'' and ''glasnost'' in the USSR. The accident provoked unprecedented openness between the USSR and the West, with Britain playing a large part in the exchanges because of its experience, albeit in separate reactor types, of large on-load fuelled graphite moderated reactor systems and pressure tube technologies. The Research and Development Institute of Power Engineering (RDIPE) had always been responsible for the design, development and safety analysis of the RBMK reactors. Since the accident it has therefore played the leading role in investigations of what went wrong and in developing the programme of RBMK safety improvements. (author)

  2. Safety indicators as a tool for operational safety evaluation of nuclear power plants

    International Nuclear Information System (INIS)

    Araujo, Jefferson Borges; Melo, Paulo Fernando Ferreira Frutuoso e; Schirru, Roberto

    2009-01-01

    Performance indicators have found a wide use in the conventional and nuclear industries. For the conventional industry, the goal is to optimize production, reducing loss of time with accidents, human error and equipment downtimes. In the nuclear industry, nuclear safety is an additional goal. This paper presents a general methodology to the establishment, selection and use of safety indicators for a two loop PWR plant, as Angra 1. The use of performance indicators is not new. The NRC has its own methodology and the IAEA presents methodology suggestions, but there is no detailed documentation about indicators selection, criteria and bases used. Additionally, only the NRC methodology performs a limited integrated evaluation. The study performed identifies areas considered critical for the plant operational safety. For each of these areas, strategic sub-areas are defined. For each strategic sub-area, specific safety indicators are defined. These proposed Safety Indicators are based on the contribution to risk considering a quantitative risk analysis. For each safety indicator, a goal, a bounded interval and proper bases are developed, to allow for a clear and comprehensive individual behavior evaluation. On the establishment of the intervals and boundaries, a probabilistic safety study, operational experience, international and national standards and technical specifications were used. Additionally, an integrated evaluation of the indicators, using expert systems, was done to obtain an overview of the plant general safety. This evaluation uses well-defined and clear rules and weights for each indicator to be considered. These rules were implemented by means of a computational language, on a friendly interface, so that it is possible to obtain a quick response about operational safety. This methodology can be used to identify situations where the plant safety is challenged, by giving a general overview of the plant operational condition. Additionally, this study can

  3. Improving the safety and reliability of Monju

    International Nuclear Information System (INIS)

    Itou, Kazumoto; Maeda, Hiroshi; Moriyama, Masatoshi

    1998-01-01

    Comprehensive safety review has been performed at Monju to determine why the Monju secondary sodium leakage accident occurred. We investigated how to improve the situation based on the results of the safety review. The safety review focused on five aspects of whether the facilities for dealing with the sodium leakage accident were adequate: the reliability of the detection method, the reliability of the method for preventing the spread of the sodium leakage accident, whether the documented operating procedures are adequate, whether the quality assurance system, program, and actions were properly performed and so on. As a result, we established for Monju a better method of dealing with sodium leakage accidents, rapid detection of sodium leakage, improvement of sodium drain facilities, and way to reduce damage to Monju systems after an accident. We also improve the operation procedures and quality assurance actions to increase the safety and reliability of Monju. (author)

  4. Safety culture in nuclear power plants

    International Nuclear Information System (INIS)

    Tanguy, P.Y.

    1996-01-01

    The development of a good safety culture within the organisation and for the individual poses a sine qua non condition for the operator in the cause of the functioning of his plant. This task must be understood as a linking of the human into a daily management of the safety and quality. Everyone has his role to play at the level of his accountability and his field of responsibility in routine operations and in crisis situations. However, success depends on how management has understood, on the one hand, to convince the staff of the importance of the efforts and the usefulness of the measures taken within the organisation and, on the other hand, to keep this same staff applied to the carrying out of these measures and thus to take account of the anxieties and proposals in the spirit of responsibility and transparency. (author)

  5. Selecting safety standards for nuclear power plants

    International Nuclear Information System (INIS)

    1981-01-01

    Today, many thousands of documents are available describing the requirements, guidelines, and industrial standards which can be used as bases for a nuclear power plant programme. Many of these documents relate to nuclear safety which is currently the focus of world-wide attention. The multitude of documents available on the subject, and their varying status and emphasis, make the processes of selection and implementation very important. Because nuclear power plants are technically intricate and advanced, particularly in relation to the technological status of many developing countries, these processes are also complicated. These matters were the subject of a seminar held at the Agency's headquarters in Vienna last December. The IAEA Nuclear Safety Standards (NUSS) programme was outlined and explained at the Seminar. The five areas of the NUSS programme for nuclear power plants cover, governmental organization, siting, design; operation; quality assurance. In each area the Agency has issued Codes of Practice and is developing Safety Guides. These provide regulatory agencies with a framework for safety. The Seminar recognized that the NUSS programme should enable developing countries to identify priorities in their work, particularly the implementation of safety standards. The ISO activities in the nuclear field are carried out in the framework of its Technical Committee 85 (ISO/TC85). The work is distributed in sub-committees. Seminar on selection and implementation of safety standards for nuclear power plants, jointly organized by the IAEA and the International Organization for Standardization (ISO), and held in Vienna from 15 to 18 December 1980 concerned with: terminology, definitions, units and symbols (SC-1), radiation protection (SC-2), power reactor technology (SC-3), nuclear fuel technology (SC-5). There was general agreement that the ISO standards are complementary to the NUSS codes and guides. ISO has had close relations with the IAEA for several years

  6. Safety and security aspects in design of digital safety I and C in nuclear power plants

    International Nuclear Information System (INIS)

    Ding, Yongjian; Waedt, Karl

    2016-01-01

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  7. Safety and security aspects in design of digital safety I and C in nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Yongjian [University of Applied Sciences Magdeburg-Stendal, Magdeburg (Germany). Inst. of Electrical Engineering; Waedt, Karl [Areva GmbH, Erlangen (Germany). PEAS-G

    2016-05-15

    The paper describes a safety objective oriented systematic design approach of digital (computerized) safety I and C in modern nuclear power plants which considers the plant safety requirements as well as cybersecurity needs. The defence in depth philosophy is applied by using different defence lines in the I and C architecture and protection zones in the plant IT environment.

  8. Improving patient safety: lessons from rock climbing.

    Science.gov (United States)

    Robertson, Nic

    2012-02-01

    How to improve patient safety remains an intractable problem, despite large investment and some successes. Academics have argued that the root of the problem is a lack of a comprehensive 'safety culture' in hospitals. Other safety-critical industries such as commercial aviation invest heavily in staff training to develop such a culture, but comparable programmes are almost entirely absent from the health care sector. In rock climbing and many other dangerous activities, the 'buddy system' is used to ensure that safety systems are adhered to despite adverse circumstances. This system involves two or more people using simple checks and clear communication to prevent problems causing harm. Using this system as an example could provide a simple, original and entertaining way of introducing medical students to the idea that human factors are central to ensuring patient safety. Teaching the buddy system may improve understanding and acceptance of other patient safety initiatives, and could also be used by junior doctors as a tool to improve the safety of their practice. © Blackwell Publishing Ltd 2012.

  9. The value of peer reviews to nuclear plant safety

    International Nuclear Information System (INIS)

    Subalusky, W.T. Jr.

    1994-01-01

    On a global basis, the nuclear utility industry has clearly demonstrated the value of peer reviews for improving nuclear safety and overall plant performance. Peer reviews are conducted by small teams of technical experts who review various aspects of plant operation, recognize strengths and recommend improvements, thereby stimulating a positive response to the recommendations. U.S. nuclear utilities initiated the operator-to-operator peer review process first through the Institute of Nuclear Power Operations (INPO). Now, voluntary peer reviews are an important activity of the World Association of Nuclear Operators (WANO). Formed just five years ago. WANO has made significant progress in its key activities of the operator-to-operator exchanges, operating experience exchange, monitoring of plant performance indicators and sharing of good practices worldwide. A fifth activity, peer review on a strictly voluntary basis, is pertinent to this paper

  10. Safety in Swiss nuclear power plants

    International Nuclear Information System (INIS)

    Cederqvist, H.

    1992-01-01

    Safety-related facilities and equipment are continuously backfitted in Swiss nuclear power plants. In the Beznau-1 and -2 nuclear generating units, the measures taken under the heading of 'Backfitting of Emergency Systems' included provisions to enhance the protection against earthquakes, airplane crash, and fire; in addition, the emergency power system was upgraded. In Muehleberg, the stack exhaust air monitoring system was optimized. The containment pressure suppression system of the plant has been designed to withstand a hypothetical accident exceeding the design basis. The BKM-Crud computer simulation model simulates steps taken to reduce radiation exposure. The power of Swiss nuclear power stations will be raised by 4% to 15% within the 'Energy 2000' action program. (orig.) [de

  11. Nuclear power plant safety under military threats

    International Nuclear Information System (INIS)

    Stritar, A.; Mavko, B.

    1993-01-01

    The Krsko nuclear power plant in Slovenia was probably the first in the world that found itself in the middle of a war region. A number of preventive measures performed during and immediately after the war in summer 1991 by the plant personnel as well as some of related activities of the Reactor Engineering Division of the 'Jozef Stefan' Institute are described. All efforts were aimed at the minimization of the possible public radiological consequences during a military attack. The critical safety functions in such conditions were checked. Concern was further devoted to maintenance of the core cooling in the core and to integrity of the spent fuel pit. The cold shutdown mode was found to be the most appropriate option. After selecting the cold shutdown mode as a most suitable operational state of the plant, further studies of the vulnerability were done. In addition to possible direct damage to structures, that would cause release of radioactivity, two important subjects were considered: AC and DC power supply and the ultimate heat sink. A quick analysis during the crisis has shown that the consequences of a military attack against the plant by jet fighters could be serious, but with the proper preventive measures and preparedness the environmental consequences could be minimized. (Z.S.) 1 fig., 1 ref

  12. Safety Second: the NRC and America's nuclear power plants

    International Nuclear Information System (INIS)

    Adato, M.; MacKenzie, J.; Pollard, R.; Weiss, E.

    1987-01-01

    In 1975, Congress created the Nuclear Regulatory Commission (NRC). Its primary responsibility was to be the regulation of the nuclear power industry in order to maintain public health and safety. On March 28, 1979, in the worst commercial nuclear accident in US history, the plant at Three Mile Island began to leak radioactive material. How was Three Mile Island possible? Where was the NRC? This analysis by the Union of Concerned Scientists (UCS) of the NRC's first decade, points specifically to the factors that contributed to the accident at Three Mile Island. The NRC, created as a watchdog of the nuclear power industry, suffers from problems of mindset, says the UCS. The commission's problems are political, not technical; it repeatedly ranks special interests above the interest of public safety. This book critiques the NRC's performance in four specific areas. It charges that the agency has avoided tackling the most pervasive safety issues; has limited public participation in decision making and power plant licensing; has failed to enforce safety standards or conduct adequate regulation investigations; and, finally, has maintained a fraternal relationship with the industry it was created to regulate, serving as its advocate rather than it adversary. The final chapter offers recommendations for agency improvement that must be met if the NRC is to fulfill its responsibility for safety first

  13. Risk-based safety performance indicators for nuclear power plants

    International Nuclear Information System (INIS)

    Chakraborty, S.; Prohaska, G.; Flodin, Y.; Grint, G.; Habermacher, H.; Hallman, A.; Isasia, R.; Melendez, E.; Verduras, E.; Karsa, Z.; Khatib-Rahbar, M.; Koeberlein, K.; Schwaeger, C.; Matahri, N.; Moravcik, I.; Tkac, M.; Preston, J.

    2003-01-01

    In a Concerted Action (CA), sponsored by the European Commission within its 5th Framework Program, a consortium of eleven partners from eight countries has reviewed and evaluated the application of Safety Performance Indicators (SPIs), which - in combination with other tools - can be used to monitor and improve the safety of nuclear power plants. The project was aimed at identification of methods that can be used in a risk-informed regulatory system and environment, and to exploit PSA techniques for the development and use of meaningful additional/alternative SPIs. The CA included the review of existing indicator systems, and the collection of information on the experience from indicator systems by means of a specific questionnaire. One of the most important and challenging issues for nuclear plant owners and/or regulators is to recognize early signs of deterioration in safety performance, caused by influences from management, organization and safety culture (MOSC), before actual events and/or mishaps take place. Most of the existing SPIs as proposed by various organizations are considered as 'lagging' indicators, that is, they are expected to show an impact only when a downward trend has already started. Furthermore, most of the available indicators are at a relatively high level, such that they will not provide useful information on fundamental weaknesses causing the problem in the first place. Regulators' and utilities' views on the use of a Safety Performance Indicator System have also been a part of the development of the CA. (author)

  14. Balance of plant improvements for future reactor projects

    International Nuclear Information System (INIS)

    Hollingshaus, H.

    1987-01-01

    Many studies have shown that improvements in portions of the plant other than the reactor systems can yield large cost savings during the design, construction, and operation of future reactor power plants. This portion is defined as the Balance of Plant which includes virtually everything except the equipment furnished by the Nuclear Steam Supply System manufacturer. It normally includes the erection of the entire plant including the NSSS. Cost of BOP equipment, engineering and construction work is therefore most of the cost of the plant. Improvements in the BOP have been identified that will substantially reduce nuclear plant cost and construction time while at the same time increasing availability and operability and improving safety. Improvements achieved through standardizatoin, simplification, three-dimensional (3D) computer-aided design, modular construction, innovative construction techniques, and applications for Artificial Intelligence Systems are described. (author)

  15. Plant control impact on IFR power plant passive safety response

    International Nuclear Information System (INIS)

    Vilim, R.B.

    1993-01-01

    A method is described for optimizing the closed-loop plant control strategy with respect to safety margins sustained in the unprotected upset response of a liquid metal reactor. The optimization is performed subject to the normal requirements for reactor startup, load change and compensation for reactivity changes over the cycle. The method provides a formal approach to the process of exploiting the innate self-regulating property of a metal fueled reactor to make it less dependent on operator action and less vulnerable to automatic control system fault and/or operator error

  16. Nuclear power plant licensing: opportunities for improvement

    International Nuclear Information System (INIS)

    1977-06-01

    On April 20, 1977, the Commission directed that recently completed licensing actions be reviewed by the staff for the purpose of identifying ways to improve the effectiveness and efficiency of NRC nuclear power plant licensing activities. This report summarizes the results of a study undertaken by an internal ad hoc Study Group established in response to that directive. The Study Group limited its considerations to safety and environmental review activities. The background, scope, assumptions and objectives of the study are discussed. A prime assumption of this study was that improvements in the efficiency should not be permitted to reduce the current quality achieved in the licensing process. This consideration underlies the conclusions and recommendations of the study

  17. Plutonium finishing plant safety systems and equipment list

    International Nuclear Information System (INIS)

    Bergquist, G.G.

    1995-01-01

    The Safety Equipment List (SEL) supports Analysis Report (FSAR), WHC-SD-CP-SAR-021 and the Plutonium Finishing Plant Operational Safety Requirements (OSRs), WHC-SD-CP-OSR-010. The SEL is a breakdown and classification of all Safety Class 1, 2, and 3 equipment, components, or system at the Plutonium Finishing Plant complex

  18. IAEA Leads Operational Safety Mission to Smolensk Nuclear Power Plant

    International Nuclear Information System (INIS)

    2011-01-01

    training facilities for radiation, fire and industrial safety; A set of handbooks for self-study are available to staff, providing them with an overview of events at plants in Russia and other countries; and There is comprehensive and fast-acting information system on the reactor status, including a detailed assessment of the neutron field in axial and radial directions. The team has also made recommendations and suggestions related to areas where operational safety of Smolensk NPP could be improved. The most significant proposals include the following: To ensure that a plant specific equipment qualification programme is developed and implemented, thus ensuring the capability of the equipment to perform its functions under postulated service conditions, including those arising from accidents; To improve the condition of cables trays and the routing of cables to ensure that the condition of cables is maintained at a high standard; To ensure that the surveillance programme for systems and equipment validates their required safety performance more effectively; and To improve the measurement methodology for the confinement system in order to ensure that the equivalent leak cross section is determined with sufficient accuracy. Smolensk NPP management expressed a determination to address all the areas identified for improvement and requested the IAEA to schedule a follow-up mission in approximately 18 months. The team handed over a draft of their recommendations, suggestions and good practices to the plant management in the form of ''Technical Notes'' for factual comments. The technical notes will be reviewed at the IAEA headquarters including any comments from Smolensk NPP and the Nuclear Regulatory Authority of the Russian Federation. The final report will be submitted to the Government of the Russian Federation within three months. This was the 165th mission of the OSART programme, which began in 1982. OSART missions were performed in the Russian Federation at Balakovo NPP in

  19. Technical Safety Appraisal of the Pinellas Plant

    International Nuclear Information System (INIS)

    1991-01-01

    This report presents the Technical Safety Appraisal (TSA) of the Pinellas Plant in Pinellas County, Florida. The plant is owned and controlled by the US Department of Energy and operated by General Electric Neutron Devices (GEND). The TSA was performed during the period January 15--31, 1989, in support of a Tiger Team Assessment which occurred during the period January 15 to February 2, 1989. The TSA provided the Safety and Health Subteam input to the Tiger Team Assessment. The completion of the assessment process includes: (1) submission of the Team's preliminary findings and concerns, in a Draft Report, to the Manager, Albuquerque Operations Office and to the site contractors at the conclusion of the onsite assessment; (2) review of the Draft Report for technical and factual accuracy; incorporation of the appropriate review comments, suggested changes, and modifications, as well as input from all interested Program Secretarial Offices; preparation of a draft Action Plan by the Albuquerque Operations Office to address the Concerns, and submittal of that Action Plan through the Program Office to ES ampersand H for their review and comment. The Secretary approved the final Action Plan on December 16, 1990, and directed its implementation. The comments and suggestions of the Program Secretarial Offices, the Operations Office, and the site contractor have been incorporated, as appropriate, in this report prior to its publication

  20. How to improve safety of laparoscopic cholecystectomy

    Directory of Open Access Journals (Sweden)

    ZHANG Yong

    2013-06-01

    Full Text Available Laparoscopic cholecystectomy (LC has become the "gold standard" of treatment for benign gallbladder disease. This paper summarizes various surgical safety measures used in recent years, and suggests an emphasis on perioperative imaging examination, preoperative prevention of risk factors, training of surgical skills, and introduction of fast-track surgery concept, so as to avoid the incidence of complications and improve the safety of LC.

  1. Operational safety improvement in OPR 1000

    International Nuclear Information System (INIS)

    Jung, Y.-E.

    2005-01-01

    Nuclear power operating experience management might be an important factor for the operational safety improvement. KHNP's nuclear information management system, called KONIS receives, distributes and manages all nuclear information from domestic and foreign, especially operating experience. Ulchin 3 and 4, the first units of OPR 1000 series operates several organizations regarding management of operating experience e.g. specialist group program, various task forces, equipment specialist system for operator, etc. Peer review is another contribution for nuclear safety. (author)

  2. Selection and verification of safety parameters in safety parameter display system for nuclear power plants

    International Nuclear Information System (INIS)

    Zhang Yuangfang

    1992-02-01

    The method and results for safety parameter selection and its verification in safety parameter display system of nuclear power plants are introduced. According to safety analysis, the overall safety is divided into six critical safety functions, and a certain amount of safety parameters which can represent the integrity degree of each function and the causes of change are strictly selected. The verification of safety parameter selection is carried out from the view of applying the plant emergency procedures and in the accident man oeuvres on a full scale nuclear power plant simulator

  3. IAEA Operational Safety Team Reviews Cattenom Nuclear Power Plant

    International Nuclear Information System (INIS)

    2011-01-01

    , transport and storage of the source; Redundant and diversified telecommunication means are deployed in the various on-site emergency response facilities; and In the event of a severe accident, the support which can be provided to plant staff through a wide range of expertise and analytical tools is commendable. The team has made recommendations and suggestions related to areas where operational safety of Cattenom NPP could be improved. Examples include: Ensuring that all management information, directives and expectations are clearly communicated to all staff and fully implemented; Enhancing the organization of the training programme in the areas of assessment, objectives and competencies; Improving the control of the plant surveillance test programme regarding scheduling and acceptance criteria; and Improving the effectiveness of the plant's Root Cause Analysis process. Cattenom management expressed a determination to address all the areas identified for improvement and requested the IAEA to schedule a follow-up mission in approximately 18 months' time. The team handed over a draft of their recommendations, suggestions and good practices to the plant management in the form of ''Technical Notes'' for factual comments. The technical notes will be reviewed at IAEA headquarters including any comments from Cattenom NPP and the French Safety Authority (ASN). The final report will be submitted to the Government of France within three months. This was the 166th mission of the OSART programme, which began in 1982, and the 23rd such mission in France. General information about OSART missions can be found on the IAEA website OSART Missions. Background An OSART mission is designed as a review of programmes and activities essential to operational safety. It is not a regulatory inspection, nor is it a design review or a substitute for an exhaustive assessment of the plant's overall safety status. Experts participating in the IAEA's June 2010 International Conference on Operational

  4. Technical Safety Appraisal of the Rocky Flats Plant

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Blake P.

    1989-01-01

    This report provides the results of a Technical Safety Appraisal (TSA) of the Rocky Flats Plant (RFP) conducted November 14 to 18 and November 28 to December 9, 1988. This appraisal covered the effectiveness and improvements in the RFP safety program across the site, evaluating progress to date against standards of accepted practice. The appraisal included coverage of the timeliness and effectiveness of actions taken in response to the recommendations/concerns in three previous Technical Safety Appraisals (TSAs) of RFP Bldg. 707 conducted in July 1986, Bldgs. 771/774 conducted in October/November 1986, and Bldgs. 776/777 conducted in January/February 1988. Results of this appraisal are given in Section IV for each of 14 technical safety areas at RFP. These results include a discussion, conclusions and any new safety concerns for each technical safety area. Appendix A contains a description of the system for categorizing concerns, and the concerns are tabulated in Appendix B. Appendix C reports on the evaluation of the contractor's actions and the current status of each of the 230 recommendations and concerns contained in the three previous TSA reports.

  5. Safety related requirements on future nuclear power plants

    International Nuclear Information System (INIS)

    Niehaus, F.

    1991-01-01

    Nuclear power has the potential to significantly contribute to the future energy supply. However, this requires continuous improvements in nuclear safety. Technological advancements and implementation of safety culture will achieve a safety level for future reactors of the present generation of a probability of core-melt of less than 10 -5 per year, and less than 10 -6 per year for large releases of radioactive materials. There are older reactors which do not comply with present safety thinking. The paper reviews findings of a recent design review of WWER 440/230 plants. Advanced evolutionary designs might be capable of reducing the probability of significant off-site releases to less than 10 -7 per year. For such reactors there are inherent limitations to increase safety further due to the human element, complexity of design and capability of the containment function. Therefore, revolutionary designs are being explored with the aim of eliminating the potential for off-site releases. In this context it seems to be advisable to explore concepts where the ultimate safety barrier is the fuel itself. (orig.) [de

  6. Knowledge Management Methodologies for Improving Safety Culture

    International Nuclear Information System (INIS)

    Rusconi, C.

    2016-01-01

    Epistemic uncertainties could affect operator’s capability to prevent rare but potentially catastrophic accident sequences. Safety analysis methodologies are powerful but fragile tools if basic assumptions are not sound and exhaustive. In particular, expert judgments and technical data could be invalidated by organizational context change (e.g., maintenance planning, supply systems etc.) or by unexpected events. In 1986 accidents like Chernobyl, the explosion of Shuttle Challenger and, two years before, the toxic release at Bhopal chemical plant represented the point of no return with respect to the previous vision of safety and highlighted the undelayable need to change paradigm and face safety issues in complex systems not only from a technical point of view but to adopt a systemic vision able to include and integrate human and organizational aspects.

  7. Safety and Radiation Protection at Swedish Nuclear Power Plants 2007

    International Nuclear Information System (INIS)

    2008-01-01

    The safety level of the plants is maintained at an acceptable level. SKI has in its regulatory supervision not found any known deficiencies in the barriers which could result in release of radioactive substances in excess of the permitted levels. SKI considers that improvements have been implemented during the year in the management, control and following up of safety work at the plants. In some cases, however, SKI has imposed requirements that improvements be made. Extensive measures are under way at the nuclear power plants to comply with the safety requirements in SKI's regulations, SKIFS 2004:2 concerning the design and construction of nuclear power reactors, and the stricter requirements regarding physical protection. Concurrently preparations are underway at eight of the ten units for thermal power increases. At the Forsmark plant considerable efforts have been during the year to correct the deficiencies in the safety culture and quality assurance system that became apparent in 2006. A programme to improve the execution of activities has been established in accordance with SKI's decision. SKI considers that the plant has developed in a positive direction but that there are further possibilities for improvement with regard to internal control. This is amongst other things concerns the areas internal auditing, independent safety review function, and working methods. SKI has had special supervision of the plant since 28 September, 2006. At the Oskarshamn plant work has been carried out to improve the organisation and routines in several areas. The plant has established routines which provide the basis to ensure that decisions are taken in a stringent manner. The quality assurance system has a clearer structure and there is a better defined division of work. Some measures remain however to be dealt with in 2008. The Ringhals plant has also worked with attitudes to routines and internal control. SKI considers that the measures have good prerequisites to provide a

  8. Safety principles and design management of Chashma Nuclear Power Plant

    International Nuclear Information System (INIS)

    Geng Qirui; Cheng Pingdong

    1997-01-01

    The basic safety consideration and detailed design principles in the design of Chashma Nuclear Power Plant is elaborated. The management within the frame setting up by 'safety culture' and 'quality culture'

  9. Biotechnological improvement of ornamental plants

    Directory of Open Access Journals (Sweden)

    Flavia Soledad Darqui

    2017-10-01

    Full Text Available The discovery of commercial transgenic varieties of orange petunias sold in Europe and the United States although they had never reached the approved status, and the consequent recommendation to destroy them, was the trigger to discuss about biotechnological improvement of ornamental plants. Inside the restricted world of 26 vegetal transgenic species, according to the ISAAA’s reports (http://www.isaaa.org, there are three ornamental species: carnation, rose and the Beijing University developed petunia; all of them with the same trait, a change in their colour. On the other hand, in 2014, the whole-genome sequence of carnation appeared which was the first and until now the only one among ornamental species. In this context, we review the publications from the last five years in petunia, rose, chrysanthemum and carnation. In these papers there are detailed descriptions of modification of the cascade of genes and transcription factors involved in stress situations, in different developmental stages and their regulation through different plant hormones. This knowledge will allow breeding for better and new varieties with changes in their abiotic or biotic stress tolerance, altered growth or yield and modified product quality as colour or fragrance.

  10. The safety of the new reprocessing plants of La Hague

    International Nuclear Information System (INIS)

    Devillers, C.; Dubois, G.

    1987-09-01

    In this document the authors show the main guiding lines on which is based the safety of the new reprocessing plant of La Hague. They are: - the objectives: to limit the impacts on workers and environment - the methods: safety analysis based on the checking and evaluation of significant risks. - the means: to make a safety plant by the use of quality assurance in the conception and in the plant construction [fr

  11. Use of PSA for improving the safety of French PWRs

    International Nuclear Information System (INIS)

    Lanore, J.M.; Chambon, J.L.

    1994-06-01

    Two French PWR Probabilistic Safety Assessment (PSA) studies were conducted for the standardized PWR series of 900 and 1300 MWe. Both PSA 900 and PSA 1300 are level 1 PSAs, that means their objective is the evaluation of core meltdown frequency. These studies have some specific features, in particular the treatment of shutdown conditions, the treatment of long term post-accidental situations, and a wide use of French experience feedback. The PSAs are used for safety improvements of the French PWRs. Following the PSA results, several modifications to plants concerning the dominant sequences were decided. (R.P.). 2 refs., 4 figs

  12. Systematic safety evaluation of old nuclear power plants

    International Nuclear Information System (INIS)

    Dredemis, G.; Fourest, B.

    1984-01-01

    The French safety authorities have undertaken a systematic evaluation of the safety of old nuclear power plants. Apart from a complete revision of safety documents (safety analysis report, general operating rules, incident and accident procedures, internal emergency plan, quality organisation manual), this examination consisted of analysing the operating experience of systems frequently challenged and a systematic examination of the safety-related systems. This paper is based on an exercise at the Ardennes Nuclear Power Plant which has been in operation for 15 years. This paper also summarizes the main surveys and modifications relating to this power plant. (orig.)

  13. Code on the safety of nuclear power plants: Design

    International Nuclear Information System (INIS)

    1988-01-01

    This Code is a compilation of nuclear safety principles aimed at defining the essential requirements necessary to ensure nuclear safety. These requirements are applicable to structures, systems and components, and procedures important to safety in nuclear power plants embodying thermal neutron reactors, with emphasis on what safety requirements shall be met rather than on specifying how these requirements can be met. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants. The document should be used by organizations designing, manufacturing, constructing and operating nuclear power plants as well as by regulatory bodies

  14. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Spanish Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  15. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2012-01-01

    This publication is a revision of Safety Requirements No. NS-R-1, Safety of Nuclear Power Plants: Design. It establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in the design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. Contents: 1. Introduction; 2. Applying the safety principles and concepts; 3. Management of safety in design; 4. Principal technical requirements; 5. General plant design; 6. Design of specific plant systems.

  16. Safety and Radiation Protection at Swedish Nuclear Power Plants 2007

    Energy Technology Data Exchange (ETDEWEB)

    2008-07-01

    The safety level of the plants is maintained at an acceptable level. SKI has in its regulatory supervision not found any known deficiencies in the barriers which could result in release of radioactive substances in excess of the permitted levels. SKI considers that improvements have been implemented during the year in the management, control and following up of safety work at the plants. In some cases, SKI has imposed requirements that improvements be made. Extensive measures are under way at the nuclear power plants to comply with the safety requirements in SKI's regulations, SKIFS 2004:2 concerning the design and construction of nuclear power reactors, and the stricter requirements regarding physical protection. Concurrently preparations are underway at eight of the ten units for thermal power increases. At the Forsmark plant considerable efforts have been during the year to correct the deficiencies in the safety culture and quality assurance system that became apparent in 2006. A programme to improve the execution of activities has been established in accordance with SKI's decision. SKI considers that the plant has developed in a positive direction but that there are further possibilities for improvement with regard to internal control. This is amongst other things concerns the areas internal auditing, independent safety review function, and working methods. SKI has had special supervision of the plant since 28 September, 2006. At the Oskarshamn plant work has been carried out to improve the organisation and routines in several areas. The plant has established routines which provide the basis to ensure that decisions are taken in a stringent manner. The quality assurance system has a clearer structure and there is a better defined division of work. Some measures remain to be dealt with in 2008. The Ringhals plant has also worked with attitudes to routines and internal control. SKI considers that the measures have good prerequisites to provide a

  17. Nuclear safety at the Paks Plant

    International Nuclear Information System (INIS)

    Bajsz, Jozsef; Vamos, Gabor

    1991-01-01

    The Paks Nuclear Power Plant is located on the Danube river 114 km south of Budapest. It consists of four PWR units of the Soviet VVER-440 design. These are of the second generation design VVER 440 (model 213) with safety features as of 1975. It should be emphasized that these are two different generations of VVER 440 units. This is not always clear, not only to the general public, but sometimes even to people working in the nuclear industry. The widespread criticism of the first generation type 230 reactors is often extended to model 213 reactors, as the differences between the two models are often not sufficiently emphasized. In this situation it is very important to provide balanced information about the advantages and disadvantages of this reactor type. This paper attempts to do that. (author)

  18. Safety of WWER type nuclear power plants - viewing from Hungary

    International Nuclear Information System (INIS)

    Voeroess, L.

    1991-01-01

    An evaluation of WWER type nuclear power plants operating in Hungary is given, relative to the safety requirements accepted internationally; how safe can they be regarded and what can be done to assure a high level of safety in all case. After an overview of general safety criteria, an overall description of WWER-440 type nuclear reactors is presented. Design safety, operational safety issues are treated in detail. Safety inspection and safety-related research and development is discussed. Regarding the future, five different issues associated with nuclear reactor safety should be considered. (R.P.) 20 refs.; 12 figs.; 3 tabs

  19. Patient Education May Improve Perioperative Safety.

    NARCIS (Netherlands)

    de Haan, L.S.; Calsbeek, H; Wolff, André

    2016-01-01

    Importance: There is a growing interest in enabling ways for patients to participate in their own care to improve perioperative safety, but little is known about the effectiveness of interventions enhancing an active patient role. Objective: To evaluate the effect of patient participation on

  20. Plant designer's view of the operator's role in nuclear plant safety

    International Nuclear Information System (INIS)

    Corcoran, W.R.; Church, J.F.; Cross, M.T.; Porter, N.J.

    1981-01-01

    The nuclear plant operator's role supports the design assumptions and equipment with four functional tasks. He must set up th plant for predictable response to disturbances, operate the plant so as to minimize the likelihood and severity of event initiators, assist in accomplishing the safety functions, and feed back operating experiences to reinforce or redefine the safety analyses' assumptions. The latter role enhances the operator effectiveness in the former three roles. The Safety Level Concept offers a different perspective that enables the operator to view his roles in nuclear plant safety. This paper outlines the operator's role in nuclear safety and classifies his tasks using the Safety Level Concept

  1. Safety guide on fire protection in nuclear power plants

    International Nuclear Information System (INIS)

    1976-01-01

    The purpose of the Safety Guide is to give specific design and operational guidance for protection from fire and explosion in nuclear power plants, based on the general guidance given in the relevant sections of the 'Safety Code of Practice - Design' and the 'Safety Code of Practice - Operation' of the International Atomic Energy Agency. The guide will confine itself to fire protection of safety systems and items important to safety, leaving the non-safety matters of fire protection in nuclear power plants to be decided upon the basis of the various available national and international practices and regulations. (HP) [de

  2. Nuclear power plants: a unique challenge to fire safety

    International Nuclear Information System (INIS)

    Nowlen, S.P.

    1992-01-01

    The evaluation of fire safety in a nuclear power plant must include the consideration of the impact of a fire on the operability of plant safety equipment and systems. This issue is not typical of the life safety and property protection issues which dominate traditional fire safety concerns. This paper provides a general discussion of the issue of nuclear power plant fire safety as it currently exists in the USA. Included is a discussion of the past history of nuclear power plant fire events, the development of nuclear industry specific fire safety guidelines, the adverse experience associated with the inadvertent operation of fire suppression systems, and the anticipated direction of fire safety requirements for future reactor designs in the USA. (Author)

  3. Safety problems in decommissioning nuclear power plants

    International Nuclear Information System (INIS)

    Auler, I.; Bardtenschlager, R.; Gasch, A.; Majohr, N.

    1975-12-01

    The safety problems at decommissioning are illustrated by the example of a LWR with 1300 MW electric power after 40 years of specified normal operation. For such a facility the radioactivity in the form of activation and contamination one year after being finally taken out of service is in the order of magnitude of 10 7 Ci, not counting the fuel assemblies. The dose rates occurring during work on the reactor vessel at nozzle level may amount to some 10 4 rem/h. After a rough estimation the accumulated dose for the decommissioning personnel during total dismantling will be about 1200 rem. During performance of the decommissioning activities the problems are mainly caused by direct radiation of the active components and systems and by the release of radioactive particles, aerosols and liquids if these components are crushed. The extent of later dismantling problems may be reduced by selecting appropriate materials as well as considering the requirements for dismantling in design and arrangement of the components already in the design stage of new facilities. Apart from plant design also the concept for the disposal of the radioactive waste from decommissioning will provide important boundary conditions. E.g. the maximum size of the pieces to be stored in the ultimate storage place will very much influence the dose expenditure for handling these parts. For complete dismantling of nuclear power plants an ultimate store must be available where large amounts of bulky decommissioning waste, containing relatively low activity, can be stored. The problems and also the cost for decommissioning may be considerably reduced by delaying complete disposal of the radioactive material >= 40 years and during this period, keeping the radioactivity enclosed within the plant in the form of a safe containment. (orig./HP) [de

  4. Safety analysis of Oi nuclear power plant

    International Nuclear Information System (INIS)

    1979-01-01

    The transient phenomena in Oi nuclear power plant were analyzed, especially on the water level fluctuation and the capability of natural circulation in the primary loop, under the assumptions that the feed water for steam generators is totally lost, and the relief valve on the pressurizer, which is actuated due to the pressure rise in the primary system, is stuck and kept open. These assumptions are related to the TMI accident. The analysing conditions are 1) the main feed water flow is totally lost suddenly during the rated power operation of the reactor, 2) two motor-driven auxiliary feed water pumps are started manually fifteen minutes after the accident initiation, 3) one relief valve on the pressurizer is opened fifteen seconds after the accident initiation and kept open, 4) the reactor is scrammed thirty three seconds after the accident initiation, 5) the turbine is tripped 33.5 seconds after the accident initiation, etc. Two cases were analysed, namely 3,800 seconds and 1,200 seconds after the accident initiation. The analytical code RELEP4/Mod5/U2/J1 was utilized for this analysis. The level fluctuation in the pressurizer after the accident initiation, the flow rate fluctuation through the pressurizer relief valve, especially that of steam, liquid single phase and two phase flows, the water level in the upper plenum in the pressure vessel, the change of flow rate at core inlet, the average pressure in the core, and the temperature fluctuation of coolant in the core, the variation of void fraction in the core, and the change of surface temperature of fuel rods are presented as the analysis results, and they are evaluated. It is recognized that the plant safety is kept under the assumed accident conditions in the Oi nuclear power plant. (Nakai, Y.)

  5. 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

  6. Inherent safety features in balance-of-plant layout

    International Nuclear Information System (INIS)

    Wattelet, P.L.; Green, K.J.

    1992-01-01

    Future nuclear units must be more economical to construct and operate, and, at the same time, clearly incorporate advances in safety over the current generation of light water reactors. To achieve these goals, the root causes of safety issues must be addressed. In this way, global, cost-effective solutions can be implemented. With simple, direct design approaches, the licensing risk is minimized and configuration control is enhanced. With proper planning in the early stages of plant design, postulated accidents and events can often be mitigated by passive features inherent in the basic structure and layout, eliminating expensive added protective structures and components often found in current designs. Korea Electric Power Corporation's Yonggwang (YGN) Units 3 and 4, shown in an artist's rendering in Figure 1, are now under construction in Korea. Engineering is more than 85% complete, and Unit 3 construction is more than 50% complete. Significant steps toward design simplification and safety enhancement have been made by addressing safety concerns very early in the design effort. The tools used to achieve this were improved symmetry and separation, isolation of potential hazards, and an improved design process

  7. Improving patient safety in radiation oncology

    International Nuclear Information System (INIS)

    Hendee, William R.; Herman, Michael G.

    2011-01-01

    Beginning in the 1990s, and emphasized in 2000 with the release of an Institute of Medicine report, healthcare providers and institutions have dedicated time and resources to reducing errors that impact the safety and well-being of patients. But in January 2010 the first of a series of articles appeared in the New York Times that described errors in radiation oncology that grievously impacted patients. In response, the American Association of Physicists in Medicine and the American Society of Radiation Oncology sponsored a working meeting entitled ''Safety in Radiation Therapy: A Call to Action''. The meeting attracted 400 attendees, including medical physicists, radiation oncologists, medical dosimetrists, radiation therapists, hospital administrators, regulators, and representatives of equipment manufacturers. The meeting was cohosted by 14 organizations in the United States and Canada. The meeting yielded 20 recommendations that provide a pathway to reducing errors and improving patient safety in radiation therapy facilities everywhere.

  8. Improvement programme of safety performance indicators (SPIs) in Korea

    International Nuclear Information System (INIS)

    Lee, S.Y.

    2001-01-01

    KINS has developed and used Safety Performance Indicators (SPIs), which are count based and composed of 10 indicators in 8 areas, to monitor the trend of performance of NPPs in Korea since 1997. However, the limited usage of SPIs and the increasing worldwide interest on SPIs became the motivation of the SPI improvement programme in Korea. Korea is planning to establish plant performance evaluation programme through analysis of SPI and result of inspection. The SPI improvement programme is a part of the plant performance evaluation programme and includes study on performance evaluation areas, indicator categories, selection and development of indicators, redefinition of indicators and introduction of graphical display system. The selected performance evaluation areas are general performance, reactor safety and radiation safety. Each area will have categories as sub-areas and a total of six categories are selected. One or two indicators for each category are determined or will be developed to make a set of Safety Performance Indicators. Also, a graphic display system will be introduced to extend the usage of SPIs. (author)

  9. Continuous improvement of the MHTGR safety and competitive performance

    International Nuclear Information System (INIS)

    Eichenberg, T.W.; Etzel, K.T.; Mascaro, L.L.; Rucker, R.A.

    1992-05-01

    An increase in reactor module power from 350 to 450 MW(t) would markedly improve the economics of the Modular High Temperature Gas-Cooled Reactor (MHTGR). The higher power level was recommended as the result of an in-depth cost reduction study undertaken to compete with the declining price of fossil fuel. The safety assessment confirms that the high level of safety, which relies on inherent characteristics and passive features, is maintained at the elevated power level. Preliminary systems, nuclear, and safety performance results are discussed for the recommended 450 MW(t) design. Optimization of plant parameters and design modifications accommodated the operation of the steam generator and circulator at the higher power level. Events in which forced cooling is lost, designated as conduction cooldowns are described in detail. For the depressurized conduction cooldown, without full helium inventory, peak fuel temperatures are significantly lowered. A more negative temperature coefficient of reactivity was achieved while maintaining an adequate fuel cycle and reactivity control. Continual improvement of the MHTGR delivers competitive performance without relinquishing the high safety margins demanded of the next generation of power plants

  10. Effect of generic issues program on improving safety

    International Nuclear Information System (INIS)

    Fard, M. R.; Kauffman, J. V.

    2010-01-01

    The U.S. Nuclear Regulatory Commission (NRC) identifies (by its assessment of plant operation) certain issues involving public health and safety, the common defense and security, or the environment that could affect multiple entities under NRC jurisdiction. The Generic Issues Program (GIP) addresses the resolution of these Generic Issues (GIs). The resolution of these issues may involve new or revised rules, new or revised guidance, or revised interpretation of rules or guidance that affect nuclear power plant licensees, nuclear material certificate holders, or holders of other regulatory approvals. U.S. NRC provides information related to the past and ongoing GIP activities to the general public by the use of three main resources, namely NUREG-0933, 'Resolution of Generic Safety Issues, ' Generic Issues Management Control System (GIMCS), and GIP public web page. GIP information resources provide information such as historical information on resolved GIs, current status of the open GIs, policy documents, program procedures, GIP annual and quarterly reports and the process to contact GIP and propose a GI This paper provides an overview of the GIP and several examples of safety improvements resulting from the resolution of GIs. In addition, the paper provides a brief discussion of a few recent GIs to illustrate how the program functions to improve safety. (authors)

  11. General design safety principles for nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide provides the safety principles and the approach that have been used to implement the Code in the Safety Guides. These safety principles and the approach are tied closely to the safety analyses needed to assist the design process, and are used to verify the adequacy of nuclear power plant designs. This Guide also provides a framework for the use of other design Safety Guides. However, although it explains the principles on which the other Safety Guides are based, the requirements for specific applications of these principles are mostly found in the other Guides

  12. Safety criteria for design of nuclear power plants

    International Nuclear Information System (INIS)

    1997-01-01

    In Finland the general safety requirements for nuclear power plants are presented in the Council of State Decision (395/91). In this guide, safety principles which supplement the Council of State Decision and which are to be used in the design of nuclear power plants are defined

  13. Code on the safety of nuclear power plants: Governmental organization

    International Nuclear Information System (INIS)

    1988-01-01

    This Code recommends requirements for a regulatory body responsible for regulating the siting, design, construction, commissioning, operation and decommissioning of nuclear power plants for safety. It forms part of the Agency's programme for establishing Codes and Safety Guides relating to land based stationary thermal neutron power plants

  14. Safety problems in fuel reprocessing plants

    International Nuclear Information System (INIS)

    Amaury, P.; Jouannaud, C.; Niezborala, F.

    1979-01-01

    The document first situates the reprocessing in the fuel cycle as a whole. It shows that a large reprocessing plant serves a significant number of reactors (50 for a plant of 1500 tonnes per annum). It then assesses the potential risks with respect to the environment as well as with respect to the operating personnel. The amounts of radioactive matter handled are very significant and their easily dispersible physical form represents very important risks. But the low potential energy likely to bring about this dispersion and the very severe and plentiful confinement arrangements are such that the radioactive risks are very small, both with respect to the environment and the operating personnel. The problems of the interventions for maintenance or repairs are mentioned. The intervention techniques in a radioactive environment are perfected, but they represent the main causes of operating personnel irradiation. The design principle applied in the new plants take this fact into account, involving a very significant effort to improve the reliability of the equipment and ensuring the provision of devices enabling the failing components to be replaced without causing irradiation of the personnel [fr

  15. IAEA Sees Safety Commitment at Spain’s Almaraz Nuclear Power Plant, Areas for Enhancement

    International Nuclear Information System (INIS)

    2018-01-01

    An International Atomic Energy Agency (IAEA) team of experts said the operator of Spain’s Almaraz Nuclear Power Plant demonstrated a commitment to the long-term safety of the plant and noted several good practices to share with the nuclear industry globally. The team also identified areas for further enhancement. The Operational Safety Review Team (OSART) today concluded an 18-day mission to Almaraz, whose two 1,050-MWe pressurized-water reactors started commercial operation in 1983 and 1984, respectively. Centrales Nucleares Almaraz-Trillo (CNAT) operates the plant, located about 200 km southwest of Madrid. OSART missions aim to improve operational safety by objectively assessing safety performance using the IAEA’s safety standards and proposing recommendations for improvement where appropriate. Nuclear power generates more than 21 per cent of electricity in Spain, whose seven operating power reactors all began operation in the 1980s.The mission made a number of recommendations to improve operational safety, including: • The plant should implement further actions related to management, staff and contractors to enforce standards and expectations related to industrial safety. • The plant should take measures to reinforce and implement standards to enhance the performance of reactivity manipulations in a deliberate and carefully-controlled manner. • The plant should improve the support, training and documented guidance for Severe Accident Management Guideline users in order to mitigate complex severe accident scenarios. The team provided a draft report of the mission to the plant’s management. The plant management and the Nuclear Safety Council (CSN), which is responsible for nuclear safety oversight in Spain, will have the opportunity to make factual comments on the draft. These will be reviewed by the IAEA and the final report will be submitted to the Government of Spain within three months. The plant management said it would address the areas

  16. Latest developments on safety analysis methodologies at the Juzbado plant

    International Nuclear Information System (INIS)

    Zurron-Cifuentes, Oscar; Ortiz-Trujillo, Diego; Blanco-Fernandez, Luis A.

    2010-01-01

    Over the last few years the Juzbado Plant has developed and implemented several analysis methodologies to cope with specific issues regarding safety management. This paper describes the three most outstanding of them, so as to say, the Integrated Safety Analysis (ISA) project, the adaptation of the MARSSIM methodology for characterization surveys of radioactive contamination spots, and the programme for the Systematic Review of the Operational Conditions of the Safety Systems (SROCSS). Several reasons motivated the decision to implement such methodologies, such as Regulator requirements, operational experience and of course, the strong commitment of ENUSA to maintain the highest standards of nuclear industry on all the safety relevant activities. In this context, since 2004 ENUSA is undertaking the ISA project, which consists on a systematic examination of plant's processes, equipment, structures and personnel activities to ensure that all relevant hazards that could result in unacceptable consequences have been adequately evaluated and the appropriate protective measures have been identified. On the other hand and within the framework of a current programme to ensure the absence of radioactive contamination spots on unintended areas, the MARSSIM methodology is being applied as a tool to conduct the radiation surveys and investigation of potentially contaminated areas. Finally, the SROCSS programme was initiated earlier this year 2009 to assess the actual operating conditions of all the systems with safety relevance, aiming to identify either potential non-conformities or areas for improvement in order to ensure their high performance after years of operation. The following paragraphs describe the key points related to these three methodologies as well as an outline of the results obtained so far. (authors)

  17. Method of safety evaluation in nuclear power plants

    International Nuclear Information System (INIS)

    Kuraszkiewicz, P.; Zahn, P.

    1988-01-01

    A novel quantitative technique for evaluating safety of subsystems of nuclear power plants based on expert estimations is presented. It includes methods of mathematical psychology recognizing the effect of subjective factors in the expert estimates and, consequently, contributes to further objectification of evaluation. It may be applied to complementing probabilistic safety assessment. As a result of such evaluations a characteristic 'safety of nuclear power plants' is obtained. (author)

  18. Strengthening of nuclear power plant construction safety management

    International Nuclear Information System (INIS)

    Yu Jun

    2012-01-01

    The article describes the warning of the Fukushima nuclear accident, and analyzes the major nuclear safety issues in nuclear power development in China, problems in nuclear power plants under construction, and how to strengthen supervision and management in nuclear power construction. It also points out that the development of nuclear power must attach great importance to the safety, and nuclear power plant construction should strictly implement the principle of 'safety first and quality first'. (author)

  19. Probabilistic safety assessment in nuclear power plant management

    International Nuclear Information System (INIS)

    Holloway, N.J.

    1989-06-01

    Probabilistic Safety Assessment (PSA) techniques have been widely used over the past few years to assist in understanding how engineered systems respond to abnormal conditions, particularly during a severe accident. The use of PSAs in the design and operation of such systems thus contributes to the safety of nuclear power plants. Probabilistic safety assessments can be maintained to provide a continuous up-to-date assessment (Living PSA), supporting the management of plant operations and modifications

  20. Operating procedure automation to enhance safety of nuclear power plants

    International Nuclear Information System (INIS)

    Husseiny, A.A.; Sabri, Z.A.; Adams, S.K.; Rodriguez, R.J.; Packer, D.; Holmes, J.W.

    1989-01-01

    Use of logic statements and computer assist are explored as means for automation and improvement on design of operating procedures including those employed in abnormal and emergency situations. Operating procedures for downpower and loss of forced circulation are used for demonstration. Human-factors analysis is performed on generic emergency operating procedures for three strategies of control; manual, semi-automatic and automatic, using standard emergency operating procedures. Such preliminary analysis shows that automation of procedures is feasible provided that fault-tolerant software and hardware become available for design of the controllers. Recommendations are provided for tests to substantiate the promise of enhancement of plant safety. Adequate design of operating procedures through automation may alleviate several major operational problems of nuclear power plants. Also, automation of procedures is necessary for partial or overall automatic control of plants. Fully automatic operations are needed for space applications while supervised automation of land-based and offshore plants may become the thrust of new generation of nulcear power plants. (orig.)

  1. Safety improvement and results of commissioning of Mochovce NPP WWER 440/213

    International Nuclear Information System (INIS)

    Lipar, M.

    1998-01-01

    Mochovce NPP is the last one of this kind and compared to its predecessors, it is characterized by several modifications which contribute to the improvement of the safety level. In addition based on Nuclear Regulatory Authority requirements and based on documents: - IAEA - Safety Issues and their ranking for NPP WWER 440/213, - IAEA - Safety Improvement of Mochovce NPP Project Review Mission, - Riskaudit - Evaluation of the Mochovce NPP Safety Improvements. Additional safety measures have been implemented before commissioning. The consortium EUCOM (FRAMATOME - SIEMENS), SKODA Praha, ENERGOPROJEKT Praha, Russian organizations and VUJE Trnava Nuclear Power Plants research institute were selected for design and implementation of the safety measures. The papers summarized, safety requirements, safety measures implemented, results of commissioning and results of safety analysis report evaluation. (author)

  2. Instrumentation and control systems important to safety in nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. It supplements Safety Standards Series No. NS-R-1: Safety of Nuclear Power Plants: Design (the Requirements for Design), which establishes the design requirements for ensuring the safety of nuclear power plants. This Safety Guide describes how the requirements should be met for instrumentation and control (I and C) systems important to safety. This publication is a revision and combination of two previous Safety Guides: Safety Series Nos 50-SG-D3 and 50-SG-D8, which are superseded by this new Safety Guide. The revision takes account of developments in I and C systems important to safety since the earlier Safety Guides were published in 1980 and 1984, respectively. The objective of this Safety Guide is to provide guidance on the design of I and C systems important to safety in nuclear power plants, including all I and C components, from the sensors allocated to the mechanical systems to the actuated equipment, operator interfaces and auxiliary equipment. This Safety Guide deals mainly with design requirements for those I and C systems that are important to safety. It expands on paragraphs of Ref in the area of I and C systems important to safety. This publication is intended for use primarily by designers of nuclear power plants and also by owners and/or operators and regulators of nuclear power plants. This Safety Guide provides general guidance on I and C systems important to safety which is broadly applicable to many nuclear power plants. More detailed requirements and limitations for safe operation specific to a particular plant type should be established as part of the design process. The present guidance is focused on the design principles for systems important to safety that warrant particular attention, and should be applied to both the design of new I and C systems and the modernization of existing systems. Guidance is provided on how design

  3. The NPPR Trnava participation in the NPP V-2 modernisation and safety improvement project

    International Nuclear Information System (INIS)

    Michal, V.; Losonsky, B.; Magdolen, J.

    1999-01-01

    The presented contribution deals with form, present state and results of Nuclear Power Plants Research Inst.e participation in the NPP V-2 Jaslovske Bohunice Modernization and Safety Improvement Project.(author)

  4. Housing improvement and home safety Effectiveness Matters

    OpenAIRE

    , Crd; Sphr@, L; , MrcSphsu

    2014-01-01

    The homes we live in impact on health, wellbeing and health inequalities. Treating illnesses directly related to living in cold, damp and dangerous homes costs the NHS £2.5 billion per year. Ensuring affordable warmth through insulation and more efficient heating can improve health and wellbeing. Home safety assessment and modification can reduce falls and risk of falling in older people. Education, promotion of exercise and wearing of appropriate footwear, environmental modifications and tra...

  5. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  6. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2005-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  7. The operating organization for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    This Safety Guide was prepared under the IAEA programme for safety standards for nuclear power plants. The present publication is a revision of the IAEA Safety Guide on Management of Nuclear Power Plants for Safe Operation issued in 1984. It supplements Section 2 of the Safety Requirements publication on Safety of Nuclear Power Plants: Operation. Nuclear power technology is different from the customary technology of power generation from fossil fuel and by hydroelectric means. One major difference between the management of nuclear power plants and that of conventional generating plants is the emphasis that should be placed on nuclear safety, quality assurance, the management of radioactive waste and radiological protection, and the accompanying national regulatory requirements. This Safety Guide highlights the important elements of effective management in relation to these aspects of safety. The attention to be paid to safety requires that the management recognize that personnel involved in the nuclear power programme should understand, respond effectively to, and continuously search for ways to enhance safety in the light of any additional requirements socially and legally demanded of nuclear energy. This will help to ensure that safety policies that result in the safe operation of nuclear power plants are implemented and that margins of safety are always maintained. The structure of the organization, management standards and administrative controls should be such that there is a high degree of assurance that safety policies and decisions are implemented, safety is continuously enhanced and a strong safety culture is promoted and supported. The objective of this publication is to guide Member States in setting up an operating organization which facilitates the safe operation of nuclear power plants to a high level internationally. The second objective is to provide guidance on the most important organizational elements in order to contribute to a strong safety

  8. Improving patient safety through quality assurance.

    Science.gov (United States)

    Raab, Stephen S

    2006-05-01

    Anatomic pathology laboratories use several quality assurance tools to detect errors and to improve patient safety. To review some of the anatomic pathology laboratory patient safety quality assurance practices. Different standards and measures in anatomic pathology quality assurance and patient safety were reviewed. Frequency of anatomic pathology laboratory error, variability in the use of specific quality assurance practices, and use of data for error reduction initiatives. Anatomic pathology error frequencies vary according to the detection method used. Based on secondary review, a College of American Pathologists Q-Probes study showed that the mean laboratory error frequency was 6.7%. A College of American Pathologists Q-Tracks study measuring frozen section discrepancy found that laboratories improved the longer they monitored and shared data. There is a lack of standardization across laboratories even for governmentally mandated quality assurance practices, such as cytologic-histologic correlation. The National Institutes of Health funded a consortium of laboratories to benchmark laboratory error frequencies, perform root cause analysis, and design error reduction initiatives, using quality assurance data. Based on the cytologic-histologic correlation process, these laboratories found an aggregate nongynecologic error frequency of 10.8%. Based on gynecologic error data, the laboratory at my institution used Toyota production system processes to lower gynecologic error frequencies and to improve Papanicolaou test metrics. Laboratory quality assurance practices have been used to track error rates, and laboratories are starting to use these data for error reduction initiatives.

  9. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  10. Safety of Nuclear Power Plants: Commissioning and Operation. Specific Safety Requirements

    International Nuclear Information System (INIS)

    2016-01-01

    This publication describes the requirements to be met to ensure the safe operation of nuclear power plants. It takes into account developments in areas such as long term operation of nuclear power plants, plant ageing, periodic safety review, probabilistic safety analysis and risk informed decision making processes. In addition, the requirements are governed by, and must apply, the safety objective and safety principles that are established in the IAEA Safety Standards Series No. SF-1, Fundamental Safety Principles. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication

  11. Seismic safety of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Katona, T.

    1993-01-01

    An extensive program is underway at Paks NPP for evaluation of the seismic safety and for development of the necessary safety increasing measures. This program includes the following five measures: investigation of methods, regulations and techniques utilized for reassessment of seismic safety of operating NPPs and promoting safety; investigation of earthquake hazards; development of concepts for creating the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept, and preliminary evaluation of the seismic safety

  12. Planning and evaluation of plant under safety aspects

    International Nuclear Information System (INIS)

    Strnad, H.

    1985-01-01

    Plant denotes a technical product characterized as being structured, complex, comprising the use of energy, and that of measuring, automatic control and monitoring systems to keep track of present, control and monitor processes. Particular attention is paid to methods of developing plant concepts, measures to exclude or detect risks, integration of safety engineering into the course of planning, safety concept and ergonomics in plant design. (DG) [de

  13. Speech Recognition Interfaces Improve Flight Safety

    Science.gov (United States)

    2013-01-01

    "Alpha, Golf, November, Echo, Zulu." "Sierra, Alpha, Golf, Echo, Sierra." "Lima, Hotel, Yankee." It looks like some strange word game, but the combinations of words above actually communicate the first three points of a flight plan from Albany, New York to Florence, South Carolina. Spoken by air traffic controllers and pilots, the aviation industry s standard International Civil Aviation Organization phonetic alphabet uses words to represent letters. The first letter of each word in the series is combined to spell waypoints, or reference points, used in flight navigation. The first waypoint above is AGNEZ (alpha for A, golf for G, etc.). The second is SAGES, and the third is LHY. For pilots of general aviation aircraft, the traditional method of entering the letters of each waypoint into a GPS device is a time-consuming process. For each of the 16 waypoints required for the complete flight plan from Albany to Florence, the pilot uses a knob to scroll through each letter of the alphabet. It takes approximately 5 minutes of the pilot s focused attention to complete this particular plan. Entering such a long flight plan into a GPS can pose a safety hazard because it can take the pilot s attention from other critical tasks like scanning gauges or avoiding other aircraft. For more than five decades, NASA has supported research and development in aviation safety, including through its Vehicle Systems Safety Technology (VSST) program, which works to advance safer and more capable flight decks (cockpits) in aircraft. Randy Bailey, a lead aerospace engineer in the VSST program at Langley Research Center, says the technology in cockpits is directly related to flight safety. For example, "GPS navigation systems are wonderful as far as improving a pilot s ability to navigate, but if you can find ways to reduce the draw of the pilot s attention into the cockpit while using the GPS, it could potentially improve safety," he says.

  14. Safety of Nuclear Power Plants: Commissioning and Operation

    International Nuclear Information System (INIS)

    2011-01-01

    The safety of a nuclear power plant is ensured by means of proper site selection, design, construction and commissioning, and the evaluation of these, followed by proper management, operation and maintenance of the plant. In a later phase, a proper transition to decommissioning is required. The organization and management of plant operations ensures that a high level of safety is achieved through the effective management and control of operational activities. This publication is a revision of the Safety Requirements publication Safety of Nuclear Power Plants: Operation, which was issued in 2000 as IAEA Safety Standards Series No. NS-R-2. The purpose of this revision was to restructure Safety Standards Series No. NS-R-2 in the light of new operating experience and new trends in the nuclear industry; to introduce new requirements that were not included in Safety Standards Series No. NS-R-2 on the operation of nuclear power plants; and to reflect current practices, new concepts and technical developments. This update also reflects feedback on the use of the standards, both from Member States and from the IAEA's safety related activities. The publication is presented in the new format for Safety Requirements publications. The present publication reflects the safety principles of the Fundamental Safety Principles. It has been harmonized with IAEA Safety Standards Series No. GS-R-3 on The Management System for Facilities and Activities. Guidance on the fulfilment of the safety requirements is provided in supporting Safety Guides. The terminology used in this publication is defined and explained in the IAEA Safety Glossary. The objective of this publication is to establish the requirements which, in the light of experience and the present state of technology, must be satisfied to ensure the safe operation of nuclear power plants. These requirements are governed by the safety objective and safety principles that are established in the Fundamental Safety Principles. This

  15. Assistance of Foreign Countries and International Organizations to Support Safety Improvements at Ignalina NPP

    International Nuclear Information System (INIS)

    Shevaldin, V.

    1997-01-01

    International cooperation and assistance for the improving safety of Ignalina NPP is described. Sweden was among the first countries which supported safety improvements at Ignalina NPP. The first project in the cooperation was BARSELINA, Probabilistic Safety Analysis of Ignalina NPP. The cooperation is still bringing significant support to the plant, including improvements in the fire protection, communications system, physical protection, and many other areas. Another one very important source of assistance was Nuclear Safety Account, administered by the EBRD. In 1993 experts of the plant, together with representatives of VATESI and SKI (Sweden) have worked out a short-term safety improvement program SIP-1, which was financed by the EBRD . Eighteen safety related projects were selected, expensive and reliable equipment was procured and installed

  16. Safety culture in the maintenance of nuclear power plants

    International Nuclear Information System (INIS)

    2005-01-01

    Safety culture is the complexity of beliefs, shared values and behaviour reflected in making decisions and performing work in a nuclear power plant or nuclear facility. The definition of safety culture and the related concepts presented in the IAEA literature are widely known to experts. Since the publication of Safety Culture, issued by the IAEA as INSAG-4 in 1991, the IAEA has produced a number of publications on strengthening the safety culture in organizations that operate nuclear power plants and nuclear facilities. However, until now the focus has been primarily on the area of operations. Apart from operations, maintenance in plants and nuclear facilities is an aspect that deserves special attention, as maintenance activities can have both a direct and an indirect effect on equipment reliability. Adverse safety effects can arise, depending upon the level of skill of the personnel involved, safety awareness and the complexity of the work process. Any delayed effects resulting from challenges to maintenance can cause interruptions in operation, and hence affect the safety of a plant or facility. Building upon earlier IAEA publications on this topic, this Safety Report reviews how challenges to the maintenance of nuclear power plants can affect safety culture. It also highlights indications of a weakening safety culture. The challenges described are in areas such as maintenance management; human resources management; plant condition assessment and the business environment. The steps that some Member States have taken to address safety culture aspects are detailed and singled out as good practices, with a view to disseminating and exchanging experiences and lessons learned. Although this report is primarily directed at plant maintenance organizations, the subject matter is applicable to a wider audience, including plant contracting organizations and regulatory authorities

  17. Inherent/passive safety in fusion power plants

    International Nuclear Information System (INIS)

    Piet, S.J.; Crocker, J.G.

    1986-01-01

    The concept of inherent or passive safety for fusion energy is explored, defined, and partially quantified. Four levels of safety assurance are defined, which range from true inherent safety to passive safety to protection via active engineered safeguard systems. Fusion has the clear potential for achieving inherent or passive safety, which should be an objective of fusion research and design. Proper material choice might lead to both inherent/passive safety and high mass power density, improving both safety and economics. When inherent or passive safety is accomplished, fusion will be well on the way to achieving its ultimate potential and to be a truly superior energy source for the future

  18. Safety philosophy for nuclear power plants in egypt

    International Nuclear Information System (INIS)

    Mervat, S.A.; Hammad, F.H.

    1988-01-01

    This work establishes the basic principles of a safety philosophy for nuclear power plants in egypt. A number of deterministic requirements stemming the multiple barriers and the defense-in-depth concept are emphasised. other requirements in the areas of siting, operational safety, safety analysis, special issues, and experience feedback are also identified. The role of international cooperation in nuclear safety technology-transfer and nuclear emergencies is highlighted. In addition probabilistic ally based guidelines are set for acceptable risk and dose limits

  19. An approach to safety problems relating to ageing of nuclear power plant components

    International Nuclear Information System (INIS)

    Conte, M.; Deletre, G.; Henry, J.Y.; Le Meur, M.

    1989-10-01

    The safety of nuclear power plants, in France, is discussed. The attention is focused on the ageing phenomena, as a potential cause of the degradation of the systems functional capabilities. The allowance for ageing in design and its importance on safety, are analyzed. The understanding of phenomena relating to ageing and the components surveillance, are considered. As the effective ageing on the components of nuclear power plants is not fully understood, technical improvements and more accurate analysis are required

  20. Safety aspects of nuclear power plant component aging

    International Nuclear Information System (INIS)

    Conte, M.; Deletre, G.; Henry, J.Y.

    1988-01-01

    The safety of nuclear plants depends on the capacity of the systems they are composed to perform the functions they were designed for. The identification and understanding of phenomena liable to degrade this operational capacity thus constitute one of the safety problems for which allowance must be made at the earliest stage of a project. Aging, a natural and hence unavoidable process affecting all the components of an installation, was identified at a very early stage as being one of these phenomena. The investigation and implementation of solutions to the safety problems associated to aging make it necessary to: defining the domain in which the consequences of aging are to be evaluated, identifying the parameters involved, identifying the components sensitive to these parameters, understanding the mechanisms which govern its evolution. The results of qualification tests, and of tests and checks carried out at different stages of construction and operation, as well as allowance for operating experience, constitute the necessary basis for establishing or improving the regulatory requirements. The procedures for validating components and systems of the installation are also drawn up on the basis of these tests. Finally, the actions initiated within the scope of research and development programmes supply the additional data necessary for such validation, and provide the indispensable support for knowledge improvement

  1. Improved obstetric safety through programmatic collaboration.

    Science.gov (United States)

    Goffman, Dena; Brodman, Michael; Friedman, Arnold J; Minkoff, Howard; Merkatz, Irwin R

    2014-01-01

    Healthcare safety and quality are critically important issues in obstetrics, and society, healthcare providers, patients and insurers share a common goal of working toward safer practice, and are continuously seeking strategies to facilitate improvements. To this end, 4 New York City voluntary hospitals with large maternity services initiated a unique collaborative quality improvement program. It was facilitated by their common risk management advisors, FOJP Service Corporation, and their professional liability insurer, Hospitals Insurance Company. Under the guidance of 4 obstetrics and gynecology departmental chairmen, consensus best practices for obstetrics were developed which included: implementation of evidence based protocols with audit and feedback; standardized educational interventions; mandatory electronic fetal monitoring training; and enhanced in-house physician coverage. Each institution developed unique safety related expertise (development of electronic documentation, team training, and simulation education), and experiences were shared across the collaborative. The collaborative group developed robust systems for audit of outcomes and documentation quality, as well as enforcement mechanisms. Ongoing feedback to providers served as a key component of the intervention. The liability carrier provided financial support for these patient safety innovations. As a result of the interventions, the overall AOI for our institutions decreased 42% from baseline (January-June 2008) to the most recently reviewed time period (July-December 2011) (10.7% vs 6.2%, p Risk Management of the American Hospital Association.

  2. Resilience improvements to UK nuclear power plants

    International Nuclear Information System (INIS)

    Pepper, Kevin

    2015-01-01

    Following the events at Fukushima, the Office for Nuclear Regulation (ONR), the UK nuclear safety regulator, undertook a series of reviews into the resilience of UK nuclear power plants to severe events. These reviews highlighted a number of areas in relation to electrical infrastructure where it considered licensees should review their arrangements, considering both onsite and offsite infrastructure as well as the ability to recover following a complete loss of site infrastructure. In response, UK licensees have been exploring four parallel approaches to improving the resilience for each of their sites. Firstly, through modifications on-site such as enhancements to the installed diesel generators and related systems. Secondly through improvements to the resilience of essential instrumentation to Station Black Out events. Thirdly, through the provision of off-site backup equipment that can be deployed to any site following a severe event. Finally, the provision of event qualified connection points on site to enable timely restoration of long term essential electrical supplies and cooling to key systems. This last item gives a central focus to the issues of switchboard availability and the resilience of the whole network to large potentially common cause internal and external hazards. This paper will discuss the electrically related findings of the ONR reviews, explore the reasoning behind those decisions, and describe the approaches being taken by UK licensees. (authors)

  3. [Improving patient safety through voluntary peer review].

    Science.gov (United States)

    Kluge, S; Bause, H

    2015-01-01

    The intensive care unit (ICU) is one area of the hospital in which processes and communication are of primary importance. Errors in intensive care units can lead to serious adverse events with significant consequences for patients. Therefore quality and risk-management are important measures when treating critically ill patients. A pragmatic approach to support quality and safety in intensive care is peer review. This approach has gained significant acceptance over the past years. It consists of mutual visits by colleagues who conduct standardised peer reviews. These reviews focus on the systematic evaluation of the quality of an ICU's structure, its processes and outcome. Together with different associations, the State Chambers of Physicians and the German Medical Association have developed peer review as a standardized tool for quality improvement. The common goal of all stakeholders is the continuous and sustainable improvement in intensive care with peer reviews significantly increasing and improving communication between professions and disciplines. Peer reviews secure the sustainability of planned change processes and consequently lead the way to an improved culture of quality and safety.

  4. System 80+ Design and Licensing : Improving Plant Reliability

    International Nuclear Information System (INIS)

    Newman, Robert E.

    1989-01-01

    The U. S. nuclear industry is striving to improve plant reliability and availability through improved plant design, component designs and plant maintenance. In an effort to improve safety and to demonstrate that commercial nuclear power is economically competitive with other energy sources, the utilities, nuclear vendors, architect engineers and constructors, and component suppliers are all participating in an industry-wide effort to develop improved Light Water Reactor (LWR) designs that are based upon the many years of successful LWR operation. In an age when the world faces the environmental pressures of the greenhouse effect and acid rain, electricity generated from nuclear energy must play an increasing role in the energy picture of Korea, the United States and the rest of the world. This paper discusses the plant availability requirement that has been established by the industry-wide effort mentioned above. After briefly describing Combustion Engineering's program for development of the System 80 Plus standard design and the participation of the Korea Advanced Energy Research Institute (KAERI) in the program, the paper then describes the design features that are being incorporated into System 80+. The industry ALRR Program has established a very ambitious criterion of 87% for the plant availability of future nuclear units. To satisfy such a requirement, the next generation of nuclear plants will include a great many design improvements that reflect the hundreds of years of operating experience that we have accrued. C-ESA's System 80+ will include a number of design changes that improve operating margins and make the plant easier to operate and maintain. Not surprisingly, there is a great deal of overlap between improved safety and improved reliability. In the end, our design will satisfy the future needs of the utilities, the regulators, and the public. C-E is very pleased that KAERI is working with US to achieve these important goals

  5. BRICS: opportunities to improve road safety.

    Science.gov (United States)

    Hyder, Adnan A; Vecino-Ortiz, Andres I

    2014-06-01

    Brazil, the Russian Federation, India, China and South Africa--the countries known as BRICS--are currently undergoing a deep epidemiological transition that is mainly driven by rapid economic growth and technological change. The changes being observed in the distribution of the burden of diseases and injuries--such as recent increases in the incidence of road traffic injuries--are matters of concern. BRICS may need stronger institutional capacity to address such changes in a timely way. In this paper, we present data on road traffic injuries in BRICS and illustrate the enormous challenge that these countries currently face in reducing the incidence of such injuries. There is an urgent need to improve road safety indicators in every country constituting BRICS. It is imperative for BRICS to invest in system-wide road safety interventions and reduce the mortality and morbidity from road traffic injuries.

  6. RBMK nuclear power plants: Generic safety issues. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-05-01

    This report has been prepared on the basis of above mentioned report and it is intended to provide information on RBMK NPPs generic safety issues. As all other insights, recommendations and conclusions resulting from the IAEA Programme, this report is intended to assist national decision makers, who have sole responsibility for the regulation and safe operation of their nuclear power plants. It also serves to focus national and international projects on priority of the RBMK safety improvements. 23 refs, 10 figs, 3 tabs

  7. Study on a quantitative evaluation method of equipment maintenance level and plant safety level for giant complex plant system

    International Nuclear Information System (INIS)

    Aoki, Takayuki

    2010-01-01

    In this study, a quantitative method on maintenance level which is determined by the two factors, maintenance plan and field work implementation ability by maintenance crew is discussed. And also a quantitative evaluation method on safety level for giant complex plant system is discussed. As a result of consideration, the following results were obtained. (1) It was considered that equipment condition after maintenance work was determined by the two factors, maintenance plan and field work implementation ability possessed by maintenance crew. The equipment condition determined by the two factors was named as 'equipment maintenance level' and its quantitative evaluation method was clarified. (2) It was considered that CDF in a nuclear power plant, evaluated by using a failure rate counting the above maintenance level was quite different from CDF evaluated by using existing failure rates including a safety margin. Then, the former CDF was named as 'plant safety level' of plant system and its quantitative evaluation method was clarified. (3) Enhancing equipment maintenance level means an improvement of maintenance quality. That results in the enhancement of plant safety level. Therefore, plant safety level should be always watched as a plant performance indicator. (author)

  8. Nuclear power plant safety. The merits of separation

    International Nuclear Information System (INIS)

    Helander, L.I.; Tiren, L.I.

    1977-01-01

    The paper illustrates how the physical separation of safety-related structures, systems and components can improve the protection of a nuclear power plant against multiple failures that may be caused by events such as fire, pipe-whip, missiles, flooding, hurricanes, lightning etc. Criteria for redundancy and separation requirements affect nuclear plant design in many areas such as building layout, arrangements for fire protection and ventilation, separation of mechanical systems and components, in particular emergency cooling systems, separation of electric equipment and cables. The implementation of the ensuing design criteria for a BWR power plant is described. This design involves the separation of emergency cooling systems into four 50% capacity systems, which are independent and separated, including the distribution network for electric power from on-site standby diesel generators and the circuitry for the reactor protection system. The plant is subdivided into a number of fire zones, each with its own independent ventilation system. The fire zones are further subdivided into a multitude of fire cells such that redundant subsystems are housed in separate cells. These design precautions with regard to fire are complemented by extensive fire fighting systems. (author)

  9. Improved safety in advanced control complexes, without side effects

    International Nuclear Information System (INIS)

    Harmon, D.L.

    1997-01-01

    If we only look for a moment at the world around us, it is obvious that advances in digital electronic equipment and Human-System Interface (HSI) technology are occurring at a phenomenal pace. This is evidenced from our home entertainment systems to the dashboard and computer-based operation of our new cars. Though the nuclear industry has less vigorously embraced these advances, their application is being implemented through individual upgrades to current generation nuclear plants and as plant-wide control complexes for advanced plants. In both venues modem technology possesses widely touted advantages for improving plant availability as well as safety. The well-documented safety benefits of digital Instrumentation and Controls (I ampersand C) include higher reliability resulting from redundancy and fault tolerance, inherent self-test and self-diagnostic capabilities which have replaced error-prone human tasks, resistance to setpoint drift increasing available operating margins, and the ability to run complex, real-time, computer-based algorithms directly supporting an operator's monitoring and control task requirements. 22 refs., 3 figs., 5 tabs

  10. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

    1985-01-01

    The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur

  11. Safety/security interface assessments at commercial nuclear power plants

    International Nuclear Information System (INIS)

    Byers, K.R.; Brown, P.J.; Norderhaug, L.R.

    1985-07-01

    The findings of the Haynes Task Force Committee (NUREG-0992) are used as the basis for defining safety/security assessment team activities at commercial nuclear power plants in NRC Region V. A safety/security interface assessment outline and the approach used for making the assessments are presented along with the composition of team members. As a result of observing simulated plant emergency conditions during scheduled emergency preparedness exercises, examining security and operational response procedures, and interviewing plant personnel, the team has identified instances where safety/security conflicts can occur. 2 refs

  12. Key asset - inherent safety of LMFBR Pool Plant

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Sevy, R.H.; Lancet, R.T.; Mills, J.C.

    1984-04-01

    The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

  13. Key asset--Inherent safety of LFMBR pool plant

    International Nuclear Information System (INIS)

    Marchaterre, J.F.; Lancet, R.T.; Mills, J.C.; Sevy, R.H.

    1984-01-01

    The safety approach used in the design of the Large Pool Plant emphasizes use of the intrinsic characteristics of Liquid Metal Fast Breeder Reactors to incorporate a high degree of safety in the design and reduce cost by providing simpler (more reliable) dedicated safety systems. Correspondingly, a goal was not to require the action of active systems to prevent significant core damage and/or provide large grace periods for all anticipated transients. The key safety features of the plant are presented and the analysis of representative flow and power transients are presented to show that the design goal has been satisfied

  14. Standardized safety management of AP1000 nuclear power plant

    International Nuclear Information System (INIS)

    Li Xingwen; Cao Zhiqiang; Cong Jiuyuan

    2011-01-01

    In 2002, China published and implemented the Law of the People's Republic of China on Work Safety and promulgated a series of guidelines and policies, which strengthened the safety management supervision. Standardization of safety, as another important step on safety supervision, comes after safety assesment and safety production licensing system, is also a permanent solution. Standardization of safety is a strategic, long term and fundamental work, which is also the basic access to achieving scientific safety management and increasing the inherent safety of an enterprise. Haiyang AP1000 nuclear power plant, adopting the modularized, 'open-top' and parallel construction means, overturned the traditional construction theory of installation work comes after the civil work and greatly shorten the construction period. At the same time, the notable increase of oversize module transportation and lifting and parallel construction raises higher demands for safety management. This article combines the characteristics and difficulties of safety management for Haiyang AP1000 nuclear power plant, puts forward ideas and methods for standardized safety management, and could also serve as reference to the safety management for other AP1000 projects. (authors)

  15. A cross-cultural study of organizational factors on safety: Japanese vs. Taiwanese oil refinery plants.

    Science.gov (United States)

    Hsu, Shang Hwa; Lee, Chun-Chia; Wu, Muh-Cherng; Takano, Kenichi

    2008-01-01

    This study attempts to identify idiosyncrasies of organizational factors on safety and their influence mechanisms in Taiwan and Japan. Data were collected from employees of Taiwanese and Japanese oil refinery plants. Results show that organizational factors on safety differ in the two countries. Organizational characteristics in Taiwanese plants are highlighted as: higher level of management commitment to safety, harmonious interpersonal relationship, more emphasis on safety activities, higher devotion to supervision, and higher safety self-efficacy, as well as high quality of safety performance. Organizational characteristics in Japanese plants are highlighted as: higher level of employee empowerment and attitude towards continuous improvement, more emphasis on systematic safety management approach, efficient reporting system and teamwork, and high quality of safety performance. The casual relationships between organizational factors and workers' safety performance were investigated using structural equation modeling (SEM). Results indicate that the influence mechanisms of organizational factors in Taiwan and Japan are different. These findings provide insights into areas of safety improvement in emerging countries and developed countries respectively.

  16. Safety issues and their ranking for WWER-440 model 213 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    1996-04-01

    The objective of this report is to present a consolidated list of generic safety concerns, called safety issues, ranked according to their safety significance and the corrective measures to improve safety. It is intended for use as a reference to facilitate the development of plant specific safety improvement programmes and to serve as a basis for reviewing their implementation. Section 2 provides and overview of the impact of all relevant issues on the main safety functions and other aspects important to overall plant safety. Section 3 presents safety issues identified in design according to the structure described below. Section 4 presents the safety issues in the area of operation, according to the same structure except that no ranking is given. At the end of Section 2, Tables 1 and 2 present a summary of all safety issues in a tabular form. 138 refs, tabs

  17. Operational characteristics of nuclear power plants - modelling of operational safety

    International Nuclear Information System (INIS)

    Studovic, M.

    1984-01-01

    By operational experience of nuclear power plants and realize dlevel of availability of plant, systems and componenst reliabiliuty, operational safety and public protection, as a source on nature of distrurbances in power plant systems and lessons drawn by the TMI-2, in th epaper are discussed: examination of design safety for ultimate ensuring of safe operational conditions of the nuclear power plant; significance of the adequate action for keeping proess parameters in prescribed limits and reactor cooling rquirements; developed systems for measurements detection and monitoring all critical parameters in the nuclear steam supply system; contents of theoretical investigation and mathematical modeling of the physical phenomena and process in nuclear power plant system and components as software, supporting for ensuring of operational safety and new access in staff education process; program and progress of the investigation of some physical phenomena and mathematical modeling of nuclear plant transients, prepared at faculty of mechanical Engineering in Belgrade. (author)

  18. Improving ICU risk management and patient safety.

    Science.gov (United States)

    Kielty, Lucy Ann

    2017-06-12

    Purpose The purpose of this paper is to describe a study which aimed to develop and validate an assessment method for the International Electrotechnical Commission (IEC) 80001-1 (IEC, 2010) standard (the Standard); raise awareness; improve medical IT-network project risk management processes; and improve intensive care unit patient safety. Design/methodology/approach An assessment method was developed and piloted. A healthcare IT-network project assessment was undertaken using a semi-structured group interview with risk management stakeholders. Participants provided feedback via a questionnaire. Descriptive statistics and thematic analysis was undertaken. Findings The assessment method was validated as fit for purpose. Participants agreed (63 per cent, n=7) that assessment questions were clear and easy to understand, and participants agreed (82 per cent, n=9) that the assessment method was appropriate. Participant's knowledge of the Standard increased and non-compliance was identified. Medical IT-network project strengths, weaknesses, opportunities and threats in the risk management processes were identified. Practical implications The study raised awareness of the Standard and enhanced risk management processes that led to improved patient safety. Study participants confirmed they would use the assessment method in future projects. Originality/value Findings add to knowledge relating to IEC 80001-1 implementation.

  19. Plant assessment system and safety culture

    International Nuclear Information System (INIS)

    Chun, Chuyoung

    1996-01-01

    The government, upon these events, keenly felt the necessity for developing the safety culture which was already forwarded in nuclear industries and started taking actions to propagate it to all parts of society. The government established a social safety director position under the Prime Minister's jurisdiction and also established a Safety Culture Promotion Headquarters in which 7 ministries and other organizations, such as Korea Economic Council, Federation of Korea Trade Union and Women's Federation Council were participating. In accordance with the government's strong will to enhance the safety consciousness of people, safety campaigns are being developed voluntarily in the private sector. The formation of non-governmental organizations, such as People's Central Council of Safety Culture Promotion, shows a good example of such movement

  20. Nuclear Power Plant (NPP) safety in Brazil

    International Nuclear Information System (INIS)

    Lederman, L.

    1980-01-01

    The multidisciplinary aspects of the activities involved in the nuclear power plant (NPP) licensing, are presented. The activities of CNEN's technical staff in the licensing of Angra-1 and Angra-2 power plants are shown. (E.G.) [pt

  1. [Karachi Nuclear Power Plant (KANUPP), Safety Management

    Energy Technology Data Exchange (ETDEWEB)

    Hasan, S M [Karachi Nuclear Power Plant (KANUPP), Karachi (Pakistan)

    1997-12-01

    The present regime for CANDU safety management in Pakistan has evolved in line with contemporary international practice, and is essential adequate to ensure the continued safety of KANUPP and other future CANDU reactors, as confirmed by international reviews as well. But the small size of Pakistan nuclear power program poses limitations in developing - expert judgment in analysis of in-service inspection data; and own methodology for CANDU safety analysis.

  2. [Karachi Nuclear Power Plant (KANUPP), Safety Management

    International Nuclear Information System (INIS)

    Hasan, S.M.

    1997-01-01

    The present regime for CANDU safety management in Pakistan has evolved in line with contemporary international practice, and is essential adequate to ensure the continued safety of KANUPP and other future CANDU reactors, as confirmed by international reviews as well. But the small size of Pakistan nuclear power program poses limitations in developing - expert judgment in analysis of in-service inspection data; and own methodology for CANDU safety analysis

  3. Seismic safety of Paks nuclear power plant

    International Nuclear Information System (INIS)

    Katona, T.

    1995-01-01

    This paper contains an overview of the results concerning the following activities: investigation of methods, regulations and techniques for reassessment of seismic safety of operating NPPs and upgrading of safety; investigation of earthquake hazards; development of concept for creation of the seismic safety location of earthquake warning system; determination of dynamic features of systems and facilities determined by the concept and preliminary evaluation of the seismic safety. It is limited on investigation of dynamic features of building structures, the building dynamical experiments and experimental investigation of the equipment

  4. Probabilistic analysis of safety in industrial irradiation plants

    International Nuclear Information System (INIS)

    Alderete, F.; Elechosa, C.

    2006-01-01

    The Argentinean Nuclear Regulatory Authority is carrying out the Probabilistic Safety Analysis (PSA) of the two industrial irradiation plants existent in the country. The objective of this presentation is to show from the regulatory point of view, the advantages of applying this tool, as well as the appeared difficulties; for it will be made a brief description of the facilities, of the method and of the normative one. Both plants are multipurpose facilities classified as 'industrial irradiator category IV' (panoramic irradiator with source deposited in pool). Basically, the execution of an APS consists of the following stages: 1. Identification of initiating events. 2. Modeling of Accidental Sequences (Event Trees). 3. Analysis of Systems (Fault trees). 4. Quantification of Accidental Sequences. The argentine normative doesn't demand to these facilities the realization of an APS, however the basic standard of Radiological Safety establishes that in the design of this type of facilities in the cases that is justified, should make sure that the annual probability of occurrence of an accidental sequence and the resulting dose in a person gives as result an radiological risk inferior to the risk limit adopted as acceptance criteria. On the other hand the design standard specifies for these irradiators it demands a maximum fault rate of 10 -2 for the related components with the systems of radiological safety. In our case, the possible initiating events have been identified that carried out to not wanted situations (about people exposure, radioactive contamination). Then, for each one of the significant initiating events, the corresponding accidental sequences were modeled and the safety systems that intervene in this sequences by means of fault trees were analyzed, for then to determine the fault probabilities of the same ones. At the moment they are completing these fault trees, but the difficulty resides in the impossibility of obtaining real data of the reliability

  5. The safety evaluation guide for laboratories and plants a tool for enhancing safety

    International Nuclear Information System (INIS)

    Lhomme, Veronique; Daubard, Jean-Paul

    2013-01-01

    The Institute for Radioprotection and Nuclear Safety (IRSN) acts as technical support for the French government Authorities competent in nuclear safety and radiation protection for civil and defence activities. In this frame, the Institute's performs safety assessments of the safety cases submitted by operators to these Authorities for each stage in the life cycle of a nuclear facility, including dismantling operations, which is subjected to a licensing procedure. In the fuel cycle field, this concerns a large variety of facilities. Very often, depending on facilities and on safety cases, safety assessment to be performed is multidisciplinary and involves the supervisor in charge of the facility and several safety experts, particularly to cover the whole set of risks (criticality, exposure to radiation, fire, handling, containment, human and organisational factors...) encountered during facility's operations. Taking these into account, and in order to formalize the assessment process of the fuel cycle facilities, laboratories, irradiators, particle accelerators, under-decommissioning reactors and radioactive waste management, the 'Plants, Laboratories, Transports and Waste Safety' Division of IRSN has developed an internal guide, as a tool: - To present the methodological framework, and possible specificities, for the assessment according to the 'Defence in Depth Concept' (Part 1); - To provide key questions associated to the necessary contradictory technical review of the safety cases (Part 2); - To capitalise on experience on the basis of technical examples (coming from incident reports, previous safety assessments...) demonstrating the questioning (Part 3). The guide is divided in chapters, each dedicated to a type of risk (dissemination of radioactive material, external or internal exposure from ionising radiation, criticality, radiolysis mechanisms, handling operations, earthquake, human or organisational factors...) or to a type

  6. Safety and Radiation Protection at Swedish Nuclear Power Plants 2004

    International Nuclear Information System (INIS)

    2005-05-01

    Government are part of the reason and will be a feature of the next ten-year period. Furthermore, in new regulations, SKI has introduced more stringent requirements on facility design and construction, based on the experience that has been accumulated since nuclear power was introduced in Sweden. At the same time, the power companies intend to implement power uprating at several of the reactors, which will require extensive safety reviews. Major power uprates require a large amount of analysis work and refurbishment at the facilities in order to take into account increased capacity requirements on safety systems and other factors. During the year, the work on the new regulations for physical protection intensified and a revised draft of regulations and general recommendations has been subjected to a formal review by licensees concerned. During the year, SKI initiated an investigation into how maintenance strategies have developed at Swedish nuclear power plants since the deregulation of the Swedish electricity market in the mid-nineties. The investigation shows that the rate of change over the past five-year period has been faster than during the previous five to ten-year period. Changes have occurred with respect to strategy and organization. Deregulation has been the most important driving force for internal improvements. One conclusion is that the changes have contributed to the learning and development of both individuals and organization. Furthermore, there is no sign of a negative impact on reactor safety. In SKI's opinion, the maintenance activity is extremely important to safety, and supervision will focus on work loads, responsibilities and roles, work satisfaction and motivation, maintenance procedures as well as the possible impact of the combination of financial pressure and a high rate of change. During the year, it was announced that Barsebaeck 2 would be closed down at the end of May 2005. SKI will continue to maintain intensified supervision, which means a

  7. Discussion on an informative system set-up for the registration and processing of reliability data on FBR components in view of its application to design and safety studies and plant exploitation improvement

    International Nuclear Information System (INIS)

    Righini, R.; Sola, P.G.; Zappellini, G.

    1990-01-01

    This report describes the set-up and management activities carried-out by ENEA-VEL in collaboration with NIER in the development of a reliability data bank on fast reactor components; this data bank consists of an informative system implemented on the IBM 3090 computer of the ENEA centre of Bologna starting from the software of the CEDB, set-up by CCR Euratom of Ispra for the registration of reliability data on thermal reactor components. This report will contain a detailed description of all the modules (engineering, operating, etc.) provided in the informative system and of the modifications introduced by ENEA in order to adapt them to the peculiarities of the fast reactors and to increase its flexibility; a short description of the available data processing methods will be also included. It will be followed by a comparison between the results obtained applying the classical methods and the particular ones set-up by ENEA: this comparison will be useful to demonstrate the importance of the method applied in order to obtain significative reliability processed data. This report will be also useful to show the importance of the set-up data bank in the improvement of the component design and of the plant safety and exploitation with particular reference to the research of the critical areas and to the definition of the best inspection and maintenance programs

  8. The safety of nuclear power plants

    International Nuclear Information System (INIS)

    Berg, J.O.

    1980-01-01

    A general presentation is given of reactor safety philosophy and risk analysis. The Rasmussen report (WASH-1400) is discussed and also the Lewis Commission's evaluation of that report. The future developments in reactor safety technology are outlined with emphasis on quality assurance. (JIW)

  9. Plant safety and performance indicators for regulatory use

    International Nuclear Information System (INIS)

    Ferjancic, M.; Nemec, T.; Cimesa, S.

    2004-01-01

    Slovenian Nuclear Safety Administration (SNSA) supervises nuclear and radiological safety of Krsko NPP. This SNSA supervision is performed through inspections, safety evaluations of plant modifications and event analyses as well as with the safety and performance indicators (SPI) which are a valuable data source for plant safety monitoring. In the past SNSA relied on the SPI provided by Krsko NPP and did not have a set of SPI which would be more appropriate for regulatory use. In 2003 SNSA started with preparation of a new set of SPI which would be more suitable for performing the regulatory oversight of the plant. New internal SNSA procedure which is under preparation will define use and evaluation of SPI and will include definitions for the proposed set of SPI. According to the evaluation of SPI values in comparison with the limiting values and/or trending, the procedure will define SNSA response and actions. (author)

  10. Safety classification of nuclear power plant systems, structures and components

    International Nuclear Information System (INIS)

    1992-01-01

    The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given

  11. Temperature and level measurements realized for Nuclear Safety Level Improvement of Slovak NPPs

    International Nuclear Information System (INIS)

    Badiar, S.; Slanina, M.; Stanc, S.; Golan, P.; Krupa, J.

    2001-01-01

    Process of continual safety improvement in the individual Slovak nuclear power plants has been in progress since the beginning of nineties with the objective to upgrade the safety level of units in operation up to the European standards. In the framework of these activities, safety instrumentation systems with 1E qualification for the control of WWER reactor coolant systems were built and added. Methods for implementation of safety instrumentation systems for monitoring temperature and level in reactor coolant systems in the particular plants in Slovakia are presented showing the objectives and methods of their implementation. (Authors)

  12. Safety upgrading of the PAKS Nuclear Plant

    International Nuclear Information System (INIS)

    Vamos, G.; Vigassy, J.

    1993-01-01

    In the last several years the net electricity from the Paks NPP represents almost half of the Hungarian total. The 4 units of Paks belong to the latest generation of the VVER-440 units, the small-sized Russian designed PWRs. Reviewing the main design features of them, the safety merits and safety concerns are summarized. Due to the conservative design and the extensive operating experience the safety merits appear to be more significant than generally believed. The VVER-440 type has two models, the 230 and 213, which have a large number of distinctive safety features. These are highlighted in the section comparisons. A quality assurance program was initiated in Paks very early. A long-term safety upgrading program was also initiated, originating from vendor recommendations, regulatory decisions, in-house operating experience and safety concerns, and independent reviews. The main areas and some examples of the measures are described. This program, like all other activities related to nuclear safety, has been under regulatory control. The specific features of the Hungarian regulatory system are described. For advanced, general and new evaluation of the safety of the units in Paks in accordance with the internationally recommended criteria of the 90's, the project AGNES has been launched with international participation. The scope of this project is summarized. International efforts as the IAEA Regional Project on safety assessment of VVER-440/213 and VVER-440/230 units are underway. Since safety is not only a question of design, but it can be significantly influenced by operations and maintenance practices, the Paks NPP has invited LAEA's OSART and ASSET missions, WANO's Pilot Peer Review

  13. Designing continuous safety improvement within chemical industrial areas

    NARCIS (Netherlands)

    Reniers, G.L.L.; Ale, B. J.M.; Dullaert, W.; Soudan, K.

    This article provides support in organizing and implementing novel concepts for enhancing safety on a cluster level of chemical plants. The paper elaborates the requirements for integrating Safety Management Systems of chemical plants situated within a so-called chemical cluster. Recommendations of

  14. Martin Marietta Energy Systems Nuclear Criticality Safety Improvement Program

    International Nuclear Information System (INIS)

    Speas, I.G.

    1987-01-01

    This report addresses questions raised by criticality safety violation at several DOE plants. Two charts are included that define the severity and reporting requirements for the six levels of accidents. A summary is given of all reported criticality incident at the DOE plants involved. The report concludes with Martin Marietta's Nuclear Criticality Safety Policy Statement

  15. Monitoring System For Improving Radiation Safety Management

    International Nuclear Information System (INIS)

    Osovizky, A.; Paran, J.; Tal, N.; Ankry, N.; Ashkenazi, B.; Tirosh, D.; Marziano, R.; Chisin, R.

    1999-01-01

    Medi SMARTS (Medical Survey Mapping Automatic Radiation Tracing System), a gamma radiation monitoring system, was installed in a nuclear medicine department. In this paper the evaluation of the system's ability to improve radiation safety management is presented. The system is based on a state of the art software that continuously collects on line radiation measurements for display, analysis and logging. Radiation is measured by GM tubes; the signal is transferred to a data processing unit and then via an RS-485 communication line to a computer. The system automatically identifies the detector type and its calibration factor, thus providing compatibility, maintainability and versatility when changing detectors. Radiation levels are displayed on the nuclear medicine department map at six locations. The system has been operating continuously for more than one year, documenting abnormal events caused by routine operation or failure incidents. In cases where abnormal working conditions were encountered, an alarm message was sent automatically to the supervisor via his tele-pager. An interesting issue observed during the system evaluation, was the inability to distinguish between high radiation levels caused by proper routine operation and those caused by safety failure incidents. The solution included examination of two parameters, radiation levels as well as their duration period. A careful analysis of the historical data, applying the appropriated combined parameters determined for each location, verified that such a system can identify abnormal events, provide alarms to warn in case of incidents and improve standard operating procedures

  16. Improving occupational safety in Kuzbass mines

    Energy Technology Data Exchange (ETDEWEB)

    Evseev, V S

    1986-08-01

    Some achievements of VostNII are listed in improving occupational safety in Kuzbass mines. Methane is a major problem: 90.6% of mines is in category III or supercategory; over 21% has an absolute methane emission of 30 m/sup 3//min or more. Another problem is spontaneous fires, which cost 2 million t of coal per year. One method of preventing these is injection of antipyrogens (urea and diammonium phosphate); another is the creation of gel (water glass, ammonium chloride and water) barriers in goaf areas. High pressure water jets are also used. Various methods of improving ventilation systems to match increased coal output are proposed, including drilling large diameter ventilation boreholes from the surface. In Leninskugol' mines the useful air is only 55.8% of the total delivered. More attention should be given to degassing (currently producing 130 million m/sup 3//y of methane). Dust levels are increasing due to the advent of narrow web cutter loaders (100% of coal cutter loaders in Kuzbass mines in 1984). Water injection and spraying are partially effective at dust suppression. Some electrical safety devices developed by VostNII are described.

  17. 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

  18. Results of research into nuclear power plant safety

    International Nuclear Information System (INIS)

    Polak, V.; Hladky, E.; Moravek, J.; Suchomel, J.; Stehlik, J.

    1976-01-01

    A survey is given of computer programmes for the safety analysis of nuclear power plants with WWER type reactors and with fast breeder reactors. The programmes solve accidents in the core, the primary circuit and the containment. A comparison is made of Czechoslovak and foreign computer programmes and their agreement proved. Also studied is the problem of radiation safety of nuclear power plants with regard to the leakage of radioactive isotopes and their detection. (J.B.)

  19. Safety criteria for siting a nuclear power plant

    International Nuclear Information System (INIS)

    2001-01-01

    The guide sets forth requirements for safety of the population and the environment in nuclear power plant siting. It also sets out the general basis for procedures employed by other competent authorities when they issue regulations or grant licences. On request STUK (Radiation and Nuclear Safety Authority of Finland) issues case-specific statements about matters relating to planning and about other matters relating to land use in the environment of nuclear power plants

  20. General view about reactor safety nuclear power plants in Brazil

    International Nuclear Information System (INIS)

    Gasparian, A.E.; Silva, D.E.; Salvatore, J.E.L.; Lima, J.M. de

    1991-01-01

    In this paper the authors describe the principles and goals that have guided, as well as the methods that have been used by the National Commission of Nuclear Energy (CNEN) to set forth measures aiming at providing safety to the Brazilian nuclear power plants. The status of the licensing process of these power plants is shown. The performance and the results obtained so far in relation to the nuclear safety are also described. (author)

  1. Trends in safety objectives for nuclear district heating plants

    Energy Technology Data Exchange (ETDEWEB)

    Brogli, R [Paul Scherrer Inst., Villigen (Switzerland)

    1997-09-01

    Safety objectives for dedicated nuclear heating plants are strongly influenced on the one hand by what is accepted for electricity nuclear stations, and on the other hand by the requirement that for economical reasons heating reactors have to be located close to population centers. The paper discusses the related trends and comes to the conclusion that on account of the specific technical characteristics of nuclear heating plants adequate safety can be provided even for highly populated sites. (author). 8 refs.

  2. Current activities on safety improvement at Ukrainian NPPs

    International Nuclear Information System (INIS)

    Stovbun, V.V.

    2000-01-01

    This report describes general development status of the national programs on safety improvement of the Ukrainian NPPs, basic approaches adopted for planning and implementation of safety improvement works, and state of implementation of principal technical activities aimed at safety improvement of Ukrainian NPPs. (author)

  3. Use of operational experience in fire safety assessment of nuclear power plants

    International Nuclear Information System (INIS)

    2000-01-01

    Fire hazard has been identified as a major contributor to a plant's operational risk and the international nuclear power industry has been studying and developing tools for defending against this hazard. Considerable progress in design and regulatory requirements for fire safety, in fire protection technology and in related analytical techniques has been made in the past two decades. Substantial efforts have been undertaken worldwide to implement these advances in the interest of improving fire safety both at new and existing nuclear power plants. To assist in these efforts, the IAEA initiated a programme on fire safety that was intended to provide assistance to Member States in improving fire safety in nuclear power plants. In order to achieve this general objective, the IAEA programme aimed at the development of guidelines and good practices, the promotion of advanced fire safety assessment techniques, the exchange of state of the art information between practitioners and the provision of engineering safety advisory services and training in the implementation of internationally accepted practices. During the period 1993-1994, the IAEA activities related to fire safety concentrated on the development of guidelines and good practice documents related to fire safety and fire protection of operating plants. One of the first tasks was the development of a Safety Guide that formulates specific requirements with regard to the fire safety of operating nuclear power plants. Several documents, which provide advice on fire safety inspection, were developed to assist in its implementation. In the period 1995-1996, the programme focused on the preparation of guidelines for the systematic analysis of fire safety at nuclear power plants (NPPs). The IAEA programme on fire safety for 1997-1998 includes tasks aimed at promoting systematic assessment of fire safety related occurrences and dissemination of essential insights from this assessment. One of the topics addressed is the

  4. Seismic safety of nuclear power plants in Eastern Europe

    International Nuclear Information System (INIS)

    Gurpinar, A.; Godoy, A.

    1995-01-01

    This paper summarizes the work performed by the International Atomic Energy Agency in the areas of safety reviews and applied research in support of programmes for the assessment and enhancement of seismic safety in WWER type nuclear power plants during the past five years. Three major topics are discussed; engineering safety review services in relation to external events, technical guidelines for the assessment and upgrading of WWER type nuclear power plants, and the Coordinated Research Programme on B enchmark study for the seismic analysis and testing of WWER type nuclear power plants . These topics are summarized in a way to provide an overview of the past and present safety situation in selected WWER type plants which are all located in Eastern European countries. Main conclusion of the paper is that although there is now a thorough understanding of the seismic safety issues in these operating nuclear power plants, the implementation of seismic upgrades to structures, systems and components are lagging behind, particularly for those cases in which the re-evaluation indicated the necessity to strengthen the safety related structures or install new safety systems. (author)

  5. Holistic safety analysis for advanced nuclear power plants

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Guimaraes, A.C.F.

    1992-01-01

    This paper reviews the basic methodology of safety analysis used in the ANGRA-I and ANGRA-II nuclear power plants, its weaknesses, the problems with public acceptance of the risks, the future of the nuclear energy in Brazil, as well as recommends a new methodology, HOLISTIC SAFETY ANALYSIS, to be used both in the design and licensing phases, for advanced reactors. (author)

  6. Radiation safety and protection on the nuclear power plants

    International Nuclear Information System (INIS)

    Nosovskij, A.V.; Bogorad, V.I.; Vasil'chenko, V.N.; Klyuchnikov, A.A.; Litvinskaya, T.V.; Slepchenko, A.Yu.

    2008-01-01

    The main issues of the radiation safety and protection provision on the nuclear power plants are considered in this monograph. The description of the basic sources of the radiation danger on NPPs, the principles, the methods and the means of the safety and radiation monitoring provision are shown. The special attention is paid to the issues of the ionizing radiation regulation

  7. Use of nondestructive evaluation methods to improve power plant availability

    International Nuclear Information System (INIS)

    Weber, R.M.

    1985-01-01

    On an ever-increasing basis, utilities are relying on nondestructive evaluation (NDE) as a management and planning tool. In addition to the conventional ASME Code and Technical Specification-required examinations, progressive utilities are utilizing NDE sampling programs to monitor existing conditions and search for potential situations affecting plant safety and reliability. Improved NDE detection and sizing procedures give management personnel the accurate information needed to make the ''go/no go'' decisions on repair programs which can significantly affect plant availability. As the burden of regulatory-imposed inspection requirements increases, plant personnel are increasingly cognizant that NDE is a significant factor in their plant's outage schedule. Whether an outage is scheduled or forced, NDE becomes part of each plant's program to assure the safety and reliability of its critical components. Knowledge and planning of NDE application is important because of the time expended in examination performance and subsequent data evaluation. Managers who are knowledgeable in NDE application can effectively improve plant availability by scheduling NDE as an integral part of their maintenance programs. Examination results can then be used in making decisions directly affecting availability

  8. Nuclear power plants. Electrical equipment of the safety system. Qualification

    International Nuclear Information System (INIS)

    2001-01-01

    This International Standard applies to electrical parts of safety systems employed at nuclear power plants, including components and equipment of any interface whose failure could affect unfavourably properties of the safety system. The standard also applies to non-electrical safety-related interfaces. Furthermore, the standard describes the generic process of qualification certification procedures and methods of qualification testing and related documentation. (P.A.)

  9. Definitive closure of nuclear power plants. Aspects concerning physical safety

    International Nuclear Information System (INIS)

    Rodriguez, C.; Puntarulo, L.; Canibano, J.

    1988-01-01

    This paper analyzes the various safety requirements that must be fulfilled by nuclear power plants for their operation without restrictions, such as safeguards, nuclear safety and physical protection. Physical protection, the subject most extensively dealt by the authors, is defined as safety measures aimed at providing protection against deliberate hostile deeds, such as robberies or non-authorized transport of radioactive materials or sabotage in nuclear facilities, performed either by individuals or by groups of individuals. (Author)

  10. Climate resilient crops for improving global food security and safety.

    Science.gov (United States)

    Dhankher, Om Parkash; Foyer, Christine H

    2018-05-01

    Food security and the protection of the environment are urgent issues for global society, particularly with the uncertainties of climate change. Changing climate is predicted to have a wide range of negative impacts on plant physiology metabolism, soil fertility and carbon sequestration, microbial activity and diversity that will limit plant growth and productivity, and ultimately food production. Ensuring global food security and food safety will require an intensive research effort across the food chain, starting with crop production and the nutritional quality of the food products. Much uncertainty remains concerning the resilience of plants, soils, and associated microbes to climate change. Intensive efforts are currently underway to improve crop yields with lower input requirements and enhance the sustainability of yield through improved biotic and abiotic stress tolerance traits. In addition, significant efforts are focused on gaining a better understanding of the root/soil interface and associated microbiomes, as well as enhancing soil properties. © 2018 The Authors Plant, Cell & Environment Published by John Wiley & Sons Ltd.

  11. Assessing nuclear power plant safety and recovery from earthquakes using a system-of-systems approach

    International Nuclear Information System (INIS)

    Ferrario, E.; Zio, E.

    2014-01-01

    We adopt a ‘system-of-systems’ framework of analysis, previously presented by the authors, to include the interdependent infrastructures which support a critical plant in the study of its safety with respect to the occurrence of an earthquake. We extend the framework to consider the recovery of the system of systems in which the plant is embedded. As a test system, we consider the impacts produced on a nuclear power plant (the critical plant) embedded in the connected power and water distribution, and transportation networks which support its operation. The Seismic Probabilistic Risk Assessment of such system of systems is carried out by Hierarchical modeling and Monte Carlo simulation. First, we perform a top-down analysis through a hierarchical model to identify the elements that at each level have most influence in restoring safety, adopting the criticality importance measure as a quantitative indicator. Then, we evaluate by Monte Carlo simulation the probability that the nuclear power plant enters in an unsafe state and the time needed to recover its safety. The results obtained allow the identification of those elements most critical for the safety and recovery of the nuclear power plant; this is relevant for determining improvements of their structural/functional responses and supporting the decision-making process on safety critical-issues. On the test system considered, under the given assumptions, the components of the external and internal water systems (i.e., pumps and pool) turn out to be the most critical for the safety and recovery of the plant. - Highlights: • We adopt a system-of-system framework to analyze the safety of a critical plant exposed to risk from external events, considering also the interdependent infrastructures that support the plant. • We develop a hierarchical modeling framework to represent the system of systems, accounting also for its recovery. • Monte Carlo simulation is used for the quantitative evaluation of the

  12. Modernization of the turbine control technique and the turbine hydraulics aimed to improved maneuverability in the load range, system safety and plant availability, plant transparency for diagnosis and long-term performance

    International Nuclear Information System (INIS)

    Baran, Detlef

    2012-01-01

    In the contribution H.Mauell GmbH presents modernization projects for the nuclear power plants Tihange-3 and Doel-4. The project volume included control technique and the turbine hydraulics for the steam turbo generating set including turbine auxiliary devices and two turbine feeding pumps. The modernizations were successfully completed in 2010 and 2011, respectively. The nuclear power plants are trouble-free operated.

  13. Study of system safety evaluation on LTO of national project. NISA safety research project on system safety of nuclear power plants

    International Nuclear Information System (INIS)

    Takizawa, Masayuki; Sekimura, Naoto; Miyano, Hiroshi; Aoyama, Katsunobu

    2012-01-01

    Japanese safety regulatory body, that is, Nuclear and Industrial Safety Agency (NISA) started a 5-year national safety research project as 'the first stage' from 2006 FY to 2010 FY whose objective is 'Improve the technical information basis in order to utilize knowledge as well as information related to ageing management and maintenance of NPPs. Fukushima disaster happened in March 2011, and the priority of research needs for ageing management dramatically changed in Japan. The second-stage national project started in October 2011 with the concept of 'system safety' of NNPs where not only ageing management on degradation phenomena of important components but also safety management on total plant systems are paid attention to. The second-stage project is so called 'Japanese Ageing Management Program for System Safety (JAMPSS)'. (author)

  14. Empirical analysis of selected nuclear power plant maintenance factors and plant safety

    International Nuclear Information System (INIS)

    Olson, J.; Osborn, R.N.; Thurber, J.A.; Sommers, P.E.; Jackson, D.H.

    1985-07-01

    This report contains a statistical analysis of the relationship between selected aspects of nuclear power plant maintenance programs and safety related performance. The report identifies a large number of maintenance resources which can be expected to influence maintenance performance and subsequent plant safety performance. The resources for which data were readily available were related statistically to two sets of performance indicators: maintenance intermediate safety indicators and final safety performance indicators. The results show that the administrative structure of the plant maintenance program is a significant predictor of performance on both sets of indicators

  15. Protein improvement in crop plants

    Energy Technology Data Exchange (ETDEWEB)

    Rabson, R

    1974-07-01

    There are compelling reasons for attempting to increase the quality and quantity of protein available in crop plants through plant breeding, despite the fact that some critics have argued that no worldwide protein shortage exists. What used to be thought of as a 'protein gap' has now come to be considered in terms of protein-calorie malnutrition. This is only right since protein and calorie nutrition are inextricable. t the moment there are still unanswered questions as to the precise protein requirements of humans as a function of age, health and ambient conditions. There are, in addition, some indications that the incidence of Kwashiorkor (protein deficiency disease) is increasing in different parts of the world. At a recent meeting of the Protein Advisory Group of the United Nations System, Dr. Jean Mayer, an eminent human nutritionist of Harvard University, U.S.A., indicated the reasons for concern for the current food situation generally, and the protein food supply in particular. These factors include: - Immoderate continuing human population increases, most pronounced in some poor developing countries. - The highly accelerated consumption of animal foods associated with increasing affluence in the richer countries of the world. The production of such foods as meat demands great expenditures of grain, which is an inefficient mode of obtaining the required calories and protein for human consumption. - The over-exploitation of many of the world's fishery resources resulting in reduced yields, perhaps irreversibly, of some fishes. - Recent price increases in petroleum and fertilizer products which have imposed a major obstacle to increasing crop production. - The apparent alteration of climates in places like Africa, Asia and other parts of the Northern hemisphere which may put significant restrictions on crop production. hey are cogent reasons to be seriously concerned about these matters. (author)

  16. Protein improvement in crop plants

    International Nuclear Information System (INIS)

    Rabson, R.

    1974-01-01

    There are compelling reasons for attempting to increase the quality and quantity of protein available in crop plants through plant breeding, despite the fact that some critics have argued that no worldwide protein shortage exists. What used to be thought of as a 'protein gap' has now come to be considered in terms of protein-calorie malnutrition. This is only right since protein and calorie nutrition are inextricable. t the moment there are still unanswered questions as to the precise protein requirements of humans as a function of age, health and ambient conditions. There are, in addition, some indications that the incidence of Kwashiorkor (protein deficiency disease) is increasing in different parts of the world. At a recent meeting of the Protein Advisory Group of the United Nations System, Dr. Jean Mayer, an eminent human nutritionist of Harvard University, U.S.A., indicated the reasons for concern for the current food situation generally, and the protein food supply in particular. These factors include: - Immoderate continuing human population increases, most pronounced in some poor developing countries. - The highly accelerated consumption of animal foods associated with increasing affluence in the richer countries of the world. The production of such foods as meat demands great expenditures of grain, which is an inefficient mode of obtaining the required calories and protein for human consumption. - The over-exploitation of many of the world's fishery resources resulting in reduced yields, perhaps irreversibly, of some fishes. - Recent price increases in petroleum and fertilizer products which have imposed a major obstacle to increasing crop production. - The apparent alteration of climates in places like Africa, Asia and other parts of the Northern hemisphere which may put significant restrictions on crop production. hey are cogent reasons to be seriously concerned about these matters. (author)

  17. Research notes : are safety corridors really safe? Evaluation of the corridor safety improvement program.

    Science.gov (United States)

    1998-08-26

    High accident frequencies on Oregons highway corridors are of concern to the Oregon Department of Transportation (ODOT). : ODOT adopted the Corridor Safety Improvement Program as part of an overall program of safety improvements using federal and ...

  18. Nuclear power plant's safety and risk (requirements of safety and reliability)

    International Nuclear Information System (INIS)

    Franzen, L.F.

    1977-01-01

    Starting out from the given safety objectives as they have evolved during the past few years and from the present legal and regulatory provisions for the construction and operation of nuclear power plants, the hazards involved in regular operation, accidents and emergency situations are discussed. In compliance with the positive safety balance of nuclear power plants in the FRG, special attention is focused on the preventive safety analysis within the frame of the nuclear licensing procedure. Reference is made to the beginnings of a comprehensive hazard concept for an unbiased plant assessment. Emergency situations are discussed from the point of view of general hazard comparisons. (orig.) [de

  19. Study on 'Safety qualification of process computers used in safety systems of nuclear power plants'

    International Nuclear Information System (INIS)

    Bertsche, K.; Hoermann, E.

    1991-01-01

    The study aims at developing safety standards for hardware and software of computer systems which are increasingly used also for important safety systems in nuclear power plants. The survey of the present state-of-the-art of safety requirements and specifications for safety-relevant systems and, additionally, for process computer systems has been compiled from national and foreign rules. In the Federal Republic of Germany the KTA safety guides and the BMI/BMU safety criteria have to be observed. For the design of future computer-aided systems in nuclear power plants it will be necessary to apply the guidelines in [DIN-880] and [DKE-714] together with [DIN-192]. With the aid of a risk graph the various functions of a system, or of a subsystem, can be evaluated with regard to their significance for safety engineering. (orig./HP) [de

  20. Applying artificial intelligence to safety and maintenance (in nuclear plants)

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Under contract from the US DoE, Odetics Inc. is developing a prototype advisory expert system which may eventually be used for the maintenance and safety of a nuclear plant. The objective of the system will be to determine off-normal conditions and to help operations personnel through a course of action which would lead to a stable plant. (UK)

  1. Safety goals for nuclear power plants: a discussion paper

    International Nuclear Information System (INIS)

    1982-02-01

    This report includes a proposed policy statement on safety goals for nuclear power plants published by the Commission for public comment and a supporting discussion paper. Proposed qualitative goals and associated numerical guidelines for nuclear power-plant accident risks are presented. The significance of the goals and guidelines, their bases and rationale, and their proposed mode of implementation are discussed

  2. B plant/WESF integrated annual safety appraisal

    International Nuclear Information System (INIS)

    Anderson, J.K.

    1990-12-01

    This report provides the results of the Fiscal Year 1990 Annual Integrated Safety Appraisal of the B Plant and Waste Encapsulation and Storage Facility in the Hanford Site 200 East Area. The appraisal was conducted in August and September 1990, by the Defense Waste Disposal Safety group, in conjunction with Health Physics and Emergency Preparedness. Reports of these three organizations for their areas of responsibility are presented. The purpose of the appraisal was to determine if the areas being appraised meet US Department of Energy (DOE) and Westinghouse Hanford Company (WHC) requirements and current industry standards of good practice. A further purpose was to identify areas in which program effectiveness could be improved. In accordance with the guidance of WHC Management Requirements and Procedures 5.6, previously identified deficiencies which are being resolved by line management were not repeated as Findings or Observations unless progress or intended disposition was considered to be unsatisfactory. The overall assessment is that there are no major safety problems associated with current operations. Programs are in place to provide the necessary safety controls, evaluations, overviews, and support. In most respects these programs are being implemented effectively. However, there are a number of deficiencies in details of program design and implementation. The appraisal identified a total of 23 Findings and 27 Observations of deficiencies. All Observations are Seriousness Category 3. Fifteen Findings were Category 2 and 8 were Category 3. Most of the Category 2 Findings were so categorized on the basis of noncompliance with mandatory DOE Orders or WHC policies and procedures, rather than potential risk to personnel

  3. Safety-related occurrences at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Reponen, H.; Viitasaari, O.

    1985-04-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STUK-B-YTO 7), which is supplemented by this report intended for experts. (author)

  4. Safety-related occurrences at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Viitasaari, O.; Rantavaara, A.

    1984-03-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Report for this period, Operation of Finnish Nuclear Power Plants (STL-B-RTO-83/7), which is supplemented by this report intended principally for experts. (author)

  5. Safety-related incidents at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Lehtinen, P.

    1985-01-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

  6. Safety-related incidents at the Finnish nuclear power plants

    International Nuclear Information System (INIS)

    Lehtinen, P.

    1986-03-01

    This report contains detailed descriptions of operating incidents and other safety-related matters at the Finnish nuclear power plants regarded as significant by the regulatory authority, the Finnish Centre for Radiation and Nuclear Safety. In this connection, an account is given of the practical actions caused by the incidents, and their significance to reactor safety is evaluated. The main features of the incidents are also described in the general Quartely Reports, Operation of Finnish Nuclear Power Plants, which are supplemented by this report intended for experts. (author)

  7. Improvement of plant reliability in PT. Badak LNG plant

    International Nuclear Information System (INIS)

    Achmad, S.; Somantri, A.

    1997-01-01

    PT. Badak's LNG sales commitment has been steadily increasing, therefore, there has been more emphasis to improve and maintain the LNG plant reliability. From plant operation historical records, Badak LNG plant experienced a high number of LNG process train trips and down time for 1977 through 1988. The highest annual number of LNG plant trips (50 times) occurred in 1983 and the longest LNG process train down time (1259 train-hours) occurred in 1988. Since 1989, PT. Badak has been able to reduce the number of LNG process train trips and down time significantly. In 1994 the number of LNG process train trips and was 18 times and the longest LNG process train down time was 377 train-hours. This plant reliability improvement was achieved by implementing plant reliability improvement programs beginning with the design of the new facilities and continuing with the maintenance and modification of the existing facilities. To improve reliability of the existing facilities, PT. Badak has been implementing comprehensive maintenance programs, to reduce the frequency and down time of the plant, such as Preventive and Predictive Maintenance as well as procurement material improvement since PT. Badak location is in a remote area. By implementing the comprehensive reliability maintenance, PT. Badak has been able to reduce the LNG process train trips to 18 and down time to 337 train hours in 1994 with the subsequent maintenance cost reduction. The average PT. Badak plant availability from 1985 to 1995 is 94.59%. New facilities were designed according to the established PT. Badak design philosophy, master plan and specification. Design of new facilities was modified to avoid certain problems from past experience. (au)

  8. Health innovation for patient safety improvement

    Directory of Open Access Journals (Sweden)

    Renukha Sellappans

    2013-01-01

    Full Text Available Medication error has been identified as a major factor affecting patient safety. Many innovative efforts such as Computerised Physician Order Entry (CPOE, a Pharmacy Information System, automated dispensing machines and Point of Administration Systems have been carried out with the aim of improving medication safety. However, areas remain that require urgent attention. One main area will be the lack of continuity of care due to the breakdown of communication between multiple healthcare providers. Solutions may include consideration of “health smart cards” that carry vital patient medical information in the form of a “credit card” or use of the Malaysian identification card. However, costs and technical aspects associated with the implementation of this health smart card will be a significant barrier. Security and confidentiality, on the other hand, are expected to be of primary concern to patients. Challenges associated with the implementation of a health smart card might include physician buy-in for use in his or her everyday practice. Training and technical support should also be available to ensure the smooth implementation of this system. Despite these challenges, implementation of a health smart card moves us closer to seamless care in our country, thereby increasing the productivity and quality of healthcare.

  9. Health innovation for patient safety improvement.

    Science.gov (United States)

    Sellappans, Renukha; Chua, Siew Siang; Tajuddin, Nur Amani Ahmad; Mei Lai, Pauline Siew

    2013-01-01

    Medication error has been identified as a major factor affecting patient safety. Many innovative efforts such as Computerised Physician Order Entry (CPOE), a Pharmacy Information System, automated dispensing machines and Point of Administration Systems have been carried out with the aim of improving medication safety. However, areas remain that require urgent attention. One main area will be the lack of continuity of care due to the breakdown of communication between multiple healthcare providers. Solutions may include consideration of "health smart cards" that carry vital patient medical information in the form of a "credit card" or use of the Malaysian identification card. However, costs and technical aspects associated with the implementation of this health smart card will be a significant barrier. Security and confidentiality, on the other hand, are expected to be of primary concern to patients. Challenges associated with the implementation of a health smart card might include physician buy-in for use in his or her everyday practice. Training and technical support should also be available to ensure the smooth implementation of this system. Despite these challenges, implementation of a health smart card moves us closer to seamless care in our country, thereby increasing the productivity and quality of healthcare.

  10. Social contention about safety of nuclear power plant

    International Nuclear Information System (INIS)

    Nemoto, Kazuyasu

    1978-01-01

    In Japan, the contentions and arguments on the safety of nuclear power generation have been active since its first introduction, and these are greatly influenced by the nation's experiences of atomic bombs in Hiroshima, Nagasaki, and Bikini. As the result, the attitude of peoples toward the acceptance of nuclear power plants is significantly different from that in other countries. The situation in Japan of social contentions about nuclear power safety is explained in two aspects: acceptance of the safety, by peoples and Japanese pattern of safety contentions. In both upstream and downstream of nuclear power generation, not only the safety but also the right or wrong for nuclear power generation itself is discussed. The problem of nuclear power safety has gone into the region beyond the technological viewpoint. The pattern of safety contentions in Japan is the entanglement of three sectors; i.e. local people, labor unions and political parties, enterprises and administration, and intellectuals. (Mori, K.)

  11. Plant improvement by known-function genes

    Directory of Open Access Journals (Sweden)

    Jesús Quiroz Chávez

    2012-09-01

    Full Text Available Plant molecular improvement by recombinant DNA technology represents an advantage to obtain new varieties or traits. This technique is promised for genetic improvement of crop plants. Lines with increased yield, quality, disease resistance, or tolerant to abiotic stress have been obtained, with clear advantages for producers, marketers and consumers. However, they have several limitations in its application to agriculture because of its risk and hazards. The aim of the document is to show the advantages and disadvantages of GM crop plant, to develop represent an opportunity to have new exotic traits.

  12. Nitrogen-system safety study: Portsmouth Gaseous Diffusion Plant

    International Nuclear Information System (INIS)

    1982-07-01

    The Department of Energy has primary responsibility for the safety of operations at DOE-owned nuclear facilities. The guidelines for the analysis of credible accidents are outlined in DOE Order 5481.1. DOE has requested that existing plant facilities and operations be reviewed for potential safety problems not covered by standard industrial safety procedures. This review is being conducted by investigating individual facilities and documenting the results in Safety Study Reports which will be compiled to form the Existing Plant Final Safety Analysis Report which is scheduled for completion in September, 1984. This Safety Study documents the review of the Plant Nitrogen System facilities and operations and consists of Section 4.0, Facility and Process Description, and Section 5.0, Accident Analysis, of the Final Safety Analysis Report format. The existing nitrogen system consists of a Superior Air Products Company Type D Nitrogen Plant, nitrogen storage facilities, vaporization facilities and a distribution system. The system is designed to generate and distribute nitrogen gas used in the cascade for seal feed, buffer systems, and for servicing equipment when exceptionally low dew points are required. Gaseous nitrogen is also distributed to various process auxiliary buildings. The average usage is approximately 130,000 standard cubic feet per day

  13. Safety-related decision making at a nuclear power plant

    International Nuclear Information System (INIS)

    Vaurio, J.K.

    1998-01-01

    The decision making environment of an operating nuclear power plant is presented. The organizations involved, their roles and interactions as well as the main influencing factors and decision criteria are described. The focus is on safety-related decisions, and the framework is based on the situation at Loviisa power station. The role of probabilistic safety assessment (PSA) is illustrated with decisions concerning plant modifications, optimization, acceptance of temporary configurations and extended repair times. Suggestions are made for rational and flexible risk-based control of allowed times to operate the plant with some components out of service. (orig.)

  14. Safety and operation of the Stade nuclear power plant

    International Nuclear Information System (INIS)

    Salcher, H.

    1991-01-01

    The concept of PreussenElektra is to continuously increase the existing safety standard of the Stade nuclear power station using experience gained from faults and operation in nuclear power stations and the progressive state of the art. Modifications to achieve the most gentle operation of the plant have been completed and other are on-going. To do so instruments were attached to those components which are susceptible to fatigue to record the transients and extensive calculatory records were kept. Although the plant has almost 20 years successful operation behind it, it can still stand up well to comparisons with more recent plants as far as safety aspects are concerned. 6 figs

  15. Safety and Radiation Protection at Swedish Nuclear Power Plants 2005

    International Nuclear Information System (INIS)

    2006-05-01

    In 2005, no severe events occurred which challenged the safety at the Swedish nuclear power plants. However, some events have been given a special focus. The 'Gudrun' storm, which occurred in January 2005, affected the operation of the reactors at Ringhals and Barsebaeck 2. At Ringhals, the switchyards were affected by salt deposits and, at Barsebaeck, the 400kV grid was subjected to interruptions. The long-term trend is that the total number of fuel defects in Swedish reactors is decreasing. The damage that occurs nowadays has mainly been caused by small objects entering the fuel via the coolant and fretting holes in the cladding. To reduce the number of defects of this type, fuel with filters is successively being introduced to prevent debris from entering the fuel assemblies and cyclone filters in the facility which cleans the coolant. Since the mid-nineties, the pressurised water reactors, Ringhals 2, 3 and 4, have had problems with fuel rod bowing in excess of the safety analysis calculations. Ringhals AB (RAB) has adopted measures to rectify the bowing. Follow-up work shows that the fuel rod bowing is decreasing. The followup in 2005 of damaged tubes in the Ringhals 4 steam generators indicates a continued slow damage propagation. Tubes with defects of such a limited extent that there are adequate margins to rupture and loosening have been kept in operation. Damaged tubes with insufficient margins have plugged. During the year, previously observed minor leakage from the reactor containment in Ringhals 2 was investigated in greater detail and repaired. The investigations showed extensive corrosion attack caused by deficiencies in connection with containment construction. The ageing of electrical cables and other equipment in the I-C systems has been examined by SKI. Regulatory supervision has so far shown that these issues are largely handled in a satisfactory manner by the licensees but that certain supplementary investigations and other measures need to be

  16. Safety and Radiation Protection at Swedish Nuclear Power Plants 2005

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-05-15

    In 2005, no severe events occurred which challenged the safety at the Swedish nuclear power plants. However, some events have been given a special focus. The 'Gudrun' storm, which occurred in January 2005, affected the operation of the reactors at Ringhals and Barsebaeck 2. At Ringhals, the switchyards were affected by salt deposits and, at Barsebaeck, the 400kV grid was subjected to interruptions. The long-term trend is that the total number of fuel defects in Swedish reactors is decreasing. The damage that occurs nowadays has mainly been caused by small objects entering the fuel via the coolant and fretting holes in the cladding. To reduce the number of defects of this type, fuel with filters is successively being introduced to prevent debris from entering the fuel assemblies and cyclone filters in the facility which cleans the coolant. Since the mid-nineties, the pressurised water reactors, Ringhals 2, 3 and 4, have had problems with fuel rod bowing in excess of the safety analysis calculations. Ringhals AB (RAB) has adopted measures to rectify the bowing. Follow-up work shows that the fuel rod bowing is decreasing. The followup in 2005 of damaged tubes in the Ringhals 4 steam generators indicates a continued slow damage propagation. Tubes with defects of such a limited extent that there are adequate margins to rupture and loosening have been kept in operation. Damaged tubes with insufficient margins have plugged. During the year, previously observed minor leakage from the reactor containment in Ringhals 2 was investigated in greater detail and repaired. The investigations showed extensive corrosion attack caused by deficiencies in connection with containment construction. The ageing of electrical cables and other equipment in the I-C systems has been examined by SKI. Regulatory supervision has so far shown that these issues are largely handled in a satisfactory manner by the licensees but that certain supplementary investigations and other measures

  17. Evaluation of seismic hazards for nuclear power plants. Safety guide

    International Nuclear Information System (INIS)

    2002-01-01

    The main objective of this Safety Guide is to provide recommendations on how to determine the ground motion hazards for a plant at a particular site and the potential for surface faulting, which could affect the feasibility of construction and safe operation of a plant at that site. The guidelines and procedures presented in this Safety Guide can appropriately be used in evaluations of site suitability and seismic hazards for nuclear power plants in any seismotectonic environment. The probabilistic seismic hazard analysis recommended in this Safety Guide also addresses the needs for seismic hazard analysis of external event PSAs conducted for nuclear power plants. Many of the methods and processes described may also be applicable to nuclear facilities other than power plants. Other phenomena of permanent ground displacement (liquefaction, slope instability, subsidence and collapse) as well as the topic of seismically induced flooding are treated in Safety Guides relating to foundation safety and coastal flooding. Recommendations of a general nature are given in Section 2. Section 3 discusses the acquisition of a database containing the information needed to evaluate and address all hazards associated with earthquakes. Section 4 covers the use of this database for construction of a seismotectonic model. Sections 5 and 6 review ground motion hazards and evaluations of the potential for surface faulting, respectively. Section 7 addresses quality assurance in the evaluation of seismic hazards for nuclear power plants

  18. Nuclear safety in fuel-reprocessing plants

    International Nuclear Information System (INIS)

    Hennies, H.H.; Koerting, K.

    1976-01-01

    The danger potential of nuclear power and fuel reprocessing plants in normal operation is compared. It becomes obvious that there are no basic differences. The analysis of possible accidents - blow-up of an evaporator for highly active wastes, zircaloy burning, cooling failure in self-heating process solutions, burning of a charged solvent, criticality accidents - shows that they are kept under control by the plant layout. (HP) [de

  19. Floating nuclear power plant safety assurance principles

    International Nuclear Information System (INIS)

    Zvonarev, B.M.; Kuchin, N.L.; Sergeev, I.V.

    1993-01-01

    In the north regions of the Russian federation and low density population areas, there is a real necessity for ecological clean energy small power sources. For this purpose, floating nuclear power plants, designed on the basis of atomic ship building engineering, are being conceptualized. It is possible to use the ship building plants for the reactor purposes. Issues such as radioactive waste management are described

  20. Nuclear safety improvement activities related to WWER-440 units in Bulgaria

    International Nuclear Information System (INIS)

    Gantchev, T.

    1998-01-01

    The systematic evaluation of the deficiencies of the original design of the WWER reactors brought to the development of a Short Term Programme for Safety Upgrading and Modernisation of Kozloduy WWER-440 units. The implementation of this Programme was completed in 1997. The strive for continuos improvement of Kozloduy Nuclear Power Plant (NPP) safety level, the new requirements of the Bulgarian Nuclear Safety Authority and the public concern initiated the development of new Complex Programme for Safety Improvement (PRG'97), now in a process of implementation. (author)

  1. Safety evaluation of the nuclear power plant at Cattenom

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    This is a systematic compilation of the material which was dealt with at the level of the German-French Commission (on questions of the safety of nuclear installations) in this discussions about the nuclear power plant at Cattenom. As a supplement to the report published already in 1982, the Commission has officially released its deliberation results on the subjects constructive safety measures, radiological effects, and precautions in case of an emergency. The allegations according to which the installation is wanting in safety are countered by the joint statement of the chairmen of GPR (Permanent Group on Reactors) and RSK (German Commission on Reactor Safety) of August 29, 1986. (HSCH) [de

  2. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (French Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  3. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Russian Edition)

    International Nuclear Information System (INIS)

    2016-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  4. Safety of Nuclear Power Plants: Design. Specific Safety Requirements (Arabic Edition)

    International Nuclear Information System (INIS)

    2017-01-01

    This publication establishes requirements applicable to the design of nuclear power plants and elaborates on the safety objective, safety principles and concepts that provide the basis for deriving the safety requirements that must be met for the design of a nuclear power plant. It will be useful for organizations involved in design, manufacture, construction, modification, maintenance, operation and decommissioning of nuclear power plants, as well as for regulatory bodies. A review of Safety Requirements publications was commenced in 2011 following the accident in the Fukushima Daiichi nuclear power plant in Japan. The review revealed no significant areas of weakness and resulted in just a small set of amendments to strengthen the requirements and facilitate their implementation, which are contained in the present publication.

  5. Research reactor management. Safety improvement activities in HANARO

    International Nuclear Information System (INIS)

    Wu, Jong-Sup; Jung, Hoan-Sung; Hong, Sung Taek; Ahn, Guk-Hoon

    2012-01-01

    Safety activities in HANARO have been continuously conducted to enhance its safe operation. Great effort has been placed on a normalization and improvement of the safety attitude of the regular staff and other employees working at the reactor and other experimental facilities. This paper introduces the activities on safety improvement that were performed over the last few years. (author)

  6. 78 FR 48029 - Improving Chemical Facility Safety and Security

    Science.gov (United States)

    2013-08-07

    ... Improving Chemical Facility Safety and Security By the authority vested in me as President by the... at reducing the safety risks and security risks associated with hazardous chemicals. However... to further improve chemical facility safety and security in coordination with owners and operators...

  7. Applications of computer based safety systems in Korea nuclear power plants

    International Nuclear Information System (INIS)

    Won Young Yun

    1998-01-01

    With the progress of computer technology, the applications of computer based safety systems in Korea nuclear power plants have increased rapidly in recent decades. The main purpose of this movement is to take advantage of modern computer technology so as to improve the operability and maintainability of the plants. However, in fact there have been a lot of controversies on computer based systems' safety between the regulatory body and nuclear utility in Korea. The Korea Institute of Nuclear Safety (KINS), technical support organization for nuclear plant licensing, is currently confronted with the pressure to set up well defined domestic regulatory requirements from this aspect. This paper presents the current status and the regulatory activities related to the applications of computer based safety systems in Korea. (author)

  8. An introduction to a new IAEA safety guide: 'ageing management for nuclear power plants'

    International Nuclear Information System (INIS)

    Pachner, J.; Inagaki, T.; Kang, K.S.

    2008-01-01

    This paper reports on a new IAEA Safety Guide entitled 'Ageing Management for Nuclear Power Plants' which is currently in an advanced draft form, awaiting approval of publication. The new Safety Guide will be an umbrella document for a comprehensive set of guidance documents on ageing management which have been issued by the IAEA. The Safety Guide first presents basic concepts of ageing management as a common basis for the recommendations on: proactive management of ageing throughout the life cycle of a nuclear power plant (NPP); systematic approach to managing ageing in the operation of NPPs; managing obsolescence; and review of ageing management for long term operation (life extension). The Safety Guide is intended to assist operators in establishing, implementing and improving systematic ageing management programs in NPPs and may be used by regulators in preparing regulatory standards and guides, and in verifying that ageing in nuclear power plants is being effectively managed. (author)

  9. A study on safety climate at nuclear power plants

    International Nuclear Information System (INIS)

    Fukui, Hirokazu; Yoshida, Michio; Yoshiyama, Naohiro

    2001-01-01

    factors as the rating scales of the safety climate. In order to study the characteristics of the safety climate at nuclear power plants, we used a causal model with safety confirmation/report' as the result and other factors as forecasting factors. As a result of the covariance structure analysis using the causal model, it was found that 'safety confirmation/report' is an action based on confidence in knowledge and skill', and is supported by 'attitude of supervisors' and 'clarity of tasks.' The analytical results also indicate that 'safety education in workplace' plays an important role in promoting the sharing of information as a medium factor. As described, 'attitude of supervisors,' 'clarity of tasks' and 'safety education in workplace,' all of which are organizational environment factors, are important forecasting factors that influence individuals' safety actions and hence considered as constituents of the safety climate. (author)

  10. Rhizosphere pseudomonads as probiotics improving plant health.

    Science.gov (United States)

    Kim, Young Cheol; Anderson, Anne J

    2018-04-20

    Many root-colonizing microbes are multifaceted in traits that improve plant health. Although isolates designated as biological control agents directly reduce pathogen growth, many exert additional beneficial features that parallel changes induced in animal and other hosts by health-promoting microbes termed probiotics. Both animal and plant probiotics cause direct antagonism of pathogens and induce systemic immunity in the host to pathogens and other stresses. They also alter host development, and improve host nutrition. The probiotic root-colonizing pseudomonads are generalists in terms of plant hosts, soil habitats and the array of stress responses that are ameliorated in the plant. This review illustrates how the probiotic pseudomonads, nurtured by the C and N sources released by the plant in root exudates, form protective biofilms on the root surface and produce the metabolites or enzymes to boost plant health. The findings reveal the multifunctional nature of many of the microbial metabolites in the plant-probiotic interplay. The beneficial effects of probiotics on plant function can contribute to sustainable yield and quality in agricultural production. This article is protected by copyright. All rights reserved. © 2018 BSPP and John Wiley & Sons Ltd.

  11. An assessment of criticality safety at the Department of Energy Rocky Flats Plant, Golden, Colorado, July--September 1989

    Energy Technology Data Exchange (ETDEWEB)

    Mattson, Roger J.

    1989-09-01

    This is a report on the 1989 independent Criticality Safety Assessment of the Rocky Flats Plant, primarily in response to public concerns that nuclear criticality accidents involving plutonium may have occurred at this nuclear weapon component fabrication and processing plant. The report evaluates environmental issues, fissile material storage practices, ventilation system problem areas, and criticality safety practices. While no evidence of a criticality accident was found, several recommendations are made for criticality safety improvements. 9 tabs.

  12. International conference on the strengthening of nuclear safety in Eastern Europe. Keynote papers. Regulatory aspects of NPP safety, status of safety improvements, status of safety analysis report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-06-01

    The Objective of the Conference was to assess the past decade of nuclear safety efforts in countries operating WWER and RBMK nuclear reactors and to address remaining safety issues which require further work. A particular focus of the Conference was on international co-operation and assistance and where such efforts should be focused in the future. All Eastern European countries that operate RBMK or WWER reactors participated in the Conference, and presented papers on three key areas of nuclear safety: Regulatory Aspects of Nuclear Power Plant Safety; Status of Safety Improvements; and Status of Safety Analysis Reports. In addition, representatives from 18 additional countries that provide financial and/or technical assistance and co-operation in the area of WWER and RBMK safety offered the most extensive commentary. Key international (IAEA, World Association of Nuclear Operators, the Nuclear Energy Agency, the G-24 NUSAC, the European Commission, and the EBRD) organizations that provide nuclear safety assistance for WWER and RBMK reactors also made presentations. There is no question that considerable progress on nuclear safety has been made in Eastern Europe. Special mention should be made of successful efforts to strengthen the independence and technical competence of the nuclear regulatory authorities. Efforts should now concentrate on improving the depth and scope of the technical abilities of the regulatory authorities. More attention by governments is needed to ensure that the regulatory authorities have the financial resources and enforcement authority to fully execute their missions. In respect to the operators of the nuclear power plants, they have demonstrated clear progress in operational safety improvements. Significant additional efforts are required to maintain and enhance an effective safety culture. Design safety improvement programmes are in place in all countries. Implementation of these programmes has varied and is particularly affected by

  13. International conference on the strengthening of nuclear safety in Eastern Europe. Keynote papers. Regulatory aspects of NPP safety, status of safety improvements, status of safety analysis report

    International Nuclear Information System (INIS)

    1999-06-01

    The Objective of the Conference was to assess the past decade of nuclear safety efforts in countries operating WWER and RBMK nuclear reactors and to address remaining safety issues which require further work. A particular focus of the Conference was on international co-operation and assistance and where such efforts should be focused in the future. All Eastern European countries that operate RBMK or WWER reactors participated in the Conference, and presented papers on three key areas of nuclear safety: Regulatory Aspects of Nuclear Power Plant Safety; Status of Safety Improvements; and Status of Safety Analysis Reports. In addition, representatives from 18 additional countries that provide financial and/or technical assistance and co-operation in the area of WWER and RBMK safety offered the most extensive commentary. Key international (IAEA, World Association of Nuclear Operators, the Nuclear Energy Agency, the G-24 NUSAC, the European Commission, and the EBRD) organizations that provide nuclear safety assistance for WWER and RBMK reactors also made presentations. There is no question that considerable progress on nuclear safety has been made in Eastern Europe. Special mention should be made of successful efforts to strengthen the independence and technical competence of the nuclear regulatory authorities. Efforts should now concentrate on improving the depth and scope of the technical abilities of the regulatory authorities. More attention by governments is needed to ensure that the regulatory authorities have the financial resources and enforcement authority to fully execute their missions. In respect to the operators of the nuclear power plants, they have demonstrated clear progress in operational safety improvements. Significant additional efforts are required to maintain and enhance an effective safety culture. Design safety improvement programmes are in place in all countries. Implementation of these programmes has varied and is particularly affected by

  14. Safety

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. These activities are reported under the headings: (1) Nuclear safety; (2) Industrial and health safety; (3) Radiation safety; and Fire protection

  15. Revision of AESJ standard 'the code of implemnetation of periodic safety review of nuclear power plants'

    International Nuclear Information System (INIS)

    Hirano, Masashi; Narumiya, Yoshiyuki

    2010-01-01

    The Periodic Safety Review (PSR) was launched in June 1992, when the Agency for Natural Resources and Energy issued a notification that required licensees to conduct comprehensive review on the safety of each existing nuclear power plant (NPP) once approximately every ten years based on the latest technical findings for the purpose of improving the safety of the NPP. In 2006, the Standard Committee of the Atomic Energy Society of Japan established the first version of 'The Standard of Implementation for Periodic Safety Review of Nuclear Power Plants: 2006'. Taking into account developments in safety regulation of PSR after the issuance of the first version, the Standard Committee has revised the Standard. This paper summarizes background on PSR, such developments are major contents of the Standard as well as the focal points of the revision. (author)

  16. Research on integrated managing system based on CIMS for nuclear power plant safety

    International Nuclear Information System (INIS)

    Zhou Gang

    2006-01-01

    In order to improve safety, economy and reliability of operation for nuclear power plant (NPP), a novel integrated managing method was proposed based on the ideas of computer and contemporary integrated manufacturing system (CIMS). The application of CIMS to nuclear power plant safety management was researched. In order to design an integrated managing system to meet the needs of NPP safety management, all work related to nuclear safety is divided into different category according to its characters. On basis of this work, general integrated managing system was designed at first. Then subsystems were designed and every subsystem implements a category of nuclear safety management work. All subsystems are independent relatively on the one hand and are interrelated on other hand by global information system. (authors)

  17. A defence in depth approach to safety assessment of existing nuclear power plant

    International Nuclear Information System (INIS)

    Butcher, P.; Holloway, N.J.

    1998-01-01

    The safety assessment of plant built to earlier standards requires an approach to prioritisation of upgrades that is based on sound engineering and safety principles. The principles of defence in depth are universally accepted and can form the basis of a prioritisation scheme for safety issues, and hence for the upgrading required to address them. The described scheme includes criteria for acceptability and issue prioritisation that are based on the number of lines of defence and the consequences of their failure. They are thus equivalent in concept to risk criteria, but are based on deterministic principles. This scheme has been applied successfully to the RBMK plant at Ignalina in Lithuania, for which a Western-style Safety Analysis Report has recently been produced and reviewed by joint Western and Eastern teams. An extended Safety Improvement Programme (SIP2) has been developed and agreed, based on prioritisations from the defence in depth assessment. (author)

  18. Most significant preliminary results of the probabilistic safety analysis on the Juragua nuclear power plant

    International Nuclear Information System (INIS)

    Perdomo, Manuel

    1995-01-01

    Since 1990 the Group for PSA Development and Applications (GDA/APS) is working on the Level-1 PSA for the Juragua-1 NPP, as a part of an IAEA Technical Assistance Project. The main objective of this study, which is still under way, is to assess, in a preliminary way, the Reactor design safety to find its potential 'weak points' at the construction stage, using a eneric data base. At the same time, the study allows the PSA team to familiarize with the plant design and analysis techniques for the future operational PSA of the plant. This paper presents the most significant preliminary results of the study, which reveal some advantages of the safety characteristics of the plant design in comparison with the homologous VVER-440 reactors and some areas, where including slight modifications would improve the plant safety, considering the level of detail at which the study is carried out. (author). 13 refs, 1 fig, 2 tabs

  19. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    1999-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  20. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2004-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  1. Decommissioning of nuclear power plants and research reactors. Safety guide

    International Nuclear Information System (INIS)

    2001-01-01

    Radioactive waste is produced in the generation of nuclear power and the use of radioactive materials in industry, research and medicine. The importance of the safe management of radioactive waste for the protection of human health and the environment has long been recognized, and considerable experience has been gained in this field. The IAEA's Radioactive Waste Safety Standards Programme aimed at establishing a coherent and comprehensive set of principles and requirements for the safe management of waste and formulating the guidelines necessary for their application. This is accomplished within the IAEA Safety Standards Series in an internally consistent set of publications that reflect an international consensus. The publications will provide Member States with a comprehensive series of internationally agreed publications to assist in the derivation of, and to complement, national criteria, standards and practices. The Safety Standards Series consists of three categories of publications: Safety Fundamentals, Safety Requirements and Safety Guides. With respect to the Radioactive Waste Safety Standards Programme, the set of publications is currently undergoing review to ensure a harmonized approach throughout the Safety Standards Series. This Safety Guide addresses the subject of decommissioning of nuclear power plants and research reactors. It is intended to provide guidance to national authorities and operating organizations for the planning and safe management of the decommissioning of such installations. This Safety Guide has been prepared through a series of Consultants and Technical Committee meetings. It supersedes former Safety Series publications Nos 52, 74 and 105

  2. Safety improvement technologies for nuclear power generation

    International Nuclear Information System (INIS)

    Nishida, Koji; Adachi, Hirokazu; Kinoshita, Hirofumi; Takeshi, Noriaki; Yoshikawa, Kazuhiro; Itou, Kanta; Kurihara, Takao; Hino, Tetsushi

    2015-01-01

    As the Hitachi Group's efforts in nuclear power generation, this paper explains the safety improvement technologies that are currently under development or promotion. As efforts for the decommissioning of Fukushima Daiichi Nuclear Power Station, the following items have been developed. (1) As for the spent fuel removal of Unit 4, the following items have mainly been conducted: removal of the debris piled up on the top surface of existing reactor building (R/B), removal of the debris deposited in spent fuel pool (SFP), and fuel transfer operation by means of remote underwater work. The removal of all spent fuels was completed in 2014. (2) The survey robots inside R/B, which are composed of a basement survey robot to check leaking spots at upper pressure suppression chamber and a floor running robot to check leaking spots in water, were verified with a field demonstration test at Unit 1. These robots were able to find the leaking spots at midair pipe expansion joint. (3) As the survey robot for reactor containment shells, robots of I-letter posture and horizontal U-letter posture were developed, and the survey on the upper part of first-floor grating inside the containment shells was performed. (4) As the facilities for contaminated water measures, sub-drain purification equipment, Advanced Liquid Processing System, etc. were developed and supplied, which are now showing good performance. On the other hand, an advanced boiling water reactor with high safety of the United Kingdom (UK ABWR) is under procedure of approval for introduction. In addition, a next-generation light-water reactor of transuranic element combustion type is under development. (A.O.)

  3. A study on safety assessment methodology for a vitrification plant

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Y. C.; Lee, G. S.; Choi, Y. C.; Kim, G. H. [Yonsei Univ., Seoul (Korea, Republic of)

    2002-03-15

    In this study, the technical and regulatory status of radioactive waste vitrification technologies in foreign and domestic plants is investigated and analyzed, and then significant factors are suggested which must be contained in the final technical guideline or standard for the safety assessment of vitrification plants. Also, the methods to estimate the stability of vitrified waste forms are suggested with property analysis of them. The contents and scope of the study are summarized as follows : survey of the status on radioactive waste vitrification technologies in foreign and domestic plants, survey of the characterization methodology for radioactive waste form, analysis of stability for vitrified waste forms, survey and analysis of technical standards and regulations concerned with them in foreign and domestic plants, suggestion of significant factors for the safety assessment of vitrification plants, submission of regulated technical standard on radioactive waste vitrification plats.

  4. Surveillance of items important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    1990-01-01

    The Guide was prepared as part of the IAEA's programme, referred to as the NUSS Programme, for establishing Codes and Safety Guides relating to nuclear power plants. THe Guide supplements the Code on the Safety of Nuclear Power Plants: Operation, IAEA Safety Series No. 50-C-O(Rev.1). The operating organization has overall responsibility for the safe operation of the nuclear power plant. Therefore, it shall ensure that adequate surveillance activities are carried out in order to verify that the plant is operated within the prescribed operational limits and conditions, and to detect in time any deterioration of structures, systems and components as well as any adverse trend that could lead to an unsafe condition. These activities can be classified as: Monitoring plant parameters and system status; Checking and calibrating instrumentation; Testing and inspecting structures, systems and components. This Safety Guide provides guidance and recommendations on surveillance activities to ensure that structures, systems and components important to safety are available to perform their functions in accordance with design intent and assumptions

  5. The experiences to improve plant performance and reliability of Ko-Ri nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Ho Weon [Korea Electric Power Corp. Ko-Ri nuclear power division, Ko-Ri (Korea, Republic of)

    1998-07-01

    This paper provides a discussion of the lessons learned from operational experience and the future plans to improve performance of the Ko-Ri plant. To operate nuclear power plants safely with good performance is the only way to mitigate the negative image of nuclear power generation to the public and to enhance the economical benefit compared to other electrical generation method. Therefore, in a continuous effort to overcome a negative challenge from outside, we have driven an aggressive 'OCTF' campaign as part of safety. As a result of our efforts, the following remarkable achievements have been accomplished. (1) 3 times of OCTF during recent three years (2) Selected twice as a top notch power plant on the list of NEI magazine in terms of plant capacity factor (3) No scram recorded in 1997 for all 4 units at Ko-Ri site. Ko-Ri is now undergoing the large scale plant betterment projects for retaking-off our operating performance to the level of new challenge target. Such improvement of critical components in the reactor coolant system and turbine system greatly contribute to increase the safety and reliability of the plant and to shortening of the planned outage period as well as to reduction of radiation exposure and radwaste. (Cho, G. S.). 5 tabs., 10 figs.

  6. The experiences to improve plant performance and reliability of Ko-Ri nuclear power plants

    International Nuclear Information System (INIS)

    Kang, Ho Weon

    1998-01-01

    This paper provides a discussion of the lessons learned from operational experience and the future plans to improve performance of the Ko-Ri plant. To operate nuclear power plants safely with good performance is the only way to mitigate the negative image of nuclear power generation to the public and to enhance the economical benefit compared to other electrical generation method. Therefore, in a continuous effort to overcome a negative challenge from outside, we have driven an aggressive 'OCTF' campaign as part of safety. As a result of our efforts, the following remarkable achievements have been accomplished. (1) 3 times of OCTF during recent three years (2) Selected twice as a top notch power plant on the list of NEI magazine in terms of plant capacity factor (3) No scram recorded in 1997 for all 4 units at Ko-Ri site. Ko-Ri is now undergoing the large scale plant betterment projects for retaking-off our operating performance to the level of new challenge target. Such improvement of critical components in the reactor coolant system and turbine system greatly contribute to increase the safety and reliability of the plant and to shortening of the planned outage period as well as to reduction of radiation exposure and radwaste. (Cho, G. S.). 5 tabs., 10 figs

  7. Safety goals for future nuclear power plants

    International Nuclear Information System (INIS)

    Todreas, Neil E.

    2001-01-01

    This talk presents technology goals developed for Generation IV nuclear energy systems that can be made available to the market by 2030 or earlier. These goals are defined in the broad areas of sustainability, safety and reliability, and economics. Sustainability goals focus on fuel utilization, waste management, and proliferation resistance. Safety and reliability goals focus on safe and reliable operation, investment protection, and essentially eliminating the need for emergency response. Economics goals focus on competitive life cycle and energy production costs and financial risk. Future reactors fall in three categories - those which are: Certified or derivatives; Designed to a reasonable extent and based on available technology; In conceptual form only with potential to most fully satisfy the GENIV goals

  8. Use of computer codes to improve nuclear power plant operation

    International Nuclear Information System (INIS)

    Misak, J.; Polak, V.; Filo, J.; Gatas, J.

    1985-01-01

    For safety and economic reasons, the scope for carrying out experiments on operational nuclear power plants (NPPs) is very limited and any changes in technical equipment and operating parameters or conditions have to be supported by theoretical calculations. In the Nuclear Power Plant Scientific Research Institute (NIIAEhS), computer codes are systematically used to analyse actual operating events, assess safety aspects of changes in equipment and operating conditions, optimize the conditions, preparation and analysis of NPP startup trials and review and amend operating instructions. In addition, calculation codes are gradually being introduced into power plant computer systems to perform real time processing of the parameters being measured. The paper describes a number of specific examples of the use of calculation codes for the thermohydraulic analysis of operating and accident conditions aimed at improving the operation of WWER-440 units at the Jaslovske Bohunice V-1 and V-2 nuclear power plants. These examples confirm that computer calculations are an effective way of solving operating problems and of further increasing the level of safety and economic efficiency of NPP operation. (author)

  9. Review of Risk Reduction Methods using Probabilistic Safety Assessment Insights and Improved Technology

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Eun-Chan; Choi, Byung-Pil [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-10-15

    As seen in the process of the periodic safety review of domestic nuclear power plants, the risk management objectives such as core damage frequency and large early release frequency are not easy to be met without continuous safety improvements and the integratoin of the improved technologies into the PSA evaluation methodologies. Because external event analyses have a protion of uncertainty factors in the current analysis methodologies, the technical efforts in various perspectives.

  10. Operational safety review programmes for nuclear power plants. Guidelines for assessment

    International Nuclear Information System (INIS)

    2002-01-01

    The IAEA has been offering the Operational Safety Review Team (OSART) programme to provide advice and assistance to Member States in enhancing the operational safety of nuclear power plants (NPPs). Simultaneously, the IAEA has encouraged self-assessment and review by Member States of their own nuclear power plants to continuously improve nuclear safety. Currently, some utilities have been implementing safety review programmes to independently review their own plants. Corporate or national operational safety review programmes may be compliance or performance based. Successful utilities have found that both techniques are necessary to provide assurance that (i) as a minimum the NPP meets specific corporate and legal requirements and (ii) management at the NPP is encouraged to pursue continuous improvement principles. These programmes can bring nuclear safety benefits to the plants and utilities. The IAEA has conducted two pilot missions to assess the effectiveness of the operational review programme. Based on these missions and on the experience gained during OSART missions, this document has been developed to provide guidance on and broaden national/corporate safety review programmes in Member States, and to assist in maximizing their benefits. These guidelines are intended primarily for the IAEA team to conduct assessment of a national/corporate safety review programme. However, this report may also be used by a country or utility to establish its own national/corporate safety review programme. The guidelines may likewise be used for self-assessment or for establishing a baseline when benchmarking other safety review programmes. This report consists of four parts. Section 2 addresses the planning and preparation of an IAEA assessment mission and Sections 3 and 4 deal with specific guidelines for conducting the assessment mission itself

  11. Safety assessment in plant layout design using indexing approach: Implementing inherent safety perspective

    International Nuclear Information System (INIS)

    Tugnoli, Alessandro; Khan, Faisal; Amyotte, Paul; Cozzani, Valerio

    2008-01-01

    Layout planning plays a key role in the inherent safety performance of process plants since this design feature controls the possibility of accidental chain-events and the magnitude of possible consequences. A lack of suitable methods to promote the effective implementation of inherent safety in layout design calls for the development of new techniques and methods. In the present paper, a safety assessment approach suitable for layout design in the critical early phase is proposed. The concept of inherent safety is implemented within this safety assessment; the approach is based on an integrated assessment of inherent safety guideword applicability within the constraints typically present in layout design. Application of these guidewords is evaluated along with unit hazards and control devices to quantitatively map the safety performance of different layout options. Moreover, the economic aspects related to safety and inherent safety are evaluated by the method. Specific sub-indices are developed within the integrated safety assessment system to analyze and quantify the hazard related to domino effects. The proposed approach is quick in application, auditable and shares a common framework applicable in other phases of the design lifecycle (e.g. process design). The present work is divided in two parts: Part 1 (current paper) presents the application of inherent safety guidelines in layout design and the index method for safety assessment; Part 2 (accompanying paper) describes the domino hazard sub-index and demonstrates the proposed approach with a case study, thus evidencing the introduction of inherent safety features in layout design

  12. Improved reliability, maintainability and safety through elastomer upgrading

    International Nuclear Information System (INIS)

    Wensel, R.; Wittich, K.C.

    1995-01-01

    Equipment in nuclear plants has historically contained whatever elastomer each component supplier traditionally used for corresponding non-nuclear service. The resulting proliferation of elastomer compounds, many of which are far from optimal for the service conditions (e.g., pressure, temperature, radiation, etc.), has multiplied the costs to provide station reliability, maintainability and safety. Cost-effective improvements are being achieved in CANDU plants by upgrading and standardizing on a handful of high performing elastomer compounds. These upgraded materials offer significant gains in service life over the materials they replace (often by factors of 2 or more). This rationalization of elastomer compounds also facilitates the EQ process for safety-related equipment. Detailed test data on aging is currently being generated for these specific elastomers, encompassing the conditions and media (air, water, oil) common in CANDU service. Two key elements characterize this testing. First, each result is specific to the compound used in the test, and second, it is specific to the tested failure mode (e.g., compression set, extrusion, fracture, etc.). Having fewer, but more thoroughly tested compounds, avoids the penalty (associated with poorly characterized materials) of having to replace parts prematurely because of conservatism, while maintaining safe, reliable service. This paper provides an overview of this approach covering: the benefits of compound rationalization; and the how and why of establishing relevant failure criteria; appropriate quality assurance to maintain EQ; procurement, storage and handling guidelines; and monitoring and predicting in-service degradation. (author)

  13. Planning and architectural safety considerations in designing nuclear power plants

    International Nuclear Information System (INIS)

    Konsowa, Ahmed A.

    2009-01-01

    To achieve optimum safety and to avoid possible hazards in nuclear power plants, considering architectural design fundamentals and all operating precautions is mandatory. There are some planning and architectural precautions should be considered to achieve a high quality design and construction of nuclear power plant with optimum safety. This paper highlights predicted hazards like fire, terrorism, aircraft crash attacks, adversaries, intruders, and earthquakes, proposing protective actions against these hazards that vary from preventing danger to evacuating and sheltering people in-place. For instance; using safeguards program to protect against sabotage, theft, and diversion. Also, site and building well design focusing on escape pathways, emergency exits, and evacuation zones, and the safety procedures such as; evacuation exercises and sheltering processes according to different emergency classifications. In addition, this paper mentions some important codes and regulations that control nuclear power plants design, and assessment methods that evaluate probable risks. (author)

  14. [Safety assessment of foods derived from genetically modified plants].

    Science.gov (United States)

    Pöting, A; Schauzu, M

    2010-06-01

    The placing of genetically modified plants and derived food on the market falls under Regulation (EC) No. 1829/2003. According to this regulation, applicants need to perform a safety assessment according to the Guidance Document of the Scientific Panel on Genetically Modified Organisms of the European Food Safety Authority (EFSA), which is based on internationally agreed recommendations. This article gives an overview of the underlying legislation as well as the strategy and scientific criteria for the safety assessment, which should generally be based on the concept of substantial equivalence and carried out in relation to an unmodified conventional counterpart. Besides the intended genetic modification, potential unintended changes also have to be assessed with regard to potential adverse effects for the consumer. All genetically modified plants and derived food products, which have been evaluated by EFSA so far, were considered to be as safe as products derived from the respective conventional plants.

  15. Survey of numerical safety targets for nuclear power plants

    International Nuclear Information System (INIS)

    Kelley, A.P. Jr.; Buttemer, D.R.

    1981-04-01

    The construction of a nuclear power plant implies, as does the construction of any major public work, the acceptance of a finite degree of risk. This risk can be reduced by an increased investment in engineered safeguards. However, at some level of risk, overinvestment in safety can render the project uneconomical. Because of the desirability of fixing safety standards on an absolute basis, there has long been an interest in establishing numerical risk criteria for the design, construction, and operation of nuclear power plants. Interest in the subject of numerical safety goals has recently been intensified by the Three Mile Island Action Plan. The USNRC has been directed by Congress to develop a national safety goal for reactor regulation. This report summarizes actions which have been historically, and are currently, taking place toward establishing national numerical risk targets for reactor regulation. Emphasis is placed upon actions taken, or currently being taken, by federal regulatory agencies and directly associated advisory bodies

  16. Uranium isotope separation by gaseous diffusion and plant safety

    International Nuclear Information System (INIS)

    Simeon, Claude; Dumas, Maurice.

    1980-07-01

    This report constitutes a safety guide for operators of uranium isotope separation plants, and includes both aspects of safety and protection. Taking into account the complexity of safety problems raised at design and during operation of plants which require specialized guides, this report mainly considers both the protection of man, the environment and goods, and the principles of occupational safety. It does not claim to be comprehensive, but intends to state the general principles, the particular points related to the characteristics of the basic materials and processes, and to set forth a number of typical solutions suitable for various human and technical environments. It is based on the French experience gained during the last fifteen years [fr

  17. Safe operation of nuclear power plants - Is safety culture an adequate management method?

    International Nuclear Information System (INIS)

    Piirto, A.

    2012-01-01

    One of the characteristics of a good safety culture is a definable commitment to the improvement of safety behaviours and attitudes at all organisational levels. A second characteristic of an organisation with excellent safety culture is free and open communication. The general understanding has been that safety culture is a part of organisation culture. In addition to safety culture thinking, proactive programmes and displays of proactive work to improve safety are required. This work needs to include, qt a minimum, actions aiming at reducing human errors, the development of human error prevention tools, improvements in training, and the development of working methods and the organisation's activities. Safety depends not only on the technical systems, but also on the organisation. There is a need for better methods and tools for organisational assessment and development. Today there is universal acceptance of the significant impact that management and organisational factors have over the safety significance of complex industrial installations such as nuclear power plants. Many events with significant economic and public impact had causes that have been traced to management deficiencies. The objective of this study is development of new methods to increase safety of nuclear power plant operation. The research has been limited to commercial nuclear power plants that are intended for electrical power generation in Finland. Their production activities, especially operation and maintenance, are primarily reviewed from a safety point of view, as well as human performance and organisational factors perspective. This defines the scope and focus of the study. The research includes studies related to knowledge management and tacit knowledge in the project management context and specific studies related to transfer of tacit knowledge in the maintenance organization and transfer of tacit knowledge between workers of old generation and young generation. The empirical results

  18. Safe operation of nuclear power plants - Is safety culture an adequate management method?

    Energy Technology Data Exchange (ETDEWEB)

    Piirto, A.

    2012-07-01

    One of the characteristics of a good safety culture is a definable commitment to the improvement of safety behaviours and attitudes at all organisational levels. A second characteristic of an organisation with excellent safety culture is free and open communication. The general understanding has been that safety culture is a part of organisation culture. In addition to safety culture thinking, proactive programmes and displays of proactive work to improve safety are required. This work needs to include, qt a minimum, actions aiming at reducing human errors, the development of human error prevention tools, improvements in training, and the development of working methods and the organisation's activities. Safety depends not only on the technical systems, but also on the organisation. There is a need for better methods and tools for organisational assessment and development. Today there is universal acceptance of the significant impact that management and organisational factors have over the safety significance of complex industrial installations such as nuclear power plants. Many events with significant economic and public impact had causes that have been traced to management deficiencies. The objective of this study is development of new methods to increase safety of nuclear power plant operation. The research has been limited to commercial nuclear power plants that are intended for electrical power generation in Finland. Their production activities, especially operation and maintenance, are primarily reviewed from a safety point of view, as well as human performance and organisational factors perspective. This defines the scope and focus of the study. The research includes studies related to knowledge management and tacit knowledge in the project management context and specific studies related to transfer of tacit knowledge in the maintenance organization and transfer of tacit knowledge between workers of old generation and young generation. The empirical

  19. Safety aspects of gas centrifuge enrichment plants

    International Nuclear Information System (INIS)

    Hansen, A.H.

    1987-01-01

    Uranium enrichment by gas centrifuge is a commercially proven, viable technology. Gas centrifuge enrichment plant operations pose hazards that are also found in other industries as well as unique hazards as a result of processing and handling uranium hexafluoride and the handling of enriched uranium. Hazards also found in other industries included those posed by the use of high-speed rotating equipment and equipment handling by use of heavy-duty cranes. Hazards from high-speed rotating equipment are associated with the operation of the gas centrifuges themselves and with the operation of the uranium hexafluoride compressors in the tail withdrawal system. These and related hazards are discussed. It is included that commercial gas centrifuge enrichment plants have been designed to operate safely

  20. Quality assurance and nuclear power plant safety

    International Nuclear Information System (INIS)

    Mullan, J.V.

    1983-01-01

    Quality assurance in the nuclear industry was born in the late 1960s. Atomic Energy Control Board staff began its regulatory practice on quality assurance during that period. In this presentation the author traces the circumstances that first led to the establishment of Canadian nuclear power plant quality assurance programmes, summarizes progress over the last decade and a half, and outlines the current regulatory approach and what has been learned so far

  1. Adapting a reactor safety assessment system for specific plants

    International Nuclear Information System (INIS)

    Ballard, T.L.; Cordes, G.A.

    1991-01-01

    The Reactor Safety Assessment System (RSAS) is an expert system being developed by the Idaho National Engineering Laboratory, the University of Maryland (UofM) and US Nuclear Regulatory Commission (NRC) for use in the NRC Operations center. RSAS is designed to help the Reactor Safety Team monitor and project core status during an emergency at a licensed nuclear power plant. Analysis uses a hierarchical plant model based on equipment availability and automatically input parametric plant information. There are 3 families of designs of pressurized water reactors and 75 plants using modified versions of the basic design. In order to make an RSAS model for each power plant, a generic model for a given plant type is used with differences being specified by plant specific files. Graphical displays of this knowledge are flexible enough to handle any plant configuration. A variety of tools have been implemented to make it easy to modify a design to fit a given plant while minimizing chance for error. 3 refs., 4 figs

  2. IAEA Concludes Safety Review at Chooz Nuclear Power Plant in France

    International Nuclear Information System (INIS)

    2013-01-01

    Full text: An IAEA-led international team of nuclear safety experts noted good practices and made recommendations to reinforce safety measures during a review of operational safety at France's Chooz Nuclear Power Plant (NPP) that concluded today. The Operational Safety Review Team (OSART) was assembled at the French Government's request. The in-depth review, which began 17 June, focused on aspects essential to the safe operation of the NPP. The team comprised experts from Switzerland, Belgium, Germany, China, India, United Kingdom, Czech Republic, Canada, Hungary and the IAEA. The review covered the areas of management, organization and administration; training and qualification of personnel; operations; maintenance; technical support; operating experience; radiation protection; chemistry; emergency planning and preparedness; and severe accident management. The conclusions of the review are based on the IAEA's Safety Standards. The OSART team identified good plant practices that will be shared with the rest of the nuclear industry for consideration. Examples include: The plant has a professional development programme as part of a joint employment effort shared by the plant and its contractors. This enables trainees to develop professional capability, understand practices and gain experience from other nuclear power plants in terms of work planning and coordination; The plant has built a strong relationship between the on-shift response team of the plant and the local fire brigade to improve firefighting and rescue operations; Self-assessment groups discuss and resolve specific issues within operations, empowering operations personnel to take ownership of improvement programmes; and The plant has improved warnings at entrances to all o range zones , areas of elevated dose rates to which only authorized staff have access. The team identified a number of improvements to operational safety at Chooz NPP. Examples include: The plant should review its process for the

  3. Applications of probabilistic safety assessment (PSA) for nuclear power plants

    International Nuclear Information System (INIS)

    2001-02-01

    This report, which compiles information on a comprehensive set of PSA applications in the areas of NPP design, operation, and accident mitigation and management, is the culmination of an IAEA project on PSA Applications and Tools to Improve NPP Safety. In this regard, the Technical Committee Meeting (TCM) held in Madrid in February 1998 allowed participants to review and provide very valuable comments for this report. Several important facts related to PSA and its applications were highlighted during this TCM: living PSAs are the basis for the risk informed approach to decision making; development and use of safety/risk monitors as tools for configuration management is spreading fast; the different uses of PSA to support NPP testing and maintenance planning and optimization are amongst the most widespread PSA applications; plant specific PSAs are being used to support the safety upgrading programmes of plants built to earlier standards; not all countries have a regulatory framework for the use of the probabilistic approach in decision making. Some countries are still far from 'risk-informed' regulation, and this means that there is still considerable work ahead, both for regulators and utilities, to clarify approaches, to establish a framework and to reach a common understanding in relation to the use of PSA in decision making. This report is based on the premise that the use of PSA can provide useful information for the decision maker. This report is intended to provide an overview of current PSA applications. Section 2 addresses the PSA application process, outlines the general requirements for PSA tools and provides a discussion on PSA aspects such as PSA level, scope and level of detail, which have to be considered when planning/performing PSA applications. Section 3 discusses the technical aspects of individual applications and is divided into three parts. Section 3.1 is dedicated to the design related PSA applications. The second part of Section 3 considers

  4. Ways of improving safety for future PWRs in France

    International Nuclear Information System (INIS)

    Gros, G.; Jalouneix, J.; Manesse, D.; Mattei, J.M.

    1994-01-01

    For the design of a new generation of nuclear power plants which could be ordered in France at the end of the nineties, there is a broad consensus on the choice of the evolutionary way, in view of the significant progress in the field of safety which appears possible with this approach, due to feedback of operating experience from a large number of reactors, results of extended safety research and development projects, general technical progress and findings from detailed probabilistic safety studies performed. This paper presents results of thinkings and studies, conducted within the Institute for Nuclear Safety and Protection (IPSN) in the various fields mentioned, in view of the definition of safety objectives and principles for future PWRs. These results contributed to the preparation of a common safety approach for future plants in France and Germany. (authors). 1 tab., 3 refs

  5. Kozloduy nuclear power plant. Units 1-4. Status of safety assessment activities. Rev. 2

    International Nuclear Information System (INIS)

    1999-01-01

    This paper presents the results of the status of safety assessment activities carried out by the Kozloduy Nuclear Power Plant (KNPP) in order to evaluate the current status of the safety of its reactor units 1-4. The steam supply system of this units is based of the reactor WWER-440/ B-230, which is a PWR of Russian design developed according to the safety standards in force in USSR in late 60-s. Now a days 10 reactor units of this type are in operation in four NPPs. Despite of efforts of the different plants to implement safety improvements measures during first 10-15 years of operation of this type of reactor its major safety problems were not eliminated and were a subject of international concern. The systematic evaluation of the deficiencies of the original design of this type of reactors have been initiated by IAEA in the beginning of 1990 and brought to developing a comprehensive list of safety problems which required urgent implementation of safety measures in all plants. To solve this problems in 1991 KNPP initiated implementation of so called 'short term' safety improvement program, developed with the help of WANO under agreement with Bulgarian Nuclear Safety Authority (BNSA) and consortium RISKAUDIT. The program was based on a stage approach and was foreseen to be implemented by tree stages in very tight time schedule in order to achieve significant and rapid improvements of the level of safety in operation of the units. The Short Tenn Program was implemented between the years 1991 and 1997 thanks of the strong safety commitment of NEK and KNPP staff and the broad international cooperation and financial support. Important part of resources were supplied under PHARE program of CEC, EBRD grant agreement and EDF support. The plant current safety level analysis has been performed using IAEA analytical methodology according to 50-SG-O12 standard 'Periodic safety review of operational nuclear power plants'. The approach and criteria for acceptable safety level

  6. Recommended general safety requirements for nuclear power plants

    International Nuclear Information System (INIS)

    1983-06-01

    This report presents recommendations for a set of general safety requirements that could form the basis for the licensing of nuclear power plants by the Atomic Energy Control Board. In addition to a number of recommended deterministic requirements the report includes criteria for the acceptability of the design of such plants based upon the calculated probability and consequence (in terms of predicted radiation dose to members of the public) of potential fault sequences. The report also contains a historical review of nuclear safety principles and practices in Canada

  7. Safety and regulatory requirements of nuclear power plants

    International Nuclear Information System (INIS)

    Kumar, S.V.; Bhardwaj, S.A.

    2000-01-01

    A pre-requisite for a nuclear power program in any country is well established national safety and regulatory requirements. These have evolved for nuclear power plants in India with participation of the regulatory body, utility, research and development (R and D) organizations and educational institutions. Prevailing international practices provided a useful base to develop those applicable to specific system designs for nuclear power plants in India. Their effectiveness has been demonstrated in planned activities of building up the nuclear power program as well as with unplanned activities, like those due to safety related incidents etc. (author)

  8. Safety assessment of emergency power systems for nuclear power plants

    International Nuclear Information System (INIS)

    1992-01-01

    This publication is intended to assist the safety assessor within a regulatory body, or one working as a consultant, in assessing the safety of a given design of the emergency power systems (EPS) for a nuclear power plant. The present publication refers closely to the NUSS Safety Guide 50-SG-D7 (Rev. 1), Emergency Power Systems at Nuclear Power Plants. It covers therefore exactly the same technical subject as that Safety Guide. In view of its objective, however, it attempts to help in the evaluation of possible technical solutions which are intended to fulfill the safety requirements. Section 2 clarifies the scope further by giving an outline of the assessment steps in the licensing process. After a general outline of the assessment process in relation to the licensing of a nuclear power plant, the publication is divided into two parts. First, all safety issues are presented in the form of questions that have to be answered in order for the assessor to be confident of a safe design. The second part presents the same topics in tabulated form, listing the required documentation which the assessor has to consult and those international and national technical standards pertinent to the topics. An extensive reference list provides information on standards. 1 tab

  9. Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-

    International Nuclear Information System (INIS)

    Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo

    1995-07-01

    This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on 'The improvement of level-1 PSA computer codes' is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author)

  10. Research on the improvement of nuclear safety -Improvement of level 1 PSA computer code package-

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Kyoo; Kim, Tae Woon; Kim, Kil Yoo; Han, Sang Hoon; Jung, Won Dae; Jang, Seung Chul; Yang, Joon Un; Choi, Yung; Sung, Tae Yong; Son, Yung Suk; Park, Won Suk; Jung, Kwang Sub; Kang Dae Il; Park, Jin Heui; Hwang, Mi Jung; Hah, Jae Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This year is the third year of the Government-sponsored mid- and long-term nuclear power technology development project. The scope of this sub project titled on `The improvement of level-1 PSA computer codes` is divided into three main activities : (1) Methodology development on the underdeveloped fields such as risk assessment technology for plant shutdown and low power situations, (2) Computer code package development for level-1 PSA, (3) Applications of new technologies to reactor safety assessment. At first, in this area of shutdown risk assessment technology development, plant outage experiences of domestic plants are reviewed and plant operating states (POS) are decided. A sample core damage frequency is estimated for over draining event in RCS low water inventory i.e. mid-loop operation. Human reliability analysis and thermal hydraulic support analysis are identified to be needed to reduce uncertainty. Two design improvement alternatives are evaluated using PSA technique for mid-loop operation situation: one is use of containment spray system as backup of shutdown cooling system and the other is installation of two independent level indication system. Procedure change is identified more preferable option to hardware modification in the core damage frequency point of view. Next, level-1 PSA code KIRAP is converted to PC-windows environment. For the improvement of efficiency in performing PSA, the fast cutest generation algorithm and an analytical technique for handling logical loop in fault tree modeling are developed. 48 figs, 15 tabs, 59 refs. (Author).

  11. Regulatory supervision of safety indicators; experience with radiation safety indicators in Dukovany nuclear power plant performance

    International Nuclear Information System (INIS)

    Urbancik, L.; Kulich, V.

    2004-01-01

    The State Office for Nuclear Safety uses three sets of indicators describing the following aspects of a favourable nuclear power plant operation: smooth operation in normal circumstances, low risk to the population, and operation with a positive safety attitude. These are three safety-related areas for assessment. Each area has its own set of indicators. Overall operational safety performance indicators were identified for each attribute. From this point, a level of strategic indicators was developed, and finally, a set of specific indicators was set up. While neither the overall indicators nor the strategic indicators are directly measurable, the specific indicators are directly measurable and are targeted during inspection. (author)

  12. Safety issues and their ranking for 'small series' WWER-1000 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

  13. Safety issues and their ranking for 'small series' WWER-1000 nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

    International Nuclear Information System (INIS)

    2000-09-01

    This report presents the safety issues in 'small series' WWER-1000 nuclear power plants (NPPs). Safety issues are deviations from current recognized safety practices in design and operation judged to be safety significant by their impact on the plants' defence in depth. This report is intended to serve as reference for the development of plant specific safety improvement programmes and for the evaluation of measures proposed and/or implemented. The identification of safety issues is based on safety studies conducted by the operators of 'small series' WWER-1000 units and by organizations dealing with these reactors, on findings of IAEA safety missions to 'small series' WWER-1000 plants in South Ukraine, at Novovoronezh and Kalinin, and on information obtained from specialists from various countries during an IAEA consultants meeting, 8-12 September 1997 in Vienna, within the framework of the Extra budgetary Programme on the Safety of WWER and RBMK NPPs. Safety issues are first presented according to their impact on the main safety functions and are then described individually. The safety issues are characterized by issue title and specified by issue clarification. Safety issues connected with plant design are followed by the ranking of the issue and ranking justification. Altogether 85 safety issues have been identified, 12 of which are in Category III (defence in depth is insufficient, immediate corrective action is necessary), 38 in Category 11 (defence in depth is degraded, action is needed to resolve the issue) and 22 in Category I (departure from international practices, to be addressed as part of actions to resolve higher priority issues). In the case of operational safety issues (13 safety issues) no ranking is provided as the available material was considered insufficient. For each safety issue, comments and recommendations are made by the IAEA; the status of corresponding measures to improve safety implemented or planned at each site are presented in the

  14. Improvement of nuclear power plant management applying the lessons learned

    International Nuclear Information System (INIS)

    Lyon, P.W.

    1987-01-01

    Active managment participation is a common thread among nuclear power plants with superior performance. Plants that benefit from the hands-on attention of senior managers are typically more reliable and can be expected to have higher margins of safety. There are numerous ways in which utilities are promoting management involvement in the day-to-day operations of their nuclear plants. The Institute of Nuclear Power Operations' (INPO) Plant Performance Indicator Program provides utilities with data in key performance areas, allowing management to monitor performance and concentrate on areas needing attention. Utilities are also setting ambitious short and long-term goals in several performance areas. This increased management attention to nuclear plant operations is reflected in improved nuclear plant performance across the coutry. For instance, over the past five years, the number of significant events per unit has declined, unplanned automatic scrams have been reduced, equivalent availability is about the same (however, many units and the industry median has improved) and collentive radiation exposure and volume of low-level waste shipped per unit are both showing a decreasing trend. (author)

  15. Safety provision for nuclear power plants during remaining running time

    International Nuclear Information System (INIS)

    Rossnagel, Alexander; Hentschel, Anja

    2012-01-01

    With the phasing-out of the industrial use of nuclear energy for the power generation, the risk of the nuclear power plants has not been eliminated in principle, but only for a limited period of time. Therefore, the remaining nine nuclear power plants must also be used for the remaining ten years according to the state of science and technology. Regulatory authorities must substantiate the safety requirements for each nuclear power plant and enforce these requirements by means of various regulatory measures. The consequences of Fukushima must be included in the assessment of the safety level of nuclear power plants in Germany. In this respect, the regulatory authorities have the important tasks to investigate and assess the security risks as well as to develop instructions and orders.

  16. Nuclear accidents and safety measures of domestic nuclear power plants

    International Nuclear Information System (INIS)

    Song Zurong; Che Shuwei; Pan Xiang

    2012-01-01

    Based on the design standards for the safety of nuclear and radiation in nuclear power plants, the three accidents in the history of nuclear power are analyzed. And the main factors for these accidents are found out, that is, human factors and unpredicted natural calamity. By combining the design and operation parameters of domestic nuclear plants, the same accidents are studied and some necessary preventive schemes are put forward. In the security operation technology of domestic nuclear power plants nowadays, accidents caused by human factors can by prevented completely. But the safety standards have to be reconsidered for the unpredicted neutral disasters. How to reduce the hazard of nuclear radiation and leakage to the level that can be accepted by the government and public when accidents occur under extreme conditions during construction and operation of nuclear power plants must be considered adequately. (authors)

  17. Improving Safety through Human Factors Engineering.

    Science.gov (United States)

    Siewert, Bettina; Hochman, Mary G

    2015-10-01

    Human factors engineering (HFE) focuses on the design and analysis of interactive systems that involve people, technical equipment, and work environment. HFE is informed by knowledge of human characteristics. It complements existing patient safety efforts by specifically taking into consideration that, as humans, frontline staff will inevitably make mistakes. Therefore, the systems with which they interact should be designed for the anticipation and mitigation of human errors. The goal of HFE is to optimize the interaction of humans with their work environment and technical equipment to maximize safety and efficiency. Special safeguards include usability testing, standardization of processes, and use of checklists and forcing functions. However, the effectiveness of the safety program and resiliency of the organization depend on timely reporting of all safety events independent of patient harm, including perceived potential risks, bad outcomes that occur even when proper protocols have been followed, and episodes of "improvisation" when formal guidelines are found not to exist. Therefore, an institution must adopt a robust culture of safety, where the focus is shifted from blaming individuals for errors to preventing future errors, and where barriers to speaking up-including barriers introduced by steep authority gradients-are minimized. This requires creation of formal guidelines to address safety concerns, establishment of unified teams with open communication and shared responsibility for patient safety, and education of managers and senior physicians to perceive the reporting of safety concerns as a benefit rather than a threat. © RSNA, 2015.

  18. Management and organization in nuclear power plant safety

    International Nuclear Information System (INIS)

    Osborn, R.N.

    1983-08-01

    In the immediate aftermath of the Three Mile Island accident, the Nuclear Regulatory Commission-sponsored investigations of the relation between human issues and safety tended to focus on individual and, at most, group level phenomena. This initial bottom up view of organizational safety has continued to be investigated by the Nuclear Regulatory Commission, as evidence by the four previous papers. Recently, however, work has begun which adopts a top down management/organization approach to nuclear power plant safety. This paper reports on the research, to date, on this focus

  19. Software important to safety in nuclear power plants

    International Nuclear Information System (INIS)

    1994-01-01

    The report provides guidance on current practices, documenting their strengths and weaknesses for dealing with the important issues of software engineering that nuclear power plant system designers, software producers and regulators are facing. The focus of the report is on safety critical applications of general purpose processors controlled by custom developed software; however, it should also have application in safety related applications and for other types of computers. In addition to system designers, software producers and regulators, the intended readership of this report includes users of software based systems, who should be aware of the relevant issues in specifying and obtaining software for systems important to safety. Refs, 1 fig., tabs

  20. Safety aspects of nuclear plants coupled with seawater desalination units

    International Nuclear Information System (INIS)

    2001-08-01

    The purpose of this publication is to address the safety and licensing aspects of nuclear power plants for which a significant portion of the heat energy produced by the reactor is intended for use in heat utilization applications. Although intended to cover the broad spectrum of nuclear heat applications, the focus of the discussion will be the desalination of sea water using nuclear power plants as the energy source for the desalination process

  1. The reliability of nuclear power plant safety systems

    International Nuclear Information System (INIS)

    Susnik, J.

    1978-01-01

    A criterion was established concerning the protection that nuclear power plant (NPP) safety systems should afford. An estimate of the necessary or adequate reliability of the total complex of safety systems was derived. The acceptable unreliability of auxiliary safety systems is given, provided the reliability built into the specific NPP safety systems (ECCS, Containment) is to be fully utilized. A criterion for the acceptable unreliability of safety (sub)systems which occur in minimum cut sets having three or more components of the analysed fault tree was proposed. A set of input MTBF or MTTF values which fulfil all the set criteria and attain the appropriate overall reliability was derived. The sensitivity of results to input reliability data values was estimated. Numerical reliability evaluations were evaluated by the programs POTI, KOMBI and particularly URSULA, the last being based on Vesely's kinetic fault tree theory. (author)

  2. Summary report on safety objectives in nuclear power plants

    International Nuclear Information System (INIS)

    1989-01-01

    The special Task Force on Safety Objectives of the Commission of the European Communities (CEC) Working Group on the Safety of Light Water Reactors reported in May 1983 on its review of existing overall safety objectives in nuclear power plants. Since then much relevant worlwide activity has taken place. This report reviews those activities that have taken place since 1983 in European Community Member States, including more recent Members, as well as in Sweden and Finland. The report confines itself to issues related to probabilistic safety objectives, and concludes that significant progress has been made in many areas. Mutual understanding of safety objectives is leading to a convergence of views and approaches, but it is noted that much work remains to be completed

  3. A concept of safety indicator system for nuclear power plants

    International Nuclear Information System (INIS)

    Lehtinen, E.

    1995-12-01

    The fundamental principle in the safety technology of nuclear power is embodied in the strategy of defence in depth. The defence lines of the strategy, completed with a PSA logic model and structure, are considered to provide an appropriate framework for identification and structuring of the operational safety performance areas for nuclear power plants. Once these areas are identified the safety indicators can be defined. Based on this approach a concept of safety indicator system was outlined. About one hundred indicator specifications have been collected, refined and related to the performance areas. The specifications enable the utilities and authorities to check the coverage of their indicators set from the operational safety point of view and select or refine indicators for testing and routine use. Finally various statistical approaches and methods for using indicators in performance evaluation are presented. (orig.) (16 refs., 2 figs., 2 tabs.)

  4. Thermophysics of safety of Nuclear Power Plants

    International Nuclear Information System (INIS)

    Klyuchnikov, A.A.; Sharaevskij, I.G.; Fialko, N.M.; Zimin, L.B.; Sharaevskij, G.I.

    2010-01-01

    This monograph presents the basic critical analysis of the vapor phase generation process and the heat exchange crisis with respect to conditions of involuntary movement of heat carrier in the steam generating core channels of the nuclear powerful reactors, as well as modern understanding on the most important current heat hydraulic peculiarities under these conditions the two-phase flows. It was suggested the series of original methods, physics mathematical models and algorithms for enhancement of thermohydraulic computation codes - the component of great importance for operational safety providing system of nuclear power units, and for giving the possibility of automatic identification for these systems in real time of nominal and pre-emergency modes of vapor-water flows current.

  5. RATU - Nuclear power plant structural safety

    International Nuclear Information System (INIS)

    Hedner, G.; Schultz, H.; Unneberg, L.

    1992-12-01

    The evaluation group is of the opinion that the work performed under the RATU programme is generally of high quality, in some areas, especially those related to water chemistry of excellent quality. The personnel gives the impression of being dedicated and enthusiastic, and the administration seems to be very effective. It is obvious that the RATU programme has taken advantage of related contracts and projects funded by different sources. It is the opinion of the valuation group that the investment and human capital have been brought to work very efficiently in all projects. The objectives of the programme and the different projects are formulated in a broad sense. The areas selected for work are however of high relevance to nuclear safety. In some projects not all aspects are addressed by the ongoing work, and further activities may be necessary to meet with the requirements of the authorities. (orig.)

  6. Quality procedure management for improved nuclear safety

    International Nuclear Information System (INIS)

    Forzano, P.; Castagna, P.

    1995-01-01

    Emergency Operating Procedures and Accident Management Procedures are the next step in the computerization of NPP control rooms. Different improvements are presently conceivable for this operator aid tool, and research activities are in development. Undergoing activities regard especially formal aspects of knowledge representation, Human-Machine interface and procedure life cycle management. These aspects have been investigated deeply by Ansaldo, and partially incorporated in the DIAM prototype. Nuclear Power Plant Procedures can be seen from essentially two viewpoints: the process and the information management. From the first point of view, it is important to supply the knowledge apt to solve problems connected with the control of the process, from the second one the focus of attention is on the knowledge representation, its structure, elicitation and maintenance, and formal quality assurance. These two aspects of procedure representation can be considered and solved separately. In particular, methodological, formal and management issues require long and tedious activities, that in most cases constitute a great barrier for procedures development and upgrade. To solve these problems, Ansaldo is developing DIAM, a wide integrated tool for procedure management to support in procedure writing, updating, usage, and documentation. One of the most challenging features of DIAM is AUTO-LAY, a CASE sub-tool that, in a complete automatical way, structures parts or complete flow diagram. This is the feature that is partial present in some other CASE products, that, anyway, do not allow complex graph handling and isomorphism between video and paper representation. AUTO-LAY has the unique prerogative to draw graphs of any complexity to section them in pages, and to automatically compose a document. This has been recognized in the literature as the most important a second-generation CASE improvement. (Author) 9 Figs., 5 Refs

  7. IR-360 nuclear power plant safety functions and component classification

    Energy Technology Data Exchange (ETDEWEB)

    Yousefpour, F., E-mail: fyousefpour@snira.co [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of); Shokri, F.; Soltani, H. [Management of Nuclear Power Plant Construction Company (MASNA) (Iran, Islamic Republic of)

    2010-10-15

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  8. Status of safety issues at licensed power plants

    International Nuclear Information System (INIS)

    1991-05-01

    As part of ongoing US Nuclear Regulatory Commission (NRC) efforts to ensure the quality and accountability of safety issue information, a program has been established whereby an annual NUREG report will be published on the status of licensee implementation and NRC verification of safety issues in major NRC requirement areas. This report, the second volume of a three-volume series, addresses the status of unresolved safety issues (USIs) at licensed plants. The data contained in these NUREG reports are a product of the NRC's Safety Issues Management System (SIMS) database, which is maintained by the Project Management Staff in the Office of Nuclear Reactor Regulation and by NRC regional personnel. The purpose of this report is to provide a comprehensive description of the status of implementation and verification of the 27 safety issues designated as USIs and to make this information available to other interested parties, including the public. A corollary purpose of this NUREG report is to serve as a follow-on to NUREG-0933, ''A Prioritization of Generic Safety Issues,'' which tracks safety issues up until requirements are approved for imposition at licensed plants. 3 figs., 4 tabs

  9. IR-360 nuclear power plant safety functions and component classification

    International Nuclear Information System (INIS)

    Yousefpour, F.; Shokri, F.; Soltani, H.

    2010-01-01

    The IR-360 nuclear power plant as a 2-loop PWR of 360 MWe power generation capacity is under design in MASNA Company. For design of the IR-360 structures, systems and components (SSCs), the codes and standards and their design requirements must be determined. It is a prerequisite to classify the IR-360 safety functions and safety grade of structures, systems and components correctly for selecting and adopting the suitable design codes and standards. This paper refers to the IAEA nuclear safety codes and standards as well as USNRC standard system to determine the IR-360 safety functions and to formulate the principles of the IR-360 component classification in accordance with the safety philosophy and feature of the IR-360. By implementation of defined classification procedures for the IR-360 SSCs, the appropriate design codes and standards are specified. The requirements of specific codes and standards are used in design process of IR-360 SSCs by design engineers of MASNA Company. In this paper, individual determination of the IR-360 safety functions and definition of the classification procedures and roles are presented. Implementation of this work which is described with example ensures the safety and reliability of the IR-360 nuclear power plant.

  10. Quantifying the effectiveness of ITS in improving safety of VRUs

    NARCIS (Netherlands)

    Silla, A.; Rämä, P.; Leden, L.; Noort, M. van; Kruijff, J. de; Bell, D.; Morris, A.; Hancox, G.; Scholliers, J.

    2017-01-01

    This paper presents the results of a safety impact assessment, providing quantitative estimates of the safety impacts of ten intelligent transport systems (ITS) which were designed to improve safety, mobility and comfort of vulnerable road users (VRUs). The evaluation method originally developed to

  11. Improved plant availability by advanced condition based inspections

    International Nuclear Information System (INIS)

    Hulshof, Harry J.M.; Noteboom, Jan W.; Welberg, Paul G.M.; Bruijn, Leo E.

    2004-01-01

    An industrial plant has to operate safely, reliably and efficiently at the lowest possible cost. Plant availability plays an important role regarding economic life optimisation. Industrial installations that are under pressure and are operating at high temperatures have a limited life due to creep and fatigue. It is, therefore, of critical importance to know the location of any possible weak spots in the installation. To avoid safety risks, unplanned plant shutdown and, as a consequence, high costs for unavailability, cycling and repair, periodic inspections and strain measurements are recommended. A Speckle Image Correlation Analysis (SPICA) system enables on-stream measurement of deformation due to creep in critical areas like the heat-affected zone in welds. Plant management and operators use the strain measurements to take action when necessary and, consequently, prevent failures. In those plants that have been provided with SPICA-technology for some years plant availability has improved significantly as a result. Another important development for yielding improved availability concerns steam drums. During some 20 years, KEMA has been performing automated ultrasonic steam drum inspections from outside. The Dutch authorities accepted this methodology in this period as an alternative (rather than an addition) after several pilot projects. An advantage of this inspection methodology is the possibility to record of the inspection results and possibility of thus trending these data. The resulting reduction of through time appeared a major benefit for plant owners. Since the authorities adopted the RBI approach during the last 10 years, another advantage of the inspection methodology became apparent: complete scanning and recording of the inspection data of circumferential and longitudinal (butt and fillet) welds, inspection of nozzle welds and inner radius as well as corrosion mapping has been covering all higher risk areas in these drums. This enhanced inspection

  12. Improved plant availability by advanced condition based inspections

    Energy Technology Data Exchange (ETDEWEB)

    Hulshof, Harry J.M.; Noteboom, Jan W; Welberg, Paul G.M.; Bruijn, Leo E

    2004-06-01

    An industrial plant has to operate safely, reliably and efficiently at the lowest possible cost. Plant availability plays an important role regarding economic life optimisation. Industrial installations that are under pressure and are operating at high temperatures have a limited life due to creep and fatigue. It is, therefore, of critical importance to know the location of any possible weak spots in the installation. To avoid safety risks, unplanned plant shutdown and, as a consequence, high costs for unavailability, cycling and repair, periodic inspections and strain measurements are recommended. A Speckle Image Correlation Analysis (SPICA) system enables on-stream measurement of deformation due to creep in critical areas like the heat-affected zone in welds. Plant management and operators use the strain measurements to take action when necessary and, consequently, prevent failures. In those plants that have been provided with SPICA-technology for some years plant availability has improved significantly as a result. Another important development for yielding improved availability conc