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Sample records for npp krsko psa

  1. NPP Krsko Living PSA Concept

    International Nuclear Information System (INIS)

    Vrbanic, I.; Spiler, J.

    2000-01-01

    NPP Krsko developed PSA model of internal and external initiators within the frame of the Individual Plant Examination (IPE) project. Within this project PSA model was used to examine the existing plant design features. In order to continue with use of this PSA model upon the completion of IPE in various risk-informed applications in support of plant operation and evaluations of design changes, an appropriate living PSA concept needed to be defined. The Living PSA concept is in NPP Krsko considered as being a set of activities pursued in order to update existing PSA model in a manner that it appropriately represents the plant design, operation practice and history. Only a PSA model which is being updated in this manner can serve as a platform for plant-specific risk informed applications. The NPP Krsko living PSA concept is based on the following major ponts. First, the baseline PSA model is defined, which is to be maintained and updated and which is to be reference point for any risk-informed application. Second, issues having a potential for impact on baseline PSA model are identified and procedure and responsibilities for their permanent monitoring and evaluation are established. Third, manner is defined in which consequential changes to baseline PSA model are implemented and controlled, together with associated responsibilities. Finally, the process is defined by which the existing version of baseline PSA model is superseded by a new one. Each time a new version of baseline PSA model is released, it would be re-quantified and the results evaluated and interpreted. By documenting these re-quantifications and evaluations of results in a sequence, the track is being kept of changes in long-term averaged risk perspective, represented by long-term averaged frequencies of core damage and pre-defined release categories. These major topics of NPP Krsko living PSA concept are presented and discussed in the paper. (author)

  2. On-line maintenance PSA support at NPP Krsko

    International Nuclear Information System (INIS)

    Prosen, R.; Vrbanic, I.; Kastelan, M.

    2000-01-01

    In 1997 Krsko NPP initiated the on-line maintenance (OLM) practice. On-line maintenance constitutes of corrective activities, preventive activities, surveillance activities, tests and inspections, as well as calibrations and modifications, taking place during the normal power operations. The on-line maintenance is a multidisciplinary process consisting of activity specification, planning, and preparation and performing of the OLM activity of interest. The primary role of the PSA group is to assess from the r isk perspective , using the plant-specific NEK PSA model, system unavailability and the impact to the plant operational risk. The intent is to support planning of the on-line maintenance activities from the risk perspective. The risk evaluation of the OLM activities is based on the probability of core damage evaluation for the defined discrete plant configuration states, determined by the OLM activities. Within this application, the optimized, plant-specific PSA model is used on Risk Spectrum platform. To perform the risk assessment of the on-line maintenance activities, first the systems to be affected are defined based on the planned OLM activities. The next important step is the assessment of the planned work schedule. To define the final schedule, the co-ordination and optimizing the planned OLM activities needs activation of all participating departments, supported also from PSA group. The P3 (i.e. Primavera) planning tool system windows are defined for different systems and groups of systems, and the activities are sorted in particular weeks according to these windows. (author)

  3. Seismic characterization of the NPP Krsko site

    International Nuclear Information System (INIS)

    Obreza, J.

    2000-01-01

    The goal of NPP Krsko PSA Project Update was the inclusion of plant changes (i.e. configuration/operational related) through the period January 1, 1993 till the OUTAGE99 (April 1999) into the integrated Internal/External Level 1/Level 2 NPP Krsko PSA RISK SPECTRUM model. NPP Krsko is located on seismotectonic plate. Highest earthquake was recorded in 1917 with magnitude 5.8 at a distance of 7-9 km. Site (founded) on Pliocene sediments which are as deep as several hundred meters. No surface faulting at the Krsko site has been observed and thus it is not to be expected. NPP Krsko is equipped with seismic instrumentation, which allows it to complete OBE (SSE). The seismic PSA successfully showed high seismic margin at Krsko plant. NPP Krsko seismic design is based on US regulations and standards

  4. Engineering safety review mission Krsko NPP external events PSA. Ljubljana, Slovenia 19-23 February 1996. Final report

    International Nuclear Information System (INIS)

    Budnitz, R.J.; Smith, P.

    1996-01-01

    Within the scope of the TC Project RER/9/035, a review mission visited Ljubljana, Slovenia, 19-23 February 1996. Two outside experts, Messrs. R.J. Budnitz (USA) and Paul Smith (USA), as well as a staff member, A. Guerpinar (ESS-NSNI) took part in the review. The purpose of the mission was to assist the Slovenian Nuclear Safety Administration to review the external events PSA prepared by Krsko NPP consultants Westinghouse Energy Systems Europe and EQE International. Another seismic safety review was performed concurrently in Ljubljana involving the investigations in relation to the tectonic stability and reassessment of the design basis ground motion characterization for the Krsko NPP site

  5. NPP Krsko decommissioning concept

    International Nuclear Information System (INIS)

    Novsak, M.; Fink, K.; Spiler, J.

    1996-01-01

    At the end of the operational lifetime of a nuclear power plant (NPP) it is necessary to take measures for the decommissioning as stated in different international regulations and also in the national Slovenian law. Based on these requirements Slovenian authorities requested the development of a site specific decommissioning plan for the NPP Krsko. In September 1995, the Nuklearna Elektrarna Krsko (NEK) developed a site specific scope and content for a decommissioning plan including the assumptions for determination of the decommissioning costs. The NEK Decommissioning Plan contains sufficient information to fulfill the decommissioning requirements identified by NRC, IAEA and OECD - NEA regulations. In this paper the activities and results of development of NEK Decommissioning Plan consisting of the development of three decommissioning strategies for the NPP Krsko and selection of the most suitable strategy based on site specific, social, technical, radiological and economic aspects, cost estimates for the strategies including the costs for construction of final disposal facilities for fuel/high level waste (fuel/HLW) and low/intermediate level waste (LLW/ILW) and scheduling of all activities necessary for the decommissioning of the NPP Krsko are presented. (author)

  6. NPP Krsko decommissioning concept

    International Nuclear Information System (INIS)

    Novsak, M.; Fink, K.; Spiler, J.

    1996-01-01

    At the end of the operational lifetime of a nuclear power plant (NPP) it is necessary to take measures for the decommissioning as stated in different international regulations and also in the national Slovenian law. Based on these requirements Slovenian authorities requested the development of a site specific decommissioning plan for the NPP KRSKO. In September 1995, the Nuklearna Elektrarna Krsko (NEK) developed a site specific scope and content for decommissioning plan including the assumptions for determination of the decommissioning costs. The NEK Decommissioning Plan contains sufficient information to fulfill decommissioning requirements identified by NRC, IAEA and OECD - NEA regulations. In this paper the activities and the results of development of NEK Decommissioning Plan consisting of the development of three decommissioning strategies for the NPP Krsko and selection of the most suitable strategy based on site specific, social, technical, radiological and economical aspects, cost estimates for the strategies including the costs for construction of final disposal facilities for fuel/high level waste (fuel/HLW) and low/intermediate level waste (LLW/ILW) and scheduling all activities necessary for the decommissioning of the NPP KRSKO are presented. (author)

  7. Parameter estimation of component reliability models in PSA model of Krsko NPP

    International Nuclear Information System (INIS)

    Jordan Cizelj, R.; Vrbanic, I.

    2001-01-01

    In the paper, the uncertainty analysis of component reliability models for independent failures is shown. The present approach for parameter estimation of component reliability models in NPP Krsko is presented. Mathematical approaches for different types of uncertainty analyses are introduced and used in accordance with some predisposed requirements. Results of the uncertainty analyses are shown in an example for time-related components. As the most appropriate uncertainty analysis proved the Bayesian estimation with the numerical estimation of a posterior, which can be approximated with some appropriate probability distribution, in this paper with lognormal distribution.(author)

  8. Operating Experience at NPP Krsko

    International Nuclear Information System (INIS)

    Kavsek, D.; Bach, B.

    1998-01-01

    Systematic analysis of operational experience by assessment of internal and industry events and the feedback of lessons learned is one of the essential activities in the improvement of the operational safety and reliability of the nuclear power plant. At NPP Krsko we have developed a document called ''Operating Experience Assessment Program''. Its purpose is to establish administrative guidance for the processing of operating events including on-site and industry events. Assessment of internal events is based on the following methods: Event and Causal Factor Charting, Change Analysis, Barrier Analysis, MORT (Management Oversight and Risk Tree Analysis) and Human Performance Evaluation. The operating experience group has developed a sophisticated program entitled ''Operating experience tracking system'' (OETS) in response to the need for a more efficient way of processing internal and industry operating experience information. The Operating Experience Tracking System is used to initiate and track operational events including recommended actions follow up. Six screens of the system contain diverse essential information which allows tracking of operational events and enables different kinds of browsing. OETS is a part of the NPP Krsko nuclear network system and can be easily accessed by all plant personnel. (author)

  9. Krsko NPP radioactive waste characteristics

    International Nuclear Information System (INIS)

    Skanata, D.; Kroselj, V.; Jankovic, M.

    2007-01-01

    In May 2005 Krsko NPP initiated the Radioactive Waste Characterization Project and commissioned its realization to the consulting company Enconet International, Zagreb. The Agency for Radwaste Management was invited to participate on the Project. The Project was successfully closed out in August 2006. The main Project goal consisted of systematization the existing and gathering the missing radiological, chemical, physical, mechanical, thermal and biological information and data on radioactive waste. In a general perspective, the Project may also be considered as a part of broader scope of activities to support state efforts to find a disposal solution for radioactive waste in Slovenia. The operational low and intermediate level radioactive waste has been structured into 6 waste streams that contain evaporator concentrates and tank sludges, spent ion resins, spent filters, compressible and non-compressible waste as well as specific waste. For each of mentioned waste streams, process schemes have been developed including raw waste, treatment and conditioning technologies, waste forms, containers and waste packages. In the paper the main results of the Characterization Project will be briefly described. The results will indicate that there are 17 different types of raw waste that have been processed by applying 9 treatment/conditioning technologies. By this way 18 different waste forms have been produced and stored into 3 types of containers. Within each type of container several combinations should be distinguished. Considering all of this, there are 34 different types of waste packages altogether that are currently stored in the Solid Radwaste Storage Facility at the Krsko NPP site. Because of these findings a new identification system has been recommended and consequently the improvement of the existing database on radioactive waste has been proposed. The potential areas of further in depth characterization are indicated. In the paper a brief description on the

  10. Quality of Industry Support to NPP Krsko

    International Nuclear Information System (INIS)

    Nemcic, K.

    2008-01-01

    NPP Krsko developed program for Supplier evaluation and performance. During the regular control of suppliers and evaluation of industry support to NPP Krsko quality problems were reported. Different quality systems were evaluated and different suppliers as: design organizations, equipment manufacturers, material vendors were audited or surveillance was performed. This paper discuss and report various cases where quality issues were problems based on audit results and present actions and efforts undertaken by the NE Krsko Quality Assurance Department to improve performance of the contractors, vendors, suppliers. New and different quality standards as approach in numerous articles are described as improvement or quality changes but also 'different opinion exist'. This paper also presents the author view and approach how to solve the possible future problems with different quality systems and organisations used by industry who support daily operation of NE Krsko and give recommendations for future nuclear projects.(author)

  11. NPP Krsko small break LOCA analysis

    International Nuclear Information System (INIS)

    Mavko, B.; Petelin, S.; Peterlin, G.

    1987-01-01

    Parametric analysis of small break loss of coolant accident for the Krsko NPP was calculated by using RELAP5/MOD1 computer code. The model that was used in our calculations has been improved over several years and was previously tested in simulation (s) of start-up tests and known NPP Krsko transients. In our calculations we modelled automatic actions initiated by control, safety and protection systems. We also modelled the required operator actions as specified in emergency operating instructions. In small-break LOCA calculations, we varied break sizes in the cold leg. The influence of steam generator tube plugging on small break LOCA accidents was also analysed. (author)

  12. Digital reactivity meter for NPP Krsko

    International Nuclear Information System (INIS)

    Glumac, B.; Vidmar, M.; Ravnik, M.

    1984-01-01

    Digital or analog reactivity meter is needed in order to perform the necessary low power physics tests after core reload in a nuclear power plant. Nuclear power plant Krsko ordered the construction of such digital reactivity meter on the basis of 'mikro-m' modular computer system that has been developed by IJS. Input signal sampling model as well as realtime reactivity calculation on the basis of the reactor inverse kinetic equation have also been developed by IJS. This digital reactivity meter has already been used to perform the start-up tests in NPP Krsko following first reload in fall of 1983. (author)

  13. Safety Culture Survey in Krsko NPP

    International Nuclear Information System (INIS)

    Strucic, M.; Bilic Zadric, T.

    2008-01-01

    The high level of nuclear safety, stability and competitiveness of electricity production, and public acceptability are the main objectives of Krsko Nuclear Power Plant. This is achievable only in environment where strong Safety Culture is taking dominant place in the way how employees communicate, perform tasks, share their ideas and attitudes, and demonstrate their concern in all aspects of work and coexistence. To achieve these objectives, behaviour of all employees as well as specific ethical values must become more transparent and that must arise from the heart of organization. Continuous ongoing and periodic self assessments of Safety Culture in Krsko NPP present major tools in implementation process of this approach. Benefits from Periodic interdisciplinary focused self assessment approach, which main intention is finding the strengths and potential areas for improvements, was used second time to assess the area of Safety Culture in Krsko NPP. Main objectives of self assessment, performed in 2006, were to increase the awareness of the present culture, to serve as a basis for improvement and to keep track of the effects of change or improvement over a longer period of time. For the purpose of effective self assessment, extensive questionnaire was used to obtain information that is representative for whole organization. Wide range of questions was chosen to cover five major characteristics of safety culture: Accountability for safety is clear, Safety is integrated into all activities, Safety culture is learning-driven, Leadership for safety is clear and Safety is a clearly recognized value. 484 Krsko NPP employees and 96 contractors were participated in survey. 70-question survey provided information that was quantified and results compared between groups. Anonymity of participant, as well as their willingness to contribute in this assessment implicates the high level of their openness in answering the questions. High number of participant made analysis of

  14. Supercompaction of radioactive waste at NPP Krsko

    International Nuclear Information System (INIS)

    Fink, K.; Sirola, P.

    1996-01-01

    The problem of radioactive waste management is both scientifically and technically complex and also deeply emotional issue. In the last twenty years the first two aspects have been mostly resolved up to the point of safe implementation. In the Republic of Slovenia, certain fundamentalist approaches in politics and the use of radioactive waste problem as a political tool, brought the final radioactive repository siting effort to a stop. Although small amounts of radioactive waste are produced in research institutes, hospitals and industry, major source of radioactive waste in Slovenia is the Nuclear Power Plant Krsko. When Krsko NPP was originally built, plans were made to construct a permanent radioactive waste disposal facility. This facility was supposed to be available to receive waste from the plant long before the on site storage facility was full. However, the permanent disposal facility is not yet available, and it became necessary to retain the wastes produced at the plant in the on-site storage facility for an extended period of time. Temporary radioactive storage capacity at the plant site has limited capacity and having no other options available NPP Krsko is undertaking major efforts to reduce waste volume generated to allow normal operation. This article describes the Radioactive Waste Compaction Campaign performed from November, 1994 through November, 1995 at Krsko NPP, to enhance the efficiency and safety of storage of radioactive waste. The campaign involved the retrieval, segmented gamma-spectrum measurement, dose rate measurement, compaction, re-packaging, and systematic storage of radioactive wastes which had been stored in the NPP radioactive waste storage building since plant commissioning. (author)

  15. Nuclear fuel reliability in NPP KRSKO

    International Nuclear Information System (INIS)

    Antolovic, A.; Kurincic, B.

    2001-01-01

    The importance of achieving and maintaining high fuel integrity comes from negative consequences of operation with failed fuel. Failed fuel has a significant effect on operating cost and performance, and increases the radiological consequences to environment. Fuel failures represent a breach in the first barrier (cladding) preventing the release of fission products. Historically NPP Krsko experienced some degradation of fuel cladding integrity. To resolve this problem and to ensure the safe, reliable and cost effective operation of nuclear fuel, NPP Krsko established 'Fuel Integrity Program'. The key elements of the Program are: continuous monitoring and trending of the fuel behaviour through operating cycle, evaluation of key performance indicators (RCS isotopes, operational parameters) to determine whether the fuel defects exist, implementation of appropriate actions to reduce and mitigate the consequences of fuel defects (four action levels), 100% examination of fuel to remove the defective fuel from operation (Ultrasonic (UT), In Mast Sipping (IMS) and visual inspection), evaluating the worldwide experience and fuel performance and, integrating the experience and knowledge into new fuel design (ZIRLO TM cladding, debris filter bottom nozzle, removable top nozzle). Since start of commercial operation fuel integrity has been evaluated considering certain aspects like operation and fuel handling, fuel rod burnup and cycle length, cladding material properties, etc. As a result of successful Fuel Integrity Program NPP Krsko has achieved high performance level in terms of fuel integrity in past four cycles. Also, NPP Krsko calculations show good matching between analytical prediction of number of failed fuel rods from primary coolant activity analysis and inspection results with the Nondestructive Testing (NDT) methods.(author)

  16. Limited Releases of Krsko NPP

    International Nuclear Information System (INIS)

    Breznik, B.; Kovac, A.

    2001-01-01

    Full text: Krsko Nuclear Power Plant is about 700 MW Pressurised Water Reactor plant located in Slovenia close to the border with Croatia. The authorised limit for the radioactive releases is basically set to 50 μSv effective dose per year to the members of the public. There is also additional limitation of total activities released in a year and concentration. The poster presents the effluents of the year 2000 and evaluated dose referring to the limits and to the natural and other sources of radiation around the plant. (author)

  17. Nuclear Oversight Function at Krsko NPP

    International Nuclear Information System (INIS)

    Bozin, B.; Kavsek, D.

    2010-01-01

    The nuclear oversight function is used at the Krsko NPP constructively to strengthen safety and improve performance. Nuclear safety is kept under constant examination through a variety of monitoring techniques and activities, some of which provide an independent review. The nuclear oversight function at the Krsko NPP is accomplished by Quality and Nuclear Oversight Division (SKV). SKV has completed its mission through a combination of compliance, performance and effectiveness-based assessments. The performance-based assessment is an assessment using various techniques (observations, interviews, walk-downs, document reviews) to assure compliance with standards and regulations, obtain insight into performance, performance trends and also to identify opportunities to improve effectiveness of implementation. Generally, the performance-based approach to oversight function is based on some essential elements. The most important one which is developed and implemented is an oversight program (procedure). The program focuses on techniques, activities and objectives commensurate with their significance to plant operational safety. These techniques and activities are: self-assessments, assessments, audits, performance indicators, monitoring of corrective action program (CAP), industry independent reviews (such as IAEA's OSART and WANO Peer Review), industry benchmarking etc. Graded approach is an inherent product of a performance based program and ranking process. It is important not only to focus on the highest ranked performance based attributes but to lead to effective utilization of an oversight program. The attributes selected for oversight need to be based on plant specific experience, current industry operating experience, supplier's performance and quality issues. Collaboration within the industry and effective utility oversight of processes and design activities are essential for achieving good plant performance. So the oversight program must integrate relevant

  18. Brief Assessment of Krsko NPP Decommissioning Costs

    International Nuclear Information System (INIS)

    Skanata, D.; Medakovic, S.; Debrecin, N.

    2000-01-01

    The first part of the paper gives a brief description of decommissioning scenarios and models of financing the decommissioning of NPPs. The second part contains a review of decommissioning costs for certain PWR plants with a brief description of methods used for that purpose. The third part of the paper the authors dedicated to the assessment of decommissioning costs for Krsko NPP. It does not deal with ownership relations and obligations ensuing from them. It starts from the simple point that decommissioning is an structure of the decommissioning fund is composed of three basic cost items of which the first refers to radioactive waste management, the second to storage and disposal of the spent nuclear fuel and the third to decommissioning itself. The assessment belongs to the category of preliminary activities and as such has a limited scope and meaning. Nevertheless, the authors believe that it offers a useful insight into the basic costs that will burden the decommissioning fund of Krsko NPP. (author)

  19. Utilization of NPP Krsko plant specific simulator

    International Nuclear Information System (INIS)

    Fifnja, I.; Pribozic, F.; Krajnc, J.

    2002-01-01

    NPP Krsko started with licensed operator training using its own plant-specific full scope simulator in April 2000. Today, two years after simulator was completed, the benefits of simulator use are visible in various fields. The simulator was effectively used to conduct licensed operator continuing training and practical examinations. Two-year continuous training program was designed to help maintain and improve operator performance. The simulator was also used to provide just-in-time training prior to plant evolutions. Together with licensed operators the non-licensed operators are also included into simulator training to provide affective team training opportunity and to foster good communication and increase scenario realism. Now, the first group of initial licensed operator training using plant-specific simulator is also almost completed. It is the first time that NPP Krsko training department conducted complete initial training and this will represent the great experience for future training. Besides training, the simulator was also utilized for procedure development and validation, operating standards development, testing of plant modifications and other activities, like emergency preparedness procedures validation and training exercises.(author)

  20. Central alarm system replacement in NPP Krsko

    International Nuclear Information System (INIS)

    Cicvaric, D.; Susnic, M.; Djetelic, N.

    2004-01-01

    Current NPP Krsko central alarm system consists of three main segments. Main Control Board alarm system (BETA 1000), Ventilation Control Board alarm system (BETA 1000) and Electrical Control Board alarm system (BETA 1100). All sections are equipped with specific BetaTone audible alarms and silence, acknowledge as well as test push buttons. The main reason for central alarm system replacement is system obsolescence and problems with maintenance, due to lack of spare parts. Other issue is lack of system redundancy, which could lead to loss of several Alarm Light Boxes in the event of particular power supply failure. Current central alarm system does not provide means of alarm optimization, grouping or prioritization. There are three main options for central alarm system replacement: Conventional annunciator system, hybrid annunciator system and advanced alarm system. Advanced alarm system implementation requires Main Control Board upgrade, integration of process instrumentation and plant process computer as well as long time for replacement. NPP Krsko has decided to implement hybrid alarm system with patchwork approach. The new central alarm system will be stand alone, digital, with advanced filtering and alarm grouping options. Sequence of event recorder will be linked with plant process computer and time synchronized with redundant GPS signal. Advanced functions such as link to plant procedures will be implemented with plant process computer upgrade in outage 2006. Central alarm system replacement is due in outage 2004.(author)

  1. NPP Krsko Containment Response Following Main Steam Line Break

    International Nuclear Information System (INIS)

    Spalj, S.; Grgic, D.; Cavlina, N.

    2000-01-01

    This paper presents the calculation of thermohydraulic environmental parameters (pressure and temperature) inside containment of Krsko NPP after postulated Main Steam Line Break (MSLB) accident. This analysis was done as a part of the ambient parameters specification in the frame of the NPP Krsko Equipment Qualification (EQ) project. The RELAP5/mod2 computer code was used for the determination of MSLB mass and energy release and computer code GOTHIC was used to calculate pressure and temperature profiles inside NPP Krsko containment. The analysis was performed for spectrum of break sizes to account for possible steam superheating during accidents with smaller break sizes. (author)

  2. NPP Krsko simulator training for operations personnel

    International Nuclear Information System (INIS)

    Pribozic, F.; Krajnc, J.

    2000-01-01

    Acquisition of a full scope replica simulator represents an important achievement for Nuclear power Plant Krsko. Operating nuclear power plant systems is definitely a set of demanding and complex tasks. The most important element in the goal of assuring capabilities for handling such tasks is efficient training of operations personnel who manipulate controls in the main control room. Use of a simulator during the training process is essential and can not be substituted by other techniques. This article gives an overview of NPP Krsko licensed personnel training historical background, current experience and plans for future training activities. Reactor operator initial training lasts approximately two and a half years. Training is divided into several phases, consisting of theoretical and practical segments, including simulator training. In the past, simulator initial training and annual simulator retraining was contracted, thus operators were trained on non-specific full scope simulators. Use of our own plant specific simulator and associated infrastructure will have a significant effect on the operations personnel training process and, in addition, will also support secondary uses, with the common goal to improve safe and reliable plant operation. A regular annual retraining program has successfully started. Use of the plant specific simulator assures consistent training and good management oversight, enhances conformity of operational practices and supports optimization of operating procedures. (author)

  3. Industry Operating Experience Process at Krsko NPP

    International Nuclear Information System (INIS)

    Bach, B.; Bozin, B.; Cizmek, R.

    2012-01-01

    Experience has shown that number of minor events and near misses, usually without immediate or significant impact to plant safety and reliability, are precursors of significant or severe events due to the same or similar root or apparent cause(s). It is therefore desirable to identify and analyze weaknesses of the precursor problems (events) in order to prevent occurrence of significant events. Theoretically, significant events could be prevented from occurring if the root cause(s) of these precursor problems could be identified and eliminated. The Operating Experience Program identifies such event precursors and by reporting them to the industry, plant specific corrective actions can be taken to prevent events at other operational plants. The intent of the Operating Experience Program is therefore to improve nuclear power plant safety and reliability of the operating nuclear power plants. Each plant develops its own Operating Experience Program in order to learn from the in-house operating experience as well as from the world community of nuclear plants. The effective use of operating experience includes analyzing both plant and industry events in order to identify fundamental weaknesses and then determining appropriate plant-specific actions that will minimize the likelihood of similar events. Learning and applying the lessons from operating experience is an integral part of station safety culture and is encouraged by managers throughout the top plant administrative programs and procedures. Krsko NPP is developed it own Operating Experience Program by using the most relevant INPO/WANO/IAEA guidelines as well as its own knowledge, skills an operating practice. The Operating Experience Program is a part of the Corrective Action Program, which is among top management programs, thus program is strongly encouraged by top management. The purpose of Operating Experience Program is to provide guidance for using, sharing, and evaluating operating experience information

  4. Plant performance monitoring program at Krsko NPP

    International Nuclear Information System (INIS)

    Bach, B.; Kavsek, D.

    2004-01-01

    A high level of nuclear safety and plant reliability results from the complex interaction of a good design, operational safety and human performance. This is the reason for establishing a set of operational plant safety performance indicators, to enable monitoring of both plant performance and progress. Performance indicators are also used for setting challenging targets and goals for improvement, to gain additional perspective on performance relative to other plants and to provide an indication of a potential need to adjust priorities and resources to achieve improved overall plant performance. A specific indicator trend over a certain period can provide an early warning to plant management to evaluate the causes behind the observed changes. In addition to monitoring the changes and trends, it is also necessary to compare the indicators with identified targets and goals to evaluate performance strengths and weaknesses. Plant Performance Monitoring Program at Krsko NPP defines and ensures consistent collection, processing, analysis and use of predefined relevant plant operational data, providing a quantitative indication of nuclear power plant performance. When the program was developed, the conceptual framework described in IAEA TECDOC-1141 Operational Safety Performance Indicators for Nuclear Power Plants was used as its basis in order to secure that a reasonable set of quantitative indications of operational safety performance would be established. Safe, conservative, cautious and reliable operation of the Krsko NPP is a common goal for all plant personnel. It is provided by continuous assurance of both health and safety of the public and employees according to the plant policy stated in program MD-1 Notranje usmeritve in cilji NEK, which is the top plant program. Establishing a program of monitoring and assessing operational plant safety performance indicators represents effective safety culture of plant personnel.(author)

  5. Corrective action program at Krsko NPP

    International Nuclear Information System (INIS)

    Skaler, F.; Divjak, G.; Kavsek, D.

    2004-01-01

    The Krsko NPP develops software that enables electronic reporting of all kind of deviations and suggestions for improvement at the plant. All the employees and permanent subcontractors have the access to the system and can report deviations. NPP has centralized decision process for the distribution of reported deviation. At this point all direct actions are electronically tracked. The immediate benefits of this new tool were: Reporting threshold has been lowered; Number of reporting people has increased; One computerized form for all processes; Decision, which process will solve the deviation, is centralized; All types of deviation are in the same environment; Our experiences of the processes are incorporated in the program; Control of work that has been done; Archiving is electronic only. Software basic data: Application system Corrective action program is a WEB application. Data is stored in Oracle 8.1.7 i database. Users access application through PL/SQL gateway on Oracle 9i Application Server 1.0.2. using Microsoft Internet Explorer browsers(Version 5 or later). Reports are implemented by Oracle Reports 6i. Menus are designed by Apycom Java Menus and Buttons v4.23. Our Presentation will include: Basic idea; Implementation change management; Demonstration of the program.(author)

  6. Intranet portal at the Krsko NPP, Slovenia

    International Nuclear Information System (INIS)

    2009-01-01

    The intranet portal (named IntraNEK) at Krsko NPP serves as a single entry point to access the internet and various plant applications and links. The front page consists of the standard internet search bar and links to various applications that can either reside within the technological computer network (TRM) or within the plant business computer network. Access to the TRM applications is read only. Some applications on the business computer network are open to all personnel who log on to the network while some applications are restricted and secured, and require additional login entries. A selected link will open in a new window. Documents will open with the appropriate software tool depending on the document file format. Some categories of documents are available in image form only (e.g. procedures, drawings etc.), while some are available in fully searchable PDF format (e.g. technical specifications, updated safety analysis reports (USARs) etc.). Plant departments (organizational units) have their own pages accessible from the front page. Their pages contain links to their own information resources or links to other resources and applications, tailored to the department needs. During recent years a number of web based applications have been developed that are connected also with a common Oracle database. Some are designed to serve for data entry and browsing while others serve for browsing only

  7. Regulatory review of NPP Krsko Periodic Safety Review

    International Nuclear Information System (INIS)

    Lovincic, D.; Muehleisen, A.; Persic, A.

    2004-01-01

    At the request of the Slovenian Nuclear Safety Administration (SNSA), Krsko NPP prepared a Periodic Safety Review (PSR) program in January 2001. This is the first PSR of NPP Krsko, the only nuclear power plant in Slovenia. The program was reviewed by the IAEA mission in May 2001 and approved by SNSA in July 2001. The program is made in accordance with the IAEA Safety Guide 'Periodic Safety Review of Operational Nuclear Power Plants' No. 50-SG-012 and with European practice. It contains a systematic review of operation of the NPP Krsko, including the review of the changes as a result of the modernization of the facility. The main tasks of PSR are review of plant status for each safety factor, development of aging and life cycle management program, review of seismic design and PSHA analysis and update of regulatory compliance program. The prioritization process of findings and action plan are also important tasks of PSR. The basic safety factors of the PSR review are: Operational Experience, Safety Assessment and Analyses, Equipment Qualification and Ageing Management, Safety Culture, Emergency Planing, Environmental Impact and Radioactive Waste, Compliance with license requirements and Prioritization. It had been agreed that SNSA will have reviewed all PSR reports generated during the PSR process. At the end of 2003 the PSR Summary Report with selected recommendations for action plan was completed and delivered to SNSA for review. The paper presents regulatory review of NPP Krsko PSR with emphasis on the evaluation of the PSR issues ranking process. (author)

  8. Disposal of spent nuclear fuel from NPP Krsko

    International Nuclear Information System (INIS)

    Mele, I.

    2004-01-01

    In order to get a clear view of the future liabilities of Slovenia and Croatia regarding the long term management of radioactive waste and spent nuclear fuel produced by the NPP Krsko, an estimation of disposal cost for low and intermediate level waste (LILW) as well as for spent nuclear fuel is needed. This cost estimation represents the basis for defining the target value for the financial resources to be accrued by the two national decommissioning and waste disposal funds, as determined in the agreement between Slovenia and Croatia on the ownership and exploitation of the NPP Krsko from March 2003, and for specifying their financial strategies. The one and only record of the NPP Krsko spent fuel disposal costs was made in the NPP Krsko Decommissioning Plan from 1996 [1]. As a result of incomplete input data, the above SF disposal cost estimate does not incorporate all cost elements. A new cost estimation was required in the process of preparation of the Joint Decommissioning and Waste Management Programme according to the provisions of the above mentioned agreement between Slovenia and Croatia. The basic presumptions and reference scenario for the disposal of spent nuclear fuel on which the cost estimation is based, as well as the applied methodology and results of cost estimation, are presented in this paper. Alternatives to the reference scenario and open questions which need to be resolved before the relevant final decision is taken, are also briefly discussed. (author)

  9. NPP Krsko Lifetime Extension - Business Impact for Hrvatska Elektroprivreda

    International Nuclear Information System (INIS)

    Vrankic, K.; Krejci, M.; Lebegner, J.

    2006-01-01

    This paper deals with the analysis of possible business impacts for HEP in the case of NPP Krsko life extension. Due to numerous reasons nuclear power plant life extension of ten to twenty years is a common procedure abroad. Having this practise in mind as well as other circumstances in Croatian and Slovenian electric power system, the extension of NPP Krsko lifetime is considered to be a possible scenario. Foreseeable impacts of this decision are evaluated primarily with consideration of its effect on HEPs projected cash flows, though other aspects will be addressed as well. Preserving a well maintained production facility with an extraordinary operational record and stable, or possibly falling overall production costs seems as a very rational choice. This is particularly true having in mind expected rise of electricity demand and energy prices in the region. Having NPP Krsko in operation beyond 2023 implies that no replacement source for NPP Krsko capacity needs to be built. This means avoiding all costs connected with the construction and operation of the replacement plant, assuming it will be fossil fuelled. Due to the high uncertainty of the future fossil fuel prices, the avoidance of replacement plant operational cost is likely to prove as highly rewarding. It should be kept in mind that avoided costs also include the replacement plant greenhouse gases emission costs, thus further enlarging the list of value adding impacts. The latter is valid anticipating the ratification of the Kyoto protocol and joining the European emission trading scheme. In addition to that, the extension of NPP Krsko lifetime would mean that the majority of costs connected with the decommissioning and final waste disposal can be postponed further down the time line. This will have very positive financial and possibly technological impact. Other value creating effects for HEP that are foreseeable as a consequence of the plant lifetime extension include: maintaining the knowledge of

  10. Prioritization of the recovery actions in the Krsko NPP SAMGs

    International Nuclear Information System (INIS)

    Basic, Ivica; Vrbanic, Ivan; Bilic-Zabric, Tea

    2006-01-01

    Replacement of the old SG (Steam Generators) and the characteristic of new ones raises the question of proper severe accident management (SAM) strategy, which leans on WOG (Westinghouse Owners Group) generic approach philosophy that repair and recovery actions have the first priority. In the current revision of NPP Krsko SAMGs (Severe Accident Management Guidelines) under DFC (Diagnostic Flow Chart) decision making process, water supply to the SG has higher priority than the re-injection of water into the core. NPP Krsko reconsideration of the current highest priority recovery action (inject water into the SG) and potential revision of Severe Accident Management Guidelines is the subject of this paper. Phenomenology Evaluations were performed by MAAP (Modular Accident Analysis Program) 4.0.5 code. (author)

  11. Internal Audits and Quality Assurance Surveillance in NPP Krsko

    International Nuclear Information System (INIS)

    Cavajda, M.; Bracic, I.

    1996-01-01

    This paper is describing establishment of the requirements for the development and execution of the Internal Audit and Quality Assurance Surveillance Program in the NPP Krsko, to identify relevant regulatory commitment and other documents, and to exhibit different functional areas, levels and work categories and factors that impact selecting and scheduling an audit or surveillance. It is not intention of this paper to explain how and by whom an audit or surveillance has to be done. (author)

  12. NPP Krsko containment environmental conditions during postulated accident

    International Nuclear Information System (INIS)

    Kozaric, M.; Cavlina, N.; Spalj, S.

    1989-01-01

    This paper presents NPP Krsko containment pressure and temperature increase during Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB). Containment environmental condition calculation was performed by CONTEMPT4/MOD4 computer code. Design accident calculations were performed by RELAP4/MOD6 and RELAP5/MOD1 computer codes. Calculational abilities and application methodology of these codes are presented. The CONTEMPT code is described in more detail. The containment pressure and temperature time distribution are presented as well. (author)

  13. Environmental Qualification Program for NPP Krsko

    International Nuclear Information System (INIS)

    Cerjak, J.; Klenovsek, P.; Pavsek, J.; Spalj, S.; Colovic, G.

    1998-01-01

    The functionality the equipment important to safety is deteriorated during its service due to ageing and harsh environment conditions. Since the environment is a potential for common cause failures, the purpose of Environmental Qualification (EQ) is to demonstrate the capability of safety-related equipment to perform its safety function in aged conditions and under extreme conditions after design bases event (DBE). EQ is one of the steps in licensing process according to US regulatory documents and standards (10CFR50.49, RG 1.89, NUREG-0588, IEEE-323). This paper presents the efforts in establishing the EQ program in the Krsko nuclear power plant. (author)

  14. NPP Krsko Severe Accident Management Guidelines Upgrade

    International Nuclear Information System (INIS)

    Mihalina, Mario; Spalj, Srdjan; Glaser, Bruno; Jalovec, Robi; Jankovic, Gordan

    2014-01-01

    Nuclear Power Plant Krsko (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry response to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG's). SAMG's are developed to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products. NEK new SAMG's revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pool (SFP) and to assess risk of core damage situation during shutdown operation. (authors)

  15. Operational and safety status of Krsko NPP

    International Nuclear Information System (INIS)

    Sirola, P.; Kavsek, D.

    1998-01-01

    Nuclear Power Plants Krsko (NEK) is producing electricity with the high level of reliability, safety and at acceptable price for 17 years. Energy is shared between both Slovenian and Croatian grid. The specifics of sharing the initial investment costs, later covering the operations costs and energy supply between Croatia and Slovenia is causing specific decision making problems about energy cost and future investments, however not influencing the plant safety, by now. NEK is continuously following the international nuclear technology practices, standards' changes and improvements and introducing them into the processes and equipment upgrades. As the member of the most important international integration, NEK is having the possibility of sharing its experience with others. Slovenian Energy Consumption and Supply Strategy is recognizing the NEK as a long term supply of energy in Slovenia being a strong decision making base for the future. According to the above mentioned Slovenian Energy Consumption and Supply Strategy the plant is obliged to keep all the radioactive waste, produced during the plant life, on site. The extensive efforts are taking place to reduce the radioactive waste production and save the area available for temporary waste deposition. The plant is licensed for the period of 40 years of commercial operation which started in 1983, so the Life Time Management is getting more and more important, including the performance tracing of the essential components, their maintenance and surveillance programs and also replacement plans of critical equipment. The major problems the NEK is confronted with at the moment are the Steam Generators which are reaching their and of life, and a very limited radioactive waste storage area. They are excerting influence on the plant availability and operations and maintenance costs. At the moment the process of Modernization is in progress, covering the Steam Generators replacement and a Plant Specific Simulators supply

  16. Evaluation of the safety margins during shutdown for NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Sadek, S.; Bajs, T.

    2004-01-01

    In the paper the results of RELAP5/mod3.3 calculations of critical parameters during shutdown for NPP Krsko are presented. Conservative evaluations have been performed at NPP Krsko to determine the minimum configuration of systems required for the safe shutdown operation. Critical parameters in these evaluations are defined as the time to start of the boiling and the time of the core dry-out. In order to have better insight into the available margins, the best estimate code RELAP5/mod3.3 has been used to calculate the same parameters. The analyzed transient is the loss of the Residual Heat Removal (RHR) system, which is used to remove decay heat during shutdown conditions. Several configurations that include open and closed Reactor Coolant System (RCS) were considered in the evaluation. The RELAP5/mod3.3 analysis of the loss of the RHR system has been performed for the following cases: 1) RCS closed and water solid, 2) RCS closed and partially drained, 3) Pressurizer manway open, Steam Generator (SG) U tubes partially drained, 4) Pressurizer and SG manways open, SG U tubes completely drained, 5) Pressurizer manway open, SGs drained, SG nozzle dams installed and 6) SG nozzle dams installed, pressurizer manway open, 1 inch break at RHR pump discharge in the loop with pressurizer. Both RHR trains were assumed in operation prior to start of the transient. The maximum average steady state temperature for all analyzed cases was limited to 333 K. (author)

  17. Investigating a possibility to implement 24-month cycle in NPP Krsko

    International Nuclear Information System (INIS)

    Bilic, T.; Pevec, D.; Smuc, T.

    1998-01-01

    The technical feasibility of a NPP Krsko reload core for a 24-month cycle has been shown. The equilibrium cycle core model for uprated NPP Krsko conditions using the computer code package FUMACS has been developed. 24-month equilibrium cycles for split feed enrichment batches of 68 and 72 fresh fuel assemblies were generated in that model. The analysis of these preliminary loading patterns for 24-month equilibrium cycles showed that it is possible to design the 24-month cycle cores for the NPP Krsko from a fuel management standpoint.(author)

  18. Prospects for the NPP Krsko Radioactive Waste Management

    International Nuclear Information System (INIS)

    Knapp, A.; Levanat, I.; Saponja-Milutinovic, D.

    2016-01-01

    Croatia adopted Strategy of radioactive waste, used sources and spent fuel management in 2014, and its Law on radiological and nuclear safety was accordingly modified in 2015. The Strategy foresees (though with some flexibility) and the Law declares decidedly that Croatia will establish a Center for radioactive waste management, in which all necessary facilities for storage and subsequent disposal of the Croatian share of the NPP Krsko radioactive waste and spent fuel will be developed. However, Slovenia and Croatia have recently agreed that a long-term dry storage for spent fuel will be established on the NPP premises by the year 2019. Therefore, only the issues of low and intermediate level waste (LILW) are addressed here. In Slovenia, the LILW repository site Vrbina in Krsko municipality was officially confirmed in 2009. Based on the 2013 investment program for a silo-type disposal facility, preparation of the repository project documentation was contracted with a national engineering company in 2014. Slovenian repository concept has been developed in two variants: one for the Slovenian LILW only, and the other intended to accept the Croatian share of LILW from the NPP as well. In the summer of 2015 Slovenia for the first time made an official offer to Croatia to use Vrbina repository for that purpose. However, the Croatian Strategy also does not preclude the option of management of all LILW from the NPP in Croatia. Therefore, present plan of activities for the third revision of the NPP radioactive waste and spent fuel management program outlines hypothetically symmetrical LILW management options: all in Slovenia, or all in Croatia, or one half in each country. So, what shall it be? This paper discusses the prospects for each of the three above mentioned options. The major problem of the Slovenian disposal plan is its high cost, mostly due to high compensations to the local community, which will be hard to finance without Croatian participation. The simplest

  19. Equipment Reliability Program in NPP Krsko

    International Nuclear Information System (INIS)

    Skaler, F.; Djetelic, N.

    2006-01-01

    Operation that is safe, reliable, effective and acceptable to public is the common message in a mission statement of commercial nuclear power plants (NPPs). To fulfill these goals, nuclear industry, among other areas, has to focus on: 1 Human Performance (HU) and 2 Equipment Reliability (EQ). The performance objective of HU is as follows: The behaviors of all personnel result in safe and reliable station operation. While unwanted human behaviors in operations mostly result directly in the event, the behavior flaws either in the area of maintenance or engineering usually cause decreased equipment reliability. Unsatisfied Human performance leads even the best designed power plants into significant operating events, which can be found as well-known examples in nuclear industry. Equipment reliability is today recognized as the key to success. While the human performance at most NPPs has been improving since the start of WANO / INPO / IAEA evaluations, the open energy market has forced the nuclear plants to reduce production costs and operate more reliably and effectively. The balance between these two (opposite) goals has made equipment reliability even more important for safe, reliable and efficient production. Insisting on on-line operation by ignoring some principles of safety could nowadays in a well-developed safety culture and human performance environment exceed the cost of electricity losses. In last decade the leading USA nuclear companies put a lot of effort to improve equipment reliability primarily based on INPO Equipment Reliability Program AP-913 at their NPP stations. The Equipment Reliability Program is the key program not only for safe and reliable operation, but also for the Life Cycle Management and Aging Management on the way to the nuclear power plant life extension. The purpose of Equipment Reliability process is to identify, organize, integrate and coordinate equipment reliability activities (preventive and predictive maintenance, maintenance

  20. Steam Generator tube plugging analysis of natural circulation conditions for NPP Krsko

    International Nuclear Information System (INIS)

    Bajs, T.; Mirkovic, D.

    1989-01-01

    Pump trip for NPP Krsko was analysed by deterministic approach. Analyses for 0% and 10% tube plugging were performed using computer code RELAP4/MOD6. The influence of steam generator tube plugging on natural circulation conditions is discussed. (author)

  1. Evaluation of rod insertion issue for NPP Krsko

    International Nuclear Information System (INIS)

    Gunstek, A.; Kurincic, B.

    1998-01-01

    The last couple of years incident with control rods sticking in lower part of the fuel assemblies have been reported of several reactor operators and fuel vendors throughout of the world. Several activities were initiated immediately to determine the root cause of incomplete rod insertion. The purpose of this activities were to collect plants trip history data and testing results, review of available worldwide experience, review of plant operation and fuel management, detailed review of manufacturing and material property and to maintain detailed mechanical model. In this paper, we will present activities in Nuclear Power Plant Krsko which have been performed after NRC initiated the Root Cause Process (NRC Bulletin 96-01). NPP Krsko has not experienced rod insertion anomaly yet but anyway the additional tests were carried out. Rod drop time measurements that were performed normally at beginning of cycle at nominal temperature and pressure (HSB mode) have been extended also to end of cycle. Rod drop time, velocity of dropped rods and magnitudes of the initial recoil bounces vs. burnup were also analyzed. Also RCCA drag test with upper internals in place and drive shafts attached to RCCAs has been performed since then. At last two outages (1997 and 1998) drag test were carried out with digital scale meter to gather additional information. In addition to that, the reload core design has been performed with new constrains on rodded fuel assembly burnup as proposed by the industry.(author)

  2. New iteration of decommissioning program for NPP Krsko

    International Nuclear Information System (INIS)

    Lokner, V.; Levanat, I.; Rapic, A.; Zeleznik, N.; Mele, I.; Jenko, T.

    2004-01-01

    As required by the paragraph 10 of the Agreement between the governments of Slovenia and Croatia on status and other legal issues related to investment, exploitation, and decommissioning of Nuclear power plant Krsko, Decommissioning program for Krsko NPP including LILW and spent fuel management was drafted. The Intergovernmental body required that the Program should be extensive revision of existing program as one of several iterations to be prepared before the final version. The purpose of the Program is to estimate the expenses of the future decommissioning, radioactive waste and spent fuel management for Krsko NPP. Costing estimation would be the basis for establishment of a special fund in Croatia and for adjustment of the annual rates for the existing decommissioning fund in Slovenia. The Program development was entrusted to specialized organizations both in Croatia and Slovenia, which formed the Project team as the operative body. Consulting firms from Croatia and Slovenia were involved as well as experts from the International Atomic Energy Agency (through short visits to Zagreb and Ljubljana) for specialized fields (e.g. economic aspects of decommissioning, pre-feasibility study for spent fuel repository in crystalline rock, etc.). The analysis was performed in several steps. The first step was to develop rational and feasible integral scenarios (strategies) of decommissioning and LILW and spent fuel management on the basis of detailed technical analysis and within defined boundary conditions. Based on technological data, every scenario was attributed with time distribution of expenses for all main activities. In the second step, financial analysis of the scenarios was undertaken aiming at estimation of total discounted expense and the related annuity (19 installments to the single fund, empty in 2003) for each of the scenarios. The third step involves additional analysis of the chosen scenarios aiming at their (technical or financial) improvements even at

  3. Integrated safety assessment of the NPP Krsko modernization

    International Nuclear Information System (INIS)

    Vrbanic, I.; Kastelan, M.; Krajnc, B.; Spiler, J.

    2000-01-01

    Nuclear power plant Krsko (NEK) replaced, as a part of the modernization project, the old Westinghouse D4 steam generators, with the new ones, designed and manufactured by consortium Siemens-Framatome. The replacement itself required a number of modifications on the secondary side (feedwater, condensate system,..), as well as others, to be implemented. Among those the most important are: installation of the inadequate core cooling monitoring instrumentation, the set of most important modifications that will improve the plant fire resistance and fighting capability, and so called 'wet cavity design' modification, etc. The SGs replacement, with some plant system modifications, as well as new generation of reactor fuel, will allow NEK to increase the reactor power for 6.3%. As the part of the same program, the NEK obtained on-site, plant specific Full Scope Simulator (KFSS). Within this modernization program, the set of consistent, comprehensive safety analyses was performed, to demonstrate that the plant could safely operate under these new conditions. Methodology selected in performing these studies, have numerous references in US, as well as in Europe. In addition to the required set of licensing analyses, NEK decided to perform the integrated safety assessment (ISA) of all plant modifications/changes, with the available plant PSA model and methodology. The starting point was complete and extensively review task of the NEK design modification/change data base, and implementation of the reviewed changes/modifications into the PSA Level 1/Level 2 analysis model, developed within IPE/IPEEE project (Individual Plant Examination for internal and external events). Paper presents results of this project. (author)

  4. Arrangement of the Krsko NPP protection scheme for the power system malfunction cases

    International Nuclear Information System (INIS)

    Omahen, P.; Pavsek, J.; Dirnbek, V.

    1996-01-01

    The Krsko NPP has been designed with the capability to reject 100% of its rated power and runback to the station electrical load. However, an adequate detection system of the outside network degradation is needed for the activation of the existing load drop anticipated (LDA) function. The Krsko NPP electrical, turbine and generator protection systems were carefully evaluated in order to redesign some of its functions. These additional functions should be able to protect and disconnect the plant from the system whenever some serious trouble of the outside electric power system is detected. On the other side, preventive measures should be introduced to avoid unnecessary plant disconnection or unnecessary power system collapse due to such disconnection. At the end, the paper presents a precise design of additional function possibilities for the Krsko NPP electrical protection system. A critical evaluation of these functions is given and the best option is proposed. (author)

  5. Some results of Krsko NPP core calculations and comparison with measurements

    International Nuclear Information System (INIS)

    Trkov, A.; Zefran, B.; Kromar, M.; Ravnik, M.; Slavic, S.

    1996-01-01

    Current status of the CORD-2 package is described. Results of the predictions of some important reactor core parameters are presented for the 12 th operation cycle of the Krsko NPP. Comparison with measurements is made to illustrate that the accuracy of the calculations is acceptable. Some comments are made on the enhancements, which are currently being implemented on the package. (author)

  6. Compensations to Local Communities in the Krsko NPP Decommissioning Program

    International Nuclear Information System (INIS)

    Levanat, I.; Knapp, A.; Lokner, V.

    2010-01-01

    In Slovenia, direct financial compensations (for 'limited land use') to local communities hosting nuclear facilities were initially specified by a government Decree from 2003. In Croatia, a possibility of direct financial compensations had been indicated in the land use plan in conjunction with the prospective RW repository siting about a decade earlier, but the topic was subsequently abandoned together with the repository project. In 2004, the joint Slovenian-Croatian Decommissioning and LILW and SF management program for NPP Krsko from 2004 (the 1st revision of the joint Program) conservatively included the compensation amounts from the Slovenian Decree into the cost estimates of LILW and SF repositories, although their location was entirely unspecified ('in Slovenia or in Croatia'). Shortly before the 2nd revision of the joint Program started in the fall of 2008, the Slovenian government had amended its Decree, practically doubling the amounts of the repository compensations. Assuming that some (or possibly all) nuclear facilities and waste, dealt with in the Program, may be located in Slovenia, the revision has adopted a conservative approach to include all compensations to local communities that may be required by the Slovenian regulations into the Program costs. This paper discusses the Slovenian government Decree, its impact on the joint Program costs, and its implications on RW and SF management in the region. The Decree suffers from the lack of self-consistency, clarity, and consistency with the more general legal provisions on which it should have been based, but it may have an important supporting role in the process of RW and SF management facilities siting. The Decree introduced significant additional costs into the joint Program, which have grown from about one hundred million eur in the 1st revision to about half a billion in this revision (depending on the Program scenario). Besides, application of the Decree in the joint Program has set a precedent

  7. System, economy and ecology viewpoints of the Krsko NPP lifetime extension

    International Nuclear Information System (INIS)

    Novsak, M.; Spiler, J.; Zagar, T.; Pirs, B.; Bole, A.; Bregar, Z.; Cuhalev, I.; Derganc, B.; Ivanjko, S.; Matvoz, D.; Sustersic, A.; Valencic, L.; Zabric, I.; Zlatarev, G.; Babuder, M.

    2007-01-01

    Krsko NPP plant life extension was analysed and evaluated with respect to system, economy and ecology viewpoints. From the system perspective it was established that also in the extended lifetime the plant will remain in operation as a base load electricity supplier. The systematic review was performed to determine its overall competitiveness against advanced coal, gas and new nuclear units. The analysis considered also hydro and renewable sources. Analysis and evaluations resulted in the conclusion that the Krsko NPP lifetime extension is the most effective alternative for base load production due to small additional capital investments, low fuel costs, no new siting requirements, lowest climate and environmental impact, and reliable and safe operation. (author)

  8. Validation of the CORD-2 System for the Nuclear Design Calculations of the NPP Krsko Core

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2016-01-01

    The CORD-2 package intended for core design calculations of PWRs has be recently updated with some improved models. Since the modifications could substantially influence the obtained results, a technical validation process is required. This paper presents comparison of some calculated and measured parameters of the NPP Krsko core needed to qualify the package. Critical boron concentrations at hot full power for selected cycle burnup points and several parameters obtained during the start-up testing at the beginning of each cycle (hot zero power critical concentration, isothermal temperature coefficient and rods worth) for all 27 finished cycles of operation are considered. In addition, assembly-wise power distribution for some selected cycles is checked. Comparison has shown very good agreement of the CORD-2 calculated values with the selected measured parameter of the NPP Krsko core.(author).

  9. New version of NPP Krsko Decommissioning program and LILW and spent fuel management

    International Nuclear Information System (INIS)

    Zeleznik, N.; Mele, I.; Jenko, T.; Lokner, V.; Levanat, I.; Rapic, A.

    2004-01-01

    According to the requirements of the bilateral agreement between Republic of Slovenia and Republic of Croatia on the legal and other obligations for Nuclear power plant (NPP) Krsko the Decommissioning program was prepared. The main purpose of the program was to estimate the overall expenses of the future decommissioning, radioactive waste and spent fuel management of the NPP Krsko in order to establish separate fund in Croatia and to correct the rate per kWh collected in the existing decommissioning fund in Slovenia. The program looked at all possible scenarios of dismantling, radioactive waste and spent fuel management and proposed the most plausible two scenarios which are technically possible and financially feasible. (author)

  10. Use of a Computerized Tool (ORAM) to Help Manage Outage Safety and Risk at NPP Krsko

    International Nuclear Information System (INIS)

    Spiler, J.; Basic, I.; Vrbanic, I.; Fifnja, I.; Kastelan, M.; Dagan, W. J.; Shanley, L. B.; Naum, T. J.

    1998-01-01

    Outage Risk Assessment and Management (ORAM) is a computerized methodology developed by the U.S. Electric Power Research Institute (EPRI) to help Nuclear Power Plant personnel manage the risk and safety associated with refueling and forced plant outages. Today, over 60 plants including NPP Krsko are using ORAM during the preparation and performance of plant outages. In fact, many plants are attributing much of the reductions in the duration of refueling outages to the use of ORAM. The success of the ORAM methodology is the capability to provide plant and management personnel with understandable results from both deterministic evaluations of plant safety and quantitative risk assessments. The Nuklearna Elektrarna Krsko (NEK) use of ORAM involves both of these approaches. The deterministic portion of ORAM is used to model the NPP Krsko Shutdown Technical Specifications and administrative considerations. The probabilistic portion of ORAM uses industry and NEK specific initiating events and other risk elements pertaining to shutdown to derive a quantitative risk assessment for various end states, including core damage and RCS boiling. This paper expands on the value of each approach and demonstrates the benefits of combining these elements in the decision-making process. Another key advantage of ORAM is the ability to apply the methodology to specific outages. Since no outage is identical, this provides tremendous benefits to plant personnel for managing the safety and risk of a particular outage. ORAM does this ba organizing all of the various plant configurations and equipment unavailability windows into numerous plant states. Furthermore, ORAM evaluations can be a utomated b y interfacing with outage scheduling software programs such as Primavera. For each plant state, the deterministic and the probabilistic logic evaluations are applied. This paper will demonstrate the ORAM evaluation for an actual NPP Krsko outage. (author)

  11. Strengthening ALARA approach in work management at Krsko NPP

    International Nuclear Information System (INIS)

    Breznik, B.; Kovac, Z.; Sirola, P.

    1999-01-01

    As Low As Reasonably Achievable (ALARA) occupational exposures at nuclear power plants should be included in work management as a concept. There are world-wide trends required by the utilities for improved design, operation and maintenance. Within the period of seventeen years of plant operation maintaining low radiation exposures requires additional efforts. The benefit of this effort should be reducing risks to nuclear workers, better work planning and performance. The Krsko Plant ALARA organisations has been revised recently and built on different levels of the hierarchy. The goal is to promote good industry practice and the management of work on primary systems. The established ALARA programme describes the objectives and defines the procedures and tools for its implementation. Brief presentation of the programme as well as organisational responsibilities of dedicated ALARA committee and working groups is the scope of this paper. The management tools and ALARA indicators are discussed to implement the programme and to evaluate the results.(author)

  12. NPP Krsko periodic safety review. Safety assessment and analyses

    International Nuclear Information System (INIS)

    Basic, I.; Spiler, J.; Thaulez, F.

    2002-01-01

    Definition of a PSR (Periodic Safety Review) project is a comprehensive safety review of a plant after ten years of operation. The objective is a verification by means of a comprehensive review using current methods that the plant remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. The overall goals of the NEK PSR Program are defined in compliance with the basic role of a PSR and the current practice typical for most of the countries in EU. This practice is described in the related guides and good practice documents issued by international organizations. The overall goals of the NEK PSR are formulated as follows: to demonstrate that the plant is as safe as originally intended; to evaluate the actual plant status with respect to aging and wear-out identifying any structures, systems or components that could limit the life of the plant in the foreseeable future, and to identify appropriate corrective actions, where needed; to compare current level of safety in the light of modern standards and knowledge, and to identify where improvements would be beneficial for minimizing deviations at justifiable costs. The Krsko PSR will address the following safety factors: Operational Experience, Safety Assessment, EQ and Aging Management, Safety Culture, Emergency Planning, Environmental Impact and Radioactive Waste.(author)

  13. Development of the NPP Krsko database for component cyclic or transient limits

    International Nuclear Information System (INIS)

    Bencik, V.; Fancev, T.; Basic, I.; Plestenjak, T.

    2004-01-01

    The safety of Nuclear Power Plants (NPPs) can be affected by aging degradation of key components or structures if the degradation is not detected prior to the loss of functional capability and if timely corrective action is not taken. Fatigue is a degradation process due to repeated loading that leads to micro damage accumulation. With an increasing number of repeating loading cycles, the existing microcracks can grow to macroscopic size. Finally, the fracture and the complete failure of the affected component can occur. For the assessment of component fatigue, the repeated transients, that are relevant for fatigue degradation, have to be counted and classified. For design limits evaluation, the transients and operational cycles were selected to represent a conservative estimate of the magnitude and frequency of the resulting temperature and pressure transients. Review and classification of NPP Krsko transients and operational cycles has been performed according to four categorization systems: 1) Design Transient Specification according to USAR NEK provides the categorization of plant events and determines the design number of occurrences over the 40-year life of the plant. The categorized transients are considered to be of magnitude and/or frequency that are significant in the component design and fatigue evaluation. 2) The American Nuclear Society (ANS) categorization of plant conditions (Regulatory Guide 1.70, Rev. 3) divides plant conditions into four categories in accordance with anticipated frequency of occurrence and potential radiological consequences to the public. This classification does not consider the fatigue evaluation. 3) Classification of transients and accidents according to NEK Technical Specifications summarizes component cyclic or transient limits into 12 categories. The components identified in the categorization system are designed and shall be maintained within the defined cyclic or transient limits. 4) Categorization of NPP Krsko transients

  14. SNSA surveillance over the ageing effects and ability for long term operation at Krsko NPP

    International Nuclear Information System (INIS)

    Savli, S.; Ferjancic, M.; Pavlin, D.; Lovincic, D.

    2007-01-01

    The paper presents the Slovenian Nuclear Safety Administration (SNSA) tools used for verification the adequacy of management of ageing effects and assuring suitability for long term operation at the Krsko NPP. In addition to tools commonly used as PSR (Periodic Safety Review), assessment of plant modifications, regular inspections, the SNSA applies some special methods like monitoring the condition of important plant structures, systems and components (SSC) through special designed software, review and assessment of important plant programmes and its own set of performance indicators

  15. Radiological Shielding Analysis of Special Building for the Old NPP Krsko Steam Generators

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    2000-01-01

    One of the NPP Krsko modernization projects is steam generator replacement, during which old steam generators have to be removed and stored on a safe location. Since old steam generators are radioactive a new, special building have to be designed and constructed at NPP Krsko site to house the old steam generators, as well as for the storage of non-compactable waste which will be produced during the replacement activities. It is the aim of this paper to estimate the dimensions of the outer walls and of the roof slab for the special building which will satisfy the radiological constraints for the unrestricted areas. Computer code ORIGEN2-PC has been used for activity estimation, while analytical expressions, QAD-CGGP and MCNP code have been used for the radiological shielding analysis. Our results indicate that 70 cm of the concrete for the wall and 60 cm for the roof provide satisfactory shield which limits the dose rates below American standards for level of radiation in unrestricted areas. (author)

  16. Impact of the measurement data on the CORD-2 nuclear design calculations of the NPP Krsko

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2004-01-01

    The CORD-2 package was developed at Jozef Stefan Institute and has been validated for the nuclear design calculations of PWR cores. It has been used for the independent verification of the NPP Krsko nuclear design for the last 6 cycles of operation. The accuracy of the package is very good fulfilling all criteria usually imposed on the design prediction of the reactor nuclear parameters. To obtain as robust package as possible and to eliminate potential systematic errors of the package, it was decided to rely on measured core power distributions. In core power measurements, which are performed each month of reactor operation, are used to obtain fuel assemblies burnup histories. Consequently, burnup distributions obtained from the power measurements of all previous cycles are taken as a starting point at the beginning of the considered cycle. Since a lot of experience has been gained with the package, it was decided to evaluate the impact of measurement data on the accuracy of the calculations. Burnup calculations of all 19 cycles of the NPP Krsko are repeated, building simultaneously the calculated library of burnup histories for all fuel assemblies. The basic reactor parameters such as HZP critical boron concentration, isothermal temperature coefficient, control rod worth and cycle length are compared to the results obtained with CORD-2 standard sequence of calculation and direct measurements.(author)

  17. The impact of NPP Krsko steam generator tube plugging on minimum DNBR at nominal conditions

    International Nuclear Information System (INIS)

    Lajtman, S.

    1996-01-01

    Typically, steam generator tube plugging (SGTP) both decreases the reactor coolant system (RCS) flow rate and the heat transfer surface area of the steam generator. At a constant thermal power and vessel outlet temperature, as tube plugging increases, the vessel average temperature, vessel inlet temperature and steam generator secondary side steam pressure decrease. This paper presents the analysis of impact of SGTP on Minimum Departure from Nucleate Boiling Ratio (MDNBR) at NPP Krsko (NEK), using the Improved Thermal Design Procedure (ITDP), WRB-1 correlation, and COBRA-III-C computer code. No credit was given to high plugging percentage region power reduction resulting from turbine volumetric flow limitations. MDNBR is found to be decreasing with increasing plugging, but not under the limiting values. (author)

  18. Radiation doses estimation for hypothetical NPP Krsko accidents without and with PCFV using RASCAL software

    International Nuclear Information System (INIS)

    Vukovic, Josip; Konjarek, Damir; Grgic, Davor

    2014-01-01

    Calculation is done using Source Term to Dose module of RASCAL (Radiological Assessment System Consequence Analysis) software to estimate projected radiation doses from a radioactive plume to the environment. Utilizing this module, it is possible to do preliminary assessment of consequences to the environment in case of adverse reactor conditions or releases from other objects containing radioactive materials before the emergency situation has happened or in the early phase. RASCAL is simple, easy to use, fast-running tool able to provide initial estimate of radiological consequences of nuclear accidents. Upon entering rather limited amount of input parameters for the Krsko NPP, mostly key plant parameters, time dependent source term calculation is executed to determine radioactive inventory release rates for different plant conditions, release paths and availability of protective measures. These rates given for each radionuclide as a function of time are used as an input to atmospheric dispersion and transport model. Together with release rates, meteorological conditions dataset serve as input to determine the behavior of the radioactive releases that is plume in the atmosphere. So as an output, RASCAL produces a 'dispersion envelope' of radionuclide concentrations in the atmosphere. These concentrations of radionuclides in the atmosphere are further used for estimating the doses to the environment and the public downwind the release point. Throughout this paper, dose assessment is performed for two distances, close-in distance and distance out to 40 km from the source, for hypothetical NPP Krsko accidents without and with Passive Containment Filtered Vent (PCFV) system used. Obvious difference is related to released radioactivity of Iodine isotopes. Results of radioactive effluents deposition in the environment are displayed via various doze parameters, radionuclide concentrations and materials exposure rates in this particular case. (authors)

  19. Relationship towards Engineering, Quality Assurance and 10CFR50.59 in the Design Change Process at the Krsko NPP

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.

    1998-01-01

    The paper discusses the relationship between the Krsko NPP design change engineering practice, quality assurance and the USA Nuclear Regulatory Commission 10 Code Federal Rule 50.59 (10CFR50.59). Together these controls ensure that plant design bases are maintained and yield a safe design. The 10CFR50.59 has been applied in Krsko NPP plant specific procedure entitled ESP-2.303 ''Authorization of Changes, Tests and Experiments'' (Safety Evaluation Screening) since 1994. All proposed changes requiring Safety Evaluations are being submitted to the SNSA (Slovenian Nuclear Safety Administration). If the proposed change is constituting an ''Unreviewed Safety Question'' the formal licensing procedure shall be completed before design change can be implemented otherwise the proposed design change is rejected. The procedure(ESP-2.303) provides the methodology to be followed in determining if a proposed activity involves an unreviewed safety question. An ''Unreviewed Safety Question'' is essentially the same as defined in 10CFR50.59(a)(2): ''A proposed change, test or experiment shall be deemed to involve an unreviewed safety question (1) if the probability of occurrence or consequences of an accident or malfunction of equipment important to safety previously evaluated in the (updated) safety analysis report may be increased; or (2) if a possibility for an accident or malfunction of a different type than any evaluated previously in the (updated) safety analysis report may be created; or (3) if the margin of safety as defined in the basis for any technical specification is reduced.'' This paper discusses the Following Krsko NPP Safety Evaluation aspects: 1. Defense in Depth Design Philosophy 2. Methodology 3. Definitions and Applicability of Terms 4. Evaluation Process Guidance and Documentation Process 5. Krsko NPP Lessons Learned. (author)

  20. Development of Tsunami PSA method for Korean NPP site

    International Nuclear Information System (INIS)

    Kim, Min Kyu; Choi, In Kil; Park, Jin Hee

    2010-01-01

    A methodology of tsunami PSA was developed in this study. A tsunami PSA consists of tsunami hazard analysis, tsunami fragility analysis and system analysis. In the case of tsunami hazard analysis, evaluation of tsunami return period is major task. For the evaluation of tsunami return period, numerical analysis and empirical method can be applied. The application of this method was applied to a nuclear power plant, Ulchin 56 NPP, which is located in the east coast of Korean peninsula. Through this study, whole tsunami PSA working procedure was established and example calculation was performed for one of real nuclear power plant in Korea

  1. Applicability of coupled code RELAP5/GOTHIC to NPP Krsko MSLB calculation

    International Nuclear Information System (INIS)

    Keco, M.; Debrecin, N.; Grgic, D.

    2005-01-01

    Usual way to analyze Main Steam Line Break (MSLB) accident in PWR plants is to calculate core and containment responses in two separate calculations. In first calculation system code is used to address behaviour of nuclear steam supply system and containment is modelled mainly as a boundary condition. In second calculation mass and energy release data are used to perform containment analysis. Coupled code R5G realized by direct explicit coupling of system code RELAP5/MOD3.3 and containment code GOTHIC is able to perform both calculations simultaneously. In this paper R5G is applied to calculation of MSLB accident in large dry containment of NPP Krsko. Standard separate calculation is performed first and then both core and containment responses are compared against corresponding coupled code results. Two versions of GOTHIC code are used, one old ver 3.4e and the last one ver 7.2. As expected, differences between standard procedure and coupled calculations are small. The performed analyses showed that classical uncoupled approach is applicable in case of large dry containment calculation, but that new approach can bring some additional insight in understanding of the transient and that can be used as simple and reliable procedure in performing MSLB calculation without any significant calculation overhead. (author)

  2. Reanalysis of some key tranzients with MAAP code for Krsko NPP after SG replacement and power uprate

    International Nuclear Information System (INIS)

    Krajnc, B.; Glaser, B.; Basic, I.; Novsak, M.; Spiler, J.

    1999-01-01

    The Krsko NPP will, as a part of the modernization project, replace old steam generators, implement required system modifications (mainly on the secondary side) as well as a number of safety improvements such as installations of inadequate core cooling monitoring instrumentation, a set of the most important modifications to improve plant fire resistant, implement the so called wet cavity modification, etc. SG replacement with some system modifications as well as new generation of fuel, will allow the Krsko NPP to increase power for 6,3 %. As a part of the risk assessment study for the above described plant changes some of the plant specific deterministic analyses performed within the Krsko NPP Individual Plant Examination for internal and external events (IPE/IPEEE) with MAAP code had to be repeated as suggested in (1). All deterministic thermohydraulic analyses within NEK IPE/IPEEE project have been performed with MAAP 3B, Ver.18. Since then (end of 1994) this analysis tool has been further improved and is now available as MAAP 4.03.5. To be able to clearly distinguish the effect of the plant changes from the effects of the code changes, it was decided that the same code version should be used. Required analyses have been performed and are documented in (2). Recently all the potentially affected analyses have been repeated with the new version of the code (MAAP 4.03.5) for both existing and future plant conditions. Analyses presented in the paper confirm the conclusions given (4) and (2). Recently all the potentially affected analyses have been repeated with the new version of the code (MAAP 4.03.5) for both existing and future plant conditions. Analyses presented in the paper confirm the conclusions given (4) and (2).(author)

  3. Analysis of the influence of steam generator tube plugging on the large break loss of coolant accident in NPP Krsko

    International Nuclear Information System (INIS)

    Bizjak, S.; Stritar, A.

    1987-01-01

    The preliminary analysis of the influence of steam generator tube plugging to the large break LOCA behaviour of the NPP Krsko was performed. If 10% of the tubes are plugged, the peak cladding temperature reached is 37 K higher than the temperature reached after LOCA if no tubes were plugged. The decrease of the maximum peaking factor from 2.34 to 2.25 would compensate the influence of 10% plugged tubes. The analysis was not fully in compliance with the requirements of the conservative methodology. (author)

  4. Modeling of containment response for Krsko NPP Full Scope Simulator verification

    International Nuclear Information System (INIS)

    Kljenak, I.; Skerlavaj, A.

    2000-01-01

    Containment responses during the first 10000 s of Anticipated Transient Without Scram and Small Break Loss-of-Coolant Accident scenarios in the Krsko two-loop Westinghouse pressurized water reactor nuclear power plant were simulated with the CONTAIN computer code. Sources of coolant were obtained from simulations with the RELAP5 code. The simulations were carried out so that the results could be used for the verification of the Krsko Full Scope Simulator. (author)

  5. Kozloduy NPP units 3 and 4 Level 2 PSA

    International Nuclear Information System (INIS)

    Vaklev, R.; Papazov, V.

    2005-01-01

    The main objectives and scope of the Kozloduy NPP units 3 and 4 Level 2 PSA are listed. The Containment Event Tree Results are given. The full power initiators risk distribution; internal events PDS risk distribution; fire PDS risk distribution and seismic PDS risk distribution are shown. The Thermal Hydraulic Analyses, including model for MELCOR code reactor nodalization; model for MELCOR code single loop with PRZ nodalization; model for MELCOR code common loop nodalization as well as model for MELCOR code JVC nodalization are presented. At the end the authors conclude that: 1) Definitive PSA-2 study results demonstrates, that the newly installed equipment dedicated to mitigation of severe accident consequences play significant role for decrease of the large early release frequency, providing the severe accident management; 2) The actions and measures prescribed by recently prepared SAMGs during Phare Project BG 01.10.0, lead to additional improvement of PSA-2 results, eventually to mitigation of severe accident consequences

  6. Experience in PSA fault tree modularization at the ASCO NPP

    International Nuclear Information System (INIS)

    Nos Llorens, V.; Frances Urmeneta, M.; Fraig Sureda, J.

    1995-01-01

    Probabilistic Safety Analysis (PSA) is a basic tool in decision-making for the optimization of back fittings, procedures and maintenance practices. ASCO NPP PSA was developed with a high level of detail in the models. This required considerable computer resources (long running time) to carry out the quantification. The quantification time had therefore to be flexible to allow continuous evaluation of the impact on the estimation and reduction of risk in the plant, and also to facilitate post-PSA applications. The most suitable way of achieving this flexibility was by compacting and reducing the detailed fault trees of the project by means of a modularization process. The purpose of the paper is to present the practical experience acquired with modularization carried out in the UTE UNITEC-INYPSA-EMPRESARIOS AGRUPADOS framework and the method applied, the support computer programs devised and their degree of effectiveness. (Author)

  7. Corrective action program at the Krsko NPP. Trending and analysis of minor events

    International Nuclear Information System (INIS)

    Bach, B.; Kavsek, D.

    2007-01-01

    understand the factors that might be responsible for such trend and to take corrective actions prior to the escalation to a significant event. Reviewed and analyzed data based on codes trending identified common problems, potential trends and common contributors, promote a good trending program. For the effective trending program, positive adverse trends identification and corrective actions that are addressed the weaknesses that have been identified, should be specified and implemented through the corrective action program. For that purpose the appropriate coding system incorporated into Corrective Action and Operating Experience Program is established at Krsko NPP. Minor events and near misses are collected and analyzed in order to aggregate detected minor problems. The different groups of codes developed include codes for direct causes and casual factors, processes and organizations, consequences, level of significance etc. For easier trending and further analysis a different code combinations were utilized in a form of graphs. For example: organisation vs. causal factors (allows particular department to trend human performance in their own organisation), direct cause vs. time (allows trending of equipment degradation), processes vs. organisation (allows trending 501.2 of processes degradation in particular organisation) any code in question vs. time (for trend confirmation) etc. The purpose of this article is to present the coding system established at the Krsko Nuclear Power Plant and variety of ways for trending by using the system. The article deals with the codes established, organization of code system, trend codes combinations and benefit for early recognizing adverse trends of lo-level events. (author)

  8. Scenario development and evaluation for the NPP Krsko revised decommissioning program

    International Nuclear Information System (INIS)

    Levanat, I.; Lokner, V.; Subasic, D.

    2004-01-01

    In this first revision, several integrated scenarios of the NPP Krsko dismantling and waste management were developed and analyzed in order to estimate the decommissioning program (DP) costs and to propose an appropriate funding plan. Most dismantling technologies and cost estimates were derived from the original decommissioning plan adopted in 1996. The LILW disposal cost estimates, however, rely on the tunnel type facility design which was developed in Slovenia a few years ago, whereas the SF repository design for this DP was adapted from the Swedish deep disposal concept. The starting assumptions for this DP were that the LILW repository would be licensed by 2013, the NPP would be permanently shut down in 2023, and the SF repository would become available in 2030. The boundary conditions also specified that DP should first re-evaluate the SID strategy from the original plan (Strategy Immediate Dismantling with immediate SF disposal, but also with a long period of on-site decay storage for the activated components, so that it actually terminates only after 96 years), and then modify it to achieve truly prompt decommissioning in which all planned activities should be completed within about 15 years after the NPP shut-down. In addition, the option of SF export to a third country should be introduced in all DP scenarios, as a realistic alternative to SF disposal into the local repository (in Slovenia or in Croatia). And finally, dry storage of SF for some 30 years before disposal or export, in an independent installation on unspecified location, should be evaluated within the DP sensitivity analysis. After a thorough analysis of the original SID strategy, it became clear that substantial modifications would be necessary in order to meet the boundary conditions while complying with the specified design and technologies of the assumed LILW and SF disposal facilities. Therefore, a systematic procedure for development and financial evaluation of feasible scenarios was

  9. Cofrentes NPP activities on PSA and severe accident analysis

    International Nuclear Information System (INIS)

    Suarez, J.; Borondo, L.; Garcia, P.J.

    1996-01-01

    Cofrentes NPP (CNPP) has developed a Level 1 PSA with the following scope: analysis of internal events, with the reactor initially operating at power, internal and external flooding risk analysis; internal fire risk analysis; reliability analysis of the containment heat removal and containment isolation systems. Level 1 CNPP-PSA results reveal that total core damage frequency in CNPP is less than other similar BWR/6 plants. The CNPP-PSA related activities and applications being carried out currently are: adjusting of MAAP 3.0B, revision 10, on VAX and PC; acquisition of MAAP 4; development of Level1/Level2-PSA interface; seismic site categorization for the IPEEE; prioritization of motor operated valves related to GL-89/10, complementary analysis for exemption to some 10CFR50 App. J requirements; Q-List grading; reliability-centered maintenance; maintenance rule support; on-line maintenance support, off-line risk-monitor development, PSA applicability to the 10CFR50 App. R requirements, analysis of the frequency of mis-oriented fuel bundle event, etc. About severe accident management, CNPP, as part of the Spanish-BWROG, is currently analyzing the generic products of the US-BWROG AMG in order to generate their specific ones. Also, in this group BWR, the development of tools to simulate accident scenarios beyond core damage will be studied and a training process oriented to warrant the optimum use of new EOP/AMG in accident scenarios will be implemented

  10. Procedure for Application of Software Reliability Growth Models to NPP PSA

    International Nuclear Information System (INIS)

    Son, Han Seong; Kang, Hyun Gook; Chang, Seung Cheol

    2009-01-01

    As the use of software increases at nuclear power plants (NPPs), the necessity for including software reliability and/or safety into the NPP Probabilistic Safety Assessment (PSA) rises. This work proposes an application procedure of software reliability growth models (RGMs), which are most widely used to quantify software reliability, to NPP PSA. Through the proposed procedure, it can be determined if a software reliability growth model can be applied to the NPP PSA before its real application. The procedure proposed in this work is expected to be very helpful for incorporating software into NPP PSA

  11. Methodology and results of the seismic probabilistic safety assessment of Krsko nuclear power plant

    International Nuclear Information System (INIS)

    Vermaut, M.K.; Monette, P.; Campbell, R.D.

    1995-01-01

    A seismic IPEEE (Individual Plant Examination for External Events) was performed for the Krsko plant. The methodology adopted is the seismic PSA (Probabilistic Safety Assessment). The Krsko NPP is located on a medium to high seismicity site. The PSA study described here includes all the steps in the PSA sequence, i.e. reassessment of the site hazard, calculation of plant structures response including soil-structure interaction, seismic plant walkdowns, probabilistic seismic fragility analysis of plant structures and components, and quantification of seismic core damage frequency (CDF). Also relay chatter analysis and soil stability studies were performed. The seismic PSA described here is limited to the analysis of CDF (level I PSA). The subsequent determination and quantification of plant damage states, containment behaviour and radioactive releases to the outside (level 2 PSA) have been performed for the Krsko NPP but are not further described in this paper. The results of the seismic PSA study indicate that, with some upgrades suggested by the PSA team, the seismic induced CDF is comparable to that of most US and Western Europe NPPs. (author)

  12. Human actions treatment in the Juragua NPP pre-operational PSA

    International Nuclear Information System (INIS)

    Ferro Fernandez, R.

    1996-01-01

    The human reliability analysis is an important part of the Probabilistic Safety Analysis (PSA). Because Juragua NPP PSA has been accomplished during construction stage of the plant, no specific operational procedures nor experience for human reliability analysis task taking into account the worlds current methodologies in this field and the actual situation of the plant. This papers describes the approach we followed

  13. Results of level 1 PSA in Trillo 1 NPP

    International Nuclear Information System (INIS)

    Gomez, F.; Lopez, C.

    1998-01-01

    In July 1991, C. N. Trillo I was requested by the Spanish Regulatory Body (CSN) to perform a PSA that should include: - Level 1 PSA at power - Internal flooding analysis - Level 2 PSA including containment capacity analysis. - External event analyses (fires, external flooding, seismic events and other external events) - Risk analysis for off power conditions (shutdown and low power) - Risk analysis due to other sources of radioactivity In 1992 the Project Plan was issued and the PSA team for the performance of Level 1 PSA was established. Before finishing the Project, it was decided to develop a Phase B to take into account some important modifications that had been accomplished in the Plant and that, probably, could affect the results. Level 1 PSA was finished in March 1998. Both the results of the study and the main conclusions derived from the importance, uncertainty and sensibility analysis performed are presented in this paper. These results de not include the internal flooding analysis conclusions and correspond to PSA revision 0 that is currently being evaluated by the Spanish Regulatory Body. (Author)

  14. Estimation of the uncertainties considered in NPP PSA level 2

    International Nuclear Information System (INIS)

    Kalchev, B.; Hristova, R.

    2005-01-01

    The main approaches of the uncertainties analysis are presented. The sources of uncertainties which should be considered in PSA level 2 for WWER reactor such as: uncertainties propagated from level 1 PSA; uncertainties in input parameters; uncertainties related to the modelling of physical phenomena during the accident progression and uncertainties related to the estimation of source terms are defined. The methods for estimation of the uncertainties are also discussed in this paper

  15. Level 1 PSA study of Mochovce unit 1 NPP (SM AA 10 and 08)

    International Nuclear Information System (INIS)

    Cillik, I.

    1997-01-01

    This paper presents genesis of Level 1 PSA project preparation for all operational modes of Mochovce NPP unit 1 including the description of its' main objectives, scope and working method. The PSA study which includes full power (FPSA) as well as shutdown and low power conditions (SPSA) Level 1 PSA has to support the nuclear safety improvements of the unit. They evaluate the basic design and the benefits of all improvements, which were found necessary to be incorporated before the start-up of the unit. The study includes internal events (transients and under-loss of coolant accident, LOCAs), internal hazards as fires and floods and selected external hazards as earthquake, influence of external industry, extreme meteorological conditions and aircraft crash.The PSA (both FPSA and SPSA) models is developed using the RISK SPECTRUM PSA code. (author)

  16. Three steam generator replacement projects in 1995: Consortium Siemens Framatome is well prepared to contribute its experience to the SGR at the Krsko NPP

    International Nuclear Information System (INIS)

    Holz, R.; Clavier, G.

    1996-01-01

    Since the companies Siemens AG and Framatome S.A. joined their experience and efforts in the field of steam generators replacements and formed a consortium in 1991, the following projects were performed in 1995: Ringhals 3, Tihange 3 and Asco 2. Further projects will follow in 1996, i.e., Doel 4 and Asco 2. Currently, this European consortium is bidding for the contract to replace the steam generators at the Krsko NPP and hopes to be awarded in 1996. An overview of the way the Consortium Siemens and Framatome approaches SG replacement projects is given based on the projects performed in 1995. Various aspects of project planning, management, licensing, personnel qualification and techniques used on site will be discussed. (author)

  17. Flooding risk reduction for the ASCO NPP PSA

    International Nuclear Information System (INIS)

    Nos Llorens, V.; Faig Sureda, J.

    1993-01-01

    Developed within the framework of the UTE (INITEC-INYPSA-Empresarios Agrupados), the Probabilistic Safety Analysis (PSA) of the Asco Nuclear Power Plant has served both as a basic tool in reducing the risk of potential internal flooding at the plant, and as a guideline for studying the optimization and feasibility of necessary plant design modifications and changes to procedures. During execution of the work, and in view of the results, a series of improvements were proposed which gave rise to design modification studies. The paper seeks to describe the effect of these modifications on reducing core damage frequency, it also includes a general description of the methodology used. Finally, it compares the results obtained in the context of similar studies performed in other PSAs. (author)

  18. A tsunami PSA methodology and application for NPP site in Korea

    International Nuclear Information System (INIS)

    Kim, Min Kyu; Choi, In-Kil

    2012-01-01

    Highlights: ► A methodology of tsunami PSA was developed in this study. ► Tsunami return period was evaluated by empirical method using historical tsunami record and tidal gauge record. ► Procedure of tsunami fragility analysis was established and target equipments and structures for investigation of tsunami fragility assessment were selected. ► A sample fragility calculation was performed for the equipment in Nuclear Power Plant. ► Accident sequence of tsunami event is developed by according to the tsunami run-up and draw down, and tsunami induced core damage frequency (CDF) is determined. - Abstract: A methodology of tsunami PSA was developed in this study. A tsunami PSA consists of tsunami hazard analysis, tsunami fragility analysis and system analysis. In the case of tsunami hazard analysis, evaluation of tsunami return period is a major task. For the evaluation of tsunami return period, numerical analysis and empirical method can be applied. In this study, tsunami return period was evaluated by empirical method using historical tsunami record and tidal gauge record. For the performing a tsunami fragility analysis, procedure of tsunami fragility analysis was established and target equipments and structures for investigation of tsunami fragility assessment were selected. A sample fragility calculation was performed for the equipment in Nuclear Power Plant. In the case of system analysis, accident sequence of tsunami event is developed by according to the tsunami run-up and draw down, and tsunami induced core damage frequency (CDF) is determined. For the application to the real Nuclear Power Plant, the Ulchin 56 NPP which located in east coast of Korean peninsula was selected. Through this study, whole tsunami PSA working procedure was established and example calculation was performed for one of real Nuclear Power Plant in Korea. But for more accurate tsunami PSA result, there are many researches needed for evaluation of hydrodynamic force, effect of

  19. Development of Krsko Severe Accident Management Database (SAMD)

    International Nuclear Information System (INIS)

    Basic, I.; Kocnar, R.

    1996-01-01

    Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. Krsko Severe Accident Management Database documents the severe accident management activities which are developed in the NPP Krsko, based on the Krsko IPE (Individual Plant Examination) insights and Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidance). (author)

  20. Severe accident management at the Loviisa NPP - Application of integrated ROAAM and PSA level 2

    International Nuclear Information System (INIS)

    Siltanen, S.; Routamo, T.; Tuomisto, H.; Lundstrom, P.

    2007-01-01

    The Risk Oriented Accident Analysis Methodology (ROAAM) was developed for assessment and management of rare, high consequence hazards. The purpose of most ROAAM applications has been to solve major, isolated severe accident issues related to early containment failure such as Mark-I Liner Attack and Direct Containment Heating. In addition to ROAAM in the issue resolution context, the so called Integrated ROAAM approach can be used to provide an overall frame of safety evaluation that allows determination of whether an adequate level of safety has been achieved for a plant. Integrated ROAAM approach brings together quantifications of probabilistic elements based on statistical inference and treatment of deterministic elements based on identification of dominant physics, for severe accident phenomenology, in a well defined and clearly structured way. Fortum, as an owner of the Loviisa NPP, used the Integrated ROAAM approach when developing and implementing a comprehensive severe accident management (SAM) strategy for the Loviisa NPP. The SAM strategy is based on unique features of this VVER-440 plant with ice condenser containment and it includes hardware modifications at the plant, substantial new I and C qualified for severe accident conditions, new SAM guidelines, a SAM Handbook, revision of emergency preparedness organization, and versatile training approaches. It could be argued that the resolution of individual severe accident issues is not sufficient for assessing the overall safety of a nuclear power plant, and thus the ROAAM (in an issue resolution context) is not performing the same function as a PSA study (level 2 included). Actually the Integrated ROAAM approach takes on even a more ambitious task than the PSA, since it determines how a balance can be achieved between accident prevention and mitigation of containment-threatening physical phenomena. Thus it provides a tool for implementing a sound diverse defence-in-depth strategy at a plant. Integrated

  1. Thermal Hydraulic Analysis of Loss of Instrument Air for PSA of Wolsong NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Bong-Jin; Han, Sang-Gu; Kim, Sook-Kwan; Kim, Hyung-Jin [ACT, Daejeon (Korea, Republic of)

    2015-10-15

    This paper describes the analysis of loss of IA using the thermal Hydraulic code for PSA of Wolsong NPP Unit 1. There are 19 events in the event tree of loss of instrument air. Loss of instrument air (IA) can occur due to compressor failures, instrument air line failures, etc. A loss of cooling to the station loads that are served by Recirculated Cooling Water (RCW) system can occur due to the loss of the RCW inventory, the loss of RCW flow and the loss of cooling to the RCW heat exchangers. Instrument air compressors are cooled by RCW. This means that a cross-link exists between RCW and instrument air. A loss of RCW can cause a loss of instrument air. These types of cross-links are either assessed during the development of the event trees or captured during the accident sequence quantification process in PSA. Reactor can be shut down safely after success of SDCS cooldown or secondary heat removal operation in event No. 1, 2, 7, 8, 13 and 14. Residual heat can be removed by ECC and Moderator Heat Sink (MHS) in the situation of pressure tubes contacting calandria tubes after failing SDCS cooldown or secondary heat removal operation.

  2. NPP Krsko Aging Management Program

    International Nuclear Information System (INIS)

    Glaser, B.; Spiler, J.

    2002-01-01

    As a part of Periodic Safety Review Program (PSR) NEK will review and perform some activities related to Equipment Qualification (EQ) and Aging Management Program (AMP). (EQ) and AMP are safety factors, which need to be assessed during PSR. The goal of PSR and AMP is to determine aging effects and give the conclusion whether the plant has been managed to control aging related degradations and that safety margins are maintained. The parallel goal is also to establish AMP for future plant operation and provide basis for possible Life Extension Program. NEK will develop NEK Aging and Life Cycle Management Program, similar by format and content to one determined by License Renewal program. The bases are in 10CFR54, and NEI 95-10 Industry Guidelines for 10 CFR 54 implementation. The process of establishment the AMP is to be done in two steps. The first step is dealing with SSC's (Systems Structures and Components) scoping and screening and identification of TLAA's (Time Limited Aging Analyses). That means, that a database of all SSC's and TLAA's will be created and then evaluated within AMP program. Based on the scope in first phase an evaluation will be performed in step two. NEK will maintain AMP program as a living program that may be also used for Life Extension and Life Cycle Management. This paper will present and describe AMP, scoping and screening process and the results achieved through the first phase of the project.(author)

  3. Krsko periodic safety review project prioritization process

    International Nuclear Information System (INIS)

    Basic, I.; Vrbanic, I.; Spiler, J.; Lambright, J.

    2004-01-01

    Definition of a Krsko Periodic Safety Review (PSR) project is a comprehensive safety review of a plant after last ten years of operation. The objective is a verification by means of a comprehensive review using current methods that Krsko NPP remains safety when judged against current safety objectives and practices and that adequate arrangements are in place to maintain plant safety. This objective encompasses the three main criteria or goals: confirmation that the plant is as safe as originally intended, determination if there are any structures, systems or components that could limit the life of the plant in the foreseeable future, and comparison the plant against modern safety standards and to identify where improvements would be beneficial at justifiable cost. Krsko PSR project is structured in the three phases: Phase 1: Preparation of Detailed 10-years PSR Program, Phase 2: Performing of 10-years PSR Program and preparing of associated documents (2001-2003), and Phase 3: Implementation of the prioritized compensatory measures and modifications (development of associated EEAR, DMP, etc.) after agreement with the SNSA on the design, procedures and time-scales (2004-2008). This paper presents the NEK PSR results of work performed under Phase 2 focused on the ranking of safety issues and prioritization of corrective measures needed for establishing an efficient action plan. Safety issues were identified in Phase 2 during the following review processes: Periodic Safety Review (PSR) task; Krsko NPP Regulatory Compliance Program (RCP) review; Westinghouse Owner Group (WOG) catalog items screening/review; SNSA recommendations (including IAEA RAMP mission suggestions/recommendations).(author)

  4. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code

    International Nuclear Information System (INIS)

    Martin, L.; Saenz Tejada, P.

    1993-01-01

    When performing a level 1 Probabilistic Safety Analysis (PSA) on a standard power plant, in order to model plant response to the potential occurrence of the various initiating events postulated in a PSA, reference documentation applicable to the type of plant in question is frequently consulted. Because of the specific design characteristics of the Jose Cabrera NPP, most of the reference documentation for the W-PWR-type power plants is not applicable to this plant. To fill in these gaps in the documentation and to construct the most realistic model of plant behaviour possible, assistance was sought from Union Fenosa by way of infrastructure, capabilities and thermohydraulic experience of the Nuclear Engineering and Fuel Group, and especially the use of calculations performed with the RELAP5/ MOD2 code. This paper will provide an overview of the general assistance rendered to the PSA by the technical experts in thermohydraulics, the calculations performed with RELAP5/MOD2 and the influence all of this has had on the development, quality and results of the Jose Cabrera NPP level 1 PSA Project. (author)

  5. Development of Krsko Severe Accident Management Guidance (SAMG)

    International Nuclear Information System (INIS)

    Cizel, F.

    1999-01-01

    In this lecture development of severe accident management guidances for Krsko NPP are described. Author deals with the history of severe accident management and implementation of issues (validation, review of E-plan and other aspects SAMG implementation guidance). Methods of Westinghouse owners group, of Combustion Engineering owners group, of Babcock and Wilcox owners group, of the BWR owners group, as well as application of US SAMG methodology in Europe and elsewhere are reviewed

  6. Overview of IAEA activities on PSA applications and IAEA references

    International Nuclear Information System (INIS)

    Gomez Cobo, A.

    1997-01-01

    The presentation describes the IAEA activities in the following areas: requirements and applications of living PSA, PSA applications and tools to improve NPP safety, use of PSA to optimize maintenance, use of PSA for regulatory decision making. 22 refs

  7. Trends in simulation and the Krsko full scope simulator

    International Nuclear Information System (INIS)

    Boire, R.; Chatlani, M.

    1998-01-01

    The nuclear power plant simulation industry is a fast-paced industry yielding continual development as a result of innovations in technology and customer requirements. This paper will discuss the current trends in simulator requirements, the status of simulation technology and the expected future developments, particularly in the context of the NPP Krsko full scope simulator. CAE Electronics has been awarded the contract for the design, construction, integration, testing and commissioning of the NPP Krsko full scope simulator (KFSS) by Nuklearna elektrarna Krsko (NEK). KFSS, as an integral part of the NPP Krsko Modernization plan, has been the subject of an extensive procurement process. KFSS will also take into account the steam generator replacement and plant uprate projects which will be delivered to provide initial training in the modernized plant configuration. As a result, the completed KFSS will meet NEK's goals for reliable training in safe plant operation as well as the licensing requirements of the Slovenian Nuclear Safety Administration. KFSS will be a state-of-the-art facility featuring high fidelity process and control models, proven technology and superior maintainability that will push the envelope of traditional simulator uses. In addition to serving its role as a high quality training vehicle, KFSS will be used for engineering purposes including procedure development and validation, optimization of plant operation and study and validation of plant modifications. KFSS models will be built for the most part with CAE's ROSE TM toolset. ROSE, is a component-based, visual programming environment for the creation, testing, integration and management of simulator models and supporting virtual panels. The NSSS will be simulated using the ANTHEM two-phase drift flux model, while be simulated using the COMET two-group, three-dimensional model. Software design and testing will be supported by an extensive series of quality procedures throughout the software

  8. Loss of Normal Feedwater analyses for Krsko Full Scope Simulator verification

    International Nuclear Information System (INIS)

    Parzer, I.; Prosek, A.; Hrvatin, S.

    2000-01-01

    The purpose of these analyses was to perform calculations of a Loss of Normal (Main) Feedwater transient for Krsko NPP. The results of calculations were used for the verification of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. To perform the thermal-hydraulic analyses, the RELAP5/MOD2 computer code and the NPP Krsko input card deck were used. In the presented paper two scenarios have been analyzed. Both of them started with a loss of normal feedwater event. Thus, a reduction or an interruption of the heat removal by the secondary system occurred. The first scenario assumed that auxiliary feedwater was available during the transient, while in the second scenario both normal and auxiliary feedwater were unavailable. The results showed that with auxiliary feedwater pumps unavailable additional operator actions would be needed to prevent overheating of the core. (author)

  9. Status of the PSA use in the Czech regulatory process

    International Nuclear Information System (INIS)

    Dusek, J.

    1994-01-01

    A review of previous probabilistic safety assessment (PSA) activities initiated by regulatory body and preparation of the preliminary PSA study and final PSA study (released in January 1994) for the nuclear power plant Dukovany with WWER-440 type 213 reactor is described. A brief information about the NPP Temelin (with WWER-1000) PSA Study, shutdown and PSA risk monitor current activities for the NPP Dukovany, next PSA activities in 1994 and about planned PSA activities in future is attached. (author). 21 refs

  10. Intranet in Nuclear Power Plant Krsko - IntraNEK

    International Nuclear Information System (INIS)

    Kocnar, R.; Krajnc, B.; Spiler, J.

    1998-01-01

    Intranet servers enable real business functionality such as publishing information, processing data and database applications, and collaboration among employees, vendors, and customers. Across all industries, intranet is rapidly reshaping company-wide communication, productivity, and innovation - and saving significant time and money in the process. IntraNEK can provide information in a way that is immediate, cost-effective, easy to use and rich in format. First Home Page of intranet in NPP Krsko was published early 1997. Until now, we published USAR-Updated Safety Analysis Report, Flow diagrams (the most frequently used series of drawings), With normal option which inernet/intranet (prompt access to any part of USAR and Technical Specification, prompt access and zooming capabilities to flow diagrams, full text searching common/advance...) and Safety Screenings Data Base. Safety Screenings Data Base is typical example of connection of Data Base (in our case popular and well know ACCESS) and PC based WWW server. Those first applications (on a one-to-many basis) show how IntraNEK brings an immediate payback to NPP Krsko, eliminating the costs of producing, printing, and shipping necessary information and reducing bulky, easily outdated paper-based documents. USAR and drawings are published on HP-UX platform (Netscape WWW Server) and Safety Screenings Data Base is published on PC platform (Microsoft WWW Server) and show us how hardware become transparent and how intranet reduce cost in client software, regardless of their choice of hardware platform. We do believe that this new technological solution should also be used in the nuclear industry, since intranet could meet and exceed all required QA and QC standards and regulations. In the paper we are going to present the examples of future applications and we will described the necessary preconditions which have to be fulfilled before IntraNET could be used as an official tool and source of information in the design

  11. Implementation of Industry Experience at Nuclear Power Plant Krsko

    International Nuclear Information System (INIS)

    Heruc, Z.; Kavsek, D.

    2002-01-01

    Being a standalone comparatively small unit NPP Krsko has adopted a business philosophy to incorporate industry experience into its daily operations. The continuos and safe operation of the unit is supported through feedback from other utilities (lessons learned) and equipment vendors and manufacturers. A permanent proactive approach in monitoring the international nuclear technology practices, standards changes and improvements, and upon feasibility review, introducing them into processes and equipment upgrades, is applied. As a member of the most important international integrations, NPP Krsko has benefited from the opportunity of sharing its experience with others (World Association of Nuclear Operators -WANO, Institute of Nuclear Power Operations - INPO, International Atomic Energy Agency - IAEA, Nuclear Operations Maintenance Information Service - NOMIS, Nuclear Maintenance Experience Exchange - NUMEX, Electric Power Research Institute - EPRI, Westinghouse Owners Group - WOG, etc.). Voluntary activities and good practices related to safety are achieved by international missions (IAEA Assessment of Safety Significant Events Team - ASSET, IAEA Operational Safety Review Team - OSART, WANO Peer Review, International Commission for Independent Safety Analysis - ICISA) and operating experience exchange programs through international organizations. These missions are promoting the highest levels of excellence in nuclear power plant operation, maintenance and support. With time, the practices described in this paper presented themselves as most contributing to safe and reliable operation of our power plant and at the same time supporting cost optimization making it a viable and reliable source of electrical energy in the more and more deregulated market. (author)

  12. Quality PSA for PSA applications

    International Nuclear Information System (INIS)

    Carska, K.; Rybar, J.

    2012-03-01

    The safety guideline defines with more precision Nuclear Regulatory Authority of the Slovak Republic requirements of the quality of probabilistic safety assessment (PSA) for PSA application. Term of quality of PSA is explained in detail. Procedure for determining the quality of PSA is provided. The categorization of PSA study according the quality of PSA is suggested. A comprehensive list of PSA applications for nuclear facilities is provided. What technical features of a PSA should be satisfied to support the PSA applications of interest is stated. (authors)

  13. Self-Assessment at Krsko Nuclear Power Plant

    International Nuclear Information System (INIS)

    Strucic, M.; Kavsek, D.; Novak, J.; Dudas, M.

    2006-01-01

    Self-assessment program in NPP Krsko is based on plant effort to identify areas for improvements, as well as strengths in various processes. The highest level tool of that program is Inter-disciplinary Self-assessment. Extensive experience in methodology from many Peer Reviews worldwide, where NPP Krsko personnel were involved, was essential contributor for successful development and implementation of Inter-disciplinary Self-assessment. Every Inter-disciplinary Self-assessment, performed by experienced NEK people, results in highly efficient and constructive action plan. It is achieved by professional approach and positive attitude of team leader and members. Typical team composition includes members from different NEK departments including their managers. They are experienced in area being assessed, as well as in Cause analysis techniques. People involved in previous Internal or Inter-discipline Self-assessments and international peer reviews are indispensable part of the team and usually team leader is one of them. Inter-disciplinary Self-assessments are planned well in advance and are approved by NEK management board. NEK directors are also involved through sponsorship. Often, they are counterparts in the interviews sessions of assessment. Methodology of carrying out Self-assessment is developed using WANO Peer reviews experience and techniques. Areas for assessment are mostly identified through Corrective action or trending processes, Internal self-assessments or Performance Indicators. Field observations, interviews with workers in the field and their superiors are reason for frequent team meetings. That process is often iterative and results in clear and precise observation reports which are separately analyzed and at the end confirmed by owner of the process. Based on analysis described in observation reports, team defines areas where generic problems are found. Team members are dedicated for particular areas, usually where they are more educated and

  14. SB LOCA thermal-hydraulic analyses for Krsko Full Scope Simulator validation

    International Nuclear Information System (INIS)

    Prosek, A.; Parzer, I.

    2005-01-01

    The Krsko nuclear power plant (NPP), which is a two-loop pressurized water reactor, Westinghouse type, before modernization in 2000 obtained plant specific full scope simulator. The purpose of the presented analyses was to perform Small Break Loss of Coolant Accident (SBLOCA) reference calculations for KFSS validation in 2004. In addition, the thermal-hydraulic response of the reactor coolant system (RCS) was studied in detail. For the thermal-hydraulic analysis the RELAP5/MOD3.3 code and input model delivered from Krsko NPP were used. The RELAP5 calculated reference results showed that the plant system response to breaks with small break area is slower compared to breaks with larger break area. The comparison of the KFSS data with calculated results suggest that the simulator validation testing in the year 2004 for this kind of accident was successful. Nevertheless, when comparing the physical phenomena and processes, the RELAP5/MOD3.3 predicted smaller core uncovery compared to the KFSS measurement. One reason is different core cycles. Finally, this finding suggests that even for simulator reference calculations the quantification of model uncertainties would be useful. (author)

  15. Living PSA

    International Nuclear Information System (INIS)

    Evans, M.G.K.

    1997-01-01

    The aim of this presentation is to gain an understanding of the requirements for a PSA to be considered a Living PSA. The presentation is divided into the following topics: Definition; Planning/Documentation; Task Performance; Maintenance; Management. 4 figs

  16. SGTR analyses for Krsko full scope simulator verification

    International Nuclear Information System (INIS)

    Parzer, I.; Koncar, B.; Prosek, A.

    2000-01-01

    The purpose of this analysis was to perform SGTR accident calculations for simulator verification of RCS thermal-hydraulic response. For the thermal-hydraulic analyses the RELAP5/MOD2 code and the input card deck for NPP Krsko was used. The analyses for SGTR accident consist of spectrum of rupture sizes, corresponding to equivalents of 0.5, 1, 2 or 3 double-ended (d-e) steam generator (SG) tubes ruptured. The rupture was supposed to be located at the hot side of the U-tubes, near the tube sheet. Sensitivity study results show the influence of rupture size on the course of the SGTR transient. Additionally, a sensitivity analysis of rupture location was performed, for the case of 1 d-e SGTR (1 SGTR). Two cases with different rupture locations were considered: the base case with the rupture located at the hot side of the U-tubes, near the tube sheet and the case where the rupture was assumed to be located at the top of U-tubes bend. (author)

  17. Radiation protection programme at Krsko nuclear power plant

    International Nuclear Information System (INIS)

    Breznik, B.

    1996-01-01

    Krsko NPP, a Westinghouse two-loop PWR of 632 M We power, is in commercial operation since 1982. Reduction of radioactive releases to the environment and the reduction of doses to workers is the basic goal in the plant radiological protection. The radiation protection programme is established to ensure that the radiation exposures to workers and members of the public are minimized according to the As Low As Reasonably Achievable approach and controlled in accordance with international safety standards and Slovenian regulations. The basis for the operational and technical measures has been provided according to the industrial good practice. The effluent control is based on the Standard Radioactive Effluent Technical Specifications, and environmental surveillance is established according to the programme defined by the regulations. The dose constraints and performance indicators are used to assure the effectiveness of the radiation protection programme and provide a convenient follow-up tool. The monitoring programme results of each year show that there is no measurable dose to the public due to radioactive releases. The commitment to the dose burden of any member of a critical group is assessed to be below the dose constraint. Individual and collective doses of the workers are within a range typical for the PWRs of a similar type. (author)

  18. Emergency preparedness exercise ''Posavje 82'' in support of the Krsko Nuclear Power Plant, Krsko, Yugoslavia

    International Nuclear Information System (INIS)

    Collins, H.E.; Emmerson, B.W.

    1983-06-01

    In October 1982, the Yugoslavian Government requested the Agency's assistance in observing and evaluating an emergency preparedness exercise (code named ''POSAVJE 82'') on 5 and 6 November 1982, to test emergency plans and arrangements supportive of the KRSKO Nuclear Power Plant. The Krsko Nuclear Power Plant is a single unit pressurized water reactor of United States (Westinghouse) design rated at 664 MWe (Gross) and is located at Krsko, Socialist Republic of Slovenia, Yugoslavia. This assistance was provided by sending a Special Assistance Mission team of experts under the general provisions of the Agency's circular letter SC/651-3 of 7 April 1981 to Member States which offered such assistance upon request. This mission was a follow-up to a previous mission requested by the Yugoslavian Government which was conducted 24 June - 1 July 1981. At that time, the mission consisted of examining the then existing arrangements for emergency planning in support of the KRSKO Nuclear Power Plant at the National, Republic, local and nuclear power plant levels and discussing with Yugoslavian authorities criteria for emergency plan development and improvement. As a result of this 1981 mission, a ''Report to the Goverment of Yugoslavia'' (IAEA TA Report 1827 of 17 September 1981) was transmitted to the Yugoslavian Government. This report set forth a number of recommendations for improving and further developing the various emergency plans and arrangements for the KRSKO Nuclear Power Plant. A summary of the major recommendations contained in the report is given in Section 2.2. The entire report is listed as Reference 1 of Annex A

  19. Modernization of the Nuclear Power Plant Krsko with new steam generators

    International Nuclear Information System (INIS)

    Holz, R.; Stach, U.; Gloaguen, C.

    2000-01-01

    The contract for the replacement of two steam generators at NPP Krsko was awarded in February 1998 to the Consortium SIEMENS AG FRAMATOME S.A.. The time frame for the replacement outage was scheduled from April to June 2000. The replacement itself started with the plant shut down on 15 th of April 2000 and the plant was back on line on 15 th of June, so that after an intensive engineering period of more than two years the plant was off line only 62 days, as scheduled. This document deals with the various aspects of the replacement phase itself and the techniques used. During the last years conference the engineering and licensing phase of the project have been presented. (author)

  20. Modernization programme at Dukovany NPP

    International Nuclear Information System (INIS)

    Trnka, M.

    2000-01-01

    The main goal of each NPP is to produce electricity safely, economically and without influence to environment. For Dukovany NPP it means to upgrade all documentation and perform the Equipment Upgrading Programme. All these activities are time and money consuming and therefore the determination of priority of all items was necessary. In the presentation there are mentioned some important changes in documentation, results of PSA studies and reason for Equipment Upgrading Programme performance. It was selected the most important item from the list of Equipment Upgrading Programme the I and C upgrading. Management has decided that Dukovany NPP will become among the best NPPs with WWER type of reactor. It seems this decision is the best way how to extend lifetime of the NPP. (author)

  1. PSA methodology

    International Nuclear Information System (INIS)

    Magne, L.

    1996-01-01

    The purpose of this text is first to ask a certain number of questions on the methods related to PSAs. Notably we will explore the positioning of the French methodological approach - as applied in the EPS 1300 1 and EPS 900 2 PSAs - compared to other approaches (Part One). This reflection leads to more general reflection: what contents, for what PSA? This is why, in Part Two, we will try to offer a framework for definition of the criteria a PSA should satisfy to meet the clearly identified needs. Finally, Part Three will quickly summarize the questions approached in the first two parts, as an introduction to the debate. 15 refs

  2. New low pressure (LP) turbines for NE Krsko

    International Nuclear Information System (INIS)

    Nemcic, K.; Novsak, M.

    2004-01-01

    During the evaluation of possible future maintenance strategies on steam turbine in very short period of time, engineering decision was made by NE Krsko in agreement with Owners to replace the existing two Low Pressure (LP) Turbines with new upgrading LP Turbines. This decision is presented with review of the various steam turbine problems as: SCC on turbine discs; blades cracking; erosion-corrosion with comparison of various maintenance options and efforts undertaken by the NE Krsko to improve performance of the original low pressure turbines. This paper presents the NEK approach to solve the possible future problems with steam turbine operation in NE Krsko as pro-active engineering and maintenance activities on the steam turbine. This paper also presents improvements involving retrofits, confined to the main steam turbine path, with major differences between original and new LP Turbines as beneficial replacement because of turbine MWe upgrading and return capital expenditures.(author)

  3. Yak experience at Nuclear Power Plant Krsko

    International Nuclear Information System (INIS)

    Mandic, D.

    2000-01-01

    In Sept. 1998, Nuclear Power Plant Krsko started Y2K (Year 2000) Readiness Assessment Program and implementation of the Y2K-NEK Project (NEK Nuklearna Elektrana Kriko). Y2K-NEK Project and the term N EK Year 2000 Readiness Assessment Program'' applies to software, or software based system or interface, whose failure due to the Y2K problem would prevent the performance of the safety function of a structure, system, or component. This project also applies to any software, or software based system or interface, whose failure due to the Y2K problem would degrade, impair, or prevent operability of the nuclear facility. It is intended to supplement and use existing NEK procedures used for software quality control, configuration management and problem reporting. The main guideline and method definition documents for Y2K-NEK Project were: NEI/NUSMG 97-07: Nuclear Utility Year 2000 Readiness (October 1997), and NEI/NUSMG 98-07; Nuclear Utility Year 2000 Readiness Contingency Planning (Aug. 1998). This paper presents project Y2K implementation experience and post Y2K transition analysis of the plant hardware/software systems behavior compared to the expected systems behavior and expected-planned scenarios based on the results of the Y2K Readiness Assessment, implemented remediations and Y2K Contingency Planning. (author)

  4. Report 1: Guidance document on practices to model and implement Seismic hazards in extended PSA. Volume 2 (implementation in Level 1 PSA)

    International Nuclear Information System (INIS)

    Prochaska, J.; Halada, P.; Pellissetti, M.; Kumar, Manorma

    2016-01-01

    The objective of this report is to provide guidance for the implementation of seismic hazards in extended L1 PSA. This report is a deliverable of the ASAMPSA-E work package 22 (WP22) - 'How to introduce hazards in L1 PSA and all possibilities of events combinations' - which aims to promote exchanges and to identify some good practices on the implementation of seismic events in L1 PSA, having as a perspective the development of extended PSA from an existing (internal events) L1 PSA (event trees). The following topics are addressed: 1) Impact on the SSCs modelled in L1 PSA event trees 2) Impact on Human Reliability Assessment modelling in L1 PSA 3) Site impact modelling in L1 PSA event trees 4) Link between external initiating events of PSA and NPP design basis conditions. (authors)

  5. Future needs in radiation protection training for NPP workers of Slovenia

    International Nuclear Information System (INIS)

    Kozelj, M.; Bogovic, T.

    1999-01-01

    Short review of history of radiation protection training for NPP workers in Slovenia and legal requirements regarding this field are presented. Courses developed in co-operation between Milan Copic Nuclear Training Centre and Krsko Nuclear Power Plant are briefly described and their implementation presented. Using available data we have predicted probable number of courses and participants in forthcoming years. Some results from inquiry on courses for regularly exposed workers are presented, enabling us to modify courses according to participants' needs.(author)

  6. Lessons learned form IRSN review of Flamanville 3 Level PSA

    International Nuclear Information System (INIS)

    Georgescu, G.; Corenwinder, F.

    2012-01-01

    In the frame of the construction and licensing of Flamanville 3 NPP the PSA (Probabilistic Safety Assessment)plays an important role for the EPR Project assessment. The PSA was used for early design verification of EPR Reactor, several design improvement being defined based on these PSA insights and following the discussions with the French and German safety authorities. IRSN, as the French Safety Authority (ASN) technical support organization, performs the review of the PSA developed by the plant operator (EDF). The paper presents the main issues regarding the using of 'design PSA', identified by IRSN following the review of the internal events Level 1 PSA transmitted by EDF in the frame of the anticipated instruction of the application for operating license of the Flamanville 3 reactor. (authors)

  7. Probabilistic analyses at Kozloduy NPP - status and plans

    International Nuclear Information System (INIS)

    Papazov, V.

    2003-01-01

    A brief review of the Kozloduy NPP units is given. The main activities and results from Units 1-4 Modernization Programs like: Jet Vortex Condenser implementation at Units 3 and 4; total update of the Units Safety Analysis Reports; implementation of symptom-based emergency operating procedures; more than 35 times reduction of confinement leakage area are listed. The Units 5 and 6 Modernization program as an unique project looking for a reasonable balance between innovation and recognized engineering practices is presented in details. The present status of PSA analyses is described as follows: for Unit 1-4: 4 revisions of PSA level 1 are issued; project for 'Living PSA' implementation has been started; second update of the shutdown PSA is made; PSA results are used for operators training program development; PSA results are used for the development and assessment of the modernization programs; Precursor event analysis is under development; PSA level 2 is started for Unit 5 and 6; PSA level 1 update is going on, next update will be after the completion of the program; Shutdown PSA is under development; PSA level 2 has been started; 'Living PSA' is planned. At the end, some results from PSA analyses are illustrated

  8. Quality assurance in the Juragua Nuclear Power Plant preoperational PSA

    International Nuclear Information System (INIS)

    Valhuerdi Debesa, C.

    1996-01-01

    Quality Assurance (QA) is nowadays an important requirement for the competence of any production or service, making possible to get the desired quality at the lowest cost In the case of PSA, which are multidisciplinary, very detailed and complex analysis, with many interfaces between analyst tasks, QA plays an important role as a tool for the analytical process management, and it is recognized as one of the PSA issues which require additional development In this paper the QA system developed for the Juragua NPP preoperational PSA, its antecedents and the experiences of its application are described

  9. NPP decontamination

    International Nuclear Information System (INIS)

    Sedov, V.M.; Senin, E.V.; Nesterenko, A.P.; Zakharova, E.V.

    1988-01-01

    Decontamination methods for NPP with LWGR and WWER reactors are considered. Circuits iof NPP with LWGR are decontaminated with solutions on the basis of oxalic acid and hydrogen peroxide with addition of nitrate-ions, whereas WWER with complexones followed treatment with strong oxidizers. As a result of decontamination γ-background of the equipment decreases 5-100 fold. The instruments and room surfaces are decontaminated using vapour-ejecting sprayers, hydromonitors and special baths with solutions on the basis of mineral acids and complexones with various additions, which can be used for stainless and carbon steels, coatings of epoxide enamels and plastifiers. Methods of dry decontamination using easily peeled off coatings on the basis of polyvinyl alcohol and other materials lately are widely used for NPP room decontamination. Small metallic parts of surface are decontaminated by electrochemical method. Different mixtures of mineral and organic acids, as well as caustic potash and boric acid, are used as electrolytes

  10. Putting PSA to work

    International Nuclear Information System (INIS)

    Gubler, R.; Gomez-Cobo, A.

    1998-01-01

    The IAEA has, during the last three years, been working intensively on PSA applications. The draft TECDOC prepared during these activities, ''PSA Applications'' is summarized in this paper. Actual events at nuclear facilities provide an important basis to compare PSAs with reality. PSA based operational event analysis therefore can be used to evaluate the importance of operational events from a risk perspective but also can contribute to validating and enhancing PSAs and to continuously check whether or not the PSA models are adequate, appropriate and complete. The work of the IAEA in this area is therefore summarized as well. In a companion paper, titled ''Towards a credible PSA fit for applications'', two specific aspects regarding the quality of the PSA to be used are discussed in detail, namely the Living PSA concept, which ensures that the PSA reflects actual design and operational features and Quality Assurance for PSA. (author)

  11. Improving PSA quality of KSNP PSA model

    International Nuclear Information System (INIS)

    Yang, Joon Eon; Ha, Jae Joo

    2004-01-01

    In the RIR (Risk-informed Regulation), PSA (Probabilistic Safety Assessment) plays a major role because it provides overall risk insights for the regulatory body and utility. Therefore, the scope, the level of details and the technical adequacy of PSA, i.e. the quality of PSA is to be ensured for the successful RIR. To improve the quality of Korean PSA, we evaluate the quality of the KSNP (Korean Standard Nuclear Power Plant) internal full-power PSA model based on the 'ASME PRA Standard' and the 'NEI PRA Peer Review Process Guidance.' As a working group, PSA experts of the regulatory body and industry also participated in the evaluation process. It is finally judged that the overall quality of the KSNP PSA is between the ASME Standard Capability Category I and II. We also derive some items to be improved for upgrading the quality of the PSA up to the ASME Standard Capability Category II. In this paper, we show the result of quality evaluation, and the activities to improve the quality of the KSNP PSA model

  12. Radioactive control of Krsko Nuclear Power Plant in the year 1995

    International Nuclear Information System (INIS)

    Lulic, S.; Miklavzic, U.; Franic, Z.; Kanduc, M.

    1996-01-01

    Regular Krsko Nuclear Power Plant (NPPK) radioactivity control comprises the supervisions of the inventory of liquid and gaseous emissions at the source, and the independent supervisions of the input of radionuclides into larger environment. The controlled environment area consist primarily of a 12 kilometers large circle around the object, where the largest values of immission could be expected, and where possible changes in the Sava river and the underground waters could first be noticed. The circle has been enlarged upon the territory of the Republic of Croatia (RC) from Jesenice on Dolenjsko until Podsused (30 km of air-line distance). As reference points relevant for the readiness in the case of accident, especially for detection of iodine and aerosol air transport, the program comprises also measuring points in the RC at larger distances (from 14 to 27 km) in the direction of Zagreb its larger western surroundings (passive Thermoluminescent (TL) dosimeters in each 42 km long). Continuous of control of emission is performed by the radiological service of (KNPP) by routine procedures, supplemented by adequate measurements from other authorized institutions. Summarized results of radioactive measurements for man-made and natural radionuclides are presented for different transfer media and exposure pathways in the form of assessed effective doses. Conservatively estimated dose burdens received by a member of the reference (critical) population group as the result of NPP emissions amount to a value of the committed effective dose equivalent smaller than 20 μSv/year. This value represents less than 1% of the annual dose received on average from natural and artificial sources by a member of the general public in the normal environment. The yearly doses from natural radioactivity, global contamination , non-nuclear industries and hospitals are also estimated from the measured data in some media. (author)

  13. Radioactive Control of Krsko Nuclear Power Plant Environment in the Year 1997

    International Nuclear Information System (INIS)

    Lulic, S.; Miklavzic, U.; Franic, Z.; Kanduc, M.

    1998-01-01

    Regular Krsko Nuclear Power Plant (NPPK) radioactivity control comprises the supervisions of the inventory of liquid and gaseous emissions at the source, and the independent supervisions of the input of radionuclides into larger environment (imisson). The controlled environment area consists primarily of a 12 kilometers large circle around the object, where the largest values of imission could be expected, and where possible changes in the Sava river and the underground waters could first be noticed. The circle has been enlarged upon the territory of the Republic of Croatia (RC) from Jesenice on Dolenjsko until Podsused (30 km of air - line distance). As reference points relevant for the readiness in the case of accident, especially for detection of iodine and aerosol air transport, the program comprises also measuring points in the RC at larger distances (from 14 to 27 km) in the direction of Zagreb its larger western surroundings (passive Thermoluminescent (TL) dosimeters in the each 42 km long). Continuous control of emission is performed by the radiological service of KNPP by routine procedures, supplemented by adequate measurements from other authorized institutions (intercomparisons, parallel measurements of representative and other samples). Summarised results of radioactive measurements for man-made and natural radionuclides are presented for different transfer media and exposure pathways in the form of assessed effective doses. Conservatively estimated dose burdens received by a member of the reference (critical) population group as the result of NPP emissions amount to a value of the committed effective dose equivalent smaller than 20 μSv/year. This value represents less than 1 % of the annual dose received on average from natural and artificial sources by a member of the general public in the normal environment. The yearly doses from natural radioactivity, global contamination (Chernobyl, atmospheric nuclear explosions), non-nuclear industries and

  14. Model of fire spread around Krsko Power Plant

    International Nuclear Information System (INIS)

    Vidmar, P.; Petelin, S.

    2001-01-01

    The idea behind the article is how to define fire behaviour. The work is based on an analytical study of fire origin, its development and spread. The study is based on thermodynamics, heat transfer and the study of hydrodynamics and combustion, which represent the bases of fire dynamics. The article shows a practical example of a leak of hazardous chemicals from a tank. Because of the inflammability of the fluid, fire may start. We have tried to model fire propagation around the Krsko power plant, and show what extended surrounding area could be affected. The model also considers weather conditions, in particular wind speed and direction. For this purpose we have used the computer code Safer Trace, which is based on zone models. That means that phenomena are described by physical and empirical equations. An imperfection in this computer code is the inability to consider ground topology. However in the case of the Krsko power plant, topology is not so important, as the plan is located in a relatively flat region. Mathematical models are presented. They show the propagation of hazardous fluid in the environment considering meteorological data. The work also shows which data are essential to define fire spread and shows the main considerations of Probabilistic Safety Assessment for external fire event.(author)

  15. 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

  16. Comparative evaluation of PSA-Density, percent free PSA and total PSA

    OpenAIRE

    Ströbel, Greta

    2010-01-01

    BACKGROUND The objective of this study was to evaluate the prostate specific antigen (PSA) density (PSAD) (the quotient of PSA and prostate volume) compared with the percent free PSA (%fPSA) and total PSA (tPSA) in different total PSA (tPSA) ranges from 2 ng/mL to 20 ng/mL. Possible cut-off levels depending on the tPSA should be established. METHODS In total, 1809 men with no pretreatment of the prostate were enrolled between 1996 and 2004. Total and free PSA were measured with t...

  17. PSA, PSA derivatives, proPSA and prostate health index in the diagnosis of prostate cancer

    OpenAIRE

    Ayyıldız, Sema Nur; Ayyıldız, Ali

    2014-01-01

    Currently, prostate- specific antigen (PSA) is the most common oncological marker used for prostate cancer screening. However, high levels of PSA in benign prostatic hyperplasia and prostatitis decrease the specificity of PSA as a cancer marker. To increase the specificity of PSA, PSA derivatives and PSA kinetics have been used. However, these new techniques were not able to increase the diagnostic specificity for prostate cancer. Therefore, the search for new molecules and derivatives of PSA...

  18. Probabilistic safety analysis forecast for Trillo 1 NPP

    International Nuclear Information System (INIS)

    Carretero Fernandino, J.A.; Martin Alvarez, L.; gomez, F.; Cuallado, G.

    1995-01-01

    The performance of Probabilistic Safety Analyses (PSA) at Trillo 1 NPP is facing a number of challenges, unprecedented in previous PSAs carried out in Spain, due to the particular design characteristics of the plant. On account of this, it has been necessary to implemented specific approaches and methodological alternatives to perform a PSA which, while maintaining detail level and requirements in line with PSAs carried out previously in Spain, offers a solution technically adapted to the characteristics of the SIEMENS-KWU design as opposed to other Spanish reactors with a basic Westinghouse-General Electric design, which are based on standard US design. The purpose of this paper is to describe the most significant characteristics of the PSA at Trillo 1 NPP and the methodology used to date, taking into account current project progress

  19. Some Aspects of Process Computers Configuration Control in Nuclear Power Plant Krsko - Process Computer Signal Configuration Database (PCSCDB)

    International Nuclear Information System (INIS)

    Mandic, D.; Kocnar, R.; Sucic, B.

    2002-01-01

    During the operation of NEK and other nuclear power plants it has been recognized that certain issues related to the usage of digital equipment and associated software in NPP technological process protection, control and monitoring, is not adequately addressed in the existing programs and procedures. The term and the process of Process Computers Configuration Control joins three 10CFR50 Appendix B quality requirements of Process Computers application in NPP: Design Control, Document Control and Identification and Control of Materials, Parts and Components. This paper describes Process Computer Signal Configuration Database (PCSCDB), that was developed and implemented in order to resolve some aspects of Process Computer Configuration Control related to the signals or database points that exist in the life cycle of different Process Computer Systems (PCS) in Nuclear Power Plant Krsko. PCSCDB is controlled, master database, related to the definition and description of the configurable database points associated with all Process Computer Systems in NEK. PCSCDB holds attributes related to the configuration of addressable and configurable real time database points and attributes related to the signal life cycle references and history data such as: Input/Output signals, Manually Input database points, Program constants, Setpoints, Calculated (by application program or SCADA calculation tools) database points, Control Flags (example: enable / disable certain program feature) Signal acquisition design references to the DCM (Document Control Module Application software for document control within Management Information System - MIS) and MECL (Master Equipment and Component List MIS Application software for identification and configuration control of plant equipment and components) Usage of particular database point in particular application software packages, and in the man-machine interface features (display mimics, printout reports, ...) Signals history (EEAR Engineering

  20. On-line maintenance at Cofrentes NPP

    International Nuclear Information System (INIS)

    Suarez, J.; Moreno, M.

    2000-01-01

    Cofrentes NPP (CNPP) has developed a Level 1 PSA with the following scope: analysis of internal events, with the reactor initially operating at power; internal and external flooding risk analysis; internal fire risk analysis; reliability analysis of the containment heat removal and containment isolation systems. Level 1 CNPP-PSA results reveal that total core damage frequency in CNPP is less than other similar BWR/6 plants. The CNPP-PSA related activities and applications being carried out currently are: prioritization of motor operated valves related to GL-89/10; complementary analysis for exemption to some 10CFR50 App. J requirements; Q-List grading; risk-informed IST program; reliability-centered maintenance; maintenance rule support; on-line maintenance support; off-line risk-monitor development; PSA applicability to the 10CFR50 App. R requirements, analysis of the frequency of miss-oriented fuel bundle event, adjusting of MAAP 3.0B, revision 10, on VAX and PC; acquisition of MAAP 4; development of Level1/Level2-PSA interface; seismic site categorization for the IPEEE; etc. (author)

  1. Decommissioning of NPP A-1

    International Nuclear Information System (INIS)

    Anon

    2009-01-01

    In this presentation the Operation history of A1 NPP, Project 'Decommissioning of A1 NPP' - I stage, Project 'Decommissioning of A1 NPP ' - II stage and Next stages of Project 'Decommissioning of A1 NPP ' are discussed.

  2. Living PSA program for VVER 440/213 in the Czech Republic

    International Nuclear Information System (INIS)

    Husak, S.; Patrik, M.

    2000-01-01

    The paper presents an overview of a Living PSA concept in the Czech Republic for the VVER 440/213 NPP Dukovany unit. The first step of PSA program was a Level 1 basic study for Unit No. 1 which was completed in 1995. The main objective of the study was to determine the risk level of full power operation and its contributors as well as to reveal the weak points of the plant. Living PSA program for a Level 1 study has been afterwards established as a framework for all activities related to risk assessment and risk based decision-making support in NPP Dukovany. The basic parts of the project are: a management of PSA models and studies to implement design and procedures, modifications or new data inputs from data collection; continuous improvement based of new analyses, experiments or more detailed models; an extensions of the scope (external events, all plant operating modes, other sources of radioactive releases). The Living PSA program in NPP Dukovany provides basis for three kinds of PSA activities: risk assessment applications, risk monitoring and risk assessment of operational. (author)

  3. Mochovce NPP simulator

    International Nuclear Information System (INIS)

    Ziakova, M.

    1998-01-01

    Mochovce NPP simulator basic features and detailed description of its characteristics are presented with its performance, certification and application for training of NPP operators as well as the training scenario

  4. Importance of the multi-modules study in PSA

    International Nuclear Information System (INIS)

    Gonzalez R, V. J.; Nelson E, P. F.

    2015-09-01

    The current approach that has taken the Probabilistic Safety Analysis (PSA) consists of doing all the APS analysis including the existence of multi-units in the nuclear power plants (NPP), this new approach seeks to analyze the risk of site, evaluating all reactors together. The main reasons for this trend are: the accident occurred on March 2011 in Fukushima Daiichi in Japan, with serious consequences in more than one reactor of the NPP and the current planning and construction of new Small Modular Reactors, which host more than one module on the same NPP and are connected to a single control room. This study analyzes how to model the risk of a multi-module NPP. In 2013, the ASME/ANS standard for advanced reactors that are not light-water reactors was published, in which the requirements to realize a PSA including multi-units or modules are shown; however, does not describe the methodology to do that. This article presents a methodology to calculate the risk of the site in a PBMR plant with two modules. This methodology consists of two models of trees of different events, one that evaluates to a single PBMR module and another that evaluates the two modules together. Both models are responsible to show their differences and compare results to finally demonstrate the need for new methodologies for risk analysis site in multi-modules and units. (Author)

  5. Approach to development and use of PSA Level 2 analysis for the Cernavoda nuclear power plant

    International Nuclear Information System (INIS)

    Turcu, I.; Deaconu, R.; Radu, G.

    1998-01-01

    This paper first describes the status of PSA activities for the Cernavoda NPP and the extension of the PSA work to include Level 2 PSA. Important characteristics of this reactor type for Level 2 PSA are outlined. Due to the specific layout of the CANDU reactor the evolution of severe accidents is considerably different to vessel type LWRs. Accidents can be roughly categorized into three categories, ''''severe accidents'''' which lead to the loss of core structural integrity, delayed loss of core structural integrity as a consequence of the loss of heat sinks, and fuel channel failures. The current work for modelling accident progression in the core region is described. The elements for the Level 2 PSA including definition of PDSs, probabilistic containment logic and source term calculation are outlined. It is pointed out that uncertainties have to be considered which are contained in the models to bridge knowledge gaps. For this purpose sensitivity studies will be carried out for key modelling assumptions. (author)

  6. Study on the risk-informed regulation of NPP

    International Nuclear Information System (INIS)

    Wang Chaogui

    2007-01-01

    The risk-informed regulation is a modern type of NPP safety management mode using both deterministic and probabilistic approaches. It is necessary to entirely and systematically study the associated regulations, standards and practices in order to promote the developments of risk-informed regulations in China. This paper introduces the risk-informed regulation, gives out the basic principles, method and acceptance risk criteria of risk-informed decision,making, discusses the PSA requirements for risk-informed decision-making and makes some suggestions about the application of risk-informed regulations in Chinese NPP. (authors)

  7. Short overview of PSA quantification methods, pitfalls on the road from approximate to exact results

    International Nuclear Information System (INIS)

    Banov, Reni; Simic, Zdenko; Sterc, Davor

    2014-01-01

    Over time the Probabilistic Safety Assessment (PSA) models have become an invaluable companion in the identification and understanding of key nuclear power plant (NPP) vulnerabilities. PSA is an effective tool for this purpose as it assists plant management to target resources where the largest benefit for plant safety can be obtained. PSA has quickly become an established technique to numerically quantify risk measures in nuclear power plants. As complexity of PSA models increases, the computational approaches become more or less feasible. The various computational approaches can be basically classified in two major groups: approximate and exact (BDD based) methods. In recent time modern commercially available PSA tools started to provide both methods for PSA model quantification. Besides availability of both methods in proven PSA tools the usage must still be taken carefully since there are many pitfalls which can drive to wrong conclusions and prevent efficient usage of PSA tool. For example, typical pitfalls involve the usage of higher precision approximation methods and getting a less precise result, or mixing minimal cuts and prime implicants in the exact computation method. The exact methods are sensitive to selected computational paths in which case a simple human assisted rearrangement may help and even switch from computationally non-feasible to feasible methods. Further improvements to exact method are possible and desirable which opens space for a new research. In this paper we will show how these pitfalls may be detected and how carefully actions must be done especially when working with large PSA models. (authors)

  8. Development of a Base Model for the New Fire PSA Training

    International Nuclear Information System (INIS)

    Kim, Kilyoo; Kang, Daeil; Kim, Wee Kyong; Do, Kyu Sik

    2013-01-01

    US NRC/EPRI issued a new fire PSA method represented by NUREG/CR 6850, and have been training many operators and inspectors to widely spread the new method. However, there is a limitation in time and efficiency for many foreigners, who generally have communication problem, to participate in the EPRI/NRC training to learn the new method. Since it is about time to introduce the new fire PSA method as a regulatory requirement for the fire protection in Korea, a simple and easy-understandable base model for the fire PSA training is required, and KAERI-KINS is jointly preparing the base model for the new fire PSA training. This paper describes how the base model is developed. Using an imaginary simple NPP, a base model of fire PSA following the new fire PSA method was developed in two ways from the internal PSA model. Since we have the base model and know the process of making the fire PSA model, the training for the new fire PSA method can be in detail performed in Korea

  9. UV/EB curable psa's

    International Nuclear Information System (INIS)

    Glotfelter, C.A.

    1995-01-01

    The author describe both water-based and 100% solids UV/EB curable PSA's (Pressure Sensitive Adhesives) and their properties. A new acrylate monomer, ethoxylated nonyl phenol acrylate, has great utility in the formulation of water-based PSA's

  10. PSA Test: What's It for?

    Science.gov (United States)

    ... PSA levels. Obesity can also lower PSA levels. Misleading results. The test doesn't always provide an ... Policy Notice of Privacy Practices Notice of Nondiscrimination Advertising Mayo Clinic is a not-for-profit organization ...

  11. Guideline level-3 PSA

    International Nuclear Information System (INIS)

    Roelofsen, P.M.; Van der Steen, J.

    1993-09-01

    For several applications of radioactive materials calculations must be executed to determine the radiation risk for the population. A guideline for the risk calculation method of two main sources: nuclear power plants, and other intended and unintended activities with radioactive materials, is given. The standards, recommendations and regulations in this report concern mainly the analysis of the radiological (external) consequences of nuclear power plant accidents, classified as level-3 PSA (Probabilistic Safety Analysis). Level-3 PSA falls within the scales 5-7 of the International Nuclear Event Scale (INES). The standards, etc., focus on the risks for groups of people and the so-called maximum individual risk. In chapter two the standards and regulations are formulated for each part of level-3 PSA: the source term spectrum, atmospheric distribution and deposition, exposure to radiation doses and calculation of radiation doses, dose-response relationships, measures to reduce the effect of radiation doses, design basis accidents, and finally uncertainty analysis. In chapter four, modelled descriptions are given of the standards and regulations, which could or should be used in a calculation program in case of level-3 PSA. In chapter three the practical execution of a probabilistic consequences analysis, the collection of input data and the presentation of the results are dealt with. 2 figs., 14 tabs., 64 refs

  12. Probabilistic safety analyses (PSA)

    International Nuclear Information System (INIS)

    1997-01-01

    The guide shows how the probabilistic safety analyses (PSA) are used in the design, construction and operation of light water reactor plants in order for their part to ensure that the safety of the plant is good enough in all plant operational states

  13. Babesiosis PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2012-04-25

    This 60 second PSA describes babesiosis, a preventable and treatable tickborne disease, including the signs and symptoms of infection and ways to prevent it.  Created: 4/25/2012 by Center for Global Health, Division of Parasitic Diseases and Malaria.   Date Released: 4/26/2012.

  14. Introduction to PSA applications

    International Nuclear Information System (INIS)

    Evans, M.G.K.

    1997-01-01

    The aim of this presentation is to show how the PSA can be used to determine the risk impact of the various deterministic processes for plant design or operational changes, and the evaluation of off normal events that occurred at the plant. The presentation is divided into the following topics: Identification of issues; Tasks/element identification; Modelling changes; Data changes. 6 figs

  15. Decommissioning of Brennilis NPP

    International Nuclear Information System (INIS)

    Baize, Jean-Marc

    1998-01-01

    This EDF press communique give information related to the decommissioning of the Brennilis NPP. The following five items are developed in this report: 1. the level-2 decommissioning operations at the Brennilis NPP; 2. the Brennilis NPP, a pilot operation from the commissioning up to the decommissioning; 3. history of the Brennilis NPP decommissioning; 4. the types of radioactive wastes generated by the Brennilis NPP decommissioning; 5. the Brennilis NPP - a yard management as a function of the wastes. The document contains also seven appendices addressing the following subjects: 1. the share of decommissioning assigned to EDF and the decommissioning steps; 2. the EDF installations in course of decommissioning; 3. the CEA decommissioned installations or in course of decommissioning; 4. regulations; 5. costs; 6. waste management - principles; 7. data on the decommissioning yard

  16. Data analysis treatment in the Juragua Nuclear Power Plant preoperational PSA

    International Nuclear Information System (INIS)

    Valhuerdi Debesa, C.

    1996-01-01

    Data Analysis is an important task within Probabilistic safety Assessment,. which usually determines the level of detail of the analysis, being the way to feed the PSA with the operational experience of the Nuclear Power Plant analysed. In this paper the role of the Data Analysis Task as part of the PSA process and the different kinds of data to be estimated are explained. A description is presented of the organization of the data Analysis in the Juragua NPP Preoperational PSA, the information sources and the criteria handled for the estimation of the different kinds of Data. The Generic Data Base adopted for equipment failures and the state of the generic data issue for VVER reactors and its prospects are also dealt with. The paper concludes with suggestions for the further development of Juragua NPP generic Data Base

  17. PSA in operator training

    International Nuclear Information System (INIS)

    Nos, V.; Faig, J.; Plesa, P.; Delgado, J. L.

    2000-01-01

    The systematic approach to training is internationally accepted as the best method to achieve and maintain the qualification and competence of power plant personnel and to guarantee the quality of their training. Following the recommendations and guidelines of international organisations competent in the field, TECNATOM SA has developed projects based on the systematic approach to training for all Spanish nuclear power plants. One of the latest projects was the systematic approach to training developed for the operation personnel of ASCO Nuclear Power Plant. In this case, certain results of the Probabilistic Safety Analysis (PSA) which complement the systematic safety and reliability criteria of the systematic approach to training process have been incorporated in the traditional processes of work and task analysis and training plan design. This incorporation provides the training manager with additional criteria based not only on safety aspects obtained through the statistical treatment of considerations of skilled technical personnel (operators, operation chief supervisors, etc), but also on the independent criterion of the PSA. The inclusion of this approach basically affects all systematics in two of its stages: During the selection process of operating practices in SMR or SGI, the possible scenarios have been associated with all those situations where human actions which lead to an initiating event or human actions to mitigate an initiating event, may take place, as defined in the PSA. During the scenario development process, the instruments involved in the performance of human actions which originate or mitigate an event taking place have been identified. This pakes it possible to reconcile the scenario event sequence with the sequence considered in the PSA study, as the most likely to provoke a more serious accident. The incorporation of these PSA results contributes to the strengthening of safety aspects in training in an objective way, and confirms that

  18. PSA in America

    International Nuclear Information System (INIS)

    Linn, M.A.; Cunningham, M.A.; Johnson, D.H.

    1996-01-01

    Although the concept of acceptable risk has always been the foundation of the nuclear industry design, the use of formal PSA (or PRA-probabilistic risk assessment) in the U.S. nuclear power industry has followed an unusual path in arriving at its current level of notability. Prior to 1975, probabilistic evaluations were limited to a few specific applications such as the evaluation of man-made (i.e., airplane crashes) and natural (i.e., earthquakes) hazards. In 1975, the industry was introduced to comprehensive PSA by the Reactor Safety Study (WASH-1400). However, the study languished in relative obscurity until the accident at Three Mile Island 2 (TMI-2) in 1979. This event significantly altered the industry's view of severe accidents in the U.S. and worldwide. Investigative committees of TMI-2 recommended that PSA techniques be more widely used to augment the traditional deterministic methods of determining nuclear plant safety. This initiated an unprecedented effort by nuclear regulators and licensees worldwide to significantly improve the state of knowledge of severe accidents at nuclear power plants. In the U.S., use of PSA began to increase as evidenced by its application in the anticipated transient without scram and station blackout rulemakings, generic issue prioritization and resolution, risk-based inspection guidelines, backfit policy, and technical specification improvements. However, broad application of probabilistic techniques to the industry as a whole was initiated in 1986 with the publication of Safety Goals for the Operation of Nuclear Power Plant; Policy Statement. This put PSA front and center in the U.S. regulatory arena by open-quotes establish[ing] goals that broadly define an acceptable level of radiological risk that might be imposed on the public as a result of nuclear power plant operation.close quotes Both qualitative safety goals and quantitative objectives were articulated in this policy statement

  19. Level 1 and 2 PSA methodology taking into account new design, operating and safety factors. Rev. 1

    International Nuclear Information System (INIS)

    Jirsa, P.; Patrik, M.

    2000-11-01

    The status of probabilistic safety assessment (PSA) is discussed (i) in relation to the expected nature of 'revolutionary' innovations and (ii) in the light of the EUR document, summarizing requirements put by European NPP operators on the future NPP design. The aims included: (1) analysis of limitations to the current PSA methodology; (2) specification of physical and operation processes the knowledge of which is necessary to ensure the safety criteria of advanced reactors; (3) summarisation of existing knowledge and description formats of the processes; (4) identification of theoretical and experimental work required to address the problem, preparation of data and computer codes, ensuring traceability to EU developmental programs. (P.A.)

  20. Suggestion of a Framework to Analyze Failure Modes and Effect of Cyber Attacks in NPP

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Young; Seong, Poong Hyun [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    The cyber security issue on NPP is inevitable issue. Unlike general cyber security, cyber-physical system like NPP can induce serious consequences such as core damage by cyber-attack. So in this paper, to find how hacker can attack the NPP, (1) PSA results were utilized to find the relationship between physical system and cyber-attack and (2) vulnerabilities on digital control systems were investigated to find how hacker can implement the possible attack. It is expected that these steps are utilized when establishing penetration test plans or cyber security drill plans.

  1. Binge Drinking PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2010-10-05

    This PSA is based on the October, 2010 CDC Vital Signs report which indicates that drinking too much, including binge drinking, causes more than 79,000 deaths in the U.S. each year and is the third leading preventable cause of death.  Created: 10/5/2010 by Centers for Disease Control and Prevention (CDC).   Date Released: 10/5/2010.

  2. Whooping Cough PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2015-01-22

    This 30 second PSA encourages pregnant women to get the whooping cough vaccine, called Tdap, during the third trimester of each pregnancy in order to pass antibodies to their babies so they are born with protection against this serious disease.  Created: 1/22/2015 by National Center for Immunization and Respiratory Diseases (NCIRD), Division of Bacterial Diseases (DBD), Meningitis and Vaccine Preventable Diseases Branch (MVPDB).   Date Released: 1/22/2015.

  3. PSA kinetics after prostate brachytherapy: PSA bounce phenomenon and its implications for PSA doubling time.

    Science.gov (United States)

    Ciezki, Jay P; Reddy, Chandana A; Garcia, Jorge; Angermeier, Kenneth; Ulchaker, James; Mahadevan, Arul; Chehade, Nabil; Altman, Andrew; Klein, Eric A

    2006-02-01

    To analyze prostate-specific antigen (PSA) kinetics in patients treated with prostate brachytherapy (PI) with a minimum of 5 years of PSA follow-up. The records of 162 patients treated with PI for localized prostate cancer with a minimum of 5 years of PSA follow-up were reviewed. A variety of pretreatment and posttreatment variables were examined. Patients were coded as having a PSA bounce if their PSA achieved a nadir, elevated at least 0.2 ng/mL greater than that nadir, and decreased to, or below, the initial nadir. Two definitions of biochemical failure (bF) or biochemical relapse-free survival (bRFS) were used: the classic American Society for Therapeutic Radiology and Oncology consensus definition of three consecutive rises (bF3) and the nadir plus 2 ng/mL definition (bFn+2). Associations between a PSA bounce and the various pre- and posttreatment factors were assessed with logistic regression analysis, and the association between a PSA bounce and bF was examined with the log-rank test. The Mann-Whitney U test was applied to test for differences in the PSA doubling time (PSADT) and the time to a PSA rise between the PSA bounce patients and the bF patients. PSADT was calculated from the nadir to the time of the first PSA rise, because this point is known first in the clinical setting. The 5-year overall bRFS rate was 87% for the bF3 definition and 96% for the bFn+2 definition. A PSA bounce was experienced by 75 patients (46.3%). Patients who experienced a PSA bounce were less likely to have a bF, regardless of the bRFS definition used (bF3: p=0.0015; bFn+2: p=0.0040). Among the pre- and posttreatment factors, only younger age predicted for a PSA bounce on multivariate analysis (p=0.0018). The use of androgen deprivation had no effect on PSA bounce. No difference was found in the PSADT between patients who had a PSA bounce and those with bF. The median PSADT for those with a PSA bounce was 8.3 months vs. 10.3 months using the bF3 definition and 8.8 months using

  4. PSA kinetics after prostate brachytherapy: PSA bounce phenomenon and its implications for PSA doubling time

    International Nuclear Information System (INIS)

    Ciezki, Jay P.; Reddy, Chandana A.; Garcia, Jorge; Angermeier, Kenneth; Ulchaker, James; Mahadevan, Arul; Chehade, Nabil; Altman, Andrew; Klein, Eric A.

    2006-01-01

    Purpose: To analyze prostate-specific antigen (PSA) kinetics in patients treated with prostate brachytherapy (PI) with a minimum of 5 years of PSA follow-up. Methods and Materials: The records of 162 patients treated with PI for localized prostate cancer with a minimum of 5 years of PSA follow-up were reviewed. A variety of pretreatment and posttreatment variables were examined. Patients were coded as having a PSA bounce if their PSA achieved a nadir, elevated at least 0.2 ng/mL greater than that nadir, and decreased to, or below, the initial nadir. Two definitions of biochemical failure (bF) or biochemical relapse-free survival (bRFS) were used: the classic American Society for Therapeutic Radiology and Oncology consensus definition of three consecutive rises (bF3) and the nadir plus 2 ng/mL definition (bFn+2). Associations between a PSA bounce and the various pre- and posttreatment factors were assessed with logistic regression analysis, and the association between a PSA bounce and bF was examined with the log-rank test. The Mann-Whitney U test was applied to test for differences in the PSA doubling time (PSADT) and the time to a PSA rise between the PSA bounce patients and the bF patients. PSADT was calculated from the nadir to the time of the first PSA rise, because this point is known first in the clinical setting. Results: The 5-year overall bRFS rate was 87% for the bF3 definition and 96% for the bFn+2 definition. A PSA bounce was experienced by 75 patients (46.3%). Patients who experienced a PSA bounce were less likely to have a bF, regardless of the bRFS definition used (bF3: p = 0.0015; bFn+2: p = 0.0040). Among the pre- and posttreatment factors, only younger age predicted for a PSA bounce on multivariate analysis (p = 0.0018). The use of androgen deprivation had no effect on PSA bounce. No difference was found in the PSADT between patients who had a PSA bounce and those with bF. The median PSADT for those with a PSA bounce was 8.3 months vs. 10.3 months

  5. PSA results and trends for Spain's NPPs

    International Nuclear Information System (INIS)

    Carretero, J.A.

    1993-01-01

    The Spain regulatory authority CSN demanded performance of PSA for all Spain nuclear power plants. The specific data analysis carried out as a part of the PSA has contributed to the realistic view on the results which could be achieved by the PSA. The main characteristics of the PSA in Spain and PSA trends in the development are presented in the paper

  6. NPP service life management

    International Nuclear Information System (INIS)

    Elagin, Yu.P.

    2001-01-01

    Problems of NPP service life management and service life prolongation are reviewed. Methods for the prolongation of the French NPP service life are discussed, priority directions of nuclear block service life management in regard to aging in the context of the European program of investigation into the materials aging are identified. Questions of the provision of the 60 years service life of the Mihama 1 block (Japan) and decision of the problem of the control equipment aging in Great Britain are discussed. Situation with the prolongation of licenses on the NPP operation in the USA and Spain is considered [ru

  7. PSA in application

    International Nuclear Information System (INIS)

    Thadani, A.C.

    1989-01-01

    This paper reports on Probabilistic Safety Assessment methodologies which are finding wide application within the nuclear power industry. Plant specific studies have been useful in identifying design vulnerabilities, providing a basis for prioritization of plant modifications and in an operations management role. PSA applications in regulatory activities have also proven very beneficial to the US NRC in assessing significance of operating plant events/issues, consideration of potential design improvements, standard plant licensing and conducting regulatory and cost-benefit analyses on proposed generic issue resolution. The benefits from such studies are concluded to be well worth their cost

  8. 3D model of steam generator of nuclear power plant Krsko

    International Nuclear Information System (INIS)

    Ravnikar, I.; Petelin, S.

    1995-01-01

    The Westinghouse Electric Corporation D4 steam generator design was analyzed from a thermal-hydraulic point of view using the 3D PHOENICS computer code. Void fraction, velocity and enthalpy distributions were obtained in the U-tube riser. The boundary conditions of primary side were provided by SMUP 1D code. The calculations were carried out for present operating conditions of nuclear power plant Krsko. (author)

  9. PSA Update Procedures, an Ultimate Need for Living PSA

    International Nuclear Information System (INIS)

    Hegedus, D.

    1998-01-01

    Nuclear facilities by their complex nature, change with time. These changes can be both physical (plant modification, etc.), operational (enhanced procedures, etc.) and organizational. In addition, there are also changes in our understanding of the plant, due to operational experience, data collection, technology enhancements, etc. Therefore, it is imperative that PSA model must be frequently up-dated or modified to reflect these changes. Over the last ten years. these has been a remarkable growth of the use of Probabilistic Safety Assessments (PSAs). The most rapidly growing area of the PSA Applications is their use to support operational decision-making. Many of these applications are characterized by the potential for not only improving the safety level but also for providing guidance on the optimal use of resources and reducing regulatory burden. To enable a wider use of the PSA model as a tool for safety activities it is essential to maintain the model in a controlled state. Moreover, to fulfill requirements for L iving PSA , the PSA model has to be constantly updated and/or monitored to reflect the current plant configuration. It should be noted that the PSA model should not only represent the plant design but should also represent the operational and emergency procedures. To keep the PSA model up-to-date several issues should be clearly defined including: - Responsibility should be divided among the PSA group, - Procedures for implementing changes should be established, and - QA requirements/program should be established to assure documentation and reporting. (author)

  10. Review process of PSA Level 2 of KBR. Concept and experience

    International Nuclear Information System (INIS)

    Andernacht, Martin; Glaser, Hendrik; Sonnenkalb, Martin

    2009-01-01

    In Germany, a periodic safety review (PSR) has to be performed every 10 years by the utility. In the past, a PSR only included a plant-specific probabilistic safety analysis (PSA) Level 1 study. For the NPP Brokdorf (KBR) PSA Level 2 project, an agreement was reached between all parties involved that the study will be performed not as a part of the PSR process, but supplementary to it. Since a revised version of the German PSA guideline has been released in 2005, these plant-specific PSAs have to include a PSA Level 2, too. This paper will focus on conclusions and findings from a ongoing parallel review process of the first full scope PSA Level 2 performed by the utility for KBR, a typical German PWR-1300. The responsible authority 'Ministerium fuer Soziales, Gesundheit, Familie, Jugend und Senioren des Landes Schleswig-Holstein (MSGF)' (Ministry of Social Affairs, Health, Family, Youth and Senior Citizens of Schleswig-Holstein) initiated this parallel review process in agreement with the utility KBR and the E.ON Kernkraft in 2006. The project will be completed soon. Such a review process allows that essential steps of the PSA will be reviewed and commented before the PSA Level 2 will be finished. So the benefit from this parallel review process is a significant enhancement of the quality and completeness of the PSA Level 2 study as the majority of the recommendations given by the review team has been taken over by the utility and the developer of the PSA, the Areva NP company. Further, a common understanding and agreement will be reached at the end between all parties involved on the major topics of the PSA Level 2 study. (orig.)

  11. Review process of PSA level 2 of KBR - Concept and Experience

    International Nuclear Information System (INIS)

    Andernacht, M.; Glaser, H.; Sonnenkalb, M.

    2013-01-01

    In Germany, a periodic safety review (PSR) has to be performed every ten years by the utility. In the past, a PSR only included a plant-specific probabilistic safety analysis (PSA) Level 1 study. Since a revised version of the German PSA guideline has been released in 2005, these plant-specific PSAs have to include a PSA Level 2, too. For the NPP Brokdorf (KBR) PSA Level 2 project, an agreement was reached between all parties involved that the study will be performed not as a part of the PSR process, but supplementary to it. This paper will focus on conclusions and findings from an ongoing parallel review process of the first full scope PSA Level 2 performed by the utility for KBR, a typical German PWR-1300. The responsible authority 'Ministerium fuer Soziales, Gesundheit, Familie, Jugend und Senioren des Landes Schleswig- Holstein' (MSGF) initiated this parallel review process in agreement with the utility KBR and the E.ON Kernkraft in 2006. The project will be completed soon. Such a review process allows that essential steps of the PSA will be reviewed and commented before the PSA Level 2 will be finished. So the benefit from this parallel review process is a significant enhancement of the quality and completeness of the PSA Level 2 study as the majority of the recommendations given by the review team has been taken over by the utility and the developer of the PSA, the AREVA NP company. Further, a common understanding and agreement will be reached at the end between all parties involved on the major topics of the PSA Level 2 study. The paper is followed by the slides of the presentation. (authors)

  12. Development of PSA workstation KIRAP

    International Nuclear Information System (INIS)

    Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

    1997-01-01

    Advanced Research Group of Korea Atomic Energy Research Institute has been developing the Probabilistic Safety Assessment(PSA) workstation KIRAP from 1992. This report describes the recent development activities of PSA workstation KIRAP. The first is to develop and improve the methodologies for PSA quantification, that are the incorporation of fault tree modularization technique, the improvement of cut set generation method, the development of rule-based recovery, the development of methodology to solve a fault tree which has the logical loops and to handle a fault tree which has several initiators. These methodologies are incorporated in the PSA quantification software KIRAP-CUT. The second is to convert PSA modeling softwares for Windows, which have been used on the DOS environment since 1987. The developed softwares are the fault tree editor KWTREE, the event tree editor CONPAS, and Data manager KWDBMAN for event data and common cause failure (CCF) data. With the development of PSA workstation, it makes PSA modeling and PSA quantification and automation easier and faster. (author). 8 refs

  13. Detection of prostate cancer with complexed PSA and complexed/total PSA ratio - is there any advantage?

    OpenAIRE

    Strittmatter F; Stieber P; Nagel D; Füllhase C; Walther S; Stief CG; Waidelich R

    2011-01-01

    Abstract Objective To evaluate the performance of total PSA (tPSA), the free/total PSA ratio (f/tPSA), complexed PSA (cPSA) and the complexed/total PSA ratio (c/tPSA) in prostate cancer detection. Methods Frozen sera of 442 patients have been analysed for tPSA, free PSA (fPSA) and cPSA. 131 patients had prostate cancer and 311 patients benign prostatic hyperplasia. Results Differences in the distribution of the biomarkers were seen as follows: tPSA, cPSA and c/tPSA were significantly higher i...

  14. Clinical significance of changes in ratio of F PSA/PSA and PSA concentration in patients with prostatic cancer

    International Nuclear Information System (INIS)

    Xiang Ming; Chen Xiuzhen

    2005-01-01

    To evaluate the clinical significance of changes in F PSA/PSA ratio and the annual changing rate of PSA concentration in prostatic cancer patients with PSA between 4 to 10 μg/L, we determined the PSA and F PSA by ELISA in different periods of time. The receptive operator character(ROC) curve was used to evaluate the prediction value of F PSA/PSA ratio and annual change of PSA concentration. The results showed that the F PSA/PSA ratio and the annual changing rate of the PSA concentration were significantly different between prostatic cancer (PC) and non-PC individuals(P 0.05), but the annual changing rate of the PSA concentration was(PC 0. 001). Our conclusion is that the F PSA/PSA ratio and the annual changing rate of PSA concentration are help ful to the diagnosis in patients with the PSA value between 4 to 10 μg/L. (authors)

  15. Analysis of specific features of digital instrumentation and control systems and possibilities of accounting for them within PSA

    International Nuclear Information System (INIS)

    Hustak, S.

    2002-10-01

    The report is structured as follows: Basic information on the peculiarities of digital technology for the I and C system at an NPP (Digital signal; Digital communication; Communication protocols; Examples of practical tools for creation of I and C digital systems); Peculiarities of the digital I and C technology from the reliability viewpoint (Software as a new component of implementation of a system function; Problems with the assessment or demonstration of reliability of software components of an I and C system); Possibilities of accounting for the specific features of digital I and C technology within PSA (Relevant PSA components; Using PSA as a supporting tool in designing new NPPs; Categorization of NPP I and C system tasks with respect to the defence-in-depth principle). (P.A.)

  16. Methodology - PSA Regulatory handbook. Comparisons to a modern PSA study

    International Nuclear Information System (INIS)

    Bostroem, Urban; Jung, Gunnar; Flodin, Yngve

    2003-03-01

    The regulatory handbook is applicable to all types of initiating events and all operating conditions. It should be noted that it does not make the traditional subdivision of PSA into internal and external events, level 1 and level 2 PSA, or power operation and shut-down. The reason for this is that this has given the regulatory handbook a more logical structure, and that this approach underlines the integrated character of PSA when it comes to creating the plan risk profile. The regulatory handbook has been structured following the requirements on a PSA for a nuclear power plant, as this is the most demanding application. However, it is applicable also to the analysis of other nuclear installations. The purpose of the comparative review presented in this report has been to, as part of a quality review establish the PSA Handbook, compare (parts of) the handbook and its criteria with a recent PSA analysis, and to identify major discrepancies. Considerable weight has also been allocated to a review of the plant model (Risk Spectrum event trees and fault trees). The results presented in the report are not based on a complete review of the PSA in question (or of the complete PSA Handbook). Following discussions between the SKI and SwedPower, and based on the experience of the SwedPower reviewers, the following issues were chosen to be the main parts of the project: 1) General comparison according to content and transparency - Levels of ambition in PSA Handbook, PSA method description and actual PSA report. 2) Detailed comparison of: Selected component failure data - Assumptions regarding room events - CCI frequencies, realism, identification, categorisation - Taking credit for non-safety classified systems - Event tree modelling - Presentation of results 3) Fault tree model, specifically - Time frame for crediting of battery capacity - Modelling of regulators - Modelling of dependencies for room events - general quality, like how the paper documentation and the logic

  17. NPP unusual events: data, analysis and application

    International Nuclear Information System (INIS)

    Tolstykh, V.

    1990-01-01

    Subject of the paper are the IAEA cooperative patterns of unusual events data treatment and utilization of the operating safety experience feedback. The Incident Reporting System (IRS) and the Analysis of Safety Significant Event Team (ASSET) are discussed. The IRS methodology in collection, handling, assessment and dissemination of data on NPP unusual events (deviations, incidents and accidents) occurring during operations, surveillance and maintenance is outlined by the reports gathering and issuing practice, the experts assessment procedures and the parameters of the system. After 7 years of existence the IAEA-IRS contains over 1000 reports and receives 1.5-4% of the total information on unusual events. The author considers the reports only as detailed technical 'records' of events requiring assessment. The ASSET approaches implying an in-depth occurrences analysis directed towards level-1 PSA utilization are commented on. The experts evaluated root causes for the reported events and some trends are presented. Generally, internal events due to unexpected paths of water in the nuclear installations, occurrences related to the integrity of the primary heat transport systems, events associated with the engineered safety systems and events involving human factor represent the large groups deserving close attention. Personal recommendations on how to use the events related information use for NPP safety improvement are given. 2 tabs (R.Ts)

  18. Applicability of PSA Issues for Risk Assessment during Optimisation of In-Service Inspection

    International Nuclear Information System (INIS)

    Kolykhanov, V.; Skalozubov, V.; Kovrigkin, Y.

    2006-01-01

    The current codes determining periodicity of in-service inspection of the NPP equipment have been formed using deterministic approaches and have an unnecessary degree of conservatism. A perspective direction of perfection of normative base is decision making on a basis of risk-informed methodologies. It allows to increase safety of NPP equipment's operation and to optimise programs on inspection of the equipment subject to limited resources by focusing efforts on the most safety significant elements of the equipment. It is internationally accepted that methodology of the probabilistic safety analysis (PSA) is the most universal and comprehensive tool focused on the general assessment of safety of NPP as a whole. By now, PSA Level 1 is fulfilled for all pilot units of the Ukrainian NPPs that is a valuable result, which should be taken into account at an assessment of reliability of the equipment. However, specificity of PSA methodology should be taken into account at the decision of the particular tasks aimed at optimisation of maintenance of the equipment within individual systems. The estimation of the contribution to core damage frequency (CDF) is a PSA issue usually used to assess the significance of consequences of failure of a system/equipment during risk-informed decision-making. This work shows that above factor is only a part of assessment of the significance of consequences as core damage can be expressed in different amount of the damaged fuel elements and, hence, severity of consequences. Besides CDF is directly affected only by active elements which failure can be an initiating event. PSA methodology uses averaged reliability factors of the equipment for all possible operating modes occurring at transitive accident process. Here, there are limited opportunities to account impact of periodicity of maintenance of the equipment on reliability and to predict impact of change of the inspection program. PSA methodology does not allow taking into account

  19. EMAS at Doel NPP

    International Nuclear Information System (INIS)

    Nolan, Peter; Thoelen, Els

    1998-01-01

    In October 1995, Doel NPP of Electrabel, Belgium opted to seek registration under the EC Eco-management and Audit Scheme (EMAS). A comprehensive environmental management system (EMS) has been introduced and implemented, encompassing all four PWRs and the supporting departments. A critical step was to seek certification from an accredited environmental auditing body against the International Standard ISO 14001. This provided the foundation for the publicly available environmental statement required by EMAS. The complications of achieving EMAS at a time when national and international standards were being re-formulated were successfully overcome and Doel NPP passed its EMAS audit in June 1997. (author)

  20. Comparison of HRA methods based on WWER-1000 NPP real and simulated accident scenarios

    International Nuclear Information System (INIS)

    Petkov, Gueorgui

    2010-01-01

    Full text: Adequate treatment of human interactions in probabilistic safety analysis (PSA) studies is a key to the understanding of accident sequences and their relative importance in overall risk. Human interactions with machines have long been recognized as important contributors to the safe operation of nuclear power plants (NPP). Human interactions affect the ordering of dominant accident sequences and hence have a significant effect on the risk of NPP. By virtue of the ability to combine the treatment of both human and hardware reliability in real accidents, NPP fullscope, multifunctional and computer-based simulators provide a unique way of developing an understanding of the importance of specific human actions for overall plant safety. Context dependent human reliability assessment (HRA) models, such as the holistic decision tree (HDT) and performance evaluation of teamwork (PET) methods, are the so-called second generation HRA techniques. The HDT model has been used for a number of PSA studies. The PET method reflects promising prospects for dealing with dynamic aspects of human performance. The paper presents a comparison of the two HRA techniques for calculation of post-accident human error probability in the PSA. The real and simulated event training scenario 'turbine's stop after loss of feedwater' based on standard PSA model assumptions is designed for WWER-1000 computer simulator and their detailed boundary conditions are described and analyzed. The error probability of post-accident individual actions will be calculated by means of each investigated technique based on student's computer simulator training archives

  1. The effect on the sensitivities of PSA and PSA-age volume score of ...

    African Journals Online (AJOL)

    Objective: The PSA-age volume (PSA-AV) score was calculated by multiplying the age and prostate volume and then dividing the total by the prebiopsy PSA level. The aim of this study was to evaluate the effect on the sensitivities of PSA and PSA-AV score of International Prostate Symptom Score (I-PSS) and nocturia in ...

  2. Development of a PSA information database system

    International Nuclear Information System (INIS)

    Kim, Seung Hwan

    2005-01-01

    The need to develop the PSA information database for performing a PSA has been growing rapidly. For example, performing a PSA requires a lot of data to analyze, to evaluate the risk, to trace the process of results and to verify the results. PSA information database is a system that stores all PSA related information into the database and file system with cross links to jump to the physical documents whenever they are needed. Korea Atomic Energy Research Institute is developing a PSA information database system, AIMS (Advanced Information Management System for PSA). The objective is to integrate and computerize all the distributed information of a PSA into a system and to enhance the accessibility to PSA information for all PSA related activities. This paper describes how we implemented such a database centered application in the view of two areas, database design and data (document) service

  3. Development of a Base Frame for the New Fire PSA Training, and Lessons Learned

    International Nuclear Information System (INIS)

    Kim, Kilyoo; Kang, DaeIl; Kim, Wee Kyoung

    2014-01-01

    US NRC/EPRI issued a new fire PSA method represented by NUREG/CR 6850, and since it is about time to introduce the new fire PSA method as a regulatory requirement for the fire protection in Korea, a simple and easy-understandable base model for the fire PSA training is required, and thus KAERI-KINS jointly prepared a base model for the new fire PSA training last year. In this year, as a base frame development, fire ignition frequencies and severity factors, which were assumed in developing of the base model, are calculated. The fire modeling is performed to get the severity factor. This paper describes how the base frame is developed. Using an imaginary simple NPP, a base frame of fire PSA following the new fire PSA method was developed, and with which two days training course was provided twice for the plant engineers and regulators. Several lessons learned from the training are described. The two methods in quantification, i.e., CCDP method and initiator method are described

  4. A study for the development of PSA integrated databse

    International Nuclear Information System (INIS)

    Kim, S. H.; Han, S. H.; Min, K. R.

    2002-01-01

    KAERI is constructing the PSA integrated database for UCN 3, 4 nuclear power plant. PSA integrated DB includes PSA model database and PSA information database. PSA model DB consists of PSA models and data which are used for PSA quantification. PSA information DB consists of the other PSA related information such as PSA reports, supporting documents, calculating sheets, etc. This paper defines the PSA integrated database and describes the data collection, classification, schema design and overall development procedure for database construction. Recently, we have developed the PSA model database schema and we are trying to design the PSA model DB browser

  5. Temelin NPP commissioning experiences

    International Nuclear Information System (INIS)

    Hanus, V.

    2002-01-01

    The Building Permit for the Temelin NPP with four VVER units was issued in 1986, which is a long time ago. Since then, however, was taken a route that is very different from what anybody imagined. Described are the legislative and design changes and given is a current condition of the power plant

  6. Survey of Dynamic PSA Methodologies

    International Nuclear Information System (INIS)

    Lee, Hansul; Kim, Hyeonmin; Heo, Gyunyoung; Kim, Taewan

    2015-01-01

    Event Tree(ET)/Fault Tree(FT) are significant methodology in Probabilistic Safety Assessment(PSA) for Nuclear Power Plants(NPPs). ET/FT methodology has the advantage for users to be able to easily learn and model. It enables better communication between engineers engaged in the same field. However, conventional methodologies are difficult to cope with the dynamic behavior (e.g. operation mode changes or sequence-dependent failure) and integrated situation of mechanical failure and human errors. Meanwhile, new possibilities are coming for the improved PSA by virtue of the dramatic development on digital hardware, software, information technology, and data analysis.. More specifically, the computing environment has been greatly improved with being compared to the past, so we are able to conduct risk analysis with the large amount of data actually available. One method which can take the technological advantages aforementioned should be the dynamic PSA such that conventional ET/FT can have time- and condition-dependent behaviors in accident scenarios. In this paper, we investigated the various enabling techniques for the dynamic PSA. Even though its history and academic achievement was great, it seems less interesting from industrial and regulatory viewpoint. Authors expect this can contribute to better understanding of dynamic PSA in terms of algorithm, practice, and applicability. In paper, the overview for the dynamic PSA was conducted. Most of methodologies share similar concepts. Among them, DDET seems a backbone for most of methodologies since it can be applied to large problems. The common characteristics sharing the concept of DDET are as follows: • Both deterministic and stochastic approaches • Improves the identification of PSA success criteria • Helps to limit detrimental effects of sequence binning (normally adopted in PSA) • Helps to avoid defining non-optimal success criteria that may distort the risk • Framework for comprehensively considering

  7. The effect on the sensitivities of PSA and PSA-age volume score of ...

    African Journals Online (AJOL)

    O. Üçer

    2017-04-27

    scorewascalculatedbymultiplyingtheageandprostatevolume and then dividing the total by the prebiopsy PSA level. The aim of this study was to evaluate the effect on the sensitivities of PSA and PSA-AV score of International Prostate Symptom Score (I-PSS) ...

  8. NPOESS preparatory project (NPP)

    Science.gov (United States)

    Murphy, R. E.; Taylor, Raynor L.; Neeck, Steven P.; Wielicki, Bruce A.; Barkstrom, Bruce R.; Crison, M.; Swenson, J. R.

    1998-12-01

    The National Aeronautics and Space Administration (NASA) is studying options for future space-based missions, building upon the measurements to be made by the first series of Earth Observing System (EOS) missions. One mission under consideration is the NPOESS Preparatory Project (NPP), a cooperative mission of NASA and the National Polar-orbiting Operational Environmental Satellite System (NPOESS). This mission would utilize new instrument technologies being developed by the NPOESS, with additional NASA requirements, to continue certain measurements from the first series of EOS missions. By flying in the 2005 time period, NPP would provide an early demonstration and validation of new instrument technologies and algorithms in support of future NPOESS missions and extend the critical time series measurements of EOS.

  9. The role of PSA in safety management

    International Nuclear Information System (INIS)

    Szikszai, T.

    1997-01-01

    The presentation discusses the following issues: defence in depth principle (the role of the barriers, how does PSA represents the barriers?); the safety management and nuclear power plants; the probabilistic and deterministic approaches; the PSA applications and safety management

  10. Reviewing PSA-based analyses to modify technical specifications at nuclear power plants

    International Nuclear Information System (INIS)

    Samanta, P.K.; Martinez-Guridi, G.; Vesely, W.E.

    1995-12-01

    Changes to Technical Specifications (TSs) at nuclear power plants (NPPs) require review and approval by the United States Nuclear Regulatory Commission (USNRC). Currently, many requests for changes to TSs use analyses that are based on a plant's probabilistic safety assessment (PSA). This report presents an approach to reviewing such PSA-based submittals for changes to TSs. We discuss the basic objectives of reviewing a PSA-based submittal to modify NPP TSs; the methodology of reviewing a TS submittal, and the differing roles of a PSA review, a PSA Computer Code review, and a review of a TS submittal. To illustrate this approach, we discuss our review of changes to allowed outage time (AOT) and surveillance test interval (STI) in the TS for the South Texas Project Nuclear Generating Station. Based on this experience gained, a check-list of items is given for future reviewers; it can be used to verify that the submittal contains sufficient information, and also that the review has addressed the relevant issues. Finally, recommended steps in the review process and the expected findings of each step are discussed

  11. Probabilistic Safety Analysis Level 2 for units 5 and 6 of the Kozloduy NPP - sensitivity analysis

    International Nuclear Information System (INIS)

    Mancheva, K.; Velev, V.

    2006-01-01

    This paper covers the results of the sensitivity analysis performed under the Probabilistic Safety Analysis (PSA) level 2 for units 5 and 6 of the Kozloduy NPP. The analysis performs the status of the unit before modernization program accomplishment. Therefore none of the measures accomplished under the modernization program is accounted in the investigation. The goal of the sensitivity analysis is to give the impact of some of the characteristics of the severe accident to the Large Early Release Frequency (LERF). (authors)

  12. Development of a PSA information management system

    International Nuclear Information System (INIS)

    Ho, Seok; Dong Kyu, Kim; Sun Koo, Kang

    2007-01-01

    In general, Probabilistic Safety Agreement (PSA) is a very complicated work that uses and generates a lot of resources such as reports, procedures, drawings, assumptions, calculation sheets, PSA model, and so on. In many PSAs, however, the data, materials and knowledge considered and generated during performing PSA are scattered in many documents so that overall structure of PSA and information relationship between documents and models cannot easily be understood. To organize and manage all documents related to PSA, to capture knowledge of analysts, and finally to improve the quality of PSA, a PSA information management system (PIMS) was developed. The PIMS can manage all the documents of a PSA in a database and connect the causal relation between one information to another in the scattered documents via link. The PIMS can manage all the assumptions and technical basis used in PSA, and it can keep the record of the design changes the revision of PSA model. It can also control the review results of PSA models. The link of the PIMS can explicitly describe and reveal the expertise of the PSA analysts, and it enables the users to capture the knowledge and to understand the structure and contents of a PSA with ease. We are planning to apply the PIMS to the PSA of Shin Kori Units 1 and 2 as feasibility study and then to all the PSAs of the nuclear power plants in Korea. The PIMS is expected to contribute to enhancing the quality and confidence of PSA and reducing the efforts and costs of maintenance and update of PSA. (authors)

  13. Development of a PSA information management system

    Energy Technology Data Exchange (ETDEWEB)

    Ho, Seok; Dong Kyu, Kim; Sun Koo, Kang [Korea Power Engineering Company, Inc (Korea, Republic of)

    2007-07-01

    In general, Probabilistic Safety Agreement (PSA) is a very complicated work that uses and generates a lot of resources such as reports, procedures, drawings, assumptions, calculation sheets, PSA model, and so on. In many PSAs, however, the data, materials and knowledge considered and generated during performing PSA are scattered in many documents so that overall structure of PSA and information relationship between documents and models cannot easily be understood. To organize and manage all documents related to PSA, to capture knowledge of analysts, and finally to improve the quality of PSA, a PSA information management system (PIMS) was developed. The PIMS can manage all the documents of a PSA in a database and connect the causal relation between one information to another in the scattered documents via link. The PIMS can manage all the assumptions and technical basis used in PSA, and it can keep the record of the design changes the revision of PSA model. It can also control the review results of PSA models. The link of the PIMS can explicitly describe and reveal the expertise of the PSA analysts, and it enables the users to capture the knowledge and to understand the structure and contents of a PSA with ease. We are planning to apply the PIMS to the PSA of Shin Kori Units 1 and 2 as feasibility study and then to all the PSAs of the nuclear power plants in Korea. The PIMS is expected to contribute to enhancing the quality and confidence of PSA and reducing the efforts and costs of maintenance and update of PSA. (authors)

  14. PSA - a tool for the nuclear safety

    International Nuclear Information System (INIS)

    Himanen, R.

    1992-01-01

    The PSA-model for BWR-type reactors of Finnish power company, Teollisuuden Voima Oy (TVO) was finished in year 1989. This basic PSA model included all safety systems, normal operating systems and auxiliary systems. Today TVO is working to enlarge the PSA to level 2 (environmental effects, for the fires, for the floodings and the outages). The TVO's experiences has been showed the PSA an useful tool for the developing the safety of BWR's (orig.)

  15. Raccoon Roundworm Infection PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2012-08-27

    This 60 second PSA describes the signs and symptoms of and ways to prevent Baylisascaris infection, a parasitic roundworm infection that is spread through raccoon feces.  Created: 8/27/2012 by Centers for Disease Control and Prevention (CDC).   Date Released: 8/28/2012.

  16. World AIDS Day PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2011-11-16

    December 1 is World AIDS Day. In this PSA, communities are encouraged to get tested for HIV.  Created: 11/16/2011 by National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention (NCHHSTP).   Date Released: 11/16/2011.

  17. Hepatitis Awareness Month PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2011-05-11

    May is National Hepatitis Awareness Month. This 30 second PSA discusses hepatitis and encourages listners to talk to their health care professional about getting tested.  Created: 5/11/2011 by National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention.   Date Released: 5/11/2011.

  18. Evolution of PSA activities in DAE, India

    International Nuclear Information System (INIS)

    Saraf, R.K.

    2006-01-01

    Probabilistic Safety Assessment (PSA) is an analytic method for identifying, analysing and evaluating, risk and show evidence that the public health and safety is protected. The world wide use of PSA within operating and regulating organizations is increasing. A substantial amount of effort has been applied in many countries to use PSA in a beneficial manner. In DAE, India regulatory body, viz., Atomic Energy Regulatory Body is now insisting on PSA results for licensing and renewal of licence though not in formal manner. Nuclear Power Corporation is also convinced that the results of PSA study are useful for decision making regarding design evaluation, back fitting, and technical specification modification. This talk covers evaluation of PSA activity in DAE, India right from its inception to the present state of art, giving emphasis on problems faced, data collection and analysis, modelling of human errors, and common cause failures. The talk also includes evaluation of external event PSA, viz., fire PSA, seismic PSA, and application of PSA towards development of Risk Monitor, Risk informed in-service inspection. The talk also outlines the future course of action like Dynamic Fault Tree Analysis, inclusion of ageing effects in PSA, online implementation of Risk Monitor, Software Reliability evaluation, and Passive System reliability. The areas where efforts are needed for effective utilization of PSA are also highlighted. (author)

  19. The PSA: Planetary Science Archive

    Science.gov (United States)

    Barthelemy, M.; Martinez, S.; Heather, D.; Vazquez, J. L.; Arviset, C.; Osuna, P.; PSA development Team

    2012-04-01

    Scientific and engineering data from ESA's planetary missions are made accessible to the world-wide scientific community via the Planetary Science Archive (PSA). The PSA consists of online services incorporating search, preview, download, notification and delivery basket functionality. Besides data from the GIOTTO spacecraft and several ground-based cometary observations, the PSA contains data from the Mars Express, Venus Express, Rosetta, SMART-1 and Huygens missions. The focus of the PSA activities is on the long-term preservation of data and knowledge from ESA's planetary missions. Scientific users can access the data online using several interfaces: - The Advanced Search Interface allows complex parameter based queries, providing the end user with a facility to complete very specific searches on meta-data and geometrical parameters. By nature, this interface requires careful use and heavy interaction with the end-user to input and control the relevant search parameters. - The Map-based Interface is currently operational only for Mars Express HRCS and OMEGA data. This interface allows an end-user to specify a region-of-interest by dragging a box onto a base map of Mars. From this interface, it is possible to directly visualize query results. The Map-based and Advanced interfaces are linked and cross-compatible. If a user defines a region-of-interest in the Map-based interface, the results can be refined by entering more detailed search parameters in the Advanced interface. - The FTP Browser Interface is designed for more experienced users, and allows for direct browsing and access of the data set content through ftp-tree search. Each dataset contains documentation and calibration information in addition to the scientific or engineering data. All data are prepared by the corresponding instrument teams, mostly located in Europe. PSA supports the instrument teams in the full archiving process, from the definition of the data products, meta-data and product labels

  20. Nuklearna Elektrarna Krsko: managing technical documentation for the 'rightsized' 90's and beyond

    International Nuclear Information System (INIS)

    Gradisar, D.; Freeland, K.R.

    1996-01-01

    Nuklearna Elektrarna Krsko (NEK) was constructed as a turnkey plant, entering commercial service in 1981. NEK operates the plant with a unique management philosophy which has adapted well to the demands of today's business environment. Contributing to this success has been the NEK Management Information System (MIS) that has assisted NEK in 'making more with less'. A key element of the MIS is the Technical Document system, which includes the DCM (Document Control Module) and QRM (Quality Records Management), enabling the successful retrofit of the existing document management and 'turnkey' plant records environment to a new, thoroughly integrated system controlling plant operation and configuration management. The system integrates the Engineering Equipment and Configuration Management data systems, providing immediate, realtime cross-reference of plant equipment and plant modifications with technical documentation. (author)

  1. Introduction of the commercial grade dedication into Nuclear Power Plant Krsko (NEK) procurement process

    International Nuclear Information System (INIS)

    Heruc, Z.; Gajsak, Z.; Nikpalj, R.

    1996-01-01

    NEK management has undertaken a set of actions to improve the ability to provide equipment, spare parts and material needed to support operation and maintenance of the Krsko plant. These actions are necessary due primarily to the fact that NEK is more and more confronted (increasing trend) with the issue that suppliers of safety-related equipment and spare parts have decided not to pursue the nuclear portion of their business, incl. specific QA systems and qualifications. The purchase orders imposing these requirements are no longer accepted. In order to continue to obtain the necessary materials at the required quality level, a 'Commercial Grade Item' (CGI) procurement and dedication program has been developed based on similar practices in the USA. (author)

  2. Transport of replaced steam generators from Port of Koper to Krsko (Slovenia)

    International Nuclear Information System (INIS)

    Kovacic, S.

    2000-01-01

    All activities regarding the transport preparation as well as preparation of the route between the Port of Koper and Krsko Site were completed within the period of 12 month. Transport configuration had been designed as a self supporting structure where two 12 axes trailers presented rolling cars. Total mass of aforesaid configuration was about 666 tons, where pulling and pushing tractors were not included. Major observations were granted to concrete structures that would be crossed during the transport. It is our pleasure to conclude, that respected design directions relevant for concrete structures (SODOC 1.0) enable rather quality bases regarding load capacity and taking over extremely heavy loads such as transport configuration itself. In manner to achieve a required safety factor and protection against permanent deformations on structures, many static analyses had been made by design engineers. (author)

  3. Technical Characteristics of the Process Information System - Nuclear Power Plant Krsko

    International Nuclear Information System (INIS)

    Mandic, D.; Smolej, M.

    1998-01-01

    process Information System (PIS) of Nuclear Power Plant Krsko (NEK) is newly installed distributed and redundant process computer system which was built in NEK (Phase I: 1991-1995) to integrate the following main functions: - Signal Data Acquisition from the technological processes and environment - Implementation of the basic SCADA functions on the real time process signals data base - Execution of complex plant specific application programs - Advanced MMI (Man Machine Interface) features for users in MCR - Process data transfer to other than Main Control Room (MCR) locations - Process data archiving and capability to retrieve same data for future analysis PIS NEK architecture consists of three hierarchically interconnected hardware platforms: - PIS Level 1, DAS (Data Acquisition System) Level - PIS Level2, Level for MMI, application programs and process data archiving - PIS Level 3, Level for distribution of process data to remote users of PIS data. (author)

  4. Introduction of the commercial grade dedication into Nuclear Power Plant Krsko (NEK) procurement process

    International Nuclear Information System (INIS)

    Heruc, Z.; Gajsak, Z.; Nikpalj, R.

    1996-01-01

    NEK management has undertaken a set of actions to improve the ability to provide equipment, spare parts and material needed to support operation and maintenance of the Krsko plant. These actions are necessary due primarily to the fact that NEK is more and more confronted (increasing trend) with the issue that suppliers of safety-related equipment and spare parts have decided not to pursue the nuclear portion of their business, incl. specific QA systems and qualifications. The purchase orders imposing these requirements are no longer accepted. In order to continue to obtain the necessary materials at the required quality level, a 'Commercial Grade Item' (CGI) procurement and dedication program has been developed based on similar practices in USA. (author)

  5. Dukovany NPP operation

    International Nuclear Information System (INIS)

    Vlcek, Jaroslav

    2010-01-01

    The topics discussed include: Dukovany NPP among CEZ Group power plants; International missions at the plant; Plant operation results; and Strategic goals and challenges. Historical data are presented in the graphical form, such as the unit capacity factor, unplanned capability loss factor, unplanned automatic scrams, fuel reliability, industrial safety accident rate, collective radiation exposure, WANO index, power generation data, and maximum achievable power by the end of year. Also discussed were the company culture and human resources, maintenance, power uprate, and related phenomena. (P.A.)

  6. Cernavoda NPP Knowledge Transfer

    International Nuclear Information System (INIS)

    Valache, C. M.

    2016-01-01

    Full text: The paper presents a description of the Knowledge Transfer (KT) process implemented at Cernavoda NPP, its designing and implementation. It is underlined that applying a KT approach should improve the value of existing processes of the organization through: • Identifying business, operational and safety risks due to knowledge gaps, • Transfer of knowledge from the ageing workforce to the peers and/or the organization, • Continually learning from successes and failures of individual or teams, • Convert tacit knowledge to explicit knowledge, • Improving operational and safety performance through creating both new knowledge and better access to existing knowledge. (author

  7. Operation safety at Ignalina NPP

    International Nuclear Information System (INIS)

    Zheltobriukh, G.

    1999-01-01

    An improvement of operational safety at Ignalina NPP covers: improvement of management structure and safety culture; symptom-based emergency operating procedures; staff training and full scope simulator; program of components ageing; metal inspection; improvement of fire safety. The first plan of Ignalina NPP Safety culture development for 1997 purposed to the SAR recommendation implementation was prepared and approved by the General Director

  8. An Analysis of Cyber-Attack on NPP Considering Physical Impact

    International Nuclear Information System (INIS)

    Lee, In Hyo; Kang, Hyun Gook; Son, Han Seong

    2016-01-01

    Some research teams performed related works on cyber-physical system which is a system that cyber-attack can lead to serious consequences including product loss, damage, injury and death when it is attacked. They investigated the physical impact on cyber-physical system due to the cyber-attack. But it is hard to find the research about NPP cyber security considering the physical impact or safety. In this paper, to investigate the relationship between physical impact and cyber-attack, level 1 PSA results are utilized in chapter 2 and cyber-attack analysis is performed in chapter 3. The cyber security issue on NPP is inevitable issue. Unlike general cyber security, cyber-physical system like NPP can induce serious consequences such as core damage by cyber-attack. So in this paper, to find how hacker can attack the NPP, (1) PSA results were utilized to find the relationship between physical system and cyber-attack and (2) vulnerabilities on digital control systems were investigated to find how hacker can implement the possible attack. It is expected that these steps are utilized when establishing penetration test plans or cyber security drill plans

  9. An Analysis of Cyber-Attack on NPP Considering Physical Impact

    Energy Technology Data Exchange (ETDEWEB)

    Lee, In Hyo; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Son, Han Seong [Joonbu University, Geumsan (Korea, Republic of)

    2016-05-15

    Some research teams performed related works on cyber-physical system which is a system that cyber-attack can lead to serious consequences including product loss, damage, injury and death when it is attacked. They investigated the physical impact on cyber-physical system due to the cyber-attack. But it is hard to find the research about NPP cyber security considering the physical impact or safety. In this paper, to investigate the relationship between physical impact and cyber-attack, level 1 PSA results are utilized in chapter 2 and cyber-attack analysis is performed in chapter 3. The cyber security issue on NPP is inevitable issue. Unlike general cyber security, cyber-physical system like NPP can induce serious consequences such as core damage by cyber-attack. So in this paper, to find how hacker can attack the NPP, (1) PSA results were utilized to find the relationship between physical system and cyber-attack and (2) vulnerabilities on digital control systems were investigated to find how hacker can implement the possible attack. It is expected that these steps are utilized when establishing penetration test plans or cyber security drill plans.

  10. Differences in the licensing requirements for Cernavoda NPP Unit 2 as compared to Cernavoda Unit 1

    International Nuclear Information System (INIS)

    Rotaru, Ioan; Dina, Dumitru; Ghita, Sorin; Stefanescu, Petre

    2000-01-01

    The main stages of Cernavoda NPP Unit 1 and Unit 2 licensing process, according to CNCAN (National Commission for Nuclear Activities) requirements, are presented comparatively. The differences occur for the following licenses: - site license; - construction license; - PIF license, regarding the loading of D 2 O in the moderator and primary circuits, fuel loading, first criticality, power increase; - trial operating license; - operating license. The paper addresses the following items: steps in licensing and the Unit 1 corresponding documentation; - the process of Unit 2 licensing; - requirements to designer; - updating the nuclear safety guides; - editing codes, guides and reference standards and implications on NPP design; - NPP behavior during severe accidents (beyond the design accident consequence); - level 2 and 3 PSA issuing prior to operation licensing; - fulfilling ISO 9000 standard by equipment components already manufactured; improving the warning/display systems in the control room

  11. Prevalence and causes of abnormal PSA recovery.

    Science.gov (United States)

    Lautenbach, Noémie; Müntener, Michael; Zanoni, Paolo; Saleh, Lanja; Saba, Karim; Umbehr, Martin; Velagapudi, Srividya; Hof, Danielle; Sulser, Tullio; Wild, Peter J; von Eckardstein, Arnold; Poyet, Cédric

    2018-01-26

    Prostate-specific antigen (PSA) test is of paramount importance as a diagnostic tool for the detection and monitoring of patients with prostate cancer. In the presence of interfering factors such as heterophilic antibodies or anti-PSA antibodies the PSA test can yield significantly falsified results. The prevalence of these factors is unknown. We determined the recovery of PSA concentrations diluting patient samples with a standard serum of known PSA concentration. Based on the frequency distribution of recoveries in a pre-study on 268 samples, samples with recoveries 120% were defined as suspect, re-tested and further characterized to identify the cause of interference. A total of 1158 consecutive serum samples were analyzed. Four samples (0.3%) showed reproducibly disturbed recoveries of 10%, 68%, 166% and 4441%. In three samples heterophilic antibodies were identified as the probable cause, in the fourth anti-PSA-autoantibodies. The very low recovery caused by the latter interference was confirmed in serum, as well as heparin- and EDTA plasma of blood samples obtained 6 months later. Analysis by eight different immunoassays showed recoveries ranging between PSA which however did not show any disturbed PSA recovery. About 0.3% of PSA determinations by the electrochemiluminescence assay (ECLIA) of Roche diagnostics are disturbed by heterophilic or anti-PSA autoantibodies. Although they are rare, these interferences can cause relevant misinterpretations of a PSA test result.

  12. PSA applications. Good practices and documentation

    International Nuclear Information System (INIS)

    Dewailly, J.; Magne, L.

    1997-10-01

    In this paper, it is shown what the condensed documentation of the main strategic choices and technical assumptions related to a PSA could contain: how to select the internal and external initiating events, how the detail the plant configuration and the general organization of the plant and operating staff, how to highlight the assumptions related to physical models, etc. The proposals in this documentation are based on the R and D D's experience with PSA (construction of PSA models, use of PSA models for operation or maintenance, PSA tools). This document also presents different types of rules or recommendations related to PSA modelling for various applications involved in nuclear power plant operating. Finally, the paper stresses the main difficulties encountered (appropriate use of uncertainties, communication of PSA results to non-specialist users) and it also outlines some prospects for the future. (author)

  13. Re-examining Prostate-specific Antigen (PSA) Density: Defining the Optimal PSA Range and Patients for Using PSA Density to Predict Prostate Cancer Using Extended Template Biopsy.

    Science.gov (United States)

    Jue, Joshua S; Barboza, Marcelo Panizzutti; Prakash, Nachiketh S; Venkatramani, Vivek; Sinha, Varsha R; Pavan, Nicola; Nahar, Bruno; Kanabur, Pratik; Ahdoot, Michael; Dong, Yan; Satyanarayana, Ramgopal; Parekh, Dipen J; Punnen, Sanoj

    2017-07-01

    To compare the predictive accuracy of prostate-specific antigen (PSA) density vs PSA across different PSA ranges and by prior biopsy status in a prospective cohort undergoing prostate biopsy. Men from a prospective trial underwent an extended template biopsy to evaluate for prostate cancer at 26 sites throughout the United States. The area under the receiver operating curve assessed the predictive accuracy of PSA density vs PSA across 3 PSA ranges (10 ng/mL). We also investigated the effect of varying the PSA density cutoffs on the detection of cancer and assessed the performance of PSA density vs PSA in men with or without a prior negative biopsy. Among 1290 patients, 585 (45%) and 284 (22%) men had prostate cancer and significant prostate cancer, respectively. PSA density performed better than PSA in detecting any prostate cancer within a PSA of 4-10 ng/mL (area under the receiver operating characteristic curve [AUC]: 0.70 vs 0.53, P PSA >10 mg/mL (AUC: 0.84 vs 0.65, P PSA density was significantly more predictive than PSA in detecting any prostate cancer in men without (AUC: 0.73 vs 0.67, P PSA increases, PSA density becomes a better marker for predicting prostate cancer compared with PSA alone. Additionally, PSA density performed better than PSA in men with a prior negative biopsy. Copyright © 2017 Elsevier Inc. All rights reserved.

  14. Programmable automation systems in PSA

    International Nuclear Information System (INIS)

    Pulkkinen, U.

    1997-06-01

    The Finnish safety authority (STUK) requires plant specific PSAs, and quantitative safety goals are set on different levels. The reliability analysis is more problematic when critical safety functions are realized by applying programmable automation systems. Conventional modeling techniques do not necessarily apply to the analysis of these systems, and the quantification seems to be impossible. However, it is important to analyze contribution of programmable automation systems to the plant safety and PSA is the only method with system analytical view over the safety. This report discusses the applicability of PSA methodology (fault tree analyses, failure modes and effects analyses) in the analysis of programmable automation systems. The problem of how to decompose programmable automation systems for reliability modeling purposes is discussed. In addition to the qualitative analysis and structural reliability modeling issues, the possibility to evaluate failure probabilities of programmable automation systems is considered. One solution to the quantification issue is the use of expert judgements, and the principles to apply expert judgements is discussed in the paper. A framework to apply expert judgements is outlined. Further, the impacts of subjective estimates on the interpretation of PSA results are discussed. (orig.) (13 refs.)

  15. On-line maintenance at Cofrentes NPP

    International Nuclear Information System (INIS)

    Roldan Vilches, J.; Moreno Matarranz, M. A.; Hermana Mendioroz, I.

    1998-01-01

    Cofrentes NPP has begun in 1997 activities related to At Power Preventive Maintenance over trains or systems which lead to a voluntary entry in a Limitative Condition of Operation (LCO) of the Technical Specifications. From others benefits, this program ha improved the risk management and the staff's knowledge over the functions and safety implications of the different systems, the better exploit of the resources, the co-ordination of the different organisations involved (Maintenance an Operation) and the reductions of works during shutdowns. Previous to each work, a feasibility study analyzes qualitative and quantitative (PSA), using the Risk Monitor, the implications on safety of all the tasks, assuring that the global safety of the Plant is always maintained. Tech. Spec. are analyzed in detail and also are analyzed situations of simultaneous unavailabilities of systems which could lead to a high risk situation. Two different risk controls are defined (punctual and accumulated) to assure that high risk situations will not be given. Finally, historical risk profile is analyzed to assure that the accumulated risk increase is not significant. Risk Monitor helps staff in the schedule and follow-up of the activities of On-Line Maintenance. Each one of the tasks are deeply planned and harshly analyzed and are carried out by high qualified workers. By the moment, this program is running with fully satisfaction on the Plant. (Author)

  16. Instrumentation database specific to Trillo I NPP

    International Nuclear Information System (INIS)

    Pereira Pagan, M.B.; Saenz de Tejada, P.; Fernandez Alvarez, A.; Haya, J.

    1997-01-01

    The analysis of data on electronic instrumentation components in the Trillo I PSA has involved and extra effort, basically due to the particular characteristics of these equipment items. This analysis has different aspects depending on the type of information used: Components whose data have been obtained from generic information sources (with or without Bayesian processing). Components whose data have been obtained from specific German studies (TUV) Components whose data have been based directly on the historical experience of Trillo I NPP Components whose data have been based on miscellaneous generic and specific sources This information can also be classified into: Micro components formed by a single module ar card Micro components: formed by set of instrumentation elements It can be further subdivided according to the operating conditions of the components: Equipment whose operation depends on the functions they perform in a particular system (eg. reactor protection system instrumentation channels) Equipment whose operation is not associated with particular conditions (eg. modules for motor-operated equipment). (Author)

  17. Turbine Control System Replacement at NPP NEK; System Specifics, Project Experience and Lessons Learned

    International Nuclear Information System (INIS)

    Mandic, D.; Zilavy, M. J.

    2010-01-01

    The main intention of this paper is to present feedback from the implementation of the new Turbine Control System (TCS) replacement project at Nuclear Power Plant (NPP) NEK - Krsko. From the plant construction time and the first plant start-up in 1981, the NPP NEK TG (Turbine-Generator) set was controlled and monitored by DEH (Digital Electro Hydraulic) Mod II Control System designed in 70's based on P2500 CPU and number of I/O controllers and modules. The P2500 CPU and associated controllers were built with discrete TTL components (TTL logic chips) and the P2500 CPU had 64k of 16 bit words of ferrite core memory. For that time, DEH Mod II had sophisticated MCR (Main Control Room) HMI (Human Machine Interface) based on digital functional keyboards, one alphanumeric black and white CRT monitor and printer. After twenty eight years of operation and because of several other reasons that are explained in the paper, NEK decided to replace the old DEH Mod II Control system with the new Emerson Ovation based DCS (Distributed Control System) on redundant platform for the control and monitoring of secondary plant systems in the NPP Krsko (NEK), and the new system was named PDEH (Programmable Digital Electro Hydraulic) TCS. In May 2007, NEK signed the turn-key contract with Westinghouse Electric Company (WEC) for the project of replacement of the TCS, Turbine Emergency Trip System (ETS), Moisture Separator Reheater (MSR) control and some other control and monitoring functions. WEC subcontracted a number of other companies for equipment delivery, AE (Architect Engineering Design) activities, specific software development tasks (changes of KFSS - Krsko Full Scope Simulator and PIS - Process Information System interface) and field installation activities. The subject project enveloped implementation of PDEH system on three application platforms: BG KFSS (Background KFSS), FG KFSS (Foreground KFSS) and PDEH system installed in the plant. The HMI for the BG KFSS platform

  18. Numerical Analysis of Loss of Residual Heal Removal System (RHRS) during Mid-Loop Operation for Hanul NPP Units 1 and 2

    International Nuclear Information System (INIS)

    Kim, Sook Kwan; Park, Seong Gyu; Han, Sang Koo

    2016-01-01

    As a part of supporting LPSD (Low Power and Shutdown) PSA (Probabilistic Safety Assessment) of Hanul NPP units 1 and 2, numerical analysis for a loss of RHRS (Residual Heat Removal system) during midloop operation was performed using RELAP5/MOD3.3 code. The one of main purpose of thermal hydraulic analysis for PSA work is to estimate times allowable for operation actions in each accident. A loss of RHRS during mid-loop operation may cause more significant results than during RCS full condition due to reduced RCS inventory. In order to perform this kind of analysis, it is particularly important to establish a steady state of mid-loop operational initial condition. Mid-loop operation corresponds to POS(Plant Operational State) 5 and 11 in the category of LPSD PSA at Hanul NPP units 1 and 2. RELAP5/MOD3.3 code was used to predict behaviors of RCS and fuels for the case of loss of RHRS during mid-loop operation at Hanul NPP units 1 and 2. The initial state of mid-loop operational condition was established by proper control of charging and letdown flow. Considering existing similar analysis results for this kind of accident, it can be concluded that RELAP5 code well predicts reasonably the behavior of RCS for loss of RHRS during mid-loop operation in Hanul NPP units 1 and 2. Thus the method developed in the analysis can be applied reasonably to support LPSD PSA

  19. Qualification of calculation aids for PSA

    International Nuclear Information System (INIS)

    Goetz, K.; Hennigs, W.; Kirstein, B.M.; Reinhardt, C.

    1998-01-01

    In Germany Probabilistic Safety Analysis (PSA) are part of the evaluation of a nuclear power plants safety. The German PSA guide stipulates that the used software must enable a PSA according to the state of the art. This software must be qualified to assure that its features, mathematic methods and its performance enable a PSA like this. In this research work specifications and requirements are developed, which allow the testing of software. A procedure was developed to qualify PSA software according to the PSA guide and the experiences of users of PSA. Setting up a procedure, a tool for a systematic and uniform examination was crated. Additionally the options, mathematic fundamentals and performance of PSA-programs were analyzed. According to this all programs that were analyzed are capable to sovle their original task, that is the calculation of the safety of high available system based on high available components. Against that the requirements of modern PSA, e.g. to handle less available functions, HRA and fire analyses, based on the use of modern software and the implementation of new developments in the field of PSA are not supported adequately by all programs. (orig.) [de

  20. THE DISCRIMINATIVE ABILITY OF PERCENT FREE PSA IN ...

    African Journals Online (AJOL)

    Blood samples were collected from all patients, and total PSA, free PSA and % free PSA were calculated in all specimens. Total PSA was measured using the Imx ... Un prélèvement de sang a été réalisé chez tous les patients avec un dosage du PSA total et de la fraction libre de PSA. Le PSA total a été mesuré par un kit ...

  1. Importance of the multi-modules study in PSA; Importancia del estudio de multi-modulos en APS

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez R, V. J.; Nelson E, P. F., E-mail: judith_gonzalez_rodriguez@outlook.es [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac 8532, 62550 Jiutepec, Morelos (Mexico)

    2015-09-15

    The current approach that has taken the Probabilistic Safety Analysis (PSA) consists of doing all the APS analysis including the existence of multi-units in the nuclear power plants (NPP), this new approach seeks to analyze the risk of site, evaluating all reactors together. The main reasons for this trend are: the accident occurred on March 2011 in Fukushima Daiichi in Japan, with serious consequences in more than one reactor of the NPP and the current planning and construction of new Small Modular Reactors, which host more than one module on the same NPP and are connected to a single control room. This study analyzes how to model the risk of a multi-module NPP. In 2013, the ASME/ANS standard for advanced reactors that are not light-water reactors was published, in which the requirements to realize a PSA including multi-units or modules are shown; however, does not describe the methodology to do that. This article presents a methodology to calculate the risk of the site in a PBMR plant with two modules. This methodology consists of two models of trees of different events, one that evaluates to a single PBMR module and another that evaluates the two modules together. Both models are responsible to show their differences and compare results to finally demonstrate the need for new methodologies for risk analysis site in multi-modules and units. (Author)

  2. The seismic reassessment Mochovce NPP

    International Nuclear Information System (INIS)

    Baumeister, P.

    2004-01-01

    The design of Mochovce NPP was based on the Novo-Voronez type WWER-440/213 reactor - twin units. Seismic characteristic of this region is characterized by very low activity. Mochovce NPP site is located on the rock soil with volcanic layer (andesit). Seismic reassessment of Mochovce NPP was done in two steps: deterministic approach up to commissioning confirmed value Horizontal Peak Ground Acceleration HPGA=0.1 g and activities after commissioning as a consequence of the IAEA mission indicate higher hazard values. (author)

  3. Workflow in Almaraz NPP

    International Nuclear Information System (INIS)

    Gonzalez Crego, E.; Martin Lopez-Suevos, C.

    2000-01-01

    Almaraz NPP decided to incorporate Workflow into its information system in response to the need to provide exhaustive follow-up and monitoring of each phase of the different procedures it manages. Oracle's Workflow was chosen for this purpose and it was integrated with previously developed applications. The objectives to be met in the incorporation of Workflow were as follows: Strict monitoring of procedures and processes. Detection of bottlenecks in the flow of information. Notification of those affected by pending tasks. Flexible allocation of tasks to user groups. Improved monitoring of management procedures. Improved communication. Similarly, special care was taken to: Integrate workflow processes with existing control panels. Synchronize workflow with installation procedures. Ensure that the system reflects use of paper forms. At present the Corrective Maintenance Request module is being operated using Workflow and the Work Orders and Notice of Order modules are about to follow suit. (Author)

  4. Incorporating Level-2 PSA Feature of CONPAS into AIMS-PSA Software

    Energy Technology Data Exchange (ETDEWEB)

    Han, Sang Hoon; Lim, Hogon; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    CONPAS (CONtainment Performance Analysis System) utilizes a methodology to treat containment phenomena in detail like APET but in simple way. In mid 2000's, KAERI has developed very fast cut set generator FTREX and PC's OS (Operating system) has changed into Windows 95. Thus, KAERI has developed new Level-1 PSA software, called AIMS-PSA (Advanced Information Management System for PSA) to replace KIRAP. Recently, KAERI has been developing an integrated PSA platform, called OCEANS (On-line Consolidator and Evaluator of All mode risk for Nuclear System), for the risk assessment of all power modes and all hazards. CONPAS for Level-2 PSA was developed in 1990's using the Visual Basic 6.0 compiler which is not supported any more. It needs to be updated for the integrated PSA software framework. This paper describes a study to incorporate the features of CONPAS into AIMS-PSA. The basic idea is to follow the approach of CONPAS, but in the integrated way. Various approaches for Level-2 PSA have been used since WASH-1400. APET approach of NUREG-1150 study would be most comprehensive and complex methodology for containment event tree analysis. CONPAS is the Level-2 PSA software to utilize an approach to treat containment phenomena in detail like APET but in simple way. But, new Level-2 PSA software is required to develop more integrated PSA framework. A modified approach of CONPAS is developed and incorporated in AIMS-PSA software that can handle Level-1 and Level-2 PSA in the integrated way (from the viewpoint of event tree and fault tree). AIMS-PSA combines whole Level-2 PSA model to produce a One Top fault tree and to generate cut sets in the same way as Level-1 PSA. Quantification results of Level-2 PSA such as frequency for each STC can be calculated from the minimal cut sets.

  5. Opium consumption is negatively associated with serum prostate-specific antigen (PSA), free PSA, and percentage of free PSA levels.

    Science.gov (United States)

    Safarinejad, Mohammad Reza; Asgari, Seyyed Alaeddin; Farshi, Alireza; Iravani, Shahrokh; Khoshdel, Alireza; Shekarchi, Babak

    2013-01-01

    Addiction to opium continues to be a major worldwide medical and social problem. The study addressing the association between opium consumption and serum prostate-specific antigen (PSA) level is lacking. We determined the effects of opium consumption on serum PSA levels in opium-addict men. Our study subjects comprised 438 opium-addict men with a mean age of 52.2 ± 6.4 years (group 1). We compared these men with 446 men who did not indicate current or past opium use (group 2). Serum total PSA (tPSA), free PSA (fPSA), % fPSA, and sex hormones were compared between the 2 groups. The mean serum tPSA level was significantly lower in group 1 (1.05 ng/mL) than in controls (1.45 ng/mL) (P = 0.001). Opium consumption was also associated with lower fPSA (P = 0.001) and % fPSA (P = 0.001). Serum free testosterone level in opium-addict patients (132.5 ± 42 pg/mL) was significantly lower than that in controls (156.2 ± 43 pg/mL) (P = 0.03). However, no significant correlation existed between tPSA and free testosterone levels (r = 0.28, 95% CI, -0.036 to 0.51, P = 0.34). Among the patients with cancer in group 1, 35% were found to have high-grade tumor (Gleason score ≥ 7) compared with 26.7% in group 2 (P = 0.02). Total PSA and fPSA were strongly correlated with duration of opium use (r = -0.06, 95% CI, -0.04 to -0.08, P = 0.0001; and r = -0.05, 95% CI, -0.03 to -0.07, P = 0.0001, respectively). Opium consumption is independently and negatively associated with serum tPSA, fPSA, and % fPSA levels.

  6. Antigenic determinants of prostate-specific antigen (PSA) and development of assays specific for different forms of PSA.

    OpenAIRE

    Nilsson, O.; Peter, A.; Andersson, I.; Nilsson, K.; Grundstr?m, B.; Karlsson, B.

    1997-01-01

    Monoclonal antibodies were raised against prostate-specific antigen (PSA) by immunization with purified free PSA, i.e. not in complex with any protease inhibitor (F-PSA) and PSA in complex with alpha1-anti-chymotrypsin (PSA-ACT). Epitope mapping of PSA using the established monoclonal antibody revealed a complex pattern of independent and partly overlapping antigenic domains in the PSA molecule. Four independent antigenic domains and at least three partly overlapping domains were exposed both...

  7. IMPROVEMENT OF THE LOCA PSA MODEL USING A BEST-ESTIMATE THERMAL-HYDRAULIC ANALYSIS

    Directory of Open Access Journals (Sweden)

    DONG HYUN LEE

    2014-08-01

    Full Text Available Probabilistic Safety Assessment (PSA has been widely used to estimate the overall safety of nuclear power plants (NPP and it provides base information for risk informed application (RIA and risk informed regulation (RIR. For the effective and correct use of PSA in RIA/RIR related decision making, the risk estimated by a PSA model should be as realistic as possible. In this work, a best-estimate thermal-hydraulic analysis of loss-of-coolant accidents (LOCAs for the Hanul Nuclear Units 3&4 is first carried out in a systematic way. That is, the behaviors of peak cladding temperature (PCT were analyzed with various combinations of break sizes, the operating conditions of safety systems, and the operator's action time for aggressive secondary cooling. Thereafter, the results of the thermal-hydraulic analysis have been reflected in the improvement of the PSA model by changing both accident sequences and success criteria of the event trees for the LOCA scenarios.

  8. Development and perspectives of PSA in Cuba

    International Nuclear Information System (INIS)

    1996-01-01

    During the last decade the GDA/PSA has carried out the pre-operational PSA task for the Juragua Nuclear Power Plant. Since 1991 the work has been accomplished in the frames of the IAEA Technical Assistance Project CUB/9/008. The paper describes the stages of this study, (concluding with the Final Report of the pre-operational Level 1 PSA Rev. O), its assumptions, limitations and the main results and concluding remarks

  9. Selection of NPP for Kazakhstan

    International Nuclear Information System (INIS)

    Zhotabaev, Zh.R.

    2003-01-01

    Commercial NPP for Kazakhstan should to meet to several main requirements: 1). Safety operation (accident probability not more than 10 -6 1/p. year). 2). High efficiency > 40 %. 3). Possibility of use for high-temperature chemistry and hydrogen production. 4). Possibility for manufacturing of considerable part of equipment in Kazakhstan. 5). Possibility for fuel production and reprocessing in Kazakhstan. 6). Independence from existence of large water-supply sources. Comparative analysis of several NPP with different reactors (WWR-1000, Candu, BREST, VG-400; graphite molten salt reactor) shows that NPP with the graphite molten salt reactor meets to all above requirements, but hydrogen production it is possible by more complete 4-stage technology, since coolant temperature is 800 Deg. C. The principle advantage is possibility of manufacturing of main equipment and fuel in Kazakhstan that reduce the cost of NPP construction and operation

  10. Psychology of NPP operation safety

    International Nuclear Information System (INIS)

    Tret'yakov, V.P.

    1993-01-01

    The book is devoted to psychologic investigations into different aspects of NPP operative personnel activities. The whole set of conditions on which successful and accident-free personnel operation depends, is analysed. Based on original engineering and socio-psychologic investigations complex psychologic support for NPP personnel and a system of training and upkeep of operative personnel skills are developed. The methods proposed have undergone a practical examination and proved their efficiency. 154 refs., 12 figs., 9 tabs

  11. Status and use of PSA in Sweden

    Energy Technology Data Exchange (ETDEWEB)

    Knochenhauer, M.

    1996-05-01

    The performance and use of PSA:s in Sweden goes back about two decades. During all of this time, the field of PSA has been developing intensively, both internationally and within Sweden. The latest years have been characterised by an increased use of PSA models and results, and by major extensions of existing PSA models. The aim of this document is to describe PSA in Sweden with respect to development, scope and maturity, as well as to the contents of the analyses and the use of results. PSA activities will be described from the point of view of both the authorities and the utilities. The report gives an overview of the development within the area of PSA in Sweden both its history and current trends. The aim has been to include a reasonable amount of detail, both on the methods and results in PSA:s performed and on the numerous supporting research programs dealing with various aspects of PSA. 39 refs 39 refs.

  12. Model engineering in a modular PSA

    International Nuclear Information System (INIS)

    Friedlhuber, Thomas

    2014-01-01

    For the purpose of PSA (Probabilistic Safety Analysis) for complex industrial systems, often PSA models in the form of fault and event trees are developed to model the risk of unwanted situations (hazards). While the recent decades, PSA models have gained high acceptance and have been developed massively. This lead to an increase in model sizes and complexity. Today, PSA models are often difficult to understand and maintain. This manuscript presents the concept of a modular PSA. A modular PSA tries to cope with the increased complexity by the techniques of modularization and instantiation. Modularization targets to treat a model by smaller pieces (the 'modules') to regain control over models. Instantiation aims to configure a generic model to different contexts. Both try to reduce model complexity. A modular PSA proposes new functionality to manage PSA models. Current model management is rather limited and not efficient. This manuscript shows new methods to manage the evolutions (versions) and deviations (variants) of PSA models in a modular PSA. The concepts of version and variant management are presented in this thesis. In this context, a model comparison and fusion of PSA models is precised. Model comparison provides important feedback to model engineers and model fusion kind of combines the work from different model engineers (concurrent model engineering). Apart from model management, methods to understand the content of PSA models are presented. The methods focus to highlight the dependencies between modules rather than their contents. Dependencies are automatically derived from a model structure. They express relations between model objects (for example a fault tree may have dependencies to basic events). To visualize those dependencies (for example in form of a model cartography) can constitute a crucial aid to model engineers for understanding complex interrelations in PSA models. Within the scope of this thesis, a software named 'Andromeda' has been

  13. Too Much Sodium PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2012-02-07

    This 60 second PSA is based on the February 2012 CDC Vital Signs report. Ninety percent of Americans age two and older eat too much sodium which can increase your risk for high blood pressure and often leads to heart disease and stroke, two leading causes of death in the US. Learn several small steps you can take to reduce the amount of sodium in your diet.  Created: 2/7/2012 by Centers for Disease Control and Prevention (CDC).   Date Released: 2/7/2012.

  14. NPP Decommissioning: the concept; state of activities

    International Nuclear Information System (INIS)

    Nemytov, S.; Zimin, V.

    2001-01-01

    The main principles of NPP decommissioning concept in Russia are given. The conditions with fulfillment of works on NPP unit pre-decommissioning and decommissioning including: development of the normative documentation, creation of special fund for financing NPP decommissioning activities, deriving the Gosatomnadzor license for decommissioning of shut down NPP units, development of the equipment and technologies for waste and spent fuel management are presented. The decommissioning cost and labour intensity of one WWER-440 unit are shown. The practical works, executed on shut down units at Beloyarsk NPP (Unit1 and 2) and Novo Voronezh NPP (Unit 1 and 2) are outlined

  15. PSA-based evaluation and rating of operational events

    International Nuclear Information System (INIS)

    Gomez Cobo, A.

    1997-01-01

    The presentation discusses the PSA-based evaluation and rating of operational events, including the following: historical background, procedures for event evaluation using PSA, use of PSA for event rating, current activities

  16. Impact of total PSA, PSA doubling time and PSA velocity on detection rates of 11C-Choline positron emission tomography in recurrent prostate cancer

    NARCIS (Netherlands)

    Rybalov, Maxim; Breeuwsma, Anthonius J.; Leliveld, Anna M.; Pruim, Jan; Dierckx, Rudi A.; de Jong, Igle J.

    PURPOSE: To evaluate the effect of total PSA (tPSA) and PSA kinetics on the detection rates of (11)C-Choline PET in patients with biochemical recurrence (BCR) after radical prostatectomy (RP) or external beam radiotherapy (EBRT). METHODS: We included 185 patients with BCR after RP (PSA >0.2 ng/ml)

  17. The dependence level analysis between the human actions in NPP Operation

    International Nuclear Information System (INIS)

    Farcasiu, M.; Nitoi, M.; Apostol, M.; Florescu, G.; Prisecaru, Ilie

    2009-01-01

    The Human Reliability Analysis (HRA) is an important method in Probabilistic Safety Assessment (PSA) studies and offers desirability for concrete improvement of the man - machine - organization interfaces, reliability and safety. An important step in HRA is the dependence level analysis between the human actions performed by the same person or between the actions performed by different persons, step in quantitative analysis of the human errors probabilities. The purpose of this paper is to develop a model to analyze the dependence level between human actions for Nuclear Power Plant (NPP) operation. The model estimates the conditional human error probabilities (CHEP) and joint human error probabilities (JHEP). The achieved sensitivity analyses determine human performance sensibility to systematic variations for dependence level between human actions. The human error probabilities estimated in this paper are adequate values for integration both in HRA and in PSA realized for NPP. This type of analysis helps in finding and analyzing the ways of reducing the likelihood of human errors, so that the impact of human factor to systems availability, reliability and safety can be realistically estimated. In order to demonstrate the usability of this model an analysis is performed upon the dependences between the necessary human actions in mitigating the consequences of LOCA events, particularly for the case of Cernavoda NPP. (authors)

  18. Human prostate specific antigen (hPSA) purification and establishment of hPSA radioimmunoassay

    International Nuclear Information System (INIS)

    Weiquiang Zhong; Li Chen; Renzhi Wang

    1996-01-01

    Human prostate specific antigen (hPSA) RIA was developed with hPSA and anti-PSA prepared in our laboratory. Its standard curve was linear with a sensitivity of 0.5 μ g/L. Serum PSA levels of 130 normal males ranged from O to 3.5 μ g/L (1.15 ± 0.93 μ g/L), which are consistent with the results of other conventinal RIA. The rcovery, intra- and inter-assay coefficients of variation conform to the demands of RIA, and the results of 41 samples obtained by both the PSA RA and PSA RIA of DPC were well correlated (γ = 0.990). PSA level of 23 patients with prostatic carcinoma was 10 - 400 μ g/L. (author). 8 refs., 3 figs

  19. Spain in South Ukraine NPP

    International Nuclear Information System (INIS)

    Ibanez, M.

    1994-01-01

    A Technical Assistance Protocol was signed between the Governments of the GIS and the Commission of the European Union (CEU) on August 2, 1991 and this was the starting point of the TACIS program. In this article, the activities described are those related to the TACIS-92/93/94 on site technical assistance to South Ukraine NPP (SUK NPP). Within the scope of the TACIS 92 Program the CEU and the Ukrainian Authorities agreed a list of projects to be implemented at South Ukraine NPP with the aim to improve the operational safety of the plant. This part of the program is called TACIS 92 on-site activities. The total budget allocated to these projects is a MECU. The European Union ''utility'' selected to lead this program at South Ukraine NPP was UNESA and the first contract to cover our activities was signed in July 1993 between the CEU (Mr. Pablo Benavides) and UNESA (Mr. Pedro Rivero). The projects will be implemented at SUK NPP but according to the contract UNESA is ''The Consultant'' and GOSKOMATON (The Ukrainian Sate Committee on Nuclear Power Utilization) is the ''Recipient Institution''. (Author)

  20. Emergency preparedness at Ignalina NPP

    International Nuclear Information System (INIS)

    Kairys, A.

    1998-01-01

    Brief review of Ignalina NPP safety upgrading and personnel preparedness to act in cases of accidents is presented. Though great activities are performed in enhancing the plant operation safety, the Ignalina NPP management pays a lot of attention to preparedness for emergency elimination and take measures to stop emergency spreading. A new Ignalina NPP emergency preparedness plan was drawn up and became operational. It is the main document to carry out organizational, technical, medical, evacuation and other activities to protect plant personnel, population, the plant and the environment from accident consequences. Great assistance was rendered by Swedish experts in drawing this new emergency preparedness plan. The plan consists of 3 parts: general part, operative part and appendixes. The plan is applied to the Ignalina NPP personnel, Special and Fire Brigade and also to other contractor organizations personnel carrying out works at Ignalina NPP. There are set the following emergency classes: incident, emergency situation, alert, local emergency, general emergency. Separate intervention level corresponds to each emergency class. Overview of personnel training to act in case of an emergency is also presented

  1. Approach to fire-PSA study of KAPS IPHWR

    International Nuclear Information System (INIS)

    Maiti, S.C.; Chatterjee, D.; Guptan, Rajee; Mohan, Nalini; Gadge, S.S.; Bajaj, S.S.

    2006-01-01

    Some safety significant fire events in nuclear power plants like Browns Ferry (U.S.A, 1975), NAPS (Narora, India, 1993) had serious implication on nuclear safety. Some of them had direct or indirect potential to cause sever reactor core damage. Therefore fire-induced core damage frequencies (FCDF) can be a significant contributor to the overall risk for a plant depending upon the design and operational characteristics of the NPP. Fire probabilistic safety assessment (Fire-PSA) studies for KAPS has been initiated using probabilistic and deterministic analysis like 'Fire Hazard Analysis' to identify the main contributing 'Fire Zone' and 'fire scenarios' to FCDF. The study is being done based on the IAEA guideline and as per international practice for carrying out Fire-PSA. Fire zone are screened out qualitatively and quantitatively based on their direct or indirect potential to Conditional Core Dantage frequency (CCDF). Fire database used for the analysis was for 25 Reactor-years of commercial operation of KAPS. Fire frequency was very low in most of the 'Fire Zone' and most of the fire were detected and suppressed at incipient stage. In most of the cases safety systems were not challenged, and all safety functions were fully available. It gives an insight to fire risk related strength- weaknesses of design and operation of IPHWR. Major contributor to fire risk i.e vulnerable area, vulnerable nuclear safety items, vulnerable fire risk action, and important fire safety components can be identified. It helps the designer and plant management to prioritize their effort in fire risk reduction. (author)

  2. Decision criteria in PSA applications

    International Nuclear Information System (INIS)

    Holmberg, J.E.; Pulkkinen, U.; Rosqvist, T.; Simola, K.

    2001-11-01

    Along with the adoption of risk informed decision making principles, the need for formal probabilistic decision rule or criteria has been risen. However, there are many practical and theoretical problems in the application of probabilistic criteria. One has to think what is the proper way to apply probabilistic rules together with deterministic ones and how the criteria are weighted with respect to each other. In this report, we approach the above questions from the decision theoretic point of view. We give a short review of the most well known probabilistic criteria, and discuss examples of their use. We present a decision analytic framework for evaluating the criteria, and we analyse how the different criteria behave under incompleteness or uncertainty of the PSA model. As the conclusion of our analysis we give recommendations on the application of the criteria in different decision situations. (au)

  3. Regulatory aspects of NPP safety

    International Nuclear Information System (INIS)

    Kastchiev, G.

    1999-01-01

    Extensive review of the NPP Safety is presented including tasks of Ministry of Health, Ministry of Internal Affairs, Ministry of Environment and Waters, Ministry of Defense in the field of national system for monitoring the nuclear power. In the frame of national nuclear safety legislation Bulgaria is in the process of approximation of the national legislation to that of EC. Detailed analysis of the status of regulatory body, its functions, organisation structure, responsibilities and future tasks is included. Basis for establishing the system of regulatory inspections and safety enforcement as well as intensification of inspections is described. Assessment of safety modifications is concerned with complex program for reconstruction of Units 1-4 of Kozloduy NPP, as well as for modernisation of Units 5 and 6. Qualification and licensing of the NPP personnel, Year 2000 problem, priorities and the need of international assistance are mentioned

  4. Safety culture at Mochovce NPP

    International Nuclear Information System (INIS)

    Markus, Jozef; Feik, Karol

    2002-01-01

    This article presents the approach of Mochovce NPP to the Safety culture. It presents activities, which have been taken by Mochovce NPP up to date in the area of Safety culture enhancement with the aim of getting the term into the subconscious of each employee, and thus minimising the human factor impact on occurrence of operational events in all safety areas. The article furthermore presents the most essential information on how the elements characterising a continuous progress in reaching the planned Safety culture goals of the company management have been implemented at Mochovce NPP, as well as the management's efforts to get among the best nuclear power plant operators in this area and to be an example for the others. (author)

  5. NPP construction cost in Canada

    International Nuclear Information System (INIS)

    Gorshkov, A.L.

    1988-01-01

    The structure of capital costs during NPP construction in Canada is considered. Capital costs comprise direct costs (cost of the ground and ground rights, infrastructure, reactor equipment, turbogenerators, electrotechnical equipment, auxiliary equipment), indirect costs (construction equipment and services, engineering works and management services, insurance payments, freight, training, operating expenditures), capital per cents for the period of construction and cost of heavy water storages. It proceeds from the analysis of the construction cost structure for a NPP with the CANDU reactor of unit power of 515, 740 and 880 MW, that direct costs make up on the average 62%

  6. Radioactive wastes management of NPP

    International Nuclear Information System (INIS)

    Klyuchnikov, A.A.; Pazukhin, Eh.M.; Shigera, Yu. M.; Shigera, V.Yu.

    2005-01-01

    Modern knowledge in the field of radiation waste management on example of the most serious man-made accident at Chernobyl NPP are illuminated. This nuclear power plant that after accident in 1986 became in definite aspect an experimental scientific ground, includes all variety of problems which have to be solved by NPP personnel and specialists from scientific organizations. This book is aimed for large sphere of readers. It will be useful for students, engineers, specialists and those working in the field of nuclear power, ionizing source and radiation technology use for acquiring modern experience in nuclear material management

  7. NPP safety and efficiency increasing. Proceedings

    International Nuclear Information System (INIS)

    2011-01-01

    The main topics of the conference are the analysis and increasing of NPP safety; efficiency improvement and extension of service life of NPP; innovation in the field of radiation and ecological safety of NPP; professional education for operating and regulatory organizations.

  8. The clinical study of serum PSA and fPSA assayed by CLIA in diagnosing prostate disease

    International Nuclear Information System (INIS)

    Xiong Jiang; Qian Xiaoyu; Ji Hong; Yang Su; Ding Ying; Zhu Ruisen; Chen Zhong

    2003-01-01

    The purpose of this study is to evaluate the clinical value of PSA (prostate specific antigen) and fPSA(free prostate specific antigen) in differentiating prostate disease. CLIA was used to quantitatively assay PSA, fPSA and fPSA/PSA in 30 cases of normal controls, 32 cases of prostate cancer patients and 76 cases of BPH patients. The result showed that if liminal value of PSA was set at 4 ng/mL, the diagnostic sensitivity and specificity of prostate cancer were 100% and 50.6% respectively. Meanwhile, if liminal value of fPSA/PSA set at 16% was added, the diagnostic sensitivity and specificity of prostate cancer were 100% and 85.3% respectively. It was concluded that the combining assay of PSA and fPSA could increase the diagnostic specificity of prostate cancer in a certain degree

  9. The diagnostic value of PSA, cPSA and bone scintigraphy for early skeletal metastasis of prostate cancer

    International Nuclear Information System (INIS)

    Xue Zhongguang

    2007-01-01

    Objective: To evaluate the value of prostate specific antigen (PSA), complexed prostate specific antigen (cPSA) and bone scintigraphic imaging in diagnosis of early skeletal metastasis of prostate cancer. Methods: 152 patients (74 with prostate cancer, 78 with benign prostate disease) and 90 controls were examined for the serum concentrations of PSA and cPSA. At the same time, the 74 patients with PCa were examined with bone scintigraphy. The cPSA/PSA ratio was calculated. Results: Serum PSA, cPSA levels and cPSA/PSA ratio of patients with prostate cancer were significantly higher than those in benign prostate patients and controls. In addition, the serum PSA, cPSA levels and cPSA/PSA ratio in prostate cancer patients with skeletal metastasis were remarkably higher than those in patients without skeletal metastasis, and the differences were significant (P 20 μg/L, cPSA>10 μg/L, cPSA/PSA>0.80, there is a high probability that skeletal metastasis of prostate cancer would be present and bone scintigraphy should be performed. (authors)

  10. Identifying measures to balance the risk profile of the Tihange 2 NPP

    International Nuclear Information System (INIS)

    D'Eer, A.M.; Monniez, J.J.

    2001-01-01

    In Belgium, each Nuclear Power Plant is subject to a periodic safety reassessment. In this context, it was found to be desirable to perform a Probabilistic Safety Assessment (PSA) in support of the ten yearly back-fitting process. The Tihange 2 NPP is a 3-loop PWR having a thermal capacity of 2905 MW. Analysis of the plant's risk profile shows that implementing feasible measures for improvement of the shutdown risk, would be beneficial. This is because a configuration leading to significant risk, namely cold pressurization when the residual heat removal system is lost during reduced primary inventory, thus can be avoided. As a result the risk between reactor shutdown and power operation will be balanced. The presentation describes the lessons learnt regarding the Tihange 2 shutdown PSA model and the expected benefits following implementation of one of the proposed measures. (author)

  11. Qualification of NPP operations personnel

    International Nuclear Information System (INIS)

    Wang Jiao.

    1987-01-01

    Competence of personnel is one of the important problems for safety operation of nuclear power plant. This paper gives a description of some aspects, such as the administration of NPP, posts, competence of personnel, training, assessing the competence and personnel management

  12. Safety aspects of NPP ageing

    International Nuclear Information System (INIS)

    1995-01-01

    Preparation of safety practices on assessment and management of aging of major NPP components important to safety, CRP on management of aging of concrete containment buildings, CRP on management of aging of in-containment instrumentation and control cables are outlined

  13. Probability of an Abnormal Screening PSA Result Based on Age, Race, and PSA Threshold

    Science.gov (United States)

    Espaldon, Roxanne; Kirby, Katharine A.; Fung, Kathy Z.; Hoffman, Richard M.; Powell, Adam A.; Freedland, Stephen J.; Walter, Louise C.

    2014-01-01

    Objective To determine the distribution of screening PSA values in older men and how different PSA thresholds affect the proportion of white, black, and Latino men who would have an abnormal screening result across advancing age groups. Methods We used linked national VA and Medicare data to determine the value of the first screening PSA test (ng/mL) of 327,284 men age 65+ who underwent PSA screening in the VA healthcare system in 2003. We calculated the proportion of men with an abnormal PSA result based on age, race, and common PSA thresholds. Results Among men age 65+, 8.4% had a PSA >4.0ng/mL. The percentage of men with a PSA >4.0ng/mL increased with age and was highest in black men (13.8%) versus white (8.0%) or Latino men (10.0%) (PPSA >4.0ng/mL ranged from 5.1% of Latino men age 65–69 to 27.4% of black men age 85+. Raising the PSA threshold from >4.0ng/mL to >10.0ng/mL, reclassified the greatest percentage of black men age 85+ (18.3% absolute change) and the lowest percentage of Latino men age 65–69 (4.8% absolute change) as being under the biopsy threshold (PPSA threshold together affect the pre-test probability of an abnormal screening PSA result. Based on screening PSA distributions, stopping screening among men whose PSA 10ng/ml has the greatest effect on reducing the number of older black men who will face biopsy decisions after screening. PMID:24439009

  14. Main Aspects and Results of Level 2 PSA for KNPP WWER-1000/B320

    International Nuclear Information System (INIS)

    Mancheva, Kaliopa

    2014-01-01

    The PSA Level 2 for Kozloduy NPP (KNPP) is an update of an older study with wider scope of analysis. The older study represented the status of the units up to 2001. The current PSA Level 2 is based on the PSA Level 1 and represents the status of the units up to 2007 year concerning the systems and procedures included in PSA level 1 and status up to 2011 for the systems and procedures (e.g. SAMG) related to containment and severe accident aspects. The study is performed after the PSA level 1 has been finished and approved by the customer. Compare to the older analysis all modes of operation for analyzed in PSA level 1 event groups as well Spent Fuel Pool accidents are investigated. The analysis consists of both deterministic and probabilistic analysis. As part of deterministic analysis a contemporary containment strength analysis and accident progression deterministic analysis using last version of MELCOR are performed. The probabilistic analysis contains of two part: Interface PSA and CET are calculated using Riskspectrum program code. Two types of models for CET have been developed: one for conditional probabilities calculations and a set of simplified CET's for each PDS group-for integral model. The purpose of the first model is to be able to perform quick calculations and for sensitivity analyses as well. The simplified CET's are used for integral calculation of the model. Source Term analysis is mainly based on the MELCOR analyses results. All characteristics of the releases have been defined, i.e. location, mass, energy of radionuclide groups and activity of the released isotopes (most important are reported only). The main goals of the study are to analyze the status of the containment, systems designed to prevent containment failure and operator action required under the severe accident and to give quantitative assessment of the risk parameter LERF (Large Early Release Frequency). This report will present main aspects, results, finding and

  15. PSA-based optimization of technical specifications for the Borssele nuclear power plant

    International Nuclear Information System (INIS)

    Seebregts, A.J.; Schoonakker, H.A.

    1996-01-01

    The Borssele Nuclear Power Plant (NPP) is a Siemens/KWU 472 MWe Pressurized Water Reactor which has been in operation since 1973. In 1989, a Probabilistic Safety Assessment (PSA) program was initiated to complement deterministic safety studies and operational experience in forming a plant safety concept. In 1993, the PSA-MER model was completed and used to determine the effects a package of proposed modifications would have on plant safety and risks to the environment. This model was used to start retrospective risks profile and allowed outage times (AOTs) analyses, which both concerned the calculation of the change in total core damage frequency (TCDF) given a change in configuration. The main problems identified and reported in this paper are: (i) How to calculate the change in TCDF (ΔTCDF)? (section 3); and (ii) How to set practical decision criteria and how to use the PSA as extension to Technical Specifications (TS) AOTs? (section 4). Finally, a pilot study was conducted in order to optimize surveillance test intervals (STIs) which are also part of the TS (section 5). (orig.)

  16. A preliminary investigation of PSA validation methods

    International Nuclear Information System (INIS)

    Unwin, S.D.

    1995-09-01

    This document has been prepared to support the initial phase of the Atomic Energy Control Board's program to review and evaluate Probabilistic Safety Assessment (PSA) studies conducted by nuclear generating station designers and licensees. The document provides (1) a review of current and prospective applications of PSA technology in the Canadian nuclear power industry; (2) an assessment of existing practices and techniques for the review or risk and hazard identification studies in the international nuclear power sector and other technological sectors; and (3) proposed analytical framework in which to develop systematic techniques for the scrutiny and evaluation of a PSA model. These frameworks are based on consideration of the mathematical structure of a PSA model and are intended to facilitate the development of methods to evaluate a model relative to intended end-uses. (author). 34 refs., 10 tabs., 3 figs

  17. Proceedings of the 10th Korea-Japan joint workshop on PSA. For Asian PSA network

    International Nuclear Information System (INIS)

    Yang, Joon-Eon; Homma, Toshimitsu

    2009-12-01

    The tenth Korea-Japan Joint Workshop on Probabilistic Safety Assessment (PSA) was held in the Jeju island of Korea, on May 18-20, 2009 organized by Korea Atomic Energy Research Institute (KAERI). The purpose of the workshop was to provide a forum for presentation and discussions on experiences and technical achievements related to PSA, risk-informed and performance-based approach, and other relevant issues in both countries. Since the first Korea-Japan Joint Workshop on PSA started in 1992, the workshops have provided an important and timely opportunity for exchange and discussion of the relevant information to all PSA practitioners and users of risk information in the industry, research, academia and regulatory arena. This was the tenth anniversary of the Joint Workshop with the main theme of 'For Asian PSA Network' and participants included those from China, Taiwan and the United States of America besides Korea and Japan. Two keynote speeches were presented by the former chairmen of this workshop, Prof. Chang-Sun Kang of Seoul National University and Prof. emeritus Shunsuke Kondo of Tokyo University. We had two special lectures, 70 papers presented by experts at 10 technical sessions related PSA, the special session on the status of PSA in Korea, Japan, China and Taiwan and panel discussion on their cooperation in PSA. This report provides the summary of each session, and all the presentation materials presented in the 10th Korea-Japan Joint Workshop on PSA. (author)

  18. IAEA work with guides for PSA quality

    International Nuclear Information System (INIS)

    Hellstroem, Per

    2004-09-01

    IAEA has a project on development of a TECDOC 'PSA Quality for Various Applications'. The project develops the guidance document in stages with intermediate meetings with exchange of ideas, thoughts and experience. Draft versions are being produced successively. The objective with the project is to use attributes to describe the quality of different elements of a PSA (Analysis of initiating events, accident progression, system, data, human reliability, etc) making the PSA suitable for application in various risk informed activities. Two of the meetings in this project took place in February 2004 and in July 2004. The February meeting discussed different aspects of PSA quality in relation to applications and a draft of the TECDOC was reviewed. The meeting made recommendations for preparation of a final document and set priorities for further work in the area. The July meeting elaborated the document further in a small working group and a new draft version was prepared. A final version is expected to be published during 2005. The project has come to the conclusion that it is a limited number of PSA element attributes that are specific for a certain application. Most of the attributes concern plant specificity, realism and level of detail in a general manner, how plant specific is the model, how realistic and how detailed? Many attributes have the characteristic that they are good to have, but not necessarily needed to do the job. This last statement is valid both for a baseline PSA and a PSA application. The IAEA project has identified a limited number of attributes that are necessary to describe characteristics needed for specific applications. The PSA scope needed for a specific application is not covered by the project/document, even though it is obvious that different applications will need different scope or approaches to handle scope limitations. The guidance on performing a PSA available today is old. It is a need to review these guides and update with regard

  19. Development of multipurpose regulatory PSA model

    International Nuclear Information System (INIS)

    Lee, Chang Ju; Sung, Key Yong; Kim, Hho Jung; Yang, Joon Eon; Ha, Jae Joo

    2004-01-01

    Generally, risk information for nuclear facilities comes from the results of Probabilistic safety assessment (PSA). PSA is a systematic tool to ensure the safety of nuclear facilities, since it is based on thorough and consistent application of probability models. In particular, the PSA has been widely utilized for risk-informed regulation (RIR), including various licensee-initiated risk-informed applications (RIA). In any regulatory decision, the main goal is to make a sound safety decision based on technically defensible information. Also, due to the increased public requests for giving a safety guarantee, the regulator should provide the visible means of safety. The use of PSA by the regulator can give the answer on this problem. Therefore, in order to study the applicability of risk information for regulatory safety management, it is a demanding task to prepare a well-established regulatory PSA model and tool. In 2002, KINS and KAERI together made a research cooperation to form a working group to develop the regulatory PSA model - so-called MPAS model. The MPAS stands for multipurpose probabilistic analysis of safety. For instance, a role of the MPAS model is to give some risk insights in the preparation of various regulatory programs. Another role of this model is to provide an independent risk information to the regulator during regulatory decision-making, not depending on the licensee's information

  20. PSA Isoforms' Velocities for Early Diagnosis of Prostate Cancer.

    Science.gov (United States)

    Heidegger, Isabel; Klocker, Helmut; Pichler, Renate; Horninger, Wolfgang; Bektic, Jasmin

    2015-06-01

    Free prostate-specific antigen (fPSA) and its molecular isoforms are suggested for enhancement of PSA testing in prostate cancer (PCa). In the present study we evaluated whether PSA isoforms' velocities might serve as a tool to improve early PCa diagnosis. Our study population included 381 men who had undergone at least one ultrasound-guided prostate biopsy whose pathologic examination yielded PCa or showed no evidence of prostatic malignancy. Serial PSA, fPSA, and proPSA measurements were performed on serum samples covering 7 years prior to biopsy using Beckmann Coulter Access immunoassays. Afterwards, velocities of PSA (PSAV), fPSA% (fPSA%V), proPSA% (proPSA%V) and the ratio proPSA/PSA/V were calculated and their ability to discriminate cancer from benign disease was evaluated. Among 381 men included in the study, 202 (53%) were diagnosed with PCa and underwent radical prostatectomy at our Department. PSAV, fPSA%V, proPSA%V as well as proPSA/PSA/V were able to differentiate significantly between PCa and non-cancerous prostate. The highest discriminatory power between cancer and benign disease has been observed two and one year prior to diagnosis with all measured parameters. Among all measured parameters, fPSA%V showed the best cancer specificity of 45.3% with 90% of sensitivity. In summary, our results highlight the value of PSA isoforms' velocity for early detection of PCa. Especially fPSA%V should be used in the clinical setting to increase cancer detection specificity. Copyright© 2015 International Institute of Anticancer Research (Dr. John G. Delinassios), All rights reserved.

  1. Proposal for a standardized PSA doubling-time calculation.

    Science.gov (United States)

    Ponholzer, Anton; Popper, Nikolaus; Breitenecker, Felix; Schmid, Hans-Peter; Albrecht, Walter; Loidl, Wolfgang; Madersbacher, Stephan; Schramek, Paul; Semjonow, Axel; Rauchenwald, Michael

    2010-05-01

    Prostate-specific antigen (PSA) doubling-time (PSA-DT) is an important indicator of progression and survival in men with prostate cancer. Three major limitations regarding PSA-DT determination may lead to inconsistent results: the variety of mathematical methods currently applied, the non-standardized handling of input variables and the potential lack of accuracy due to PSA variability. The aim of this project was to develop a reproducible PSA-DT determination tool which simultaneously provides a PSA-DT error estimation. An internet-based PSA-DT calculation tool via nonlinear optimization implementing the least squares error method using the most recent three PSA values was developed. PSA-DT calculation error is estimated via randomly disturbed measurement data streams (n=65) based on a variable (5-25%) PSA variability. According to a simulation in five men, PSA-DT was calculated to be between 1.7 and 15 month (mean: 6.3 month) and determined with another standard tool between 1.3 and 14.5 month (mean: 4.2 month). We present a defined, open and reproducible PSA-DT calculation and PSA-DT error estimation tool based on a standardized PSA data input. This tool is not better compared to other methods but is scientifically standardized and freely accessible via the following internet address: http://adam.drahtwarenhandlung.at/webapp/mg2008/chapter_prostata4/example_psa.

  2. SERIAL PERCENT-FREE PSA IN COMBINATION WITH PSA FOR POPULATION-BASED EARLY DETECTION OF PROSTATE CANCER

    Science.gov (United States)

    Ankerst, Donna Pauler; Gelfond, Jonathan; Goros, Martin; Herrera, Jesus; Strobl, Andreas; Thompson, Ian M.; Hernandez, Javier; Leach, Robin J.

    2016-01-01

    PURPOSE To characterize the diagnostic properties of serial percent-free prostate-specific antigen (PSA) in relation to PSA in a multi-ethnic, multi-racial cohort of healthy men. MATERIALS AND METHODS 6,982 percent-free PSA and PSA measures were obtained from participants in a 12 year+ Texas screening study comprising 1625 men who never underwent biopsy, 497 who underwent one or more biopsies negative for prostate cancer, and 61 diagnosed with prostate cancer. Area underneath the receiver-operating-characteristic-curve (AUC) for percent-free PSA, and the proportion of patients with fluctuating values across multiple visits were determined according to two thresholds (under 15% versus 25%) were evaluated. The proportion of cancer cases where percent-free PSA indicated a positive test before PSA > 4 ng/mL did and the number of negative biopsies that would have been spared by percent-free PSA testing negative were computed. RESULTS Percent-free PSA fluctuated around its threshold of PSA tested positive earlier than PSA in 71.4% (34.2%) of cancer cases, and among men with multiple negative biopsies and a PSA > 4 ng/mL, percent-free PSA would have tested negative in 31.6% (65.8%) instances. CONCLUSIONS Percent-free PSA should accompany PSA testing in order to potentially spare unnecessary biopsies or detect cancer earlier. When near the threshold, both tests should be repeated due to commonly observed fluctuation. PMID:26979652

  3. Evaluation of total PSA assay on vitros ECi and correlation with Kryptor-PSA assay.

    Science.gov (United States)

    Cassinat, B; Wacquet, M; Toubert, M E; Rain, J D; Schlageter, M H

    2001-01-01

    An increasing number of multiparametric immuno-analysers for PSA assays are available. As different immuno-assays may vary in their analytical quality and their accuracy for the follow-up of patients, expertise is necessary for each new assay. The PSA assay on the Vitros-ECi analyser has been evaluated and compared with the PSA assay from the Kryptor analyser. Variation coefficients were 0.91 to 1.98% for within-run assays, and 4.2% to 5.4% for interassay (PSA levels = 0.8 microgram/L to 33.6 micrograms/L). Dilution tests showed 93 to 136% recovery until 70 micrograms/L PSA. Functional sensitivity was estimated at 0.03 microgram/L. Equimolarity of the test was confirmed. Correlation of PSA levels measured with Vitros-ECi and Kryptor analysers displayed a correlation coefficient r2 of 0.9716. The half-lives and doubling times of PSA were similar using both methods. Vitros-ECi PSA assay meets the major criteria for the management of prostate cancer patients.

  4. Purification of PSA from human semen

    International Nuclear Information System (INIS)

    Venkatesh, M.

    1997-01-01

    Full text: 1. Human seminal plasma collected from many volunteers are pooled and passed through a column of phenyl sepharose equilibrated with 1.25 M ammonium sulphate. Elution is carried out with 1.25 M ammonium sulphate initially, to remove the bulk non-adsorbing proteins. Gradient elution of the absorbed proteins with 0.01 M Tris-HCl, 0.25 M NaCl, pH 7.0 buffer gives a sharp peak containing PSA. At each stage, PSA has to be identified by an independent method such as immunodiffusion or an immunoassay. 2. The absorbed protein peak containing PSA is then lyophilised, redissolved in Tris-HCl buffer and chromatographed in a Superdex-75 or Sephadex-75 column. The absorbed proteins elute out as multiple peaks and PSA is eluted as a sharp peak.At each stage, PSA has to be identified by an independent method such as immunodiffusion or an immunoassay. 3. Step 2 is repeated for better purity. 4. The PSA peak is lyophilised, dissolved in Tris-HCl buffer without NaCl and further purified on an ion exchange column (either anion or cation exchange columns such as DEAE Sephadex or CM-Sephadex or Mono Q). Gradient elution using Tris-HCl buffer without NaCl and Tris-HCl buffer with 0.25 M NaCl resulted in a sharp pure PSA peak (homogenous, sharp single band on SDS-PAGE). This procedure is based on that reported by Wang et al., Oncology, 39,1,1982

  5. Retrofitting of NPP Computer systems

    International Nuclear Information System (INIS)

    Pettersen, G.

    1994-01-01

    Retrofitting of nuclear power plant control rooms is a continuing process for most utilities. This involves introducing and/or extending computer-based solutions for surveillance and control as well as improving the human-computer interface. The paper describes typical requirements when retrofitting NPP process computer systems, and focuses on the activities of Institute for energieteknikk, OECD Halden Reactor project with respect to such retrofitting, using examples from actual delivery projects. In particular, a project carried out for Forsmarksverket in Sweden comprising upgrade of the operator system in the control rooms of units 1 and 2 is described. As many of the problems of retrofitting NPP process computer systems are similar to such work in other kinds of process industries, an example from a non-nuclear application area is also given

  6. ASSET experience at Paks NPP

    International Nuclear Information System (INIS)

    Szabo, I.

    1997-01-01

    At Paks NPP special attention has been paid to international reviews since the very beginning of operation. Several international teams visited Paks in order to provide independent assessment of plant performance, conditions and safety. Paks NPP Management has the further intention to invite international reviews regularly (yearly) in the future as well. The experience gained during these reviews helped to establish a unified process of preparation for the reviews, performing them and handling the results. The Safety Department is in charge of organization of the whole process. All these reviews have their specific features and they are focused on different areas. The ASSET reviews provides the assessment of plant performance and safety through the analysis of safety significant events, which have occurred at the nuclear power plant. This approach makes this review specific and different from the other ones

  7. Vergleichende Einschätzung der diagnostischen Aussagekraft der Kenngrößen freies PSA, Alpha1-Antichymotrypsin-PSA und komplexiertes PSA in der Diagnostik des Prostatakarzinoms

    OpenAIRE

    Baumgart E; Deger S; Jung K; Lein M; Loening SA; Schnorr D

    2001-01-01

    Ziel der Studie war die vergleichende Einschätzung der diagnostischen Aussagekraft von Gesamt-PSA (tPSA), freiem PSA (fPSA), alpha1-Antichymotrypsin-PSA (ACT-PSA) und komplexiertem PSA (cPSA) sowie der entsprechenden Quotienten zur Differenzierung zwischen einem Prostatakarzinom (PCa) und einer Benignen Prostatahyperplasie (BPH). Die Bestimmung erfolgte bei insgesamt 324 Männern (PCa: n = 144; BPH: n = 89; Kontrollen: n = 91). Die tPSA- und cPSA-Konzentrationen wurden mit dem Bayer Immuno 1 S...

  8. Can PSA Reflex Algorithm be a valid alternative to other PSA-based prostate cancer screening strategies?

    Science.gov (United States)

    Caldarelli, G; Troiano, G; Rosadini, D; Nante, N

    2017-01-01

    The available laboratory tests for the differential diagnosis of prostate cancer, are represented by the total PSA, the free PSA, and the free/total PSA ratio. In Italy most of doctors tend to request both total and free PSA for their patients even in cases where the total PSA doesn't justify the further request of free PSA, with a consequent growth of the costs for the National Health System. The aim of our study was to predict the saving in Euro (due to reagents) and reduction in free PSA tests, applying the "PSA Reflex" algorithm. We calculated the number of total PSA and free PSA exams performed in 2014 in the Hospital of Grosseto and, simulating the application of the "PSA Reflex" algorithm in the same year, we calculated the decrease in the number of free PSA requests and we tried to predict the Euro savings in reagents, obtained from this reduction. In 2014 in the Hospital of Grosseto 25,955 total PSA tests have been performed: 3,631 (14%) resulted greater than 10 ng / ml; 7,686 (29.6%) between 2 and 10 ng / ml; 14,638 (56.4%) lower than 2 ng / ml. The performed free PSA tests were 16904. Simulating the use of "PSA Reflex" algorithm, the free PSA tests would be performed only in cases with total PSA values between 2 and 10 ng / mL with a saving of 54.5% of free PSA exams and of 8,971 euros, only for reagents. Our study showed that the "PSA Reflex" algorithm is a valid alternative leading to a reduction of the costs. The estimated intralaboratory savings, due to the reagents, seem to be modest, however, they are followed by the additional savings due to the other diagnostic processes for prostate cancers.

  9. Licensing of the Ignalina NPP

    International Nuclear Information System (INIS)

    Kutas, S.

    1999-01-01

    Since 1991 State Nuclear Power Safety Inspectorate (VATESI) has regulated Ignalina NPP operation by issuing annual operating permits. Those have been issued following submission of specified documents by the Ignalina NPP that have been reviewed by VATESI. However, according to to the procedures that are now established in the Law on Nuclear Energy and subordinate regulations the use of nuclear energy in the Republic of Lithuania is subject to strict licensing. Therefore a decision about the licence for continued operation of unit 1 should be taken. Licence would be granted by VATESI in cooperation with the Ministry of Health, Ministry of Environment and the institutions of local authorities. Ignalina NPP presented to the VATESI safety analysis report (SAR) with other documents. SAR was made mainly by foreign experts and financed by European Bank for Reconstruction and Development (EBRD). VATESI in this process is supported by western regulators. A special project LAP - Licensing Assistance Project was launched to help VATESI perform licensing according western practices

  10. Insulin promotes cell migration by regulating PSA-NCAM

    International Nuclear Information System (INIS)

    Monzo, Hector J.; Coppieters, Natacha; Park, Thomas I.H.; Dieriks, Birger V.; Faull, Richard L.M.; Dragunow, Mike; Curtis, Maurice A.

    2017-01-01

    Cellular interactions with the extracellular environment are modulated by cell surface polysialic acid (PSA) carried by the neural cell adhesion molecule (NCAM). PSA-NCAM is involved in cellular processes such as differentiation, plasticity, and migration, and is elevated in Alzheimer's disease as well as in metastatic tumour cells. Our previous work demonstrated that insulin enhances the abundance of cell surface PSA by inhibiting PSA-NCAM endocytosis. In the present study we have identified a mechanism for insulin-dependent inhibition of PSA-NCAM turnover affecting cell migration. Insulin enhanced the phosphorylation of the focal adhesion kinase leading to dissociation of αv-integrin/PSA-NCAM clusters, and promoted cell migration. Our results show that αv-integrin plays a key role in the PSA-NCAM turnover process. αv-integrin knockdown stopped PSA-NCAM from being endocytosed, and αv-integrin/PSA-NCAM clusters co-labelled intracellularly with Rab5, altogether indicating a role for αv-integrin as a carrier for PSA-NCAM during internalisation. Furthermore, inhibition of p-FAK caused dissociation of αv-integrin/PSA-NCAM clusters and counteracted the insulin-induced accumulation of PSA at the cell surface and cell migration was impaired. Our data reveal a functional association between the insulin/p-FAK-dependent regulation of PSA-NCAM turnover and cell migration through the extracellular matrix. Most importantly, they identify a novel mechanism for insulin-stimulated cell migration. - Highlights: • Insulin modulates PSA-NCAM turnover through upregulation of p-FAK. • P-FAK modulates αv-integrin/PSA-NCAM clustering. • αv-integrin acts as a carrier for PSA-NCAM endocytosis. • Cell migration is promoted by cell surface PSA. • Insulin promotes PSA-dependent migration in vitro.

  11. Insulin promotes cell migration by regulating PSA-NCAM

    Energy Technology Data Exchange (ETDEWEB)

    Monzo, Hector J.; Coppieters, Natacha [Centre for Brain Research, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Department of Anatomy and Medical Imaging, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Park, Thomas I.H. [Centre for Brain Research, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Department of Pharmacology, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Dieriks, Birger V.; Faull, Richard L.M. [Centre for Brain Research, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Department of Anatomy and Medical Imaging, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Dragunow, Mike [Centre for Brain Research, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Department of Pharmacology, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Curtis, Maurice A., E-mail: m.curtis@auckland.ac.nz [Centre for Brain Research, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand); Department of Anatomy and Medical Imaging, Faculty of Medical and Health Sciences, The University of Auckland, Private Bag, 92019, Auckland (New Zealand)

    2017-06-01

    Cellular interactions with the extracellular environment are modulated by cell surface polysialic acid (PSA) carried by the neural cell adhesion molecule (NCAM). PSA-NCAM is involved in cellular processes such as differentiation, plasticity, and migration, and is elevated in Alzheimer's disease as well as in metastatic tumour cells. Our previous work demonstrated that insulin enhances the abundance of cell surface PSA by inhibiting PSA-NCAM endocytosis. In the present study we have identified a mechanism for insulin-dependent inhibition of PSA-NCAM turnover affecting cell migration. Insulin enhanced the phosphorylation of the focal adhesion kinase leading to dissociation of αv-integrin/PSA-NCAM clusters, and promoted cell migration. Our results show that αv-integrin plays a key role in the PSA-NCAM turnover process. αv-integrin knockdown stopped PSA-NCAM from being endocytosed, and αv-integrin/PSA-NCAM clusters co-labelled intracellularly with Rab5, altogether indicating a role for αv-integrin as a carrier for PSA-NCAM during internalisation. Furthermore, inhibition of p-FAK caused dissociation of αv-integrin/PSA-NCAM clusters and counteracted the insulin-induced accumulation of PSA at the cell surface and cell migration was impaired. Our data reveal a functional association between the insulin/p-FAK-dependent regulation of PSA-NCAM turnover and cell migration through the extracellular matrix. Most importantly, they identify a novel mechanism for insulin-stimulated cell migration. - Highlights: • Insulin modulates PSA-NCAM turnover through upregulation of p-FAK. • P-FAK modulates αv-integrin/PSA-NCAM clustering. • αv-integrin acts as a carrier for PSA-NCAM endocytosis. • Cell migration is promoted by cell surface PSA. • Insulin promotes PSA-dependent migration in vitro.

  12. A methodology for PSA model validation

    International Nuclear Information System (INIS)

    Unwin, S.D.

    1995-09-01

    This document reports Phase 2 of work undertaken by Science Applications International Corporation (SAIC) in support of the Atomic Energy Control Board's Probabilistic Safety Assessment (PSA) review. A methodology is presented for the systematic review and evaluation of a PSA model. These methods are intended to support consideration of the following question: To within the scope and depth of modeling resolution of a PSA study, is the resultant model a complete and accurate representation of the subject plant? This question was identified as a key PSA validation issue in SAIC's Phase 1 project. The validation methods are based on a model transformation process devised to enhance the transparency of the modeling assumptions. Through conversion to a 'success-oriented' framework, a closer correspondence to plant design and operational specifications is achieved. This can both enhance the scrutability of the model by plant personnel, and provide an alternative perspective on the model that may assist in the identification of deficiencies. The model transformation process is defined and applied to fault trees documented in the Darlington Probabilistic Safety Evaluation. A tentative real-time process is outlined for implementation and documentation of a PSA review based on the proposed methods. (author). 11 refs., 9 tabs., 30 refs

  13. An application of PSA techniques to on-line maintenance program. A feasibility study of maintenance strategy during plant operation using PSA techniques

    International Nuclear Information System (INIS)

    Onoue, Akira; Kojima, Shigeo; Aoi, Sadanori

    1998-01-01

    This study confirms an on-line maintenance possibility for a Japanese NPP from a plant risk management point of view. One-line maintenance has already been planned and executed in some NPPs in the United States. The problem preceding on-line maintenance is how to keep a certain level of plant safety with out of service condition, safety-related systems and components, during maintenance. Deterministically, the technical specification defines unplanned maintenance periods of safety related standby systems. Probabilistically, this principle is able to be verified by utilizing PSA technology. In this study a PSA method for Japanese PWR plants is applied to improve plant availability, and to investigate safety issues associated with on-line maintenance strategy, system configuration, management and financial advantage regarding Japanese plant constraints. On-line maintenance is one of the risk management activities to improve safety and manpower workload during shutdown. As a result of this study, core damage frequency can show that on-line maintenance of some safety related systems is able to be performed and that of other safety related systems, not. On-line maintenance for Japanese NPPs utilizing risk method, will be introducing maintenance programs during power operation, like as performed in the United States. It will be important to prepare guidelines for on-line maintenance principles for further risk management programs. (author)

  14. Proteolytic activity of prostate-specific antigen (PSA towards protein substrates and effect of peptides stimulating PSA activity.

    Directory of Open Access Journals (Sweden)

    Johanna M Mattsson

    Full Text Available Prostate-specific antigen (PSA or kallikrein-related peptidase-3, KLK3 exerts chymotrypsin-like proteolytic activity. The main biological function of PSA is the liquefaction of the clot formed after ejaculation by cleavage of semenogelins I and II in seminal fluid. PSA also cleaves several other substrates, which may explain its putative functions in prostate cancer and its antiangiogenic activity. We compared the proteolytic efficiency of PSA towards several protein and peptide substrates and studied the effect of peptides stimulating the activity of PSA with these substrates. An endothelial cell tube formation model was used to analyze the effect of PSA-degraded protein fragments on angiogenesis. We showed that PSA degrades semenogelins I and II much more efficiently than other previously identified protein substrates, e.g., fibronectin, galectin-3 and IGFBP-3. We identified nidogen-1 as a new substrate for PSA. Peptides B2 and C4 that stimulate the activity of PSA towards small peptide substrates also enhanced the proteolytic activity of PSA towards protein substrates. Nidogen-1, galectin-3 or their fragments produced by PSA did not have any effect on endothelial cell tube formation. Although PSA cleaves several other protein substrates, in addition to semenogelins, the physiological importance of this activity remains speculative. The PSA levels in prostate are very high, but several other highly active proteases, such as hK2 and trypsin, are also expressed in the prostate and may cleave protein substrates that are weakly cleaved by PSA.

  15. Proteolytic activity of prostate-specific antigen (PSA) towards protein substrates and effect of peptides stimulating PSA activity.

    Science.gov (United States)

    Mattsson, Johanna M; Ravela, Suvi; Hekim, Can; Jonsson, Magnus; Malm, Johan; Närvänen, Ale; Stenman, Ulf-Håkan; Koistinen, Hannu

    2014-01-01

    Prostate-specific antigen (PSA or kallikrein-related peptidase-3, KLK3) exerts chymotrypsin-like proteolytic activity. The main biological function of PSA is the liquefaction of the clot formed after ejaculation by cleavage of semenogelins I and II in seminal fluid. PSA also cleaves several other substrates, which may explain its putative functions in prostate cancer and its antiangiogenic activity. We compared the proteolytic efficiency of PSA towards several protein and peptide substrates and studied the effect of peptides stimulating the activity of PSA with these substrates. An endothelial cell tube formation model was used to analyze the effect of PSA-degraded protein fragments on angiogenesis. We showed that PSA degrades semenogelins I and II much more efficiently than other previously identified protein substrates, e.g., fibronectin, galectin-3 and IGFBP-3. We identified nidogen-1 as a new substrate for PSA. Peptides B2 and C4 that stimulate the activity of PSA towards small peptide substrates also enhanced the proteolytic activity of PSA towards protein substrates. Nidogen-1, galectin-3 or their fragments produced by PSA did not have any effect on endothelial cell tube formation. Although PSA cleaves several other protein substrates, in addition to semenogelins, the physiological importance of this activity remains speculative. The PSA levels in prostate are very high, but several other highly active proteases, such as hK2 and trypsin, are also expressed in the prostate and may cleave protein substrates that are weakly cleaved by PSA.

  16. Proteolytic Activity of Prostate-Specific Antigen (PSA) towards Protein Substrates and Effect of Peptides Stimulating PSA Activity

    Science.gov (United States)

    Mattsson, Johanna M.; Ravela, Suvi; Hekim, Can; Jonsson, Magnus; Malm, Johan; Närvänen, Ale; Stenman, Ulf-Håkan; Koistinen, Hannu

    2014-01-01

    Prostate-specific antigen (PSA or kallikrein-related peptidase-3, KLK3) exerts chymotrypsin-like proteolytic activity. The main biological function of PSA is the liquefaction of the clot formed after ejaculation by cleavage of semenogelins I and II in seminal fluid. PSA also cleaves several other substrates, which may explain its putative functions in prostate cancer and its antiangiogenic activity. We compared the proteolytic efficiency of PSA towards several protein and peptide substrates and studied the effect of peptides stimulating the activity of PSA with these substrates. An endothelial cell tube formation model was used to analyze the effect of PSA-degraded protein fragments on angiogenesis. We showed that PSA degrades semenogelins I and II much more efficiently than other previously identified protein substrates, e.g., fibronectin, galectin-3 and IGFBP-3. We identified nidogen-1 as a new substrate for PSA. Peptides B2 and C4 that stimulate the activity of PSA towards small peptide substrates also enhanced the proteolytic activity of PSA towards protein substrates. Nidogen-1, galectin-3 or their fragments produced by PSA did not have any effect on endothelial cell tube formation. Although PSA cleaves several other protein substrates, in addition to semenogelins, the physiological importance of this activity remains speculative. The PSA levels in prostate are very high, but several other highly active proteases, such as hK2 and trypsin, are also expressed in the prostate and may cleave protein substrates that are weakly cleaved by PSA. PMID:25237904

  17. Ignalina NPP: living and working conditions

    International Nuclear Information System (INIS)

    Chiuzhas, A.

    1998-01-01

    The conference was devoted to discuss the social problems related with the operation of Ignalina NPP. The main topics are the following: analysis of public opinion of surrounding region of Ignalina NPP including neighbouring Daugavpils district in Latvia, environment impact evaluation of Daugavpils district, assessment of the influence of Ignalina NPP operation to the development of business in the region, investigation of problems of Visaginas town - residence of Ignalina NPP personnel. The specificity of Visaginas (former Sniechkus) is defined by the majority of non-native Lithuanians living there. Cultural transformation and political organization of the region were surveyed as well

  18. Metamorphosis of NPP A1, V1, V2

    International Nuclear Information System (INIS)

    Dobak, D.; Moncekova, M.

    2005-01-01

    In this book the history of construction, commissioning and exploitation of NPP A1, NPP V1 and NPP V2 in Jaslovske Bohunice is presented on documentary photos. Vicinity around of these NPPs is presented, too

  19. PSA-PSMA profiles and their impact on sera PSA levels and angiogenic activity in hyperplasia and human prostate cancer.

    Science.gov (United States)

    Ben Jemaa, A; Bouraoui, Y; Sallami, S; Banasr, A; Nouira, Y; Oueslati, R

    2014-06-01

    The relevance of prostate specific antigen (PSA)-prostate specific membrane antigen (PSMA) profiles in pathologic prostate (hyperplasia and cancer) has not been fully understood. The aim of this study is to investigate the impact of PSA-PSMA profiles on sera PSA levels and angiogenic activity in benign prostate hyperplasia (BPH) and prostate carcinoma (PC). The study has been carried out in 6 normal prostate (NP), 29 BPH and 33 PC with dominant Gleason grade>8. Immunohistochemical analysis has been performed. Monoclonal antibodies 3E6 and ER-PR8 have been used to assess PSMA and PSA expression respectively. The evaluation of angiogenesis has been made by CD34 immune marker. Serum levels of PSA have been assayed by Immulite autoanalyser. The study of each protein separately among sera PSA levels showed that PSMA expression and angiogenic activity have the highest intensity in PC patients with serum PSA levels>20 ng/mL. Nevertheless, the lowest tissue PSA expression was found in PC patients with this latter sera PSA group. The most relevant results showed that in PC patients (PSA+, PSMA+) and (PSA-, PSMA+) profile were found to be inversely related to sera PSA levels. In PC patients, a high immunoexpression of (PSA+, PSMA+) profile has detected in the sera PSA group>20 ng/mL; whereas a high immunoexpression of (PSA-, PSMA+) profile was detected in the sera PSA group between 0 and 4 ng/mL. The highest angiogenic activity was found in PC patients with (PSA+, PSMA+) profile. Our findings clearly have supported the feasibility of PSA-PSMA profiles to improve in vivo diagnostic and therapeutic approaches in prostate cancer patients. Copyright © 2014. Published by Elsevier SAS.

  20. Design review of SPWR with PSA methodology

    International Nuclear Information System (INIS)

    Oikawa, Tetsukuni; Muramatsu, Ken; Iwamura, Takamichi; Tone, Tatsuzo; Kasahara, Takeo; Mizuno, Yoshio

    1993-01-01

    This paper presents the procedures and results of a PSA (Probabilistic Safety Assessment) of the SPWR (System-Integrated PWR), which is being developed at the Japan Atomic Energy Research Institute (JAERI) as a medium sized innovative passive safe reactor, to assist in the design improvement of the SPWR by reviewing the design and identifying the design weaknesses. This PSA was performed in four steps: (1) identification of initiating events by the failure mode effect analysis and other methods, (2) delineation of accident sequences for three selected initiating events using accident progression flow charts and event trees, (3) quantification of event trees based on the review of past PSAs for LWRs, and (4) sensitivity analysis and interpretation of results. Qualitative and quantitative results of PSA provided very useful information for decision makings of design improvement and recommendations for further consideration in the process of detailed design

  1. Regulatory requirements on PSA level 2: Review, aspects and applications

    International Nuclear Information System (INIS)

    Husarcek, J.

    2003-01-01

    The general requirements concerning utility obligations, probabilistic safety criteria (CDF should not exceed 1.0E-4/reactor year and LERF should not exceed 1.0E-5/reactor year), documentation and results, living PSA requirements and major steps in level 2 PSA are presented. PSA developments in Slovakia, collection and assembly of information, plant damage states, containment performance and failure modes, severe accident progression analyses, containment failure modes and source terms as a part of performed level 2 PSA are discussed. The PSA applications in design and operation evaluation, support to plant upgrade and modifications are also described. At the end, the following conclusion is made: more extensive PSA application needs to foster the exchange of experience and communication between PSA specialists, non-PSA engineers, designers, and the regulatory body staff responsible for safety assessment, inspection and enforcement

  2. Prostate Cancer Screening: Should You Get a PSA Test?

    Science.gov (United States)

    ... Staff Cancer screening tests — including the prostate-specific antigen (PSA) test to look for signs of prostate cancer — ... to the person undergoing the testing. Prostate-specific antigen (PSA) is a protein produced by both cancerous (malignant) ...

  3. Seminal plasma PSA in spinal cord injured men

    DEFF Research Database (Denmark)

    Brasso, K; Sønksen, J; Sommer, P

    1998-01-01

    The aim of the study was to evaluate the impact of spinal cord injury on seminal plasma PSA concentration.......The aim of the study was to evaluate the impact of spinal cord injury on seminal plasma PSA concentration....

  4. Molecular Form Differences Between Prostate-Specific Antigen (PSA) Standards Create Quantitative Discordances in PSA ELISA Measurements

    Science.gov (United States)

    McJimpsey, Erica L.

    2016-02-01

    The prostate-specific antigen (PSA) assays currently employed for the detection of prostate cancer (PCa) lack the specificity needed to differentiate PCa from benign prostatic hyperplasia and have high false positive rates. The PSA calibrants used to create calibration curves in these assays are typically purified from seminal plasma and contain many molecular forms (intact PSA and cleaved subforms). The purpose of this study was to determine if the composition of the PSA molecular forms found in these PSA standards contribute to the lack of PSA test reliability. To this end, seminal plasma purified PSA standards from different commercial sources were investigated by western blot (WB) and in multiple research grade PSA ELISAs. The WB results revealed that all of the PSA standards contained different mass concentrations of intact and cleaved molecular forms. Increased mass concentrations of intact PSA yielded higher immunoassay absorbance values, even between lots from the same manufacturer. Standardization of seminal plasma derived PSA calibrant molecular form mass concentrations and purification methods will assist in closing the gaps in PCa testing measurements that require the use of PSA values, such as the % free PSA and Prostate Health Index by increasing the accuracy of the calibration curves.

  5. Molecular Form Differences Between Prostate-Specific Antigen (PSA) Standards Create Quantitative Discordances in PSA ELISA Measurements

    Science.gov (United States)

    McJimpsey, Erica L.

    2016-01-01

    The prostate-specific antigen (PSA) assays currently employed for the detection of prostate cancer (PCa) lack the specificity needed to differentiate PCa from benign prostatic hyperplasia and have high false positive rates. The PSA calibrants used to create calibration curves in these assays are typically purified from seminal plasma and contain many molecular forms (intact PSA and cleaved subforms). The purpose of this study was to determine if the composition of the PSA molecular forms found in these PSA standards contribute to the lack of PSA test reliability. To this end, seminal plasma purified PSA standards from different commercial sources were investigated by western blot (WB) and in multiple research grade PSA ELISAs. The WB results revealed that all of the PSA standards contained different mass concentrations of intact and cleaved molecular forms. Increased mass concentrations of intact PSA yielded higher immunoassay absorbance values, even between lots from the same manufacturer. Standardization of seminal plasma derived PSA calibrant molecular form mass concentrations and purification methods will assist in closing the gaps in PCa testing measurements that require the use of PSA values, such as the % free PSA and Prostate Health Index by increasing the accuracy of the calibration curves. PMID:26911983

  6. Summary review of PSA topics in connection with Romanian regulatory body activities

    International Nuclear Information System (INIS)

    Stoian, Alexandru

    2000-01-01

    This presentation includes the Romanian regulatory body guide; PSA level 1 requirements; PSA level 1 project status; PSA level 2 support activities; Scenario for PSA activities; external and internal cooperation

  7. The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

    International Nuclear Information System (INIS)

    Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

    2005-01-01

    The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code

  8. Radiation monitoring in the NPP environment, control of radioactivity in NPP-environment system

    International Nuclear Information System (INIS)

    Egorov, Yu.A.

    1987-01-01

    Problems of radiation monitoring and control of the NPP-environment system (NPPES) are considered. Radiation control system at the NPP and in the environment provides for the control of the NPP, considered as the source of radioactive releases in the environment and for the environmental radiation climate control. It is shown, that the radiation control of the NPP-environment system must be based on the ecological normalization principles of the NPP environmental impacts. Ecological normalization should be individual for the NPP region of each ecosystem. The necessity to organize and conduct radiation ecological monitoring in the NPP regions is pointed out. Radiation ecological monitoring will provide for both environmental current radiation control and information for mathematical models, used in the NPPES radiation control

  9. Safety culture in Ignalina NPP, regulatory view

    International Nuclear Information System (INIS)

    Maksimovas, G.

    1997-01-01

    The presentation describes how success on the way to a high level Safety Culture in Ignalina NPP may be achieved by daily, well motivated activities with good attitude and proper management participation, ensuring the development and proper implementation of Safety Culture principles within the activities of Operational organization of Ignalina NPP

  10. Methodic issues of NPP personnel professional selection

    International Nuclear Information System (INIS)

    Kuznetsova, Zh.Ya.

    1983-01-01

    Basic methodic principles related to the NPP personnel selection are considered. Suggestions on the selection organization are given as well as some psycho-physiological methods of the personnel professional fitness estimating. The personnel distribution over the working places with respect to psycho-phisiological abilities has been shown to represent a way to improvement of NPP radiation safety [ru

  11. Emergency preparedness at Barsebaeck NPP in Sweden

    International Nuclear Information System (INIS)

    Olsson, R.; Lindvall, C.

    1998-01-01

    On-site emergency preparedness plan at Barsebaeck NPP is presented. In an emergency the responsibility of the NPP is to alarm the emergency organizations, spend all efforts to restore safe operation, assess the potential source term as to size and time, protect their own personnel, inform personnel and public. Detailed emergency procedures overview is provided

  12. NPP Bohunice experience with ASSET services

    International Nuclear Information System (INIS)

    Klimo, J.

    1996-01-01

    The general description of Bohunice NPP ASSET experience history was given at the last annual workshop in 1995. In my short presentation I would like to pay attention to the progress in this area which was achieved at our NPP during the last year. (author)

  13. Chernobyl NPP accident. Overcoming experience. Acquired lessons

    International Nuclear Information System (INIS)

    Nosovskij, A.V.; Vasil'chenko, V.N.; Klyuchnikov, A.A.; Prister, B.S.

    2006-01-01

    This book is devoted to the 20 anniversary of accident on the Chernobyl NPP unit 4. History of construction, causes of the accident and its consequences, actions for its mitigation are described. Modern situation with Chernobyl NPP decommissioning and transferring of 'Ukryttya' shelter into ecologically safe system are mentioned. The future of Chernobyl site and exclusion zone was discussed

  14. STD Awareness PSA - Male Announcer 2 (:30)

    Centers for Disease Control (CDC) Podcasts

    2010-04-22

    This PSA encourages listeners to get tested for STDs. Target - Men who have sex with other men.  Created: 4/22/2010 by Centers for Disease Control and Prevention (CDC).   Date Released: 4/22/2010.

  15. STD Awareness PSA - College 2 (:30)

    Centers for Disease Control (CDC) Podcasts

    2010-04-22

    This PSA, targeted to college-aged youth and young adults, encourages listeners to get tested for STDs.  Created: 4/22/2010 by Centers for Disease Control and Prevention (CDC).   Date Released: 4/22/2010.

  16. STD Awareness PSA - College 1 (:30)

    Centers for Disease Control (CDC) Podcasts

    2010-04-22

    This PSA, targeted to college-aged youth and young adults, encourages listeners to get tested for STDs.  Created: 4/22/2010 by Centers for Disease Control and Prevention (CDC).   Date Released: 4/22/2010.

  17. Screen for Life: Meryl Streep PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2013-05-16

    In this 60 second PSA, Academy Award®-winning actress Meryl Streep urges viewers to get screened for colorectal cancer.  Created: 5/16/2013 by Centers for Disease Control and Prevention (CDC).   Date Released: 5/16/2013.

  18. Screen for Life: Meryl Streep PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2013-05-16

    In this 30 second PSA, Academy Award®-winning actress Meryl Streep urges viewers to get screened for colorectal cancer.  Created: 5/16/2013 by Centers for Disease Control and Prevention (CDC).   Date Released: 5/16/2013.

  19. Secundaire analyses organisatiebeleid psychosociale arbeidsbelasting (PSA)

    NARCIS (Netherlands)

    Kraan, K.O.; Houtman, I.L.D.

    2016-01-01

    Hoe het organisatiebeleid rond psychosociale arbeidsbelasting (PSA) eruit ziet anno 2014 en welke samenhang er is met ander beleid en uitkomstmaten, zijn de centrale vragen in dit onderzoek. De resultaten van deze verdiepende analyses kunnen ten goede komen aan de lopende campagne ‘Check je

  20. Interoperability in the Planetary Science Archive (PSA)

    Science.gov (United States)

    Rios Diaz, C.

    2017-09-01

    The protocols and standards currently being supported by the recently released new version of the Planetary Science Archive at this time are the Planetary Data Access Protocol (PDAP), the EuroPlanet- Table Access Protocol (EPN-TAP) and Open Geospatial Consortium (OGC) standards. We explore these protocols in more detail providing scientifically useful examples of their usage within the PSA.

  1. The use of PSA in the French regulatory practice

    International Nuclear Information System (INIS)

    Mennesiez, H.

    1994-01-01

    The presentation gives a description of fundamental documents (since 1977-1978) through which have been set up in France probabilistic objectives, and PSAs, including shutdown states, performed for 900-1300 MWe PWR-type nuclear power plants. PSA developments and use, including fire PSA, level 2 and PSA for the future French-German European Pressurized Reactor (EPR) are also discussed

  2. Development of a Probabilistic Tsunami Hazard Analysis Method and Application to an NPP in Korea

    International Nuclear Information System (INIS)

    Kim, M. K.; Choi, Ik

    2012-01-01

    A methodology of tsunami PSA was developed in this study. A tsunami PSA consists of tsunami hazard analysis, tsunami fragility analysis and system analysis. In the case of tsunami hazard analysis, evaluation of tsunami return period is a major task. For the evaluation of tsunami return period was evaluated with empirical method using historical tsunami record and tidal gauge record. For the performing a tsunami fragility analysis, procedure of tsunami fragility analysis was established and target equipment and structures for investigation of tsunami fragility assessment were selected. A sample fragility calculation was performed for the equipment in a Nuclear Power Plant. For the system analysis, accident sequence of tsunami event was developed according to the tsunami run-up and draw down, and tsunami induced core damage frequency (CDF) is determined. For the application to the real nuclear power plant, the Ulchin 56 NPP which is located on the east coast of Korean peninsula was selected. Through this study, whole tsunami PSA (Probabilistic Safety Assessment) working procedure was established and an example calculation was performed for one nuclear power plant in Korea

  3. The probabilistic safety analysis of Jose Cabrera NPP in the Context of the Periodic safety review

    International Nuclear Information System (INIS)

    Garcia, A.; Lupianez, J. M.; Ortega, P.; Gallo, J.; Saiz, J.; Gomez, C.

    2000-01-01

    In July 1989, the Spanish Nuclear Safety Council (CSN) called on Jose Cabrera NPP (JCNPP) to perform a probabilistic safety analysis (PSA). Edition 1 of this PSA was presented in July 1993. Edition 2 was delivered to the CSN, along with the database of items pending from the evaluation of Edition 1, December 1997. In October 1998, the CSN and JCNPP agreed on the appropriateness of having a PSA approved for use in the evaluation of the Periodic Safety Review (PSR) and in the renewal process of the Provisional Operating Permit (October 1999). This involved a great effort on the part of both parties, who established a joint calendar of actions to be taken, setting strict deadlines. The deadline for delivering Edition 3 (models, data and quantification programmes was set for 15 june 1999. This was complemented by the preparation of applications on licensing-related issues, and a document reflecting the resolution of pending items. Subsequently, In April, JCNPP was required to prepare additional applications. (Author)

  4. Time to PSA rise differentiates the PSA bounce after HDR and LDR brachytherapy of prostate cancer.

    Science.gov (United States)

    Burchardt, Wojciech; Skowronek, Janusz

    2018-02-01

    To investigate the differences in prostate-specific antigen (PSA) bounce (PB) after high-dose-rate (HDR-BT) or low-dose-rate (LDR-BT) brachytherapy alone in prostate cancer patients. Ninety-four patients with localized prostate cancer (T1-T2cN0), age ranged 50-81 years, were treated with brachytherapy alone between 2008 and 2010. Patients were diagnosed with adenocarcinoma, Gleason score ≤ 7. The LDR-BT total dose was 144-145 Gy, in HDR-BT - 3 fractions of 10.5 or 15 Gy. The initial PSA level (iPSA) was assessed before treatment, then PSA was rated every 3 months over the first 2 years, and every 6 months during the next 3 years. Median follow-up was 3.0 years. Mean iPSA was 7.8 ng/ml. In 58 cases, PSA decreased gradually without PB or biochemical failure (BF). In 24% of patients, PB was observed. In 23 cases (24%), PB was observed using 0.2 ng/ml definition; in 10 cases (11%), BF was diagnosed using nadir + 2 ng/ml definition. The HDR-BT and LDR-BT techniques were not associated with higher level of PB (26 vs. 22%, p = 0.497). Time to the first PSA rise finished with PB was significantly shorter after HDR-BT then after LDR-BT (median, 10.5 vs. 18.0 months) during follow-up. Predictors for PB were observed only after HDR-BT. Androgen deprivation therapy (ADT) and higher Gleason score decreased the risk of PB (HR = 0.11, p = 0.03; HR = 0.51, p = 0.01). The higher PSA nadir and longer time to PSA nadir increased the risk of PB (HR 3.46, p = 0.02; HR 1.04, p = 0.04). There was no predictors for PB after LDR-BT. HDR-BT and LDR-BT for low and intermediate risk prostate cancer had similar PB rate. The PB occurred earlier after HDR-BT than after LDR-BT. ADT and higher Gleason score decreased, and higher PSA nadir and longer time to PSA nadir increased the risk of PB after HDR-BT.

  5. Lessons learned from current Qinshan CANDU project and the impact on future NPP's

    International Nuclear Information System (INIS)

    Hedges, K. R.; Didsbury, R.; Yu, S. K. W.

    2000-01-01

    AECL has adopted an evolutionary approach to the development of the CANDU 6 and CANDU 9 Nuclear Power Plant (NPP) designs. Each new NPP project benefits from previous projects and contains an increasing number of fully proven enhancements. In accordance with this evolutionary design approach, AECL has built on the Wolsong and Qinshan successes and the solid performance of the reference CANDU stations to define, review and implement the enhancements for the CANDU 9 NPP. Some of these enhancements include fully integrated project information systems and databases, safety enhancements coming from PSA studies and licensing activities, distributed control systems for plant-wide control and an advanced control center which addresses human factors engineering concepts. Examples of the Qinshan CANDU project delivery enhancements are the utilization of electronic engineering tools for the complete plant, and the linking of these tools with the project material management system and document management systems. The project information is reviewed and approved at the engineering office in Canada and then transmitted to site electronically. Once the electronic data is at site the information packages are extracted as necessary to enable construction and facilitate contract needs with minimum effort. This paper will provide details of the CANDU Qinshan project experiences as well as describing some of the corresponding CANDU 9 enhancements. (author)

  6. PSA testing anxiety, psychological morbidity, and PSA utility in the management of prostate cancer.

    OpenAIRE

    Micsunescu, Anamaria Elia

    2017-01-01

    Anecdotal reports from urologists and medical oncologists have suggested that patients with prostate cancer (PCa) often present with anxiety related to ongoing monitoring of their PSA levels as part of their disease management. The purpose of the current study, therefore, was to determine the prevalence and severity of prostate specific antigen (PSA) testing anxiety in a population of patients with either localised or metastatic PCa living in Australia. Other aspects of psychological morbidit...

  7. Almaraz I and II NPP

    International Nuclear Information System (INIS)

    Bernaldo de Quiros, J. M.

    2005-01-01

    At the beginning of the year, both Units were connected to the grid and operating stably at 100% power. During 2004, 16,351 million kWh were generated, which is the highest gross annual production ever achieved in Almaraz NPP. Between the two units the year's accumulated unit capability factor was greater than 95.5% (99.89%in Unit I and 91.48% in Unit II). The operating factor was 100% in Unit I and 92.01% in Unit II. As the operating factor indicates, Unit I remained online without interruption throughout the entire year and operated stably from the beginning of its seventeenth operating cycle (october 27, 2003). Unit I generated a total of 8,522 million kWh, which is the highest ever per-unit annual gross production in Almaraz NPP. Unit II generated a total of 7,830 million kWh, it had a planned shutdown in January to replace one of the main transformers (phase S), and the fifteenth refueling outage took place in October. The gross electric energy accumulated at source is 155.957 million kWh in Unit I and 150,937 million kWh in Unit II. At year end both Units were operating at 100% power, with Unit I supplying 980 MWe and Unit II 985 MWe II. (Author)

  8. Generation of monoclonal antibodies against prostate specific antigen (PSA) for the detection of PSA and its purification

    International Nuclear Information System (INIS)

    Acevedo Castro, Boris Ernesto

    2012-01-01

    The prostate cancer in Cuba is a problem of health (2672 diagnosed cases and 2769 deaths in 2007). Various diagnostic methods have been implemented for the detection and management of this disease, emphasizing among them (PSA) prostate-specific antigen serological determination. At this work was generated and characterized a panel of 11 antibodies (AcMs) monoclonal IgG1 detected with high affinity described major epitopes of the PSA, both in solution and attached to the test plate. From the panel obtained AcMs was the standardization of an essay type ELISA for the detection of serum total PSA (associated and free) equimolar, based on antibody monoclonal CB-PSA.4 in the coating and the CB-PSA.9 coupled with biotin as liner, with a detection limit of 0.15 ng/mL. Similarly, standardized system for detection in serum free PSA, based on the AcMs CB-PSA.4 (coating) and CB-PSA.2 coupled with biotin (liner), with a detection limit of 0.5 ng/mL. Finally, with the purpose of using PSA as standard in trials type ELISA, developed a simple method of inmunopurificación based on the AcM, CB-PSA.2, which was obtained the PSA with a purity exceeding 90%. Immunoassay Centre on the basis of the AcMs panel and the results of this study, developed and recorded two diagnostic systems for the detection of PSA in human serum. (author)

  9. Personalized prostate cancer screening: improving PSA tests with genomic information.

    Science.gov (United States)

    Witte, John S

    2010-12-15

    The use of a prostate-specific antigen (PSA) test to screen for prostate cancer is controversial because of its modest predictive value and the potential overdiagnosis and over-treatment of the disease. A research article in this issue of Science Translational Medicine describes single-nucleotide polymorphisms (SNPs) in or near six genes that are independently associated with serum PSA concentrations and that help to explain interindividual PSA variation. Three of these SNPs are also associated with prostate biopsy outcomes. These findings are an important step toward incorporating genetic markers into PSA screening, with the ultimate goal of devising personalized PSA tests for use in the clinic.

  10. Development of Integrated PSA Database and Application Technology

    International Nuclear Information System (INIS)

    Han, Sang Hoon; Kang, Dae Il; Park, Jin Hee; Kim, Seung Hwan; Choi, Sun Yeong; Jung, Woo Sik; Ha, Jae Joo; Ahn, Kwang Il

    2007-06-01

    The high quality of PSA is essential for the risk informed regulation and applications. The main elements of PSA are the model, methodology, reliability data, and tools. The purpose of the project is to develop the reliability database for the Korean nuclear power plants and PSA analysis and management system. The reliability database system has been developed and the reliability data has been collected for 4 types of reliability data such as the reactor trip, the piping, the component and the common cause failure. The database provides the reliability data for PSAs and risk informed applications. The FTREX software is the fastest PSA quantification engine in the world. The license agreement between KAERI and EPRI is made to sell FTREX to the members of EPRI. The advanced PSA management system AIMS- PSA has been developed. The PSA model is stored in the database and solved by clicking one button. All the information necessary for the KSNP Level-1 and 2 PSA is stored in the PSA information database. It provides the PSA users a useful mean to review and analyze the PSA

  11. A Joint Report on PSA for New and Advanced Reactors

    International Nuclear Information System (INIS)

    2013-01-01

    This report addresses the application of Probabilistic Safety Assessment (PSA) to new and advanced nuclear reactors. As far as advanced reactors are concerned, the objectives were to characterize the ability of current PSA technology to address key questions regarding the development, acceptance and licensing of advanced reactor designs, to characterize the potential value of advanced PSA methods and tools for application to advanced reactors, and to develop recommendations for any needed developments regarding PSA for these reactors. As far as the design and commissioning of new nuclear power plants is concerned, the objectives were to identify and characterize current practices regarding the role of PSA, to identify key technical issues regarding PSA, lessons learned and issues requiring further work; to develop recommendations regarding the use of PSA, and to identify future international cooperative work on the identified issues. In order to reach these objectives, questionnaires had been sent to participating countries and organisations

  12. Does climate directly influence NPP globally?

    Science.gov (United States)

    Chu, Chengjin; Bartlett, Megan; Wang, Youshi; He, Fangliang; Weiner, Jacob; Chave, Jérôme; Sack, Lawren

    2016-01-01

    The need for rigorous analyses of climate impacts has never been more crucial. Current textbooks state that climate directly influences ecosystem annual net primary productivity (NPP), emphasizing the urgent need to monitor the impacts of climate change. A recent paper challenged this consensus, arguing, based on an analysis of NPP for 1247 woody plant communities across global climate gradients, that temperature and precipitation have negligible direct effects on NPP and only perhaps have indirect effects by constraining total stand biomass (Mtot ) and stand age (a). The authors of that study concluded that the length of the growing season (lgs ) might have a minor influence on NPP, an effect they considered not to be directly related to climate. In this article, we describe flaws that affected that study's conclusions and present novel analyses to disentangle the effects of stand variables and climate in determining NPP. We re-analyzed the same database to partition the direct and indirect effects of climate on NPP, using three approaches: maximum-likelihood model selection, independent-effects analysis, and structural equation modeling. These new analyses showed that about half of the global variation in NPP could be explained by Mtot combined with climate variables and supported strong and direct influences of climate independently of Mtot , both for NPP and for net biomass change averaged across the known lifetime of the stands (ABC = average biomass change). We show that lgs is an important climate variable, intrinsically correlated with, and contributing to mean annual temperature and precipitation (Tann and Pann ), all important climatic drivers of NPP. Our analyses provide guidance for statistical and mechanistic analyses of climate drivers of ecosystem processes for predictive modeling and provide novel evidence supporting the strong, direct role of climate in determining vegetation productivity at the global scale. © 2015 John Wiley & Sons Ltd.

  13. Early detection of prostate cancer in Syria using T.PSA and F.PSA

    International Nuclear Information System (INIS)

    Adel, M.; Abu Daher, D.

    2009-12-01

    The aim of the current study is performing an initial prostate cancer screening test using PSA and F PSA tumour markers. A total of 3000 men in 40-75 years of age were participated in this study. Demographic and clinical data for subjects were collected by the programme staff. Total PSA and free PSA assays were determined using the ImunoTech total and free PSA assay kits, based on IRMA technique (kindly provided by the International Atomic Energy Agency). Criteria for participating in this study included : 1) men of age 50-75 (men of age as low as 40 were included in case of positive family history). 2) No previous history of prostate cancer. The following parameters were followed to refer the suspicious cases to a specialized hospital specific tests: 1)PSA>3 ng/ml . 2)High PSA value according to the participant age group. 3) Low F/TPSA ratio. In the hospital the following tests were performed:1) Complete clinical exam including DRE.2)TRUS in some cases.3) Biopsy for highly suspicious cases. 4)The low suspicious cases were retested in six months. Out of 338 cases referred to a specialized hospital, 264 cases were shown prostatic benign prostatic hyperplasia (BPH),while 36 cases proved to be prostatic cancer. However, the contact was lost in 36 cases because of changing the phone number or travelling outside the country . The detection rate of prostate cancer among all participating cases in this study was 1.2%, while this ratio was 10.7% among the referred cases. F/TPSA ratio has shown a good ability to discriminate between prostate cancer and benign prostatic hyperplasia. (author)

  14. Inspection Qualification Centre in NPP 'Kozloduy'

    International Nuclear Information System (INIS)

    Mikhovski, M.

    2000-01-01

    In May 1999 according to the working plan of the IAEA project RER 4/020 and the decision of the NPP the Inspection Qualification Centre (IQC) has been established in order to provide examination services in the NPP. During year 1999 IVC (AEA Technology) in the framework of the DTI project provides consulting and technical assistance to the NPP, IQC, Bulgarian Academy of Sciences and Regulatory Authorities in setting up the IQC infrastructure. Now IQC work as an independent inspection body B type. The IQC activities for the period 1999-2000 are presented and further tasks are outlined

  15. Improved technical specifications for Korean NPP

    International Nuclear Information System (INIS)

    Ryu, J. D.; Lee, D. H.; Seong, C. K.

    2002-01-01

    PWRs use Technical Specifications(Tech. Spec.) to ensure safe operation of the plant. Recently, many efforts were made to improve Tech. Spec. and as a result, Improved Standard Technical Specifications(ISTS) have been developed. Korean NPP technical specifications were converted to ISTS format. KAERI also provided supporting documents for technical specification conversion including mark-up's and description of changes. This paper describes and summarizes the results of implementation of ISTS for Korean NPP. The new Tech. Spec. will improve safety of Korean NPP

  16. Management and organisational factors in PSA

    International Nuclear Information System (INIS)

    Balfanz, H.P.

    1999-01-01

    The constraints of PSA are increasingly considered with increasing application of PSA for the safety management of nuclear power plants (see US-NRC, 'Risk Informed Regulation', NRC-1). There is a vivid international discourse about the applicability of the variables of plant management and organisation in PSAs, which has lead to a great variety of research activities into this matter (see PSAM 4). This paper here summarizes the current state of progress of research work and discusses the applicability of results. The studies for comparative assessment of methodology and results were performed by the TUeV Nord under the roof of the BMU/BfS-sponsored project SR 2260, ''Further development of probabilistic methods for nuclear power plant safety assessment. (orig./CB) [de

  17. Towards a PSA harmonization French-Belgian comparison of the level 1 PSA for two similar PWR types

    International Nuclear Information System (INIS)

    Dupuy, P.; Corenwinder, F.; Lanore, J.M.; Gryffroy, D.; Gelder, P. de; Hulsmans, M.

    2002-06-01

    In the framework of the cooperation between French and Belgian regulatory authorities, a PSA (Probabilistic Safety Assessment) comparison exercise has been carried out for several years. This comparison deals with two PSA level 1 studies for internal events, performed for both power and shutdown states: the French PSA of the 900 MWe-series PWR, and the Belgian PSA of the Tihange 1 PWR, which both concern PWRs with a similar Framatome design. The purpose of this paper is to describe the PSA comparison methodology and to present, in a qualitative way, an overview of the insights obtained up to now. It also shows that such an 'a posteriori' benchmark exercise turns out to be a step towards PSA harmonization, and gives more confidence in the results of plant specific PSA when used for applications like precursor analysis or evaluations of importance to safety. (authors)

  18. A Development of Interfacing System between Level 2 and Level 3 PSA for Integrated Analysis of Full Scope PSA

    International Nuclear Information System (INIS)

    Han, Seok Jung

    2011-01-01

    An integrated assessment of full scope and entire level PSA including level 1, 2 and 3 PSA is an essential issue of the current PSA implementation for operating and developing nuclear power plants. In order to perform an integrated assessment under restricted resources for PSA, integrated and automation assessment tools are essentially required. For this purpose, KAERI is in the development of an integrated PSA assessment software package named by OCEANS. As a part of OCEANS, an interfacing system linked between level 2 and level 3 PSA was developed. The purpose of this paper is to introduce an overview of the currently developing interfacing system with a concept of link method. This interfacing system was designed as a subsidiary tool of SARA program which is a supporting utility of level 3 PSA with Microsoft Window based- program

  19. Take Charge. Take the Test. PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2012-03-07

    As part of the Take Charge. Take the Test. campaign, this 30 second PSA encourages African American women to get tested for HIV. Locations for a free HIV test can be found by visiting hivtest.org/takecharge or calling 1-800-CDC-INFO (1-800-232-4636).  Created: 3/7/2012 by National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention (NCHHSTP).   Date Released: 3/7/2012.

  20. Safer Food Saves Lives PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2015-11-03

    This 60 second PSA is based on the November 2015 CDC Vital Signs report. Contaminated food sent to several states can cause multistate outbreaks of foodborne illness and make a lot of people seriously ill. Learn what can be done to prevent and stop outbreaks.  Created: 11/3/2015 by National Center for Emerging and Zoonotic Infectious Diseases (NCEZID).   Date Released: 11/3/2015.

  1. Communication Can Save Lives PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2015-08-04

    This 60 second public service announcement (PSA) is based on the August 2015 CDC Vital Signs report. Antibiotic-resistant germs cause at least 23,000 deaths each year. Learn how public health authorities and health care facilities can work together to save lives.  Created: 8/4/2015 by National Center for Emerging and Zoonotic Infectious Diseases (NCEZID).   Date Released: 8/4/2015.

  2. Stop C. difficile Infections PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2012-03-06

    This 60 second PSA is based on the March 2012 CDC Vital Signs report. C. difficile is a germ that causes diarrhea linked to 14,000 deaths in the US each year. This podcast helps health care professionals learn how to prevent C. difficile infections.  Created: 3/6/2012 by Centers for Disease Control and Prevention (CDC).   Date Released: 3/6/2012.

  3. More Adults Are Walking PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2012-07-31

    This 60 second PSA is based on the August 2012 CDC Vital Signs report. While more adults are walking, only half get the recommended amount of physical activity. Listen to learn how communities, employers, and individuals may help increase walking.  Created: 7/31/2012 by Centers for Disease Control and Prevention (CDC).   Date Released: 8/7/2012.

  4. Study of Serum Total PSA and Free PSA as an Oncological Marker in Breast Tumour.

    Science.gov (United States)

    Jahir, Elteza Tahjiba; Devi, Runi; Borthakur, Bibhuti Bhushan

    2017-03-01

    Breast Cancer (BC) cases are rising alarmingly all over the world and India is not an exception. This rising trend is due to an increased age at first child birth, decreased breast feeding, and the changing lifestyle mostly in urban India. With the advent of more sensitive methodologies and research works in this field, it has been suggested that Prostate Specific Antigen (PSA) plays an important role in the pathogenesis of breast cancer besides other established tumour markers. To study the molecular forms of PSA-total and free PSA in benign and malignant tumours and to analyse their association with the tumour burden. The present study was conducted in collaboration with Gauhati Medical College and Hospital and Dr B Borooah Cancer Institute, Guwahati, Assam, India. Women in the age group of 18-65 years with recently diagnosed tumour (benign/malignant) in the breast were included in the study. Women taking Oral Contraceptive Pill (OCP), hormone replacement therapy, with past/present history of gynaecological/other malignancy and chronic endocrine disease like diabetes, thyroid disorders were excluded. The case group comprised of 50 female subjects with newly diagnosed Benign Breast Disease (BBD) and 50 subjects with BC, while 50 age matched healthy females without any signs and symptoms of breast discomfort were included in the control group. Laboratory tests done were Serum Total PSA (TPSA), Free PSA (FPSA), Fasting Blood Glucose (FBS), serum urea, serum creatinine and fasting lipid profile. TPSA and FPSA was measured again in both the test groups after 10-14 days of surgery/therapy. A fall in postoperative value of total and free PSA in BC case group was noticed. In Grade I tumours the mean value of total PSA (1.813 ng/ml) and free PSA (1.149 ng/ml) were higher than those with Grade III tumours (TPSA-1.07 ng/ml and FPSA-1.002 ng/ml). Mean value of Fasting Blood Sugar (FBG), total cholesterol and Low Density Lipoprotein (LDL) in BC case group was higher than the

  5. PSA application on the Tokai Reprocessing Plant

    International Nuclear Information System (INIS)

    Ishida, Michihiko; Nakano, Takafumi; Morimoto, Kazuyuki; Nojiri, Ichiro

    2003-01-01

    The Periodic Safety Review (PSR) of the Tokai Reprocessing Plant (TRP) has been carrying out to obtain an overall view of actual plant safety. As a part of the PSR, Probabilistic Safety Assessment (PSA) methodology has been applied to evaluate the relative importance of safety functions that prevent the progress of events causing to postulated accidents. Based on the results of the safety reassessments of the TRP that was carried out in 1999, event trees were developed to model sequences of postulated accidents. Event trees were quantified by using the results of fault tree analysis and human reliability analysis. In the quantification, the reliability data generally used in PSA of nuclear power plants were mainly used. Operating experiences of the TRP were also utilized to evaluated both component/system reliability and human reliability. The relative importance of safety functions was evaluated by using two major importance measures, Fussell-Vesely and Risk Achievement Worth both generally used in PSA of nuclear power plants. Through these evaluations, some useful insights into the safety of the TRP have been obtained. The results of the relative importance measures would be utilized to qualify TRP component/equipment important to the safety. (author)

  6. The organizational factor in PSA framework

    Energy Technology Data Exchange (ETDEWEB)

    Farcasiu, Mita, E-mail: mmfarcasiu@yahoo.com; Nitoi, Mirela

    2015-11-15

    The goals of the Man–Machine–Organization system analysis are to develop the suitable studies and techniques to identify, prevent and predict the cause of system unavailability. A descriptive concept of man–machine–organization system was developed. MMOS is defined in probability theory in the attempt to find ways for its qualitative and quantitative quantification in a PSA framework. The need for this study was demonstrated by analysis of variance of the complex system unavailability in relation to human error probability (HEP) and organizational error probability (OMP). The PSA model proposed in this paper assesses the organizational factor in MMOS by observing its influence on the human factor and equipment. Thus the influence of organizational factors is evaluated not only on component but also on the human performance. The study highlights the need to improve the understanding of the influence of organizational factors on the safe operation of nuclear installations. Using MMOS concept in PSA could identify any serious deficiencies of human and equipment performance which can sometime be corrected by improvement of the organizational factor.

  7. NPP Grassland: NPP Estimates from Biomass Dynamics for 31 Sites, 1948-1994, R1

    Data.gov (United States)

    National Aeronautics and Space Administration — This data set includes monthly grassland biomass data, net primary productivity (NPP) estimates, and climate (rainfall amounts and temperature) data for multiple...

  8. Radiation ecological monitoring in NPP region

    International Nuclear Information System (INIS)

    Egorov, Yu.A.; Kazakov, S.V.

    1985-01-01

    The known principle of sanitary-hygienic regulation of NPP radiation effect on man and environment is analyzed. An ecological approach is required to optimize NPP relations with the environment and to regulate radioactivity of the NPP - environment system. The ecological approach envisages the development of standards of permissible concentrations of radioactive and chemical substances (as well as heat) in natural environment, taking into account their synergism, corresponding to ecologically permissible response reactions of biota to their effect. The ecological approach also comprises the sanitary-hygienic principle of radiation protection of man. Attention is paid to ecological monitoring in NPP region, comprising consideration of factors, affecting the environment, evaluation of the actual state of the environment, prediction of the environmental state, evaluation of the expected environmental state

  9. 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

  10. Environmental impact assessment of NPP decommissioning

    International Nuclear Information System (INIS)

    Hinca, R.

    2009-01-01

    In this presentation the following potential impacts of decommissioning of NPP are discussed: - Impacts on population; Impacts on natural environment; Land impacts; Impacts on urban complex and land utilisation; Possible impacts on area as a result of failure.

  11. Cuba: Juragua NPP. Project Control. Annex 5

    International Nuclear Information System (INIS)

    Serradet, M.A.

    1999-01-01

    This annex deals with project control. The long suspension of Juragua NPP has affected personnel, equipment and site structures. Efforts are being made to revive the plant and to protect existing resources (assets). An action plan has been prepared. (author)

  12. NPP A-1 decommissioning - Phase I

    International Nuclear Information System (INIS)

    Krstenik, A.; Blazek, J.

    2000-01-01

    Nuclear power plant A-1 with output 150 MW e , with metallic natural uranium fuelled, CO 2 cooled and heavy water moderated reactor had been prematurely finally shut down in 1977. It is necessary to mention that neither operator nor regulatory and other authorities have been prepared for the solution of such situation. During next two consecutive years after shutdown main effort of operator focused on technical and administrative activities which are described in the previous paper together with approach, condition and constraints for NPP A-1 decommissioning as well as the work and research carried out up to the development and approval of the Project for NPP A-1 decommissioning - I. phase. Subject of this paper is description of: (1) An approach to NPP A -1 decommissioning; (2) An approach to development of the project for NPP A-1 decommissioning; (3) Project - tasks, scope, objectives; (4) Mode of the Project realisation; (5) Progress achieved up to the 1999 year. (authors)

  13. Insulin promotes cell migration by regulating PSA-NCAM.

    Science.gov (United States)

    Monzo, Hector J; Coppieters, Natacha; Park, Thomas I H; Dieriks, Birger V; Faull, Richard L M; Dragunow, Mike; Curtis, Maurice A

    2017-06-01

    Cellular interactions with the extracellular environment are modulated by cell surface polysialic acid (PSA) carried by the neural cell adhesion molecule (NCAM). PSA-NCAM is involved in cellular processes such as differentiation, plasticity, and migration, and is elevated in Alzheimer's disease as well as in metastatic tumour cells. Our previous work demonstrated that insulin enhances the abundance of cell surface PSA by inhibiting PSA-NCAM endocytosis. In the present study we have identified a mechanism for insulin-dependent inhibition of PSA-NCAM turnover affecting cell migration. Insulin enhanced the phosphorylation of the focal adhesion kinase leading to dissociation of αv-integrin/PSA-NCAM clusters, and promoted cell migration. Our results show that αv-integrin plays a key role in the PSA-NCAM turnover process. αv-integrin knockdown stopped PSA-NCAM from being endocytosed, and αv-integrin/PSA-NCAM clusters co-labelled intracellularly with Rab5, altogether indicating a role for αv-integrin as a carrier for PSA-NCAM during internalisation. Furthermore, inhibition of p-FAK caused dissociation of αv-integrin/PSA-NCAM clusters and counteracted the insulin-induced accumulation of PSA at the cell surface and cell migration was impaired. Our data reveal a functional association between the insulin/p-FAK-dependent regulation of PSA-NCAM turnover and cell migration through the extracellular matrix. Most importantly, they identify a novel mechanism for insulin-stimulated cell migration. Copyright © 2017 Elsevier Inc. All rights reserved.

  14. Training courses for the staff of the nuclear power station KRSKO conducted at the TRIGA reactor center in Ljubljana

    International Nuclear Information System (INIS)

    Pregl, G.; Najzer, M.

    1976-01-01

    The training program for the Nuclear Power Station Krsko was divided into two modules: fundamentals of nuclear engineering and specialized training according to duties that candidates are supposed to take at the power station. Basic training was organized at the TRIGA Reactor Center in Ljubljana in two different versions. The first version intended for plant operators and all engineers lasted for six months and included about 500 hours of classroom lessons and seminars and 31 laboratory experiments. The educational program was conventional. The following topics were covered: nuclear and atomic physics, reactor theory, reactor dynamics, reactor instrumentation and control, heat transfer in nuclear power plants, nuclear power plant systems, reactor materials, reactor safety, and radiation protection. Until now, two groups, consisting of 37 candidates altogether, have attended this basic course. Plans have been made to conduct two additional courses of about 20 students each for technicians other than operators. The program of this second version will be reduced, with the emphasis on reactor core physics and radiation protection. Classroom lessons will be strongly supported by laboratory experiments. (author)

  15. Knowledge Management and Organizational Proficiency with NPP

    International Nuclear Information System (INIS)

    Marler, M.

    2016-01-01

    Full text: The pace of new NPP construction, startup, and operation is straining the supply of proficient operators, technicians, and engineers. This technical brief explains an approach implemented by a US nuclear utility to capture and transfer knowledge possessed by proficient workers to new workers using the VISION learning content management system. This approach could also be used to accelerate worker proficiency in new NPP organizations. (author

  16. Ignalina NPP Safety Analysis: Models and Results

    International Nuclear Information System (INIS)

    Uspuras, E.

    1999-01-01

    Research directions, linked to safety assessment of the Ignalina NPP, of the scientific safety analysis group are presented: Thermal-hydraulic analysis of accidents and operational transients; Thermal-hydraulic assessment of Ignalina NPP Accident Localization System and other compartments; Structural analysis of plant components, piping and other parts of Main Circulation Circuit; Assessment of RBMK-1500 reactor core and other. Models and main works carried out last year are described. (author)

  17. Radionuclide localization at the Chernobyl' NPP territory

    International Nuclear Information System (INIS)

    Mamaev, L.A.; Galkin, G.A.; Khrabrov, S.L.; Polyakov, A.S.; Mikhejkin, S.V.

    1989-01-01

    Experience is generalized of using different dust suppression (DS) compounds during Chernobyl' accident consequence elimination. Polymer DS compounds were used at the NPP operating site; the compounds kept dust-like radioactive contaminations during 1-2 months. DS at the country was realized by means of the compound on base of latex. The conclusion is made that the DS measures improved radiation situation in the NPP zone. 7 refs

  18. Training human resource for NPP in Vietnam

    International Nuclear Information System (INIS)

    Nguyen, Trung Tinh; Dam, Xuan Hiep

    2008-01-01

    Vietnam will establish the first NPP in the near future. With us the first important thing is the human resource, but now there is no university in Vietnam training nuclear engineers. In EPU (Electric Power University), now we are preparing for training nuclear engineers. In this paper, we review the nuclear man power and the way to train the high quality human resource for NPP and for other nuclear application in Vietnam. (author)

  19. Serum Prostate-Specific Antigen (PSA) Concentration, PSA Mass, and Obesity: A Mathematical Analysis.

    Science.gov (United States)

    Vollmer, Robin T

    2018-02-17

    To provide a mathematical background for understanding the phenomenon of analyte hemodilution using a kinetic analysis. The first assumption for this analysis is that change in concentration of any analyte, such as prostate-specific antigen (PSA), is due to the flux of the analyte from an organ into the blood minus its flux from the blood. What results is a relatively simple differential equation that emphasizes the importance of plasma volume, organ mass, and two rate constants. The analyses demonstrate how serum PSA can be affected by plasma volume as well as body mass and how hemodilution due to obesity can be at least partly corrected for by expressing PSA in units of total mass or total mass density. At a time when obesity is prevalent, expressing analytes in units of total mass may make them relate more closely to disease status and prognosis.

  20. Development of PSA audit guideline and regulatory PSA model for SMART

    International Nuclear Information System (INIS)

    Cho, Namchul; Lee, Chang-Ju; Kim, I.S.

    2012-01-01

    SMART is under development for dual purposes of power generation and seawater desalination in Korea. It is an integral reactor type with a thermal power output of 330 MW and employs advanced design features such as a passive system for the removal of residual heat and also the setting of all the components of the primary system inside the reactor pressure vessel. It is essential to develop new probabilistic safety assessment (PSA) validation guidance for SMART. For the purpose of regulatory verification to the risk level of SMART, the insights and key issues on the PSA are identified with referring some worldwide safety guides as well as its design characteristics. Regulatory PSA model under the development for the design confirmation and its preliminary result are also described. (authors)

  1. PSA-operations synergism for the advanced test reactor shutdown operations PSA

    International Nuclear Information System (INIS)

    Atkinson, S.A.

    1996-01-01

    The Advanced Test Reactor (ATR) Probabilistic Safety Assessment (PSA) for shutdown operations, cask handling, and canal draining is a successful example of the importance of good PSA-operations synergism for achieving a realistic and accepted assessment of the risks and for achieving desired risk reduction and safety improvement in a best and cost-effective manner. The implementation of the agreed-upon upgrades and improvements resulted in the reductions of the estimated mean frequency for core or canal irradiated fuel uncovery events, a total reduction in risk by a factor of nearly 1000 to a very low and acceptable risk level for potentially severe events

  2. Elevation of PSA after prostate radiotherapy: Rebound or biochemical recurrence?

    International Nuclear Information System (INIS)

    Toledano, A.; Kanoui, A.; Chiche, R.; Lamallem, H.; Beley, S.; Thibault, F.; Sebe, P.

    2008-01-01

    The fact that external beam radiotherapy and brachytherapy are now considered to be curative techniques has led to major review of the modalities of follow-up after radiotherapy for prostate cancer. The problem concerns both the diagnosis of recurrence, rapidly announced by elevation of prostatic-specific antigen (PSA), usually at a subclinical stage, and the validity of criteria of biochemical recurrence to allow comparison of various study. Physicians involved in follow-up should be aware of the potential of bounce in PSA follow-up after external beam radiotherapy or brachytherapy. The PSA bounce phenomenon was defined by a rise of PSA values (+ 0.1 -0.8 ng/ml) with a subsequent fall. Biochemical failure after external beam radiotherapy or brachytherapy (with or without hormonotherapy) was defined by Phoenix criteria by a rise of 2 ng/ml above an initial PSA nadir. This definition was more correlated to PSA bounce phenomenon. (authors)

  3. Workshop on PSA for New and Advanced Reactors

    International Nuclear Information System (INIS)

    2012-01-01

    This workshop was organized by the NEA Working Group on Risk Assessment (WGRISK). The key objective of the workshop was to share the current state-of-the art on the PSA (Probabilistic Safety Assessment) applied for new reactors and advanced reactors. Fifty experts from 13 countries and one international organization (IAEA) participated in the present workshop, and 35 technical papers were presented. The main topics of interest, discussed during the workshop, included the followings: regulatory aspects, risk-informed methods, technical aspects of the PSA for new and advanced reactors, hazards of PSA (internal and external), severe accident/source term/Level 2 PSA, and consequence analysis/Level 3 PSA. Among the technical aspects of the PSA, the assessment of the reliability of passive safety systems appears to be a recurrent issue

  4. A Study on the Risk Reduction Effect by MLCS (Mid-loop Level Control System) of EUAPR using the Low-Power and Shutdown PSA Result

    International Nuclear Information System (INIS)

    Lee, Keunsung; Choi, Sunmi; Kim, Eden

    2016-01-01

    The EU-APR design has been developed in order to expand and diversify the global nuclear power market of APR1400. For the improvement of shutdown risk for the EUAPR, the mid-loop level control system (MLCS) is considered during mid-loop operation for the EU-APR, which is not incorporated into SKN 3 and 4 (APR1400 Type) in Korea. Commonly, the risk associated with the NPP can be identified through the PSA. Thus, this paper discusses the low power and shutdown (LPSD) risk reduction effect by MLCS using the Low-Power and Shutdown PSA Result. LPSD level 1 PSA models for EU-APR have been developed. The risk reduction effect by MLCS is discussed. Because the loss of shutdown cooling function during mid-loop is one of the most vulnerable events, the MLCS have a significant influence on CDF in LPSD PSA. The shutdown risk of domestic power plants would likely be reduced if the MLCS is adopted in all operating NPPs in Korea during the mid-loop operation. It is expected that this work will contribute to reduce shutdown risk of domestic power plants

  5. Pre-Study of Off-site Consequence Analysis in Level 3 PSA of Wolsong Unit 1

    International Nuclear Information System (INIS)

    Kim, Won-Jik; Yang, Ho-Chang; Choi, Seong-Soo

    2015-01-01

    In order to perform level 3 PSA, MACCS II (MELCOR Accident Consequence Code System 2) is needed. MACCS II is used in PSA for plants in order to evaluate population dose that is the effects on health and environment caused by released radioisotopes after an accident. In this study, Steam Generator Tube Rupture (SGTR) event in CANDU-6 plants is evaluated population dose that is the effects on health and environment caused by released radioisotopes after an accident. In this study, Steam Generator Tube Rupture (SGTR) event has been evaluated by using Level 1 PSA result and Level 2 PSA result(ISSAC) and MACCS II. As a result, We are obtained the following conclusion. - Early maximum early fatalities is 5.35E+02 equal to latent maximum early fatalities.(99.5%) - Early and latent maximum cancer fatalities are 2.33E+03 and 1.11E+04, respectively. (99.5%) - Early and latent maximum population doses are 1.25 and 5.00 person-rem/yr, respectively. (99.5%) Other study has shown that MACCS II was performed evaluation for Wolsong NPP. Small Break Loss of Coolant Accident(SBLOCA) event is selected by other study. The results of early and cancer fatalities applied similar assumption were 3.02E+00 and 1.89E+03, respectively. This study's results are higher than other study's result. Because, basis input data is different each studies, and event frequency are different (This study : 2.10E-07/ Other study : 4.93E-09)

  6. Regulatory use of risk information - initial developments at Slovenian Nuclear Safety Administration

    International Nuclear Information System (INIS)

    Muehleisen, A.; Koncar, M.; Vojnovic, D.; Persic, A.

    2004-01-01

    Similarly to other regulators worldwide, the SNSA intends to enhance the use of PSA and risk insights in its activities in order to ensure a better and more focused regulatory oversight as well as improved interface with a licensee. The main aim of the SNSA is to establish PSA as a standard tool to complement the deterministic based regulation for a variety of regulatory tasks. The PSA applications should, in particular, support the decision making process as well as the interactions with the Krsko NPP. As a first step in the internal use of PSA, PSA event analysis and risk based performance indicators are being introduced. In 2004, the SNSA will start introducing risk follow up and risk informed inspections. By mid 2005 the legal basis for the use of PSA will be also established in Slovenian legislation. (author)

  7. Decommissioning study of Forsmark NPP

    International Nuclear Information System (INIS)

    Anunti, Aake; Larsson, Helena; Edelborg, Mathias

    2013-06-01

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for the Forsmark NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding

  8. Decommissioning Study of Oskarshamn NPP

    International Nuclear Information System (INIS)

    Larsson, Helena; Anunti, Aake; Edelborg, Mathias

    2013-06-01

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for Oskarshamn NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding

  9. Decommissioning study of Forsmark NPP

    Energy Technology Data Exchange (ETDEWEB)

    Anunti, Aake; Larsson, Helena; Edelborg, Mathias [Westinghouse Electric Sweden AB, Vaesteraas (Sweden)

    2013-06-15

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for the Forsmark NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding.

  10. Decommissioning Study of Oskarshamn NPP

    Energy Technology Data Exchange (ETDEWEB)

    Larsson, Helena; Anunti, Aake; Edelborg, Mathias [Westinghouse Electric Sweden AB, Vaesteraas (Sweden)

    2013-06-15

    By Swedish law it is the obligation of the nuclear power utilities to satisfactorily demonstrate how a nuclear power plant can be safely decommissioned and dismantled when it is no longer in service as well as calculate the estimated cost of decommissioning of the nuclear power plant. Svensk Kaernbraenslehantering AB (SKB) has been commissioned by the Swedish nuclear power utilities to meet the requirements of current legislation by studying and reporting on suitable technologies and by estimating the costs of decommissioning and dismantling of the Swedish nuclear power plants. The present report is an overview, containing the necessary information to meet the above needs, for Oskarshamn NPP. Information is given for the plant about the inventory of materials and radioactivity at the time for final shutdown. A feasible technique for dismantling is presented and the waste management is described and the resulting waste quantities are estimated. Finally a schedule for the decommissioning phase is given and the costs associated are estimated as a basis for funding.

  11. ORSEC technological risks, Blayais NPP

    International Nuclear Information System (INIS)

    2016-01-01

    The Particular intervention plan (PPI in French) is an emergency plan which foresees the measures and means to be implemented to address the potential risks of the presence and operation of a nuclear facility. This plan is implemented and developed by the Prefect in case of nuclear accident (or incident leading to a potential accident), the impact of which extending beyond the facility perimeter. It represents a special section of the organisation plan for civil protection response (ORSEC plan). The PPI foresees the necessary measures and means for crisis management during the first hours following the accident and is triggered by the Department Prefect according to the information provided by the facility operator. Its aim is to protect the populations leaving within 10 km of the facility against a potential radiological hazard. The PPI describes: the facility, the intervention area, the protection measures for the population, the conditions of emergency plan triggering, the crisis organisation, the action forms of the different services, and the post-accident stage. This document is the public version of the Particular intervention plan of the Blayais NPP (Gironde, France)

  12. 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

  13. Serum PSA levels in the Indian population: Is it different?

    Science.gov (United States)

    Agrawal, Amit; Karan, Shailesh Chandra

    2017-04-01

    Serum prostate-specific antigen (PSA) is an important tumour, marker which is widely used to trigger trans-rectal ultrasound (TRUS)-guided prostate biopsy. However, the PSA levels vary with race and ethnicity. Therefore, there is a need to have an Indian reference range. All adult male patients meeting the inclusion and exclusion criteria were enrolled in this study. They were subjected to assessment of serum total PSA, digital rectal examination and trans-abdominal ultrasound. If any one or more of these were found abnormal, then a TRUS-guided 12-core prostate biopsy was done. Patients who were detected to have prostatic cancer were excluded from the final analysis. The data so obtained was grouped among the following three age groups: 40-49, 50-59 and 60-70 years, and the age-specific PSA values, prostatic volume and PSA density were found. A total of 1772 patients were analysed. The mean serum total PSA was 1.76 ng/ml with a standard deviation of 2.566 ng/ml. Group-wise age distribution of the mean serum total PSA was 1.22, 1.97 and 2.08 ng/ml in 40-49, 50-59 and 60-70 years age groups. The mean total PSA and the age-specific PSA range tend to be lower in the Indians than the Western population.

  14. Operation Aspect of the Main Control Room of NPP

    International Nuclear Information System (INIS)

    Sahala M Lumbanraja

    2009-01-01

    The main control room of Nuclear Power Plant (NPP) is operational centre to control all of the operation activity of NPP. NPP must be operated carefully and safely. Many aspect that contributed to operation of NPP, such as man power whose operated, technology type used, ergonomic of main control room, operational management, etc. The disturbances of communication in control room must be anticipated so the high availability of NPP can be achieved. The ergonomic of the NPP control room that will be used in Indonesia must be designed suitable to anthropometric of Indonesia society. (author)

  15. Assessment of the state of modernization of NPP Kozloduy units 3 and 4

    International Nuclear Information System (INIS)

    Horstmann, R.

    2002-01-01

    The status of the implemented modernization programmes for the Kozloduy NPP is presented. The Three Stage Term Modernization Program for units 1-4 has been implemented between 1991 and 1997 and includes the installation of new safety systems and components such as pressurized safety valves, main steam safety valves, complementary emergency feedwater system, second fire fighting system etc. The total investment od the Program amounted to 129.1 mill. ECU. The Complex Modernization Program for units 1-4 has been developed 1996 -1997 and further updated in 2000. The total investment necessary for the implementation are assessed at about 66 mill. USD. The safety assessment shows that due to the modernization programs the units have been upgraded to additional accident management capabilities. The reactor confinement has been fundamentally improved by the Jet Vortex Condenser System. PSA has been also conducted for the units 3 and 4

  16. Development of kits for total PSA monitoring

    International Nuclear Information System (INIS)

    Suprarop, P.

    1999-01-01

    The development of kits for Total PSA assay has shown promising results. All essential components of the assay were prepared with reproducibility and used to optimize the assay. By choosing two steps method, we could avoid the hook effect and obtain satisfactory Q.C. parameters of the standard curve i.e. blank = 0.8%, maximum binding = 65%. If reference material for calibration of the standard is agree upon, the validation could then be carried out with total confidence. Our final goal is to reduce the step of incubation to just one step with no interference from hook effect

  17. Changing NPP consumption patterns in the Holocene: from Megafauna "liberated" NPP to "ecological bankruptcy"

    Science.gov (United States)

    Doughty, C.

    2015-12-01

    There have been vast changes in how net primary production (NPP) is consumed by humans and animals during the Holocene beginning with a potential increase in availability following the Pleistocene megafauna extinctions. This was followed by the development of agriculture which began to gradually restrict availability of NPP for wild animals. Finally, humans entered the industrial era using non-plant based energies to power societies. Here I ask the following questions about these three energy transitions: 1. How much NPP energy may have become available following the megafauna extinctions? 2. When did humans, through agriculture and domestic animals, consume more NPP than wild mammals in each country? 3. When did humans and wild mammals use more energy than was available in total NPP in each country? To answer this last question I calculate NPP consumed by wild animals, crops, livestock, and energy use (all converted to units of MJ) and compare this with the total potential NPP (also in MJ) for each country. We develop the term "ecological bankruptcy" to refer to the level of consumption where not all energy needs can be met by the country's NPP. Currently, 82 countries and a net population of 5.4 billion are in the state of ecologically bankruptcy, crossing this threshold at various times over the past 40 years. By contrast, only 52 countries with a net population of 1.2 billion remain ecologically solvent. Overall, the Holocene has seen remarkable changes in consumption patterns of NPP, passing through three distinct phases. Humans began in a world where there was 1.6-4.1% unclaimed NPP to consume. From 1700-1850, humans began to consume more than wild animals (globally averaged). At present, >82% of people live in countries where not even all available plant matter could satisfy our energy demands.

  18. Integrated ageing management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos; Sabransky, Mario

    2013-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  19. Integrated ageing management of Atucha NPP

    Energy Technology Data Exchange (ETDEWEB)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos, E-mail: ranalli@cnea.gov.ar [Gerencia Coordinacion Proyectos CNEA-NASA, Comision Nacional de Energia Atomica, Buenos Aires (Argentina); Sabransky, Mario, E-mail: msabransky@na-sa.com.ar [Departamento Gestion de Envejecimiento, Central Nuclear Atucha I-II Nucleoelectrica Argentina S.A., Provincia de Buenos Aires (Argentina)

    2013-07-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  20. Integrated Ageing Management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, J.M.; Marchena, M.H.; Zorrilla, J.R.; Sabransky, M.

    2012-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor twice as big as Atucha I finishing a delayed construction . With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  1. NPP Tropical Forest: Pasoh, Malaysia, 1971-1973, R1

    Data.gov (United States)

    National Aeronautics and Space Administration — This data set contains four ASCII data files (.txt format), one providing net primary production (NPP) component data and three providing climate data. The NPP...

  2. NPP Tropical Forest: Atherton, Australia, 1974-1985, R1

    Data.gov (United States)

    National Aeronautics and Space Administration — This data set contains eight data files (.txt format): three net primary productivity (NPP) data files and five climate data files. The NPP estimates are based on...

  3. NPP Tropical Forest: Kade, Ghana, 1957-1972, R1

    Data.gov (United States)

    National Aeronautics and Space Administration — This data set contains one NPP data file and two climate data files (ASCII .txt format). The NPP file contains above- and below- ground biomass, litterfall, standing...

  4. Reserves of labour content reduction in NPP construction

    International Nuclear Information System (INIS)

    Bekerman, R.E.

    1986-01-01

    Specific labour contents when constructing NPP with RBMK-1000 and WWER-1000 type reactors are presented. Factors affecting labour content of NPP construction are shown. Measures aimed at labour content decrease are suggested

  5. ISLSCP II IGBP NPP Output from Terrestrial Biogeochemistry Models

    Data.gov (United States)

    National Aeronautics and Space Administration — ABSTRACT: This data set contains modeled annual net primary production (NPP) for the land biosphere from seventeen different global models. Annual NPP is defined as...

  6. ISLSCP II IGBP NPP Output from Terrestrial Biogeochemistry Models

    Data.gov (United States)

    National Aeronautics and Space Administration — This data set contains modeled annual net primary production (NPP) for the land biosphere from seventeen different global models. Annual NPP is defined as the net...

  7. New trends in designing NPP control boards

    International Nuclear Information System (INIS)

    Kondrat'ev, V.V.

    1981-01-01

    A short analytical summary of the latest developments and future trends in designing NPP control boards is given. The designs of the Westinghause and the Hynkley-Point NPP control boards are described in detail. The essence of the advanced control board concept consists , firstly, in expanded use of computer-controlled displays for the sake of reducing the content of unimportant information presented to an operator, and, secondary, in better account of human possibilities to convert the NPP operation information into a more suitable form. An enlarged use of the direct digital reactor control utilizing microprocessors is expected. Besides, the employment of full-scale control board mock-ups and information desks as well as testing newly-developed control boards at computer reactor simulators are concluded to be used at all-growing rate [ru

  8. Qinshan NPP large break LOCA safety analysis

    International Nuclear Information System (INIS)

    Shi Guobao; Tang Jiahuan; Zhou Quanfu; Wang Yangding

    1997-01-01

    Qinshan NPP is the first nuclear power plant in the mainland of China, a 300 MW(e) two-loop PWR. Large break LOCA is the design-basis accident of Qinshan NPP. Based on available computer codes, the own analysis method which complies with Appendix k of 10 CFR 50 has been established. The RELAP4/MOD7 code is employed for the calculations of blowdown, refill and reflood phase of the RCS respectively. The CONTEMPT-LT/028 code is used for the containment pressure and temperature analysis. The temperature transient in the hot rod is calculated using the FRAP-6T code. Conservative initial and functional assumptions were adopted during Qinshan NPP large break LOCA analysis. The results of the analysis show the applicable acceptance criteria for the loss-of-coolant accident are met

  9. What is living PSA. [Probalistic safety analysis at Nuclear Electric

    Energy Technology Data Exchange (ETDEWEB)

    Freeman, R.I.; Moir, G.R. (Nuclear Electric plc. (United Kingdom))

    1993-12-01

    The paper describes the pioneering work into living probabilistic safety analysis (PSA) within Nuclear Electric plc as part of its strategic development programme. This programme is targeted to provide risk management tools to assist in optimizing the testing and maintenance arrangements for the essential safety systems within the company's nuclear power plants. Three broad categories of living PSA development are described with their advantages. The first is a Stage 1 living PSA which is an off-line PSA which can be upgraded on a regular basis. Then a Stage 2 living PSA which is an on-line time-independent PSA, in which standby component failure probabilities are modelled at the end of their inspection period, and is updated by the plant operator as and when changes in plant configuration (i.e. opening/closing sectioning valves) or plant availability occur. Finally a Stage 3 living PSA which is an on-line dynamic PSA, in which all standby components are modelled with respect to their last inspection time, and is updated by the plant operator as and when changes in plant configuration and plant availability occur and when testing takes place. (author).

  10. Quality of the current low power and shutdown PSA practice

    International Nuclear Information System (INIS)

    Jang, Seung Cheol; Park, Jin Hee; Lim, Ho Gon; Kim, Tae Woon

    2004-01-01

    A probabilistic safety assessment (PSA) for the low-power and shutdown (LPSD) modes in a Korea standard nuclear power plant (KSNP) has been performed for the purpose of estimating the LPSD risk and identifying the vulnerabilities of LPSD operations. Both the operational experience and PSA results indicate that the risks from LPSD operations could be comparable with those from power operations. However, the application of the LPSD risk insights to risk-informed decision making has been slow to be adopted in practice. It is largely due to the question of whether the current LPSD PSA practice is appropriate for application to risk-informed decision making or not. Such a question has to do with the quality of the current LPSD PSA practice. In this paper, we have performed self-assessment of the KSNP LPSD PSA quality based on the ANS Standard (draft as of 13 Sep. 2002). The aims of the work are to find the LPSD PSA technical areas insufficient for application to risk-informed decision making and to efficiently allocate the limited research resources to improve the LPSD PSA model quality. Many useful findings regarding the current LPSD PSA quality are presented in this paper

  11. Association between steroid hormone receptors and PSA gene ...

    African Journals Online (AJOL)

    The prostate specific antigen (PSA) gene is a member of the human kallikrein gene family and is known that to be tightly regulated by androgens in the male prostate The presence of PSA is strongly associated with presence of steroid hormone receptors. The aim of this research was to show differential expression and ...

  12. Level 2 PSA methodology and severe accident management

    International Nuclear Information System (INIS)

    1997-01-01

    The objective of the work was to review current Level 2-PSA (Probabilistic Safety Assessment) methodologies and practices and to investigate how Level 2-PSA can support severe accident management programmes, i.e. the development, implementation, training and optimisation of accident management strategies and measures. For the most part, the presented material reflects the state in 1996. Current Level 2 PSA results and methodologies are reviewed and evaluated with respect to plant type specific and generic insights. Approaches and practices for using PSA results in the regulatory context and for supporting severe accident management programmes by input from level 2 PSAs are examined. The work is based on information contained in: PSA procedure guides, PSA review guides and regulatory guides for the use of PSA results in risk informed decision making; plant specific PSAs and PSA related literature exemplifying specific procedures, methods, analytical models, relevant input data and important results, use of computer codes and results of code calculations. The PSAs are evaluated with respect to results and insights. In the conclusion section, the present state of risk informed decision making, in particular in the level 2 domain, is described and substantiated by relevant examples

  13. Use of PSA in the development of SMRs

    International Nuclear Information System (INIS)

    Maioli, A.; Finnicum, D.J.; Lichtenstein, R.H.; Harsche, S.Y.

    2012-01-01

    This paper reviews the potential new scenarios where PSA (probabilistic safety assessment) may be of significant support to design and operation of SMRs (Small Modular Reactors); it reviews Westinghouse's experience and lessons learned in this endeavour and will discuss related challenges and what the PSA community is currently developing to address them. (authors)

  14. Analysis of Paks NPP Personnel Activity during Safety Related Event Sequences

    International Nuclear Information System (INIS)

    Bareith, A.; Hollo, Elod; Karsa, Z.; Nagy, S.

    1998-01-01

    Within the AGNES Project (Advanced Generic and New Evaluation of Safety) the Level-1 PSA model of the Paks NPP Unit 3 was developed in form of a detailed event tree/fault tree structure (53 initiating events, 580 event sequences, 6300 basic events are involved). This model gives a good basis for quantitative evaluation of potential consequences of actually occurred safety-related events, i.e. for precursor event studies. To make these studies possible and efficient, the current qualitative event analysis practice should be reviewed and a new additional quantitative analysis procedure and system should be developed and applied. The present paper gives an overview of the method outlined for both qualitative and quantitative analyses of the operator crew activity during off-normal situations. First, the operator performance experienced during past operational events is discussed. Sources of raw information, the qualitative evaluation process, the follow-up actions, as well as the documentation requirements are described. Second, the general concept of the proposed precursor event analysis is described. Types of modeled interactions and the considered performance influences are presented. The quantification of the potential consequences of the identified precursor events is based on the task-oriented, Level-1 PSA model of the plant unit. A precursor analysis system covering the evaluation of operator activities is now under development. Preliminary results gained during a case study evaluation of a past historical event are presented. (authors)

  15. Substantiation of the safety in the technical project of Belene NPP

    International Nuclear Information System (INIS)

    Boyadzhiev, A.

    1990-01-01

    The chapter contains an evaluation of the safety of Belene NPP project, based on an experts study of the corresponding volume of the Technical Project documentation of the main contractor and also on other related documents. The authors state that most of the remarks are constitutive, part of them requiring additional information or research. The general explicit conclusion is that the materials on the safety substantiation provided in the project are insufficient for making final statements on the safety of the NPP and there is a need for a detailed analysis and expertise. There are 12 topical conclusion paragraphs and each of them comprises a number of remarks. Among the remarks there are some related to the reactivity coefficient values in certain modes of operation, the problem of the mechanical safety and control system efficiency, the unacceptable operation at nominal power in case of stringent safety rules enforcement, the insufficiency of the PSA, the automatic control systems and the software codes not standing up to the contemporary requirements. (R.Ts.)

  16. Flooding PSA with Plant Specific Operating Experiences of Korean PWRs

    International Nuclear Information System (INIS)

    Choi, Sun Yeong; Yang, Joon Yull

    2006-01-01

    The purpose of this paper is to update the flooding PSA with Korean plant specific operating experience data and the appropriate estimation method for the flooding frequency to improve the PSA quality. The existing flooding PSA used the NPE (Nuclear Power Experience) database up to 1985 for the flooding frequency. They are all USA plant operating experiences. So an upgraded flooding frequency with Korean specific plant operation experience is required. We also propose a method of only using the PWR (Pressurized Water Reactor) data for the flooding frequency estimation in the case of the flooding area in the primary building even though the existing flooding PSA used both PWR and BWR (Boiled Water Reactor) data for all kinds of plant areas. We evaluate the CDF (Core Damage Frequency) with the modified flooding frequency and compare the results with that of the existing flooding PSA method

  17. Updating the Psoriatic Arthritis (PsA) Core Domain Set

    DEFF Research Database (Denmark)

    Orbai, Ana-Maria; de Wit, Maarten; Mease, Philip J

    2017-01-01

    OBJECTIVE: To include the patient perspective in accordance with the Outcome Measures in Rheumatology (OMERACT) Filter 2.0 in the updated Psoriatic Arthritis (PsA) Core Domain Set for randomized controlled trials (RCT) and longitudinal observational studies (LOS). METHODS: At OMERACT 2016, research...... conducted to update the PsA Core Domain Set was presented and discussed in breakout groups. The updated PsA Core Domain Set was voted on and endorsed by OMERACT participants. RESULTS: We conducted a systematic literature review of domains measured in PsA RCT and LOS, and identified 24 domains. We conducted...... and breakout groups at OMERACT 2016 in which findings were presented and discussed. The updated PsA Core Domain Set endorsed with 90% agreement by OMERACT 2016 participants included musculoskeletal disease activity, skin disease activity, fatigue, pain, patient's global assessment, physical function, health...

  18. Is PSA density still useful in diagnosing prostate cancer?

    Science.gov (United States)

    Pepe, Pietro; Candiano, Giuseppe; Fraggetta, Filippo; Galia, Antonio; Grasso, Giuseppe; Allegro, Rosalinda; Aragona, Francesco

    2009-12-01

    To evaluate the concordance between the PSAD (PSA density) values calculated using the actual prostate weight and the PSAD values calculated by using the dimensions of the gland given by the pathologist when freshly excised (volume 1) or using TRUS measures (volume 2). Diagnostic accuracy of PSAD in diagnosing PCa (prostate cancer) was evaluated and compared with accuracy obtained using PSA free/total (FIT). 102 consecutive patients with PSA included between 2 and 10 ng/mL underwent radical prostatectomy. Indications to perform prostate biopsy were: abnormal digital rectal examination, PSA 0.10 whereas, with a cut-off > 0.15, a diagnostic accuracy of 36.9% (PSAD1), 58.6% (PSAD2) and 60.8% (PSAD3) was reported. No concordance between the actual prostate weight and the estimated volume was found; moreover PSAD accuracy was of poor value in diagnosing PCa in comparison with PSA F/T.

  19. Safety upgrading program in NPP Mochovce

    International Nuclear Information System (INIS)

    Baumeister, P.

    1999-01-01

    EMO interest is to operate only nuclear power plants with high standards of nuclear safety. This aim EMO declare on preparation completion and commissioning of Mochovce Nuclear Power Plant. Wide co-operation of our company with International Atomic Energy Agency and west European Inst.ions and companies has been started with aim to fulfil the nuclear safety requirements for Mochovce NPP. Set of 87 safety measures was implemented at Mochovce Unit 1 and is under construction at Unit 2. Mochovce NPP approach to safety upgrading implementation is showed on chosen measures. This presentation is focused on the issues category III.(author)

  20. Establishment of ''Internal Rules'' and EDMS - Electronic Document Management System at NPP NEK

    International Nuclear Information System (INIS)

    Mandic, D.

    2012-01-01

    The main purpose of this paper is to present NPP's plans regarding the on-going project that started in November 2011, and that is related to the establishment of ''Internal Rules'' and EDMS - Electronic Document Management System.The term ''Internal Rules'' has been directly translated from Slovenian language (''Notranja pravila'') and adopted from the translated version of appropriate Slovenian national codes (ZVDAGA [1] in Slovenian language or PDAAIA [2] in English version). ''Internal Rules on capture and storage of materials in digital form'' refer to the rules adopted by a person as his/her internal act with reference to storage of his/her material. The main purpose for the establishment of the Internal Rules is to be able to justify that Krsko NPP is organized in compliance with the national codes covering that subject and strictly performing according to those Internal Rules. Once a Slovenian company achieves recognized and registered status in accordance with the Internal Rules document that has been certified and approved by the ARS (Archives of the Republic Slovenia), such company can utilize e-documents in the same way as they would utilize physical documents. Furthermore, a Slovenian company with approved Internal Rules can use e-documents in any legal aspect associated with the document's life cycle and the document's content as they would use the physical document or an authorized and approved copy of the physical document. Related to the nuclear regulatory background, NEK operates in compliance with the Slovenian legislation and also the US codes, regulations and guidelines; therefore, regarding the NPP specific documents, the Internal Rules and EDMS must also be in compliance with them. Since early 1990's, NEK has implemented document/records management system oriented towards supporting storage and management of physical documents/records and controlling distribution of active document copies. Document/records management system was supported by

  1. Guidelines for reliability analysis of digital systems in PSA context. Phase 1 status report

    International Nuclear Information System (INIS)

    Authen, S.; Larsson, J.; Bjoerkman, K.; Holmberg, J.-E.

    2010-12-01

    Digital protection and control systems are appearing as upgrades in older nuclear power plants (NPPs) and are commonplace in new NPPs. To assess the risk of NPP operation and to determine the risk impact of digital system upgrades on NPPs, quantitative reliability models are needed for digital systems. Due to the many unique attributes of these systems, challenges exist in systems analysis, modeling and in data collection. Currently there is no consensus on reliability analysis approaches. Traditional methods have clearly limitations, but more dynamic approaches are still in trial stage and can be difficult to apply in full scale probabilistic safety assessments (PSA). The number of PSAs worldwide including reliability models of digital I and C systems are few. A comparison of Nordic experiences and a literature review on main international references have been performed in this pre-study project. The study shows a wide range of approaches, and also indicates that no state-of-the-art currently exists. The study shows areas where the different PSAs agree and gives the basis for development of a common taxonomy for reliability analysis of digital systems. It is still an open matter whether software reliability needs to be explicitly modelled in the PSA. The most important issue concerning software reliability is proper descriptions of the impact that software-based systems has on the dependence between the safety functions and the structure of accident sequences. In general the conventional fault tree approach seems to be sufficient for modelling reactor protection system kind of functions. The following focus areas have been identified for further activities: 1. Common taxonomy of hardware and software failure modes of digital components for common use 2. Guidelines regarding level of detail in system analysis and screening of components, failure modes and dependencies 3. Approach for modelling of CCF between components (including software). (Author)

  2. Guidelines for reliability analysis of digital systems in PSA context. Phase 1 status report

    Energy Technology Data Exchange (ETDEWEB)

    Authen, S.; Larsson, J. (Risk Pilot AB, Stockholm (Sweden)); Bjoerkman, K.; Holmberg, J.-E. (VTT, Helsingfors (Finland))

    2010-12-15

    Digital protection and control systems are appearing as upgrades in older nuclear power plants (NPPs) and are commonplace in new NPPs. To assess the risk of NPP operation and to determine the risk impact of digital system upgrades on NPPs, quantitative reliability models are needed for digital systems. Due to the many unique attributes of these systems, challenges exist in systems analysis, modeling and in data collection. Currently there is no consensus on reliability analysis approaches. Traditional methods have clearly limitations, but more dynamic approaches are still in trial stage and can be difficult to apply in full scale probabilistic safety assessments (PSA). The number of PSAs worldwide including reliability models of digital I and C systems are few. A comparison of Nordic experiences and a literature review on main international references have been performed in this pre-study project. The study shows a wide range of approaches, and also indicates that no state-of-the-art currently exists. The study shows areas where the different PSAs agree and gives the basis for development of a common taxonomy for reliability analysis of digital systems. It is still an open matter whether software reliability needs to be explicitly modelled in the PSA. The most important issue concerning software reliability is proper descriptions of the impact that software-based systems has on the dependence between the safety functions and the structure of accident sequences. In general the conventional fault tree approach seems to be sufficient for modelling reactor protection system kind of functions. The following focus areas have been identified for further activities: 1. Common taxonomy of hardware and software failure modes of digital components for common use 2. Guidelines regarding level of detail in system analysis and screening of components, failure modes and dependencies 3. Approach for modelling of CCF between components (including software). (Author)

  3. The open PSA standard as a framework for migration of probabilistic models. Experiences with the KKB PSA

    International Nuclear Information System (INIS)

    Becker, G.; Hussels, U.; Epstein, S.; Rauzy, A.; Schubert, B.

    2008-01-01

    In its present state, the open PSA standard is helpful to determine capabilities of PSA approaches, which have been taken into account by those who formulated it. As soon, as tools come up, which can automatically bring a given PSA into the standard form, the data will be accessible by other software tools, which either are supplementary to the original one, or they may act in the context of quality control. Taking into account, that a PSA model represents a value of some two to ten person years (dependent on level of completeness and level of detail), it is important to have the data in a transparent way, which does not depend on proprietary formats, and can thus be used for more purposes than those, which are implemented in given PSA codes. (orig.)

  4. Cancer of the prostate - role of PSA

    International Nuclear Information System (INIS)

    Shittu, O.B.

    1999-02-01

    Since 1979 when prostate specific antigen (PSA), found in the cytoplasm of benign and malignant prostatic cells, was first purified, it has attained world wide popularity in prostate cancer detection. It is also a sensitive test for skeletal meta states from carcinoma of the prostate. Prostate cancer has become the number one cancer in men and constitutes 11% of all cancers. Approximately 50% of men over 50 years have symptoms referable to the lower urinary tract. 50% or more of patients at Ibadan present an advanced stage of the disease and are therefore not curable. Thus, lacking the skill to manage advanced manifestations, early detection and screening programs are the best means to reduce mortality due to prostate cancer

  5. Policy and System Approach (PSA: A primer

    Directory of Open Access Journals (Sweden)

    Chandrakant Lahariya

    2017-01-01

    Full Text Available A number of public health challenges have emerged at global and national level in the last two decades. The response to these challenges has rarely been swift and often “knee-jerk.” The national and state level program officials responsible for the activities often apportion the blame on weak health systems or fragmented health service delivery mechanisms, amongst other. In India, the viral illnesses (including those due to dengue and chikungunya are becoming the increasing realities. The Public health response of early identification, disease surveillance, reporting and the preventive and curative measures, remains suboptimal. The health challenges which require multidimensional interventions are usuallyattempted to be resolved through piece meal solutions. This article proposes “policy and system approach (PSA,” combining concepts of “Health in all policies” for intersectoral coordination and “health system approach” for intra-sectoral tackling of the emerging and existing health challenges.

  6. Daily Pill Can Prevent HIV PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2015-11-24

    This 60 second public service announcement (PSA) is based on the November 24, 2015 CDC Vital Signs report. Preexposure prophylaxis, or PrEP, is a daily medicine that can be used to prevent getting HIV. PrEP is for people who don’t have HIV but who are at very high risk for getting it from sex or injection drug use. Unfortunately, many people who can benefit from PrEP aren’t taking it.  Created: 11/24/2015 by National Center for HIV/AIDS, Viral Hepatitis, STD, and TB Prevention (NCHHSTP).   Date Released: 11/24/2015.

  7. Probabilistic precursor analysis - an application of PSA

    International Nuclear Information System (INIS)

    Hari Prasad, M.; Gopika, V.; Sanyasi Rao, V.V.S.; Vaze, K.K.

    2011-01-01

    Incidents are inevitably part of the operational life of any complex industrial facility, and it is hard to predict how various contributing factors combine to cause the outcome. However, it should be possible to detect the existence of latent conditions that, together with the triggering failure(s), result in abnormal events. These incidents are called precursors. Precursor study, by definition, focuses on how a particular event might have adversely developed. This paper focuses on the events which can be analyzed to assess their potential to develop into core damage situation and looks into extending Probabilistic Safety Assessment techniques to precursor studies and explains the benefits through a typical case study. A preliminary probabilistic precursor analysis has been carried out for a typical NPP. The major advantages of this approach are the strong potential for augmenting event analysis which is currently carried out purely on deterministic basis. (author)

  8. Regulatory review of probabilistic safety assessment (PSA) Level 2

    International Nuclear Information System (INIS)

    2001-07-01

    Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from deterministic analysis. Many regulatory authorities consider the current state of the art in PSA to be sufficiently well developed for results to be used centrally in the regulatory decision making process-referred to as risk informed regulation. For these applications to be successful, it will be necessary for the regulatory authority to have a high degree of confidence in the PSA. However, at the 1994 IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997, the IAEA and OECD Nuclear Energy Agency agreed to produce, in cooperation, guidance on Regulatory Review of PSA. This led to the publication of IAEA-TECDOC-1135 on the Regulatory Review of Probabilistic Safety Assessment (PSA) Level 1, which gives advice for the review of Level 1 PSA for initiating events occurring at power plants. This TECDOC extends the coverage to address the regulatory review of Level 2 PSA.These publications are intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable level of quality so that it can be used as the

  9. A framework for a quality assurance programme for PSA

    International Nuclear Information System (INIS)

    1999-08-01

    Reviews organized by the IAEA of probabilistic safety assessments (PSAs) of nuclear facilities have, in the past years, shown significant progress in the technical methods and data used for these studies. The IAEA has made a considerable effort to support the development of technical capabilities for PSA in Member States and in writing technical procedures for carrying out PSAs. However, the reviews have also shown significant deficiencies in quality assurance (QA) for PSAs, ranging from no QA at all to inappropriate, inefficient or unbalanced QA. As a PSA represents a very complex model which describes the risk associated with a nuclear facility, an appropriate and efficient QA programme is crucial to obtain a quality PSA. Historically, in the first integral PSAs, many of the PSA elements were handled by independent groups. These elements were finally integrated and put together in the overall model. Many of the interfaces between the elements or tasks were handled as appropriate by exchanging information in oral or written form. Since WASH-1400, the first integral PSA, the process of constructing the PSA model has been further developed. PSA elements previously considered separately can now be handled together with the capable software developed in recent years. Software has made interface control and data transfer easier to perform, but also permits the development of more detailed and complex models. Previously, QA for PSA projects was organized in an ad hoc manner and was sometimes very limited. In recent years, increasingly comprehensive QA programmes have been developed and implemented for PSA projects. Today, a comprehensive, effective and performance-oriented QA is considered to be essential for a reliable and credible PSA. This report describes the framework for developing an adequate QA programme for PSA studies. The framework is based on and is in accordance with the related QA guidelines of the IAEA for safety in nuclear power plants and other nuclear

  10. Akkuyu NPP – the first Turkish NPP. The new history of the project

    International Nuclear Information System (INIS)

    Tzocheva, V.

    2012-01-01

    An overview is given to the Turkish energy sector and nuclear power plans. The project for the construction of the first NPP in Turkey is presented. The general parameters of the Project are: CAPEX: $ 20 bln; Project design: NPP-2006; (VVER- 1200); Number of units: 4; Total capacity: 4 800 MW; Construction period: 2014 – 2023; PPA period; 15 years, fixed price terms. An account of the activities during 2011, the Worley Parsons participation are presented and a tentative project schedule is given

  11. Prostate specific antigen (PSA) in diagnosis of polycystic ovarian syndrome - a new insight.

    Science.gov (United States)

    Rudnicka, Ewa; Radowicki, Stanislaw; Suchta, Katarzyna

    2016-11-01

    Polycystic ovary syndrome (PCOS) is the commonest endocrine disorder and cause of androgen excess in women. Prostate specific antigen (PSA) could be a new marker of hyperandrogenism in PCOS. The aim of the study was to assess the concentration PSA (total PSA - TPSA and free PSA - fPSA) in 165 patients with PCOS and 40 healthy female controls, the relationship between PSA (TPSA and fPSA) and hormonal parameters and to determine the performance of PSA in diagnosis of PCOS. Total PSA was higher in PCOS group versus controls. The fPSA was below the lower detection levels among all patients. The median value of FAI was 4.31 in PCOS patients versus 1.79 in controls, p PSA serum levels in diagnostic of PCOS.

  12. Principles of tariff determination for NPP electric power generation

    International Nuclear Information System (INIS)

    Ratnikov, B.E.; Gitel'man, L.D.; Artemov, Yu.N.; Fiantsev, V.S.

    1988-01-01

    Foundations of price-setting and order of accounting arrangement for NPP electric power are considered. NPP tariffs are established proceeding from standard costs of power generation. The standards are differentiated as to NPP groups, depending on technical, regional and natural geographic factors, taking into account the facility type, unit capacity and the number of similar NPP units. The conclusion is made that under conditions of NPP economic independence expansion and creation of prerequisites for going over to self-financing principles and also due to the qualitatively new stage of nuclear power generation development the level of efficiency, forseen by the tariffs, should be increased

  13. Biotic elements of NPP techno-ecosystem

    International Nuclear Information System (INIS)

    Protasov, A.A.; Silaeva, A.A.

    2013-01-01

    Specific features of biotic elements in the NPP techno-ecosystems were considered and compared with natural ecosystems. Relationships between biotic communities and environmental factors that are specific to the techno-ecosystems were discussed, and the problems of limitation of biological hindrances in operation of equipment, principles of hydrobiological and environmental monitoring were considered.

  14. NPP Mochovce - a project of extraordinay significance

    International Nuclear Information System (INIS)

    Chwolik, I.; Debru, M.

    2000-01-01

    In this paper and in this presentation the reactor safety upgrading of two blocks of the NPP V-1 Bohunice, some results of participation on safety upgrading by the German-French consortium EUCOM (Framatome and Siemens-KWU) are presented. (author)

  15. Cernavoda NPP training programs The paper presents a general assessment of Cernavoda NPP personnel training programs,

    International Nuclear Information System (INIS)

    Valache, Cornelia

    2008-01-01

    The paper presents a general assessment of Cernavoda NPP personnel training programs, highlighting the role of training in human performance improvement. Cernavoda NPP Personnel Training and Authorization Department (PTAD) is responsible for the training of CNE Cernavoda NPP personnel and its contractors. PTAD is structured in a manner ensuring the support and response to all plant training, qualification and authorization requirements. The training of personnel is continuously adapted based on IAEA Guides and INPO/WANO recommendations, to keep with world standards, based on the internal and external reviews. At Cernavoda NPP the Training Concept and the Training Programs are based on SAT - Systematic Approach to Training. The Training Concept is established on a set of training documents (RD's, SI's, IDP's), which address all the SAT phases: Analysis, Design, Development, Implementation and Evaluation. The Training Programs are structured on the initial and continuing personnel training. Their content and goals are responding to the training specific needs for each plant major job family. In order to successfully support NPP training programs, CNPP training center has upgraded classrooms with new presentation facilities and there are plans to expand the space of the building, to develop additional operator and maintenance skills facilities. By responding in a timely and completely manner to all plant training requirements PTAD will help in rising human performance of Cernavoda NPP personnel, supporting the safe, efficient and cost effective production of power. (author)

  16. Cernavoda NPP simulator - next generation

    International Nuclear Information System (INIS)

    Tatar, F.; Ionescu, T.; Dascalu, M.

    2003-01-01

    Demand for extending the amount of training and scope for Cernavoda Unit 1 as well as the new trend in the simulator owners world, led to a change in the Romanian philosophy of simulator specification. Up to now the training was conducted on a Full Scope simulator, a 1:1 replica of Cernavoda Unit 1 reference plant. The present task is to define the simulation facilities and structure capable to meet the requirements for training, qualification and licensing of personnel for both Cernavoda Unit 1 and Unit 2. Obviously, the Cernavoda Unit 2 belongs to the same technological family but has rather different control room layout. Since this target requires a new simulator the costs would be rather high in accordance to the degree of automation of Cernavoda NPP. Therefore, depending on training requirements and financing, the Cernavoda Unit 1 simulator modernization, which also provides an alternative to full scope control room simulator, may be a viable option. Therefore the solution that with discuss for Cernavoda training extension is the migration of Cernavoda Unit 1 simulator to state-of-the-art. Consequently, the Cernavoda Unit 1 simulator modernization task will be organized as project including the following major items: 1. Rehost existing U1 simulation software from VAX 4500 to: - Best commercial multi-processor server for simulation server (HP, O/S Linux); - Best commercial single processor PC for I/O communications (HP, O/S Linux); 2. Replace DCC with enhanced emulated version: Best commercial individual PC for DCC emulation (HP, O/S Windows); Support for actual keyboards; Replacement of RAMTEK System and CONRAC Monitors with X terminals or PC's; 3. Conversion of AutoCAD-based panel graphic pages to RAVE-based; 4. Install the required software tools for developing enhanced simulation modules; 5. Replace the simulation modules with advanced modules; 6. Replace the present Windows Instructor Facilities with ISIS; 7. Development of a selection of MCR-U1 virtual

  17. Methods and results of a PSA level 2 for a German BWR of the 900 MWe class

    International Nuclear Information System (INIS)

    Loffler, H.; Sonnenkalb, M.

    2006-01-01

    On behalf of the federal Ministry for Environment, Nature Conservation and Reactor Safety (BMU) GRS has performed a PSA level 2 for a BWR type 69 NPP of the 900 MWe class, equipped with a N 2 inerted steel containment and a pressure suppression system. Integral deterministic accident analyses have been performed with the computer code MELCOR 1.8.5. Additional analyses have been done for those events and phenomena which are not or not sufficiently covered by MELCOR. The probabilistic event tree analysis begins with the core damage states received from PSA level 1, and it ends with the definition of release categories and the determination of their frequencies. Uncertainties about the frequency of core damage states and about events during the accident progression are taken into account by means of Monte Carlo simulations. If there is a core damage state there is a high probability (>50 %) for a very high and rapid release of radionuclides into the environment. This high conditional probability is due to the very low probability to retain a partly destroyed core inside the reactor pressure vessel (RPV) and because the containment almost certainly fails at the bottom of the control rod drives room after melt release from the failed RPV. (authors)

  18. Mutational analysis of photosystem I polypeptides in the cyanobacterium Synechocystis sp. PCC 6803. Targeted inactivation of psaI reveals the function of psaI in the structural organization of psaL

    Science.gov (United States)

    Xu, Q.; Hoppe, D.; Chitnis, V. P.; Odom, W. R.; Guikema, J. A.; Chitnis, P. R.; Spooner, B. S. (Principal Investigator)

    1995-01-01

    We cloned, characterized, and inactivated the psaI gene encoding a 4-kDa hydrophobic subunit of photosystem I from the cyanobacterium Synechocystis sp. PCC 6803. The psaI gene is located 90 base pairs downstream from psaL, and is transcribed on 0.94- and 0.32-kilobase transcripts. To identify the function of PsaI, we generated a cyanobacterial strain in which psaI has been interrupted by a gene for chloramphenicol resistance. The wild-type and the mutant cells showed comparable rates of photoautotrophic growth at 25 degrees C. However, the mutant cells grew slower and contained less chlorophyll than the wild-type cells, when grown at 40 degrees C. The PsaI-less membranes from cells grown at either temperature showed a small decrease in NADP+ photoreduction rate when compared to the wild-type membranes. Inactivation of psaI led to an 80% decrease in the PsaL level in the photosynthetic membranes and to a complete loss of PsaL in the purified photosystem I preparations, but had little effect on the accumulation of other photosystem I subunits. Upon solubilization with nonionic detergents, photosystem I trimers could be obtained from the wild-type, but not from the PsaI-less membranes. The PsaI-less photosystem I monomers did not contain detectable levels of PsaL. Therefore, a structural interaction between PsaL and PsaI may stabilize the association of PsaL with the photosystem I core. PsaL in the wild-type and PsaI-less membranes showed equal resistance to removal by chaotropic agents. However, PsaL in the PsaI-less strain exhibited an increased susceptibility to proteolysis. From these data, we conclude that PsaI has a crucial role in aiding normal structural organization of PsaL within the photosystem I complex and the absence of PsaI alters PsaL organization, leading to a small, but physiologically significant, defect in photosystem I function.

  19. Application of PSA in risk informed decision making

    International Nuclear Information System (INIS)

    Hari Prasad, M.; Vinod, Gopika; Saraf, R.K.; Ghosh, A.K.; Kushwaha, H.S.

    2006-01-01

    Probabilistic Safety Assessment (PSA) models have been successfully employed during design evaluation to assess weak links and carry out design modifications to improve system reliability and safety. Recently, studies are directed towards applying PSA in various decision making issues concerned with plant operations and safety regulations. This necessitates development of software tools like Living PSA, Risk Monitor etc. Risk Monitor is a PC based tool developed to assess the risk, based on the actual status of systems and components. Such tools find wide application with plant personnel and regulatory authorities since they can provide solutions to various plant issues and regulatory decision making issues respectively. (author)

  20. Regulatory review of probabilistic safety assessment (PSA) level 1

    International Nuclear Information System (INIS)

    2000-02-01

    Probabilistic safety assessment (PSA) is increasingly being used as part of the decision making process to assess the level of safety of nuclear power plants. The methodologies in use are maturing and the insights gained from the PSAs are being used along with those from the deterministic analysis. Many regulatory authorities consider that the current state of the art in PSA (especially Level 1 PSA) is sufficiently well developed that it can be used centrally in the regulatory decision making process - referred to as 'risk informed regulation'. For these applications to be successful, it will be necessary for regulatory authorities to have a high degree of confidence in PSA. However, at the IAEA Technical Committee Meeting on Use of PSA in the Regulatory Process in 1994 and at the OECD Nuclear Energy Agency Committee for Nuclear Regulatory Activities (CNRA) 'Special Issues' Meeting in 1997 on Review Procedures and Criteria for Different Regulatory Applications of PSA, it was recognized that formal regulatory review guidance for PSA did not exist. The senior regulators noted that there was a need to produce some international guidance for reviewing PSAs to establish an agreed basis for assessing whether important technological and methodological issues in PSAs are treated adequately and to verify that conclusions reached are appropriate. In 1997 the IAEA and OECD Nuclear Energy Agency agreed to produce in co-operation a technical document on the regulatory review of PSA. This publication is intended to provide guidance to regulatory authorities on how to review the PSA for a nuclear power plant to gain confidence that it has been carried out to an acceptable standard so that it can be used as the basis for taking risk informed decisions within a regulatory decision making process. The document gives guidance on how to set about reviewing a PSA and on the technical issues that need to be addressed. This publication gives guidance for the review of Level 1 PSA for

  1. Characteristics Studies of 125I- and total PSA antibody's Binding with prostate specific antigen (PSA) in Human Uterus Tumors

    International Nuclear Information System (INIS)

    Al-Mudaffar, S.; Al-Salihi, J.

    2005-01-01

    Two groups of uterus tumors (benign and malignant) postmenopausal patients were used to investigate the presence of prostate specific antigen (PSA). Preliminary experiments were performed to follow the binding of '1 25 I-anti total PSA antibody with PSA in uterus tissues homogenates of the two groups with their corresponding antigen and found to be (8.8,7.1%) for benign and malignant tumors, respectively. An Immuno Radio Metric Assay (IRMA) procedure was developed for measuring PSA in benign and malignant uterus tumors homogenates. The optimum conditions of the binding of 125 I-anti total PSA antibody with PSA were as follows: PSA concentration (150,200 μg protein),tracer antibody concentration (125,250 μg protein), p H (7.6,7.2), temp (15,25?C) and time (1.5 hrs) for postmenopausal benign and malignant uterus tumors tissue homogenates, respectively. The use of different concentrations of Na + and Mg 2+ ions were shown to cause an increase in the binding at concentration of (125,75 mΜ) of Na 1+ ions (75,225 mΜ) of Mg 2+ ions for benign and malignant uterus tumors homogenates, respectively, while the use of different concentrations of urea and polyethylene glycol (PEG) Caused a decrease in the binding with the increase in the concentration of each of urea and PEG in the both cases

  2. Detection of prostatic carcinoma: the role of TRUS, TRUS guided biopsy, digital rectal examination, PSA and PSA density.

    Science.gov (United States)

    Men, S; Cakar, B; Conkbayir, I; Hekimoglu, B

    2001-12-01

    The purpose of this study was to evaluate the efficacy of various diagnostic tests including transrectal ultrasound (TRUS), TRUS guided biopsy, digital rectal examination (DRE), prostate specific antigen (PSA), and prostate specific antigen density (PSAD) in detecting prostatic carcinomas. One hundred and thirty-four men underwent TRUS guided random, or directed and random sonographic biopsies of the prostate. The mean age was 64.67 (range, 31- 88) years. Indications for biopsy were abnormal findings suggesting prostatic carcinoma on DRE or increased levels of PSA, defined as 4.0 ng/ml or greater in a monoclonal antibody assay. PSAD was calculated by dividing the serum PSA in ng/ml to the volume of the entire prostate in cm3. The biopsy results were grouped as benign, malign and, prostatitis. The patients were also divided into three groups according to their PSA values. Of the 134 patients evaluated, 31 (23.1%) had prostate adenocarcinoma, 89 (66.4%) had benign prostatic tissue, hyperplasia or prostatic intraepithelial neoplasia, and 14 (10.4%) had prostatitis. The mean PSA and PSAD of the carcinoma group were significantly higher than those of the noncancer group. In the group of patients with PSA levels between 4 and 10 ng/ml, abnormal TRUS or DRE increased cancer detection rate, where neither PSA nor PSAD was capable of discriminating the patients with and without cancer. PSAD did not prove to be superior to the other diagnostic tests in this study. We recommend biopsy when either TRUS or DRE is abnormal in patients with PSA levels between 4 and 10 ng/ml. In the patients with PSA levels greater than 10 ng/ml, biopsy is indicated whatever the findings on TRUS or DRE are, since cancer detection rate is high.

  3. From prostate-specific antigen (PSA) to precursor PSA (proPSA) isoforms: a review of the emerging role of proPSAs in the detection and management of early prostate cancer.

    Science.gov (United States)

    Hori, Satoshi; Blanchet, Jean-Sebastien; McLoughlin, John

    2013-10-01

    Despite the popularity of PSA blood testing for prostate cancer, there are a number of important limitations of this popular serum marker including the limited ability to accurately distinguish patients with and without prostate cancer and those who harbour an aggressive form of the disease. This is especially true when the total PSA is PSA (proPSAs), with a special emphasis on [-2]proPSA in the detecion and management of early prostate cancer. The clinical utility of Prostate Health Index (phi) is also discussed. Despite the overall success of prostate-specific antigen (PSA) blood test, its use as a serum marker for prostate cancer has been limited due to the lack of specificity, especially in men presenting with a total PSA (tPSA) level of PSA testing has also resulted in an increase in the number of patients being diagnosed with low-grade, potentially clinically insignificant prostate cancer. There is therefore an urgent need for new markers that can accurately detect as well as differentiate patients with aggressive vs unaggressive prostate cancer. In this review, we discuss the emerging role of precursor forms of PSA (proPSAs) and the Prostate Health Index (phi) measurement in the detection and management of early stage prostate cancer. A literature search was conducted using PubMed® to identify key studies. Studies to date suggest that [-2]proPSA, a truncated form of proPSA is the most cancer-specific form of all, being preferentially expressed in cancerous prostatic epithelium and being significantly elevated in serum of men with prostate cancer. There is evidence to suggest that %[-2]proPSA measurement ([-2]proPSA/free PSA [fPSA] × 100) improves the specificity of both tPSA and fPSA in detecting prostate cancer. phi incorporating [-2]proPSA, fPSA and tPSA measurements has also yielded promising results and appears superior to tPSA and fPSA in predicting those patients with prostate cancer. Increased phi levels also seem to preferentially detect patients

  4. PSA Velocity Does Not Improve Prostate Cancer Detection

    Science.gov (United States)

    A rapid increase in prostate-specific antigen (PSA) levels is not grounds for automatically recommending a prostate biopsy, according to a study published online February 24, 2011, in the Journal of the National Cancer Institute.

  5. PSA LEVEL 3 DAN IMPLEMENTASINYA PADA KAJIAN KESELAMATAN PWR

    Directory of Open Access Journals (Sweden)

    Pande Made Udiyani

    2015-03-01

    Full Text Available Kajian keselamatan PLTN menggunakan metodologi kajian probabilistik sangat penting selain kajian deterministik. Metodologi kajian menggunakan Probabilistic Safety Assessment (PSA Level 3 diperlukan terutama untuk estimasi kecelakaan parah atau kecelakaan luar dasar desain PLTN. Metode ini banyak dilakukan setelah kejadian kecelakaan Fukushima. Dalam penelitian ini dilakukan implementasi PSA Level 3 pada kajian keselamatan PWR, postulasi kecelakan luar dasar desain PWR AP-1000 dan disimulasikan di contoh tapak Bangka Barat. Rangkaian perhitungan yang dilakukan adalah: menghitung suku sumber dari kegagalan teras yang terjadi, pemodelan kondisi meteorologi tapak dan lingkungan, pemodelan jalur paparan, analisis dispersi radionuklida dan transportasi fenomena di lingkungan, analisis deposisi radionuklida, analisis dosis radiasi, analisis perlindungan & mitigasi, dan analisis risiko. Kajian menggunakan rangkaian subsistem pada perangkat lunak PC Cosyma. Hasil penelitian membuktikan bahwa implementasi metode kajian keselamatan PSA Level 3 sangat efektif dan komprehensif terhadap estimasi dampak, konsekuensi, risiko, kesiapsiagaan kedaruratan nuklir (nuclear emergency preparedness, dan manajemen kecelakaan reaktor terutama untuk kecelakaan parah atau kecelakaan luar dasar desain PLTN. Hasil kajian dapat digunakan sebagai umpan balik untuk kajian keselamatan PSA Level 1 dan PSA Level 2. Kata kunci: PSA level 3, kecelakaan, PWR   Reactor safety assessment of nuclear power plants using probabilistic assessment methodology is most important in addition to the deterministic assessment. The methodology of Level 3 Probabilistic Safety Assessment (PSA is especially required to estimate severe accident or beyond design basis accidents of nuclear power plants. This method is carried out after the Fukushima accident. In this research, the postulations beyond design basis accidentsof PWR AP - 1000 would be taken, and simulated at West Bangka sample site. The

  6. [Prostate Specific Antigen (PSA) use in a national health department].

    Science.gov (United States)

    Panach-Navarrete, Jorge; Carratalá-Calvo, Arturo; Valls-González, Lorena; Sales-Maicas, María Ángeles; Martínez-Jabaloyas, José María

    2015-10-01

    PSA is a frequently used marker in the daily clinical practice for the diagnosis and management of prostate cancer. We analysed the use of PSA in our health department in patients with and without prostate cancer diagnosis. The registry of all PSA petitions in our health department during 2011 and 2012 was used. Demographic data were used to establish each year's population and the data corresponding to the prevalence of prostate cancer patients, performing a descriptive study. Thus, the use of PSA in patients with or without prostate cancer was studied. 25.700 PSA petitions are issued annually in our department over a total of 67.000 males older than 45. This entails a cost of 332.815 Euros annually. Within the group of patients with no prostate cancer diagnosis, it was noticed that the percentage of individuals with at least one annual PSA petition per decade of age is of 23% in males in their fifties, 40% in their sixties, 46% in their seventies, and 36% in their eighties or successive decades. Furthermore, in these cancer-free patients, around 3.800 annual petitions fall on individuals over 75 and with PSA under 4 ng/ml, from which 20% are repeated petitions over the same individual in the same year. Over 1100 males under 45 have an annual PSA. Regarding the average PSA value for decade of age in cancer-free patients, it is of 0.89 +/- 0.4 ng/ml in the forties decade, 1.26 +/- 1.07 ng/ml in the fifties, 1.67 +/- 1.38 ng/ml in the sixties, 1.96 +/- 1.78 ng/ml in the seventies, and 2.24 +/- 2.16 ng/ml in the eighties. We ascertained, also, that for every 144 PSA petitions one prostate cancer case is diagnosed. Regarding the use of this marker in cancer patients, 1.800 petitions are destined to patients follow up annually, and over 200 fall on the newly diagnosed cases. Even though annually less than 50% of males get PSA petitions in any decade of age, its use is sometimes incorrect, including repeated petitions in a short period of time or in individuals of

  7. Influence of PSA, PSA velocity and PSA doubling time on contrast-enhanced 18F-choline PET/CT detection rate in patients with rising PSA after radical prostatectomy

    International Nuclear Information System (INIS)

    Schillaci, Orazio; Calabria, Ferdinando; Tavolozza, Mario; Caracciolo, Cristiana Ragano; Orlacchio, Antonio; Danieli, Roberta; Simonetti, Giovanni; Agro, Enrico Finazzi; Miano, Roberto

    2012-01-01

    To evaluate the accuracy of contrast-enhanced 18 F-choline PET/CT in restaging patients with prostate cancer after radical prostatectomy in relation to PSA, PSA velocity (PSAve) and PSA doubling time (PSAdt). PET/CT was performed in 49 patients (age range 58-87 years) with rising PSA (mean 4.13 ng/ml) who were divided in four groups according to PSA level: ≤1 ng/ml, 1 to ≤2 ng/ml, 2 to ≤4 ng/ml, and >4 ng/ml. PSAve and PSAdt were measured. PET and CT scans were interpreted separately and then together. PET/CT diagnosed relapse in 33 of the 49 patients (67%). The detection rates were 20%, 55%, 80% and 87% in the PSA groups ≤1, 1 to ≤2, 2 to ≤4 and >4 ng/ml, respectively. PET/CT was positive in 7 of 18 patients (38.9%) with a PSA ≤2 ng/ml, and in 26 of 31 (83.9%) with a PSA >2 ng/ml. PET/CT was positive in 7 of 25 patients (84%) with PSAdt ≤6 months, and in 12 of 24 patients (50%) with PSAdt >6 months, and was positive in 26 of 30 patients (86%) with a PSAve >2 ng/ml per year, and in 7 of 19 patients (36.8%) with PSAve ≤2 ng/ml per year. PET alone was positive in 31 of 49 patients (63.3%), and of these 31 patients, CT was negative in 14 but diagnosed bone lesions in 2 patients in whom PET alone was negative. CT with the administration of intravenous contrast medium did not provide any further information. Detection rate of 18 F-choline imaging is closely related to PSA and PSA kinetics. In particular, 18 F-choline PET/CT is recommended in patients with PSA >2 ng/ml, PSAdt ≤6 months and PSAve >2 ng/ml per year. CT is useful for detecting bone metastases that are not 18 F-choline-avid. The use of intravenous contrast agent seems unnecessary. (orig.)

  8. [Use of prostatic specific antigen in primary care (PSA)].

    Science.gov (United States)

    Panach-Navarrete, J; Gironés-Montagud, A; Sánchez-Cano, E; Doménech-Pérez, C; Martínez-Jabaloyas, J M

    2017-04-01

    In the literature it is shown that the use of PSA is occasionally wrong, by requesting this marker in very young or very old men, and repeated measurements in short periods of time. The main objective of this study was to describe the use of PSA in daily practice by primary care physicians in our area, dealing with aspects such as the importance of patient age, the value in the screening for prostate cancer, or the subjective beliefs about its usefulness. A secondary objective was the comparison of use, and beliefs among doctors who claim to know PSA well, and those who do not. A descriptive and comparative study was conducted using questionnaires that were handed to primary care doctors in all health centres in our area. A descriptive analysis was performed and response rates among doctors who thought they had enough information about PSA, and those who did not, were compared using the Chi-squared test. A total of 103 questionnaires were received from the physicians, with 83.5% claiming to have sufficient knowledge about the PSA. The professionals in this latter group request PSA at an earlier age (P=.029), with a higher frequency (P=.011) and have more doubts about its usefulness (P=.009) than those with less knowledge. Almost half (49.5%) said they request less than 50 determinations per year, and 33% between 50 and 100. More than half (53.4%) of doctors would not request the first PSA on a patient until their 50s, and up to 49% request it up to 80 years. The true value of PSA has been established many times by 64.1% of requesters, and 29.1% believe it is unhelpful in the diagnosis of cancer. In our study, 64% of primary care physicians have considered the true value of the PSA several times, and 29% believe it to be of little use in the diagnosis of prostate cancer. In addition, some data suggest it has limited use due to the fact that 50% made less than 50 PSA requests per years, and 28% of the professionals would never request it on a male without urinary

  9. Review of APR+ Level 2 PSA. Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Lehner, John R. [Brookhaven National Lab. (BNL), Upton, NY (United States); Mubayi, Vinod [Brookhaven National Lab. (BNL), Upton, NY (United States); Pratt, W. Trevor [Brookhaven National Lab. (BNL), Upton, NY (United States); Kim, Do Sam [Korea Institute of Nuclear Safety (KINS), Daejeon (Korea, Republic of); Cho, Yong Jin [Korea Institute of Nuclear Safety (KINS), Daejeon (Korea, Republic of); Cho, Sang Jin [Korea Institute of Nuclear Safety (KINS), Daejeon (Korea, Republic of); Kim, In Goo [Korea Institute of Nuclear Safety (KINS), Daejeon (Korea, Republic of)

    2012-02-17

    Brookhaven National Laboratory (BNL) assisted the Korea Institute of Nuclear Safety (KINS) in reviewing the Level 2 Probabilistic Safety Assessment (PSA) of the APR+ Advanced Pressurized Water Reactor (PWR) prepared by the Korea Hydro & Nuclear Power Co., Ltd (KHNP) and KEPCO Engineering & Construction Co., Inc. (KEPCO-E&C). The work described in this report involves a review of the APR+ Level 2 PSA submittal [Ref. 1]. The PSA and, therefore, the review is limited to consideration of accidents initiated by internal events. As part of the review process, the review team also developed three sets of Requests for Additional Information (RAIs). These RAIs were provided to KHNP and KEPCO-E&C for their evaluation and response. This final detailed report documents the review findings for each technical element of the PSA and includes consideration of all of the RAIs made by the reviewers as well as the associated responses. This final report was preceded by an interim report [Ref. 2] that focused on identifying important issues regarding the PSA. In addition, a final meeting on the project was held at BNL on November 21-22, 2011, where BNL and KINS reviewers discussed their preliminary review findings with KHNP and KEPCO-E&C staffs. Additional information obtained during this final meeting was also used to inform the review findings of this final report. The review focused not only on the robustness of the APR+ design to withstand severe accidents, but also on the capability and acceptability of the Level 2 PSA in terms of level of detail and completeness. The Korean nuclear regulatory authorities will decide whether the PSA is acceptable and the BNL review team is providing its comments for KINS consideration. Section 2.0 provides the basis for the BNL review. Section 3.0 presents the review of each technical element of the PSA. Conclusions and a summary are presented in Section 4.0. Section 5.0 contains the references.

  10. Child Injury: What You Need to Know PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2012-04-16

    This 60 second PSA is based on the April 2012 CDC Vital Signs report. Many childhood deaths and injuries are preventable, including those caused by crashes, suffocation, poisoning, drowning, fires, and falls. The PSA discusses ways to help prevent these deaths and injuries.  Created: 4/16/2012 by Centers for Disease Control and Prevention (CDC).   Date Released: 4/16/2012.

  11. Incorporation of organizational failures in PSA

    International Nuclear Information System (INIS)

    Davoudian, K.; Wu, Jya-Syin; Apostolakis, G.

    1993-01-01

    A large portion of the work performed at nuclear power plants follows standardized flow paths. For example, although components on which the maintenance crew works differ from one assignment to the next, all assignments basically follow the same process: requesting, reviewing, planning, scheduling, executing, testing, and documenting the work. In general, the term open-quotes work processclose quotes is used to refer to a standardized sequence of tasks designed within the operational environment of an organization to achieve a specific goal. The predictable nature of work processes suggests that a systematic analysis can be conducted to identify the desirable design of the process and to develop performance measures with respect to the strengths and weaknesses in the process. Furthermore, because of the close relationship of the work process to plant performance and plant safety, it is believed that such an analysis will facilitate the inclusion of organizational factors into probabilistic safety assessment (PSA) methodology. The Work Process Analysis Model (WPAM) has been developed with these goals in mind

  12. Development of PSA procedure for a criticality in reprocessing facilities

    International Nuclear Information System (INIS)

    Endo, Shigeki; Takanashi, Mitsuhiro; Ueda, Yoshinori

    2012-08-01

    Utilization of risk information for the nuclear safety regulation is being discussed in Japan. The development of probabilistic safety assessment (PSA) procedure is indispensable for the utilization of risk information. The Japan Nuclear Energy Safety Organization (JNES) has been conducting trial PSA to a model plant for major events, i.e. hydrogen explosion, solution boiling, rapid decomposition of TBP complexes, criticality, solvent fire, leakage of molten glass, leakage of high active concentrated liquid waste, loss of all AC electricity, drop of a fuel assembly, for the purpose of developing the PSA procedure for reprocessing facilities. For criticality events results of trial PSA were summarized as a report in which how to evaluate an amount of radioactive materials released from a facility and a health effect on the public were emphasized. Therefore, for criticality events the results of trial PSA were summarized in this report to emphasize procedures from making event progression scenarios to quantifying event sequences, which were not handled in the previous report, in a style of a document describing PSA procedures. (author)

  13. Generation of risk importance information from severe accident PSA model

    International Nuclear Information System (INIS)

    Seo, Mi Ro; Kim, Hyeong Taek; Moon, Chan Kook

    2012-01-01

    One of the important objects conducting Probabilistic Safety Assessment (PSA) is the relative evaluation of importance of the component or function that is greatly affected to the plant safety. This evaluation is performed by the importance assessment methods such as Risk Reduction Worth, Risk Achievement Worth, and Fuss el Vessley method from the aspect of core damage frequency (CDF). In the Level 1 PSA model, the importance of each component can be evaluated since the CDF is calculated by the combination of the branch probability of event tree and the component failure probability in the fault tree. But, the Level 2 PSA model in order to assess the containment integrity cannot evaluate the risk importance by the above methods because the model is consisted of 3 parts, plant damage status, containment event tree, and source term category. So, in the field that the Level 2 PSA risk importance information should be reflected, such as maintenance rule program, risk importance has been determined by the subjective judgment of the model developer. This study was performed in order to generate the risk importance information more objectively and systematically in the Level 2 PSA model, focused on the containment event tree in the domain PHWR Level 2 PSA model

  14. Radiotherapy for men with PSA failure following radical prostatectomy

    International Nuclear Information System (INIS)

    Shiota, Masaki; Noma, Hideya; Yamaguchi, Akito

    2007-01-01

    We evaluated the efficacy of salvage external beam radiotherapy (RT) to the prostate bed for men with prostate-specific antigen (PSA) failure following radical prostatectomy. Fourteen patients underwent RT for PSA failure following radical prostatectomy between 1999 and 2000. Median follow-up was 24 months. Median PSA level before RT was 0.51 ng/ml. Radiation dose was 60 Gy or 61.4 Gy. The 3-year actuarial biochemical disease-free survival (bDFS) rate was 40%. The biochemical effectiveness of RT was better in cases with a PSA level of less than 1 ng/ml compared to that in cases with a level higher than 1 ng/ml. The PSA level before RT and surgical margin involvement were identified as prognostic factors for bDFS. No patients experienced grade 3 toxicity. RT for PSA failure following radical prostatectomy seems to be very effective and was only slightly toxic during a limited follow-up period. (author)

  15. Prostate specific antigen (PSA) kinetics after 125I seed implantation (permanent Brachytherapy) for localized prostate cancer

    International Nuclear Information System (INIS)

    Ebara, Shin; Katayama, Norihisa; Manabe, Daisuke

    2007-01-01

    Prostatic specific antigen (PSA) bounce (over 0.1 ng/ml) was observed in 25.7% of patients (18 of 70) within 30 month after brachytherapy in our series. Several reports demonstrated that PSA bounce was observed in 30-50% of patients, observed within 2 years after brachytherapy and continued following 1 year. PSA bounce should be considered when assessing a patient with a rising PSA level before PSA nadir was achieved 4-5 years after brachytherapy. (author)

  16. Intelligent system for accident identification in NPP

    International Nuclear Information System (INIS)

    Hernandez, J.L.

    1998-01-01

    Accidental situations in NPP are great concern for operators, the facility, regulatory bodies and the environmental. This work proposes a design of intelligent system aimed to assist the operator in the process of decision making initiator events with higher relative contribution to the reactor core damage occur. The intelligent System uses the results of the pre-operational Probabilistic safety Assessment and the Thermal hydraulic Safety Analysis of the NPP Juragua as source for building its knowledge base. The nucleus of the system is presented as a design of an intelligent hybrid from the combination of the artificial intelligence techniques fuzzy logic and artificial neural networks. The system works with variables from the process of the first circuit, second circuit and the containment and it is presented as a model for the integration of safety analyses in the process of decision making by the operator when tackling with accidental situations

  17. SAT for NPP personnel training in USA

    International Nuclear Information System (INIS)

    Spinney, R.

    1995-01-01

    This discussion addressed the experience with the application of SAT at USA NPPs. In particular, the transition of NPP training processes, staff composition, and reporting structure from the TMI accident to present. As well, oversight and guidance activities of the INPO and more intensive inspection by the NRC began during this period. The average NPP training staff grew to 30-40 per unit, along with a change in reporting line from plant to corporate management. With the reduction of resources occurring in the late 1980s, overall training staff size decreased, the composition changed, and reporting line reverted to plant management. The overall lessons-learned for application of the SAT consisted of the need for simplification, management involvement, and exploitation of the technology

  18. NPP pipeline devices improvement by unconventional means

    International Nuclear Information System (INIS)

    Ionajtis, R.R.

    1998-01-01

    The main nontraditional approaches to NPP pipeline devices improvement including complex use of the set of principles and engineering criteria assuring reliability increase, introduction of stepwise throttling, application of alloys with shape memory, creation of principally new valves available for repair with removable (small-sized) drives are considered. It is shown that the problems of valves for NPP may be successfully solved by application of alloys and devices with shape memory which provide high applied forced and deformations, and as a results of which the compactness and complex functionality will be considerably (by the factor of 8-10) improved, i.e. realization of control, transmission, amplifying and actuation functions in a single device is possible. That is of great importance when designing passive safety means

  19. Modernization of the oldest Swedish NPP

    International Nuclear Information System (INIS)

    Hagberth, Ronald

    1998-01-01

    OKG operates three BWR units of ABB design: Oskarshamn 1 with a net capacity of 440' MW, Oskarshamn 2 of 600 MW and Oskarshamn 3 of 1160 MW. Oskarshamn 1 NPP was commissioned in 1972 as the first commercial nuclear unit in Sweden. After more than twenty years of successful operation, the unit is now also the first reactor in Sweden to undergo a large safety modernization program. In the year 2000 the Oskarshamn 1 NPP will be modernized to a high level of safety standard and ready for operation for another period of at least 20 years. Experience gained can be used when modernizing other NPPs. The investment program for life extension is reasonable and shows that NPPs can be operated with an expected life span of more than 40 years at an ever-increasing safety level and still be very competitive in a deregulated market. (author)

  20. Mass spectrometry measurements of prostate-specific antigen (PSA) peptides derived from immune-extracted PSA provide a potential strategy for harmonizing immunoassay differences.

    Science.gov (United States)

    Klee, Eric W; Bondar, Olga P; Goodmanson, Marcia K; Trushin, Sergey A; Singh, Ravinder J; Anderson, N Leigh; Klee, George G

    2014-04-01

    Harmonization of prostate-specific antigen (PSA) immunoassays is important for good patient care. The specificity of the antibodies used to detect circulating PSA could cause differences in the PSA measurements. We used mass spectrometry (MS) to quantitate the concentration of five peptides cleaved from trypsin digestion of PSA and compared these measurements with six automated immunoassays. Linear regression and a mixed-effects model were used to analyze the results. PSA measurements from the immunoassays and the five MS peptide assays were highly correlated (R(2) > 0.99), but the recovery of the World Health Organization standard and the regression slopes differed across assays. The same relative patterns of immunoassay differences were seen in comparing their results with each of the five MS peptide measurements from different parts of the circulating PSA molecules. Mass spectrometry quantitation of peptides derived from trypsin digestion of immune-extracted PSA could be used to harmonize PSA immunoassays.

  1. Aspects of accident management in Cernavoda NPP

    International Nuclear Information System (INIS)

    Dascalu, N.

    1999-01-01

    As a general conclusion, the accident management system as implemented at Cerna voda NPP is expected to be appropriate for handling a severe accident, should it occur, in such a way that the environmental radiological consequences would be insignificant and radiation exposure of the personnel be within recommendations. It is recognized, however, that continued development and verification of the system as well as effective personnel training programs are essential to maintain the safety level achieved. (author)

  2. Organizational aspects of NPP operator training

    International Nuclear Information System (INIS)

    Vel'chinskij, V.I.

    1992-01-01

    The main points of the document regulating the selection, prepation, permission for work and in-service control of NPP personnel developed on the basis of the IAEA requirements are considered. The specialists engaged for work are subjected to qualification, medical, professional, psychological and psychophysiological selections. The scheduled monthly instructive lessons are conducted during the work. The antiaccident and fire-fighting trainings are organized not rarely than twice in three months

  3. Current status of NPP generation IV

    International Nuclear Information System (INIS)

    Yohanes Dwi Anggoro; Dharu Dewi; Nurlaila; Arief Tris Yuliyanto

    2013-01-01

    Today development of nuclear technology has reached the stage of research and development of Generation IV nuclear power plants (advanced reactor systems) which is an innovative development from the previous generation of nuclear power plants. There are six types of power generation IV reactors, namely: Very High Temperature Reactor (VHTR), Sodium-cooled Fast Reactor (SFR), Gas-cooled Fast Reactor (GFR), Lead-cooled Fast Reactor (LFR), Molten Salt Reactor (MSR), and Super Critical Water-cooled Reactor (SCWR). The purpose of this study is to know the development of Generation IV nuclear power plants that have been done by the thirteen countries that are members of the Gen IV International Forum (GIF). The method used is review study and refers to various studies related to the current status of research and development of generation IV nuclear power. The result of this study showed that the systems and technology on Generation IV nuclear power plants offer significant advances in sustainability, safety and reliability, economics, and proliferation resistance and physical protection. In addition, based on the research and development experience is estimated that: SFR can be used optimally in 2015, VHTR in 2020, while NPP types GFR, LFR, MSR, and SCWR in 2025. Utilization of NPP generation IV said to be optimal if fulfill the goal of NPP generation IV, such as: capable to generate energy sustainability and promote long-term availability of nuclear fuel, minimize nuclear waste and reduce the long term stewardship burden, has an advantage in the field of safety and reliability compared to the previous generation of NPP and VHTR technology have a good prospects in Indonesia. (author)

  4. 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

  5. [The incidence of the variability of the free PSA/total PSA ratio on the early diagnosis of prostate cancer].

    Science.gov (United States)

    de La Taille, A; Houlgatte, A; Houdelette, P; Berlizot, P; Fournier, R; Ricordel, I

    1997-06-01

    In patients with moderate elevation of total PSA, the use of the Free PSA/Total PSa ratio (F+T PSA) has been shown to be useful in the diagnosis of impalpable prostatic cancer. However, the cut-off values proposed in the literature vary from study to study and according to the immunoassay kit used. Our prospective study was designed to compare 3 different kits (Tandem, Cis Bio and Immunocorp) on the same series of patients in order to determine on the basis of these results and a review of the literature, the optimal ratio for which prostatic biopsies should be indicated in the presence of an isolated elevation of Total PSA. Serum samples from 141 patients (43 cancers and 98 cases of histologically confirmed BPH) were included. Assays were performed concomitantly with histological examination of the prostate, using Tandem, Cis Bio and Immunocorp kits. In the overall patient population, Total PSA and Free PSA assays were statistically different for the 3 kits (p digital rectal examination and a F/T PSA ratio less than 0.10, the cancer detection rate (number of biopsies required to diagnose 1 cancer) was 1.66 for Tandem, 1 for Cis Bio and 1.87 for Immunocorp versus 7.7 while, when the F/T PSA ratio was greater than 0.22, the cancer detection rate was infinity (no cancer beyond this limit), 12.5 and 23, respectively. The F/T PSA ratio increases the specificity of prostatic cancer detection in patients with a total PSA between 4 and 10.0 ng/ml with a non-suspicious digital rectal examination, therefore resulting in a reduction of the useless biopsy rate and defining more relevant indications for biopsies in the case of periodic follow-up. The systematic indication of this ratio in the screening context cannot be recommended, but it can be useful to demonstrate stage TIC tumours in order to avoid numerous useless biopsies.

  6. The NPP Isar comprehensive Aging Management Program

    International Nuclear Information System (INIS)

    Zander, Andre; Ertl, Stefan

    2012-01-01

    The majority of System, Structure and Components (SSC) in a nuclear power plants are designed to experience a service life, which is far above the intended design life. In most cases, only a small percentage of SSCs are subject to significant aging effects, which may affect the integrity or the function of the component. The process of aging management (AM) has the objective to monitor and control degradation effects which may compromise safety functions of the plant. And furthermore, to ensure, that testing and maintenance programs sufficiently provide preventive measures to control degradation effects. Safety-related aspects and the targeted high availability of the power plant as well as the requirements stipulated by German regulatory authorities prompted the operator of NPP ISAR to introduce an aging surveillance program. The NPP Isar as well as the German NPPs has to be following in the scope of aging management the KTA 1403 guideline. The NPP Isar surveillance program based on the KTA 1403 guideline covers the following aspects: - Scoping and screening of safety relevant Systems, Structures and Components (SSC); - Identification of possible degradation mechanisms for safety relevant SSC; - Ensure, that testing and maintenance programs sufficiently provide preventive measures to control degradation effects; - Transferability check of industry experience (internal and external events); - Annual preparation of an AM status report. (author)

  7. Optimization of radiation protection at Bohunice NPP

    International Nuclear Information System (INIS)

    Dobis, L.; Svitek, J.

    2003-01-01

    Bohunice Nuclear Power Plant is situated in south - western part of Slovakia about 50 km away from Bratislava. There are four PWR reactors 440 MW e each - two units with reactors WWER - 230 (V1 NPP) and two units with WWER - 213 (V2 NPP). requirements for the optimization process are given in the mentioned Code No.12 of Ministry of Health. Code 12 stipulates the technical and organizational requirements for proving the Rational Achievable Level (RAL) of radiation protection. This level can be proved by means of the comparison of the dose distribution to the costs of protection. An example of two figures of dose constraints is: collective dose 20 man mSv for the specific task; individual exposure 1 mSv per day. The values of the financial equivalents of personal exposure - so called the alpha coefficients - are used for the calculation of the benefit of proposed measures. Impact of legislative changes into Bohunice NPP and optimization process are presented. Apparently the new law and the associate code created a base of transparent and understandable policy of radiation protection and optimization in Slovak Republic. The radiation protection legislative was implemented into the praxis and persons became familiar with it. Defining clear and unambiguous terms facilitated the communication between users and the regulatory body - State Health Institute. Optimization was generally accepted by the workers and managers and began to be a part of safety culture of operation at nuclear power plants. (authors)

  8. Knowledge management in the NPP domain

    International Nuclear Information System (INIS)

    Nilsen, Svein; Bisio, Rossella; Ludvigsen, Jan Tore

    2004-03-01

    This report gives an outlook on Knowledge Management (KM) activities within NPP related establishments as of today. There may be less activity in the NPP world as compared to many other industrial sectors. Still there is an awakening within the NPP industry demanding that KM should be attended to at a larger scale. The most notable reason for this is maybe an imminent increase in the number of people going into retirement. The types of establishments involved cover the major kinds such as utilities, research institutes and worldwide nuclear organizations. The report sums up a few of those efforts that are presently being implemented. Moreover the report looks at general advancements within the field of knowledge management. Simply stated the endeavours belong to either one of two classes. The first class emphasize the use of technology to solve knowledge management problems. The second class regard knowledge management as a problem pertaining to human factors and organizational issues. This report maintain that knowledge management initiatives should make due considerations to both perspectives. This report also sums up the Halden Reactor Project short term KM initiative. (Author)

  9. Safety parameter display system for Kalinin NPP

    International Nuclear Information System (INIS)

    Andreev, V.I.; Videneev, E.N.; Tissot, J.C.; Joonekindt, D.; Davidenko, N.N.; Shaftan, G.I.; Dounaev, V.G.; Neboyan, V.T.

    1995-01-01

    The paper discusses the safety parameter display system (SPDS), which is being designed for Kalinin NPP. The assessment of the safety status of the plant is done by the continuous monitoring of six critical safety functions and the corresponding status trees. Besides, a number of additional functions are realized within the scope of KlnNPP, aimed at providing the operator and the safety engineer in the main control room with more detailed information in accidental situation as well as during the normal operation. In particular, these functions are: archiving, data logs and alarm handling, safety actions monitoring, mnemonic diagrams indicating the state of main technological equipment and basic plant parameters, reference data, etc. As compared with the traditional scope of functions of this kind of systems, the functionality of KlnNPP SPDS is significantly expanded due to the inclusion in it the operator support function ''computerized procedures''. The basic SPDS implementation platform is ADACS of SEMA GROUP design. The system architecture includes two workstations in the main control room: one is for reactor operator and the other one for safety engineer. Every station has two CRT screens which ensures computerized procedures implementation and provides for extra services for the operator. Also, the information from the SPDS is transmitted to the local crisis center and to the crisis center of the State utility organization concern ''Rosenergoatom''. (author). 3 refs, 6 figs, 1 tab

  10. Summary report of already existing guidance on the implementation of External Hazards in extended Level 1 PSA

    International Nuclear Information System (INIS)

    Klug, J.; Kumar, M.; Prochaska, J.; Brac, P.; Vasseur, D.; Brinkman, H.; Kahia, S.; Nitoi, Mirela; Apostol, M.; Georgescu, G.; Volkanovski, Andrija; Mustoe, J.; Alzbutas, R.; La Rovere, S.

    2015-08-01

    The report provides a summary of already existing guidance on the implementation of external hazards in extended level 1 PSA. It summarized the lessons learnt from existing standards, existing gaps and possibility for future development within the work-package WP22 'How to introduce hazards in L1 PSA and all possibilities of events combinations'. The report is focused on the four following areas, for several hazards: 1) Impact on the SSCs modelled in L1 PSA event trees; 2) Impact on Human Reliability Assessment modelling in L1 PSA; 3) Site impact modelling in L1 PSA event trees; 4) Link between external initiating events of PSA and NPP design basis conditions. During the review of existing guidance, it appeared that many of the references form a suitable basis to introduce external hazards in L1 PSA including event combination. Available guidelines provide usable recommendations to evaluate failure probabilities of SSCs depending on the influence of single hazard or events combination. The most detailed guidelines are devoted to the seismic events and fires. Even if these guidelines deal only with single event impact, they can be also used for combined events purpose to evaluate particular effects induced by analyzed external hazards. Guidelines provide general systematic framework how to determine the scope of SSCs for extended PSA and failure modes (develop an extended list of components). In general available guidelines provide detailed framework for analysis of seismic event. The other external hazards are not always covered so deeply. This is probably caused by specific site nature of these hazards like external floods, fires etc. In case of HRA, more detailed information and HRA models are available for seismic events or fire events. For the other external hazards, the literature with regard to HRA is not well developed. The PSA for external hazards should take account the potential for human response to be affected by the external event. More

  11. Financial and organizational models of NPP construction projects

    International Nuclear Information System (INIS)

    Ivanov, Timur

    2010-01-01

    The recent evolution of financial and organizational models of NPP projects can be truly reputed to open a new page of the world market of NPP construction. The definition of the concrete model is based mostly on specific cooperation backgrounds and current terms and conditions under which the particular NPP project is being evolved. In this article the most commonly known strategies and schemes of financing structuring for export NPP construction projects are scrutinized. Special attention is paid to the analysis of BOO/BOT models which are based on the public-private partnership. Most BOO/BOT projects in the power sector has Power Purchase Agreements (PPA) as an integral part of them. The PPA key principles are studied here as well. The flexibility and adaptability of the public-private partnership models for financing and organization of the NPP projects contributes substantially to the competitiveness of the NPP projects especially under current economic conditions. (orig.)

  12. Predictor of response to salvage radiotherapy in patients with PSA recurrence after radical prostatectomy. The usefulness of PSA doubling time

    International Nuclear Information System (INIS)

    Numata, Kousaku; Azuma, Koji; Hashine, Katsuyoshi; Sumiyoshi, Yoshiteru

    2005-01-01

    We assessed predictors of response to salvage radiotherapy (sRT) in patients with prostate-specific antigen (PSA) recurrence after radical prostatectomy. A total of 21 patients receiving sRT for PSA recurrence without systemic progression after radical prostatectomy had medical records available for retrospective review. We defined sRT as external beam radiotherapy for patients with a continuous increase in PSA level≥0.2 ng/ml after radical prostatectomy. Response was defined as achievement of a PSA nadir of ≤0.1 ng/ml. Various pre-treatment parameters were evaluated retrospectively. The median follow-up period after sRT was 38 months. Of the 21 patients, 15 were good responders (71%). The only predictive factor was PSA doubling time (PSADT). Age and PSA level at diagnosis, Gleason score and surgical margin status were not significant predictors of response. The median PSADT in responders was 6.2 months versus 1.9 months in non-responders (P=0.019). The patients with a PSADT of ≥5 months were all responders. PSADT appears to be a good predictor of response to sRT. sRT was especially effective when PSADT was ≥5 months. (author)

  13. On psichological problem of NPP operation and control

    International Nuclear Information System (INIS)

    Mashin, V.A.

    1994-01-01

    The role of psichological factor as a reserve for increasing NPP safety connected with human factor is discussed. It is emphasized that the process of NPP personnel professional training should not be restricted by formation of a certain set of knowledge, skills and experience. It is necessary to initiate ability for constant self-developing. Control for assurance of effective interaction of the whole NPP personnel is an important problem

  14. Murine Polyomavirus Virus-Like Particles Carrying Full-Length Human PSA Protect BALB/c Mice from Outgrowth of a PSA Expressing Tumor

    Science.gov (United States)

    Eriksson, Mathilda; Andreasson, Kalle; Weidmann, Joachim; Lundberg, Kajsa; Tegerstedt, Karin

    2011-01-01

    Virus-like particles (VLPs) consist of capsid proteins from viruses and have been shown to be usable as carriers of protein and peptide antigens for immune therapy. In this study, we have produced and assayed murine polyomavirus (MPyV) VLPs carrying the entire human Prostate Specific Antigen (PSA) (PSA-MPyVLPs) for their potential use for immune therapy in a mouse model system. BALB/c mice immunized with PSA-MPyVLPs were only marginally protected against outgrowth of a PSA-expressing tumor. To improve protection, PSA-MPyVLPs were co-injected with adjuvant CpG, either alone or loaded onto murine dendritic cells (DCs). Immunization with PSA-MPyVLPs loaded onto DCs in the presence of CpG was shown to efficiently protect mice from tumor outgrowth. In addition, cellular and humoral immune responses after immunization were examined. PSA-specific CD4+ and CD8+ cells were demonstrated, but no PSA-specific IgG antibodies. Vaccination with DCs loaded with PSA-MPyVLPs induced an eight-fold lower titre of anti-VLP antibodies than vaccination with PSA-MPyVLPs alone. In conclusion, immunization of BALB/c mice with PSA-MPyVLPs, loaded onto DCs and co-injected with CpG, induces an efficient PSA-specific tumor protective immune response, including both CD4+ and CD8+ cells with a low induction of anti-VLP antibodies. PMID:21858228

  15. Seismic margin assessment for nuclear facilities of Kozloduy NPP

    International Nuclear Information System (INIS)

    Marinova, Bozhana

    2014-01-01

    In accordance with the decision of the European Commission and ENSREG Declaration of 13 May 2011, all nuclear power plants in the European Union were subjected to a stress test. The stress test is defined as a targeted reassessment of the safety margins of nuclear power plants in the light of the events which occurred at Fukushima: extreme natural events challenging the plant safety functions and leading to a severe accident. Seismic margins assessment is based on the analysis of the seismic resistance of the equipment, which is important for safety and participates in mitigation of accident scenarios. Seismic margin is determined on the basis of the prescribed limits of seismic accelerations that any nuclear facility can withstand without severe fuel damage and radioactive release into the environment. The determining of the weak points and boundary effects in case of seismic action is done based on the data from the seismic PSA Level 1. Based on the calculated median values of the probable seismic deviations, the ranges of probable seismic action have been determined for which the resistance of the different nuclear facilities is assessed. The safety margins re-assessment should define the nuclear facility ultimate capacity, i.e. to determine the values of accelerations for which the SSC failures would result in non-availability of the safety functions, and fuel damage would be inevitable. The assessment of this acceleration value is done using the data of the seismic analyses as performed on design stage, accounting for the dynamic response and actual spatial dimensions of the civil structures and for the materials properties. For the purpose of the KNPP nuclear facilities safety re-assessment, the seismic capacity is accepted to be determine by the value of seismic acceleration, for which it can be ascertained with 95% certainty that the safety factor obtained at the respective seismic acceleration is not lower than 1. The purpose of this report is to present

  16. Chernobyl NPP decommissioning efforts - Past, Present and Future. Decommissioning Efforts on Chernobyl NPP site - Past, Present

    International Nuclear Information System (INIS)

    Kuchinskiy, V.

    2017-01-01

    Two unique large-scale projects are underway at the moment within the Chernobyl - Exclusion zone - Shelter object transformation into ecologically safe system and the decommissioning of 3 Chernobyl NPP Units. As a result of beyond design accident in 1986 the entire territory of the industrial site and facilities located on it was heavily contaminated. Priority measures were carried out at the damaged Unit under very difficult conditions to reduce the accident consequences and works to ensure nuclear and radiation safety are continuous, and the Unit four in 1986 was transformed into the Shelter object. Currently, works at the Shelter object are in progress. Under assistance of the International Community new protective construction was built above the existing Shelter object - New Safe Confinement, which will ensure the SO Safety for the long term - within up to 100 years. The second major project is the simultaneous decommissioning of Chernobyl NPP Units 1, 2 and 3. Currently existing Chernobyl NPP decommissioning Strategy has been continuously improved starting from the Concept of 1992. Over the years the following was analyzed and taken into account: the results of numerous research and development works, international experience in decommissioning, IAEA recommendations, comments and suggestions from the governmental and regulatory bodies in the fields of nuclear energy use and radioactive waste management. In 2008 the final decommissioning strategy option for Chernobyl NPP was approved, that was deferred gradual dismantling (SAFSTOR). In accordance with this strategy, decommissioning will be carried out in 3 stages (Final Shutdown and Preservation, Safe Enclosure, Dismantling). The SAFSTOR strategy stipulates: -) the preservation of the reactor, the primary circuit and the reactor compartment equipment; -) the dismantling of the equipment external in relation to the reactor; -) the safe enclosure (under the supervision); -) the gradual dismantling of the primary

  17. Probabilistic assessment of NPP safety under aircraft impact

    International Nuclear Information System (INIS)

    Birbraer, A.N.; Roleder, A.J.; Arhipov, S.B.

    1999-01-01

    Methodology of probabilistic assessment of NPP safety under aircraft impact is described below. The assessment is made taking into account not only the fact of aircraft fall onto the NPP building, but another casual parameters too, namely an aircraft class, velocity and mass, as well as point and angle of its impact with the building structure. This analysis can permit to justify the decrease of the required structure strength and dynamic loads on the NPP equipment. It can also be especially useful when assessing the safety of existing NPP. (author)

  18. From the chronicle of training of Dukovany NPP staff

    International Nuclear Information System (INIS)

    2005-01-01

    The long way the Dukovany NPP had to go before the plant staff was fully qualified and skilled is described. First the training concept was prepared, then the necessary training facilities were set up, lecturers and instructors were hired and trained, training programmes and training materials were developed, and ultimately the first training course was launched in 1979. A training NPP was constructed and a full-scope simulator of the Dukovany NPP was set up. The current status of organization of NPP staff training by the CEZ utility is highlighted. (author)

  19. Predisposal of Radioactive Waste from NPP 1000 MWe

    International Nuclear Information System (INIS)

    Suryantoro

    2007-01-01

    Predisposal of radioactive waste from NPP 1000 MW which was planned to be operated in 2016 has been conducted. In this study NPP applying PWR type was assumed. This assessment comprises all aspects of radioactive waste coming from NPP. One through cycle was chosen consequently no reprocessing step will be conducted. The assessment shows that technologically all radioactive waste treatment process rising from NPP operation has similarities to the existing radioactive waste process conducted by RWI which has lower scale of waste amount. (author)

  20. Romania: Cernavoda 2 NPP. Licensing requirements. Annex 9

    International Nuclear Information System (INIS)

    Biro, L.

    1999-01-01

    The annex deals with the Romanian Nuclear Safety Authority (National Commission for Nuclear Activities Control - CNCAN) licensing requirements for Cernavoda 2 NPP. These licensing requirements are in accordance with the Regulation Policy Statement applicable for Cernavoda NPP and contain the general aspects deriving from laws, regulations and regulation practices included in norms and in specific documentation. The licensing requirements issued by CNCAN in May 1997 takes into consideration the fact that Cernavoda 2 is a delayed NPP. This annex provides only those key elements, which are relevant to illustrate the regulatory requirements for Cernavoda 2 as a delayed NPP. More details are presented in the original document issued by CNCAN. (author)

  1. Training center of Rovenskaya NPP. The experience of creation

    International Nuclear Information System (INIS)

    Fedorov, O.M.; Aristov, B.N.

    1991-01-01

    Experience in creation of a teaching-training centre at the Rovno NPP, which uses means available at unified NPPs, at most is discussed. The centre hardware complex functions include the event filing and providing for user-friendly interface with NPP technical personnel under training. The system of personnel training at the Rovno NPP teaching-training centre gives an opportunity to analyze accidents and emergency conditions more completely and carefully. The taching analysis of failures and accidents by a NPP operators using the complex of the teaching-training centre hardware sufficiently improves knowledge of particular accidents

  2. Comparative Study on Atmospheric Dispersion Module of Level 3 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Dahye; Jang, Misuk; Kang, Hyun Sik; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

    2016-10-15

    Some regulation documents such as Regulatory Guides and NUREG publications from the U.S. Nuclear Regulatory Commission (NRC) have influences on domestic radiation environmental analyses. As renewal versions of NUREG-0800 and NUREG-1555 have issued lately, the assessment for Severe Accident (SA) with Probabilistic Safety Assessment (PSA) should be added to Safety Analysis Report (SAR) and Radiation Environmental Report (RER). Because these reports are the required documents for obtaining the construction permit and operating license, it is important to understand the PSA methodology and it needs to improve the site-specific input data of L3PSA codes for SA. First, our review focuses on the atmospheric dispersion and deposition related input data of L3PSA code in this paper. Then we will continue to review the improvements of other input data. Two atmospheric dispersion models, which are PAVAN developed for design basis accident and ATMOS of MACCS2 code developed for SA, were reviewed in this paper. L3PSA deals with the effects of severe accidents and basically includes the evaluation of both short- and long-term effects. Therefore, both the deposition effects and nuclide information(type, amount, and chemical characteristics of released radionuclide) would be considered as the input parameters of atmospheric dispersion model for L3PSA. Additionally, the meteorological data would be sampled randomly to meet the purpose of probabilistic method. However, the sampling method would be selected according to analysis purpose. After review, ATMOS module and its input data are suitably developed for the atmospheric dispersion analysis of L3PSA. However, ATMOS module was developed using the site-specific terrain and environment characteristics. For the domestic application, it needs to study the input data reflecting the Korean terrain and environment characteristics. It would be also continuously improved in response to the time- and site-specific changes of weather

  3. Seismic PSA of nuclear power plants a case study

    International Nuclear Information System (INIS)

    Hari Prasad, M.; Dubey, P.N.; Reddy, G.R.; Saraf, R.K.; Ghosh, A.K.

    2006-07-01

    Seismic Probabilistic Safety Assessment (Seismic PSA) analysis is an external event PSA analysis. The objective of seismic PSA for the plants is to examine the existence of plant vulnerabilities against postulated earthquakes by numerically assessing the plant safety and to take appropriate measures to enhance the plant safety. Seismic PSA analysis integrates the seismic hazard analysis, seismic response analysis, seismic fragility analysis and system reliability/ accident sequence analysis. In general, the plant consists of normally operating and emergency standby systems and components. The failure during an earthquake (induced directly by excessive inertial stresses or indirectly following the failure of some other item) of an operating component will lead to a change in the state of the plant. In that case, various scenarios can follow depending on the initiating event and the status of other sub-systems. The analysis represents these possible chronological sequences by an event tree. The event trees and the associated fault trees model the sub-systems down to the level of individual components. The procedure has been applied for a typical Indian nuclear power plant. From the internal event PSA level I analysis significant contribution to the Core Damage Frequency (CDF) was found due to the Fire Water System. Hence, this system was selected to establish the procedure of seismic PSA. In this report the different elements that go into seismic PSA analysis have been discussed. Hazard curves have been developed for the site. Fragility curve for the seismically induced failure of Class IV power has been developed. The fragility curve for fire-water piping system has been generated. Event tree for Class IV power supply has been developed and the dominating accident sequences were identified. CDF has been estimated from these dominating accident sequences by convoluting hazard curves of initiating event and fragility curves of the safety systems. (author)

  4. International network on incorporation of ageing effects into PSA

    International Nuclear Information System (INIS)

    Kirchsteiger, C.; Patrik, M.

    2006-01-01

    This paper describes the background and status of a new International Network on ''Incorporating Ageing Effects into Probabilistic Safety Assessment''. The Joint Research Centre (JRC) of the European Commission organized in September 2004 the kickoff meeting of this Network at JRC's Institute for Energy in Petten, Netherlands, with the aims to open the APSA Network, to start discussion of ageing issues in relation to incorporating ageing effects into PSA tools and to come to consensus on objectives and work packages of the Network, taking into account the specific expectations of potential Network partners. The presentations and discussions at the meeting confirmed the main conclusion from the previously organized PSAM 7 pre-conference workshop on ''Incorporating PSA into Ageing Management'', Budapest, June 2004, namely that incorporating ageing effects into PSA seems to be more and more a hot topic particularly for risk assessment and ageing management of nuclear power plants operating at advanced age (more than 25-30 years) and for the purpose of plant life extension. However, it also appeared that, especially regarding the situation in Europe, at present there are several on-going feasibility or full studies in this area, but not yet a completed Ageing PSA leading to applications. The project's working method is a NETWORK of operators, industry, research, academia and consultants with an active interest in the area (physical networking via a series of workshops and virtual networking via the Internet). The resulting knowledge should help PSA developers and users to incorporate the effects of equipment ageing into current PSA tools and models, to identify and/or develop most effective corresponding methods, to focus on dominant ageing contributors and components and to promote the use of PSA for ageing management of Nuclear Power Plants. (orig.)

  5. The CEC NRWG task force on regulatory actions related to PSA

    International Nuclear Information System (INIS)

    Campbell, J.F.

    1994-01-01

    The aims and objectives of the Task Force were: Assessment of how regulatory bodies currently estimate the value of PSA results, including the understanding of benefits and limitations with regard to methods and approaches; to show the current use of the PSA tool at the regulatory level, and to establish differences in the approach in various countries, for example whether PSA is part of the licensing process or is used outside the licensing process as a supplementary tool by the utility; specific requirements in PSA methodology and guides by regulatory bodies on PSA procedures. Also guidelines on how to review a PSA once it has been submitted by the licensee; contributions made by the regulatory body in PSA with regard to review of the PSA and the evaluation of the numerical values in the PSA

  6. ASAMPSA-E guidance for level 2 PSA Volume 2. Implementing external Events modelling in Level 2 PSA

    International Nuclear Information System (INIS)

    Cazzoli, E.; Vitazkova, J.; Loeffler, H.; Burgazzi, L.

    2016-01-01

    The objective of the present document is to provide guidance on the implementation of external events into an 'extended' L2 PSA. It has to be noted that L2 PSA addresses issues beginning with fuel degradation and ending with the release of radionuclides into the environment. Therefore, the present document may touch upon, but does not evaluate explicitly issues that involve events or phenomena which occur before the fuel begins to degrade. Following the accident at Fukushima Dai-ichi, the nuclear safety community has realized that much attention should be given to the areas of operator interventions and accidents that may develop at the same time in more than one unit if they are initiated by one or more common external events. For this reason and to fulfill the PSA end-users' wish list (as reflected by an ASAMPSA-E survey), the attention is mostly focused on interface between L1 and L2 PSA, fragility analysis, human response analysis and some consideration is given to L2 PSA modeling of severe accidents for multiple unit sites, even though it is premature to provide extensive guidance in this area. The following recommendations, mentioned in various sections within this document, are summarized here: 1. Vulnerability/fragility analyses should be performed with respect to all external hazards and all structures, systems and components potentially affected that could be relevant to L2 PSA, 2. Importance should be given to the assessment of human performance following extreme external events; for extreme circumstances with high stress level, low confidence is justified for SAM human interventions and for such conditions, human interventions could be analyzed as sensitivity cases only in L2 PSA, 3. Results presentation should include assessment of total risk measures compared with risk targets able to assess all contributions to the risk and to judge properly the safety, 4. Total risk measures shall be associated to appropriate information on all

  7. Comparison of SKIFS 2004:1 and Tillsynshandbok PSA against the ASME PRA Standard and European requirements on PSA; Jaemfoerelse av SKIFS 2004:1 och Tillsynshandbok PSA mot ASME PRA Standard och Europeiska krav paa PSA

    Energy Technology Data Exchange (ETDEWEB)

    Hellstroem, Per

    2005-04-15

    Requirements on PSA for risk informed applications are expressed in different international documents. The ASME PRA standard published in spring 2002 is one such document, PSA requirements are also expressed in the European Utility Requirements (EUR) for new reactors. The Swedish PSA requirements are provided in the Swedish regulators (SKI) statutes SKIFS 2004:1. SKI also has a review handbook for PSA activities (SKI report 2003:48). The review handbook is a support during review of the utilities PSA activities and the PSAs themselves. The review handbook expresses SKIs expectations by providing so called important aspects for both the PSA work and the PSAs, A comparison of SKIFS requirements and the important aspects in the Review handbook, on one side, and the requirements on PSA in EUR and ASME on the other side, is presented. The comparison shows a large difference in the level of detail in the different documents, where ASME is most detailed and specific. This is expected since the SKI review handbook not is a 'PSA guide' in the same way as the ASME PRA standard. A direct comparison of the ASME PRA standard requirements with the important aspects in the review handbook cannot answer the question which ASME capacity level that is achieved by a PSA meeting all important aspects. The conclusion is that it is not likely to achieve capacity level 2 and 3, since very few ASME level 3 attributes are explicitly expressed as important aspects, though many are expressed in general terms. The review handbook important aspects that are most similar to the ASME capacity level 1 attributes are initiating events, sequence analysis, and system analysis while less similarity is found for analysis of operator actions data analysis, quantification and containment analysis (level 2). Less similarity is found for capacity level 2 and 3. However, the number of additional ASME attributes on capacity level 2 and 3 are few. There are also important aspects in the review

  8. Review of updated design of SPWR with PSA methodology

    International Nuclear Information System (INIS)

    Oikawa, Tetsukuni; Muramatsu, Ken

    1995-01-01

    This paper presents the procedures and results of a PSA (Probabilistic Safety Assessment) of the SPWR (System-Integrated PWR), which is being developed at the Japan Atomic Energy Research Institute (JAERI) as a medium sized innovative passive safe reactor, to assist in the design improvement of the SPWR at the basic or conceptual design phase by reviewing the design and identifying the design vulnerability. The first phase PSA, which was carried out in 1991, was a scoping analysis in order to understand overall plant characteristics and to search for general design weakness. After discussing the results of the first phase PSA, the SPWR designer group changed some designs of the SPWR. The second phase PSA of the SPWR was performed for the modified design in order to identify the design vulnerability as well as to grasp its overall safety level. Special items of these PSAs are as follows: (1) systematic identification of initiating events related to newly designed systems by the failure mode effect analysis (FMEA), (2) delineation of accident sequences for the internal initiating events using accident progression flow charts which is cost effective for conceptual design phase, (3) quantification of event trees based on many engineering judgement, and (4) lots of sensitivity analyses to examine applicability of data assignment. Qualitative and quantitative results of PSA provided very useful information for decision makings of design improvement and recommendations for further consideration in the process of detailed design. (author)

  9. Risk Metrics and Measures for an Extended PSA

    International Nuclear Information System (INIS)

    Wielenberg, A.; Loeffler, H.; Hasnaoui, C.; Burgazzi, L.; Cazzoli, E.; Jan, P.; La Rovere, S.; Siklossy, T.; Vitazkova, J.; Raimond, E.

    2016-01-01

    This report provides a review of the main used risk measures for Level 1 and Level 2 PSA. It depicts their advantages, limitations and disadvantages and develops some more precise risk measures relevant for extended PSAs and helpful for decision-making. This report does not recommend or suggest any quantitative value for the risk measures. It does not discuss in details decision-making based on PSA results neither. The choice of one appropriate risk measure or a set of risk measures depends on the decision making approach as well as on the issue to be decided. The general approach for decision making aims at a multi-attribute approach. This can include the use of several risk measures as appropriate. Section 5 provides some recommendations on the main risk metrics to be used for an extended PSA. For Level 1 PSA, Fuel Damage Frequency and Radionuclide Mobilization Frequency are recommended. For Level 2 PSA, the characterization of loss of containment function and a total risk measure based on the aggregated activity releases of all sequences rated by their frequencies is proposed. (authors)

  10. Design of NPP of new generation being constructed at the Novovoronezh NPP site

    International Nuclear Information System (INIS)

    Afrov, A.; Berkovich, V.; Generalov, V.; Dragunov, Yu.; Krushelnitsky, V.

    1999-01-01

    The design of a new generation NPP is described, underscoring advances in physical attributes and passive safety systems based on experiences with earlier designs at operating NPPs. This paper elaborates on systems for handling and storing radioactive wastes, on refinements in containment measures and on experimental and analytic validation of critical design factors. (author)

  11. Support to NPP operation and maintenance technology risk management. A concept for establishing criteria and procedure for the selection of components with respect to their importance. Stage 3.1. NPP equipment reliability management

    International Nuclear Information System (INIS)

    Stvan, F.

    2003-12-01

    A proposal was developed for a procedure using the deterministic approach to the assessment of components from the operational point of view and other aspects that cannot be directly and readily quantified and of the probabilistic approach for the assessment of component importance with respect to nuclear safety. A specific PSA study performed for the Dukovany NPP was employed. The structure of the report is as follows: (1) Aspects of component selection; (2) Introductory procedure; (3) Criteria for the selection of components with respect to their importance (4) Assessing the priority of use of the assets - effect on production, safety, and profit; (5) Assessment of the risk aspect of the assets - effect on major processes; (6) Assessment of the level of use of the assets; (7) Assessment of the structure of the assets - optimal structure for maintenance in relation to the major processes; (8) Assessment of the criteria for estimating the importance of the components; (9) Probabilistic assessment of importance from the safety aspect by means of PSA; and (10) Deterministic assessment of importance from the safety aspect. (P.A.)

  12. Association of PSA, free-PSA and testosteron levels in serum of patients with benign prostate hyperplasia (BPH) and prostate cancer

    International Nuclear Information System (INIS)

    Wiwin Mailana; Kristina Dwi P; Sri Insani WW; Puji Widayati

    2015-01-01

    Prostate cancer screening can be done by measuring the concentration levels of PSA, free-PSA and testosterone in serum that examined with radioimmunoassay (RIA). A total of 30 patients of 45-81 years old had enrolled in this study and were taken their venous blood. The aim of research is to know the relationship between PSA and testosterone free-PSA with BPH and prostate cancer. Results showed that there was no correlation between age with BPH and prostate cancer (p = 0.06), but there is a relationship between PSA with BPH and prostate cancer (p = 0.002), the relationship between free-PSA with BPH and prostate cancer (p = 0.001). No correlation was found between PSA ratio with BPH and prostate cancer as well as the absence of a relationship between testosterone with BPH and prostate cancer (p = 0.924). (author)

  13. Jpss System Architecture Npp to the Future

    Science.gov (United States)

    Furgerson, J.; Trumbower, G.

    2012-12-01

    The National Oceanic and Atmospheric Administration (NOAA) is acquiring the next-generation weather and environmental satellite system, named the Joint Polar Satellite System (JPSS). The National Aeronautics and Space Administration (NASA) serves as the acquisition and development agent. JPSS replaces the current Polar-orbiting Operational Environmental Satellites (POES) managed by NOAA in the 1330 local time of ascending node (LTAN) orbit. The Suomi National Polar-orbiting Partnership (NPP) was launched into the 1330 LTAN orbit on October 28, 2011, and carries advanced sensors which will be featured on JPSS. It serves as a bridge mission and provides continuity for the NASA Earth Observation System and the POES. JPSS-1 is scheduled to launch in 2017. The Defense Meteorological Satellite Program (DMSP) managed by the DoD is operating in the 1730 LTAN orbit. The DoD is developing the Defense Weather Satellite Follow-on (WSF) system which will continue in the 1730 orbit. NASA is developing the Common Ground System (CGS) with the capability to process data from both the JPSS and WSF constellations. The CGS will be operated by NOAA. This poster will provide a top level status update of the program, as well as an overview of the JPSS system architecture. The space segment carries a suite of sensors that collect meteorological, oceanographic, and climatological observations of the earth and atmosphere. The system design allows centralized mission management and delivers high quality environmental products to military, civil and scientific users through a Command, Control, and Communication Segment (C3S). The data processing for NPP/JPSS is accomplished through an Interface Data Processing Segment (IDPS)/Field Terminal Segment (FTS) that processes NPP/JPSS satellite data to provide environmental data products to NOAA and DoD processing centers as well as remote terminal users.

  14. Corporate portal system at PAKS NPP, Hungary

    International Nuclear Information System (INIS)

    2016-01-01

    The new Corporate Portal System (CPS) of Paks NPP was launched in November 2006. The portal is based on one of the latest technologies, Plumtree Enterprise WEB 5.0. The main purpose of the installation of the new technology was to serve the working culture change, to give a platform to access all information and applications including the integrated process model used at the NPP. The new technology also supports those goals which were defined in the organization development programme: e.g. to improve internal communication with the establishment of communities of practice. Installation of the CPS has provided a powerful tool for knowledge management; it is possible to share and find all information through a controlled access in documents from various sources, to have links to people, portlets and different communities. Document management of the Paks NPP is supported by the integration of the Document 5 application, as the new Electronic Data Management System (EDMS) and the CPS. Depending on their access rights, all users of the CPS, through Microsoft Internet Explorer, can access technical, economic and human resources documents which are stored anywhere on the internal network (file servers, EDMS, old INRANET). The CPS is also accessible from the internet through a secure connection. The main concept is the integration of all applications to one platform and to help users to find all information they need. An access control list specifies which users and groups have access to an object (and what kind of access privileges they have such as read, select, edit, admin)

  15. Social consequences of closing the Ignalina NPP

    International Nuclear Information System (INIS)

    Baubinas, R.; Burneika, D.

    2001-01-01

    The possible social consequences of closing the Ignalina Nuclear Power Plant are studied. The social and economical situation in Visaginas and in the Utena region as a precondition for possible social consequences is shown. Also, two main groups of factors that can possibly influence the situation in the labour market are analysed. The problems of the enterprises that create working places and of the inhabitants of Visaginas whose possible behaviour can affect the situation in the labour market are discussed. Also, some proposals to neutralize the social costs of closing the Ignalina NPP are made. (author)

  16. Seismic fragility analysis for NPP structural components

    International Nuclear Information System (INIS)

    Casciati, F.; Faravelli, L.

    1984-01-01

    Relationship between probability of failure and seismic intensity (fragility curves) summarize the results of the vulnerability analysis of a NPP structural component. The appropriate operative procedures for the evaluation of these fragility curves are discussed. The paper illustrates a general purpose computer code which includes two different probabilistic models for two different levels of sophistication in the fragility analysis. Attention is then focussed (i) on the mechanical aspects associated with the local strength estimation (ii) on the simulation of suitable artificial ground motions and (iii) on the definition of an appropriate measure of the seismic intensity. (Author) [pt

  17. 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)

  18. Students education and training for Slovak NPP

    International Nuclear Information System (INIS)

    Lipka, J.; Slugen, V.; Hascik, J.; Miglierini, M.

    2004-01-01

    Slovak University of Technology is the largest and also the oldest university of technology in Slovakia. Surely more than 50% of high-educated technicians who work nowadays in nuclear industry have graduated from this university. The Department of Nuclear Physics and Technology of the Faculty of Electrical Engineering and Information Technology as a one of seven faculties of this University feels responsibility for proper engineering education and training for Slovak NPP operating staff. The education process is realised via undergraduate (Bc.), graduate (MSc.) and postgraduate (PhD.) study as well as via specialised training courses in a frame of continuous education system. (author)

  19. Safe 15 Terawatt of Temelin NPP

    International Nuclear Information System (INIS)

    Sula, M.

    2010-01-01

    In this work author presents a project Safe 15 Terawatt realised on the Temelin NPP. This project is one of the eight key projects of the CEZ group, associated in the 'Programme of efficiency'. The project started in June 2007 with long-term goals for horizon of year 2012. The safety indicators will be reached of the first quarter level of world's nuclear power plant - by the end of the first decade. By the end of year 2012 we will have achieved annual production of 15 billion kWh - in the Czech Republic: 15 Terawatt.

  20. Baltic NPP Project specifics and current status

    International Nuclear Information System (INIS)

    2011-01-01

    Project overview: 2 x 1194 MW Units (AES-2006 series); Location in Kaliningrad region of the; Russian Federation; Operation dates: Unit 1 – Oct 2016; Unit 2 – Apr 2018; Site preparatory works ongoing. This is first NPP project in the Russian Federation providing opportunity for participation of foreign investors. Foreign investors may acquire up to 49% share. Cross-border transmission lines developed under separate project with participation of foreign investors. Conclusion: At the selected set of assumptions, the project is financially feasible in all scenarios

  1. Simulation and optimization of an industrial PSA unit

    Directory of Open Access Journals (Sweden)

    Barg C.

    2000-01-01

    Full Text Available The Pressure Swing Adsorption (PSA units have been used as a low cost alternative to the usual gas separation processes. Its largest commercial application is for hydrogen purification systems. Several studies have been made about the simulation of pressure swing adsorption units, but there are only few reports on the optimization of such processes. The objective of this study is to simulate and optimize an industrial PSA unit for hydrogen purification. This unit consists of six beds, each of them have three layers of different kinds of adsorbents. The main impurities are methane, carbon monoxide and sulfidric gas. The product stream has 99.99% purity in hydrogen, and the recovery is around 90%. A mathematical model for a commercial PSA unit is developed. The cycle time and the pressure swing steps are optimized. All the features concerning with complex commercial processes are considered.

  2. Uncertainty and sensitivity methods in support of PSA level 2

    International Nuclear Information System (INIS)

    Devictor, N.; Bolado Lavin, R.

    2007-01-01

    Dealing with uncertainties in PSA level 2 requires using a set of statistical techniques to assess input uncertainty, to propagate uncertainties in an efficient way, to characterize appropriately output uncertainty and to get information from computer code runs through an intelligent use of sensitivity analysis techniques. The purpose of this paper is to give an overview of statistical and probabilistic methods and tools to answer to these topics, and to provide some guidance about their suitability and limitations to be used in a PSA level 2. Our position about their implementation in L2 PSA software has been written; it could be noticed that a lot of these methods are very time-consuming, and seem more suitable for the analysis of submodels or for focusing on specific questions. (authors)

  3. PROSTATE CANCER SCREENING: PSA TEST AWARENESS AMONG ADULT MALES.

    Science.gov (United States)

    Obana, Michael; O'Lawrence, Henry

    2015-01-01

    The overall purpose of this study was to determine whether visits to the doctor in the last 12 months, education level, and annual household income for adult males increased the awareness of prostate-specific antigen (PSA) tests. The effect of these factors for the knowledge of PSA exams was performed using statistical analysis. A retrospective secondary database was utilized for this study using the questionnaire in the California Health Interview Survey from 2009. Based on this survey, annual visits to the doctor, higher educational levels attained, and greater take-home pay were statistically significant and the results of the study were equivalent to those hypothesized. This also reflects the consideration of marketing PSA blood test screenings to those adult males who are poor, uneducated, and do not see the doctor on a consistent basis.

  4. Exponential rise in prostate-specific antigen (PSA) during anti-androgen withdrawal predicts PSA flare after docetaxel chemotherapy in patients with castration-resistant prostate cancer.

    Science.gov (United States)

    Han, Kyung Seok; Hong, Sung Joon

    2015-03-01

    To investigate the relationship between rising patterns of prostate-specific antigen (PSA) before chemotherapy and PSA flare during the early phase of chemotherapy in patients with castration-resistant prostate cancer (CRPC). This study included 55 patients with CRPC who received chemotherapy and in whom pre-treatment or post-treatment PSA levels could be serially obtained. The baseline parameters included age, performance, Gleason score, PSA level, and disease extent. PSA doubling time was calculated using the different intervals: the conventional interval from the second hormone manipulation following the nadir until anti-androgen withdrawal (PSADT1), the interval from the initial rise after anti-androgen withdrawal to the start of chemotherapy (PSADT2), and the interval from the nadir until the start of chemotherapy (PSADT3). The PSA growth patterns were analyzed using the ratio of PSADT2 to PSADT1. There were two growth patterns of PSA doubling time: 22 patients (40.0%) had a steady pattern with a more prolonged PSADT2 than PSADT1, while 33 (60.0%) had an accelerating pattern with a shorter PSADT2 than PSADT1. During three cycles of chemotherapy, PSA flare occurred in 11 patients (20.0%); of these patients, 3 were among 33 (9.1%) patients with an accelerating PSA growth pattern and 8 were among 22 patients (36.4%) with a steady PSA growth pattern (p=0.019). Multivariate analysis showed that only PSA growth pattern was an independent predictor of PSA flare (p=0.034). An exponential rise in PSA during anti-androgen withdrawal is a significant predictor for PSA flare during chemotherapy in CRPC patients.

  5. Sociological investigations on Ignalina NPP and within its surroundings

    International Nuclear Information System (INIS)

    Chiuzhas, A.

    1998-01-01

    The purpose of this study was to determine the impact of Ignalina NPP and Visaginas town on the social territorial processes in the region and to reveal the impact of Ignalina NPP on the regional economic, social, demographic, political and cultural processes in the context of ecological and psychological affect. According to the results of this research three quarters of the inhabitants and the functionaries of local administration hold an opinion that operation of Ignalina NPP posses threat for the population and environment. Meanwhile they are sure that danger of Ignalina NPP is not critical. 21 - 35 % of the local administrators speak for the closure of Ignalina NPP , whereas half of Visaginas residents and three quarters of the local administrators indicate that operation of reactors is expedient. Over 90% of the population do not have sufficient information on the operation of Ignalina NPP. In the opinion of the rest Lithuanian people Ignalina NPP zone is related with the physical danger and the image of Visaginas residents as the 'others', 'strangers'. More than 90% of Ignalina NPP employees are Russian speaking, not native Lithuanians. The social relations of Visaginas with the environment are poor as a result of the situation of the town, lack of communications and cultural self isolation. (author)

  6. Using the digital reactor control systems at NPP

    International Nuclear Information System (INIS)

    Schirl, G.; Hertel, J.

    2006-01-01

    A conception of application of the digital reactor control systems (RCS) at NPP is presented. The digital RCS architecture and safety ensuring are considered. The strategy and algorithm of the operating NPP equipping with the new digital RCS are given too [ru

  7. Unit Commissioning of “Belene” NPP (Bulgaria)

    International Nuclear Information System (INIS)

    2009-01-01

    This presentations gives detailed information about the following topics about commissioning: principles of NPP commissioning; phases of NPP commissioning; organization of commissioning activities; duties and responsibilities of the parties for carrying out unit commissioning activities; responsibility and obligations of the sides during commissioning of power unit; documentation required for power unit commissioning; quality assurance for commissioning activities

  8. Balancing preventive and corrective maintenance in Cernavoda Unit 1 NPP

    International Nuclear Information System (INIS)

    Riedel, M.; Marinescu, S.

    2000-01-01

    The paper presents a short reminder of Romania's Cernavoda NPP entering commercial operation and a brief description of the CANDU-6 project on which Unit 1 is based. The short term objectives of the maintenance management, the status of the existing maintenance programmes as well as future predictable maintenance programmes are outlined together with the Government plan to complete the balance of NPP. (author)

  9. Structuring a risk-informed and performance-based process for optimization of regulation for Laguna Verde NPP

    International Nuclear Information System (INIS)

    Rodriguez-Hernandez, A.

    2001-01-01

    This work describes the plan for a process to incorporate into the regulatory activities the risk information derived from probabilistic risk assessments, as well as information generated by the periodic evaluation of the Maintenance Rule (MR, 10CFR50.65). The current status of the Laguna Verde NPP (LVNPP) risk analysis, PSA Level 1, allows determining in a reliable way the accident scenarios and the involved systems having significant impact on safety. The determination of system's risk significance allows carrying out a prioritization of safety issues to be evaluated and inspected; for example, operational events, changes to technical specifications, design modifications, inspection priorities, etc. In addition, complementary and basic information are the results generated by the performance monitoring of structures, systems and components (SSCs) under the scope of the MR. The SSCs performance trends are indicatives to focus evaluation and inspection activities on important issues. Then, with the reportability in short periods the performance evaluations of SSCs and the incorporation of a process of risk management, the evaluation and inspection activities will be directed to those risk significant systems showing degraded performance. Therefore, based on systems performance results and risk information, it is feasible to have certain flexibility or a better balance between the regulatory requirements. Inside this process, a consensus is needed with the utility to establish quality attributes for the plant-specific PSA, as well as the rules to be followed in the use of this tool and the kind of information to be reported for MR results. (author)

  10. PSA modeling of long-term accident sequences

    International Nuclear Information System (INIS)

    Georgescu, Gabriel; Corenwinder, Francois; Lanore, Jeanne-Marie

    2014-01-01

    In the context of the extension of PSA scope to include external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the accident sequences induced by initiators which affect a whole site containing several nuclear units (reactors, fuel pools,...). These methodological improvements represent an essential prerequisite for the development of external hazards PSA. However, it has to be noted that in French PSA, even before Fukushima, long term accident sequences were taken into account: many insight were therefore used, as complementary information, to enhance the safety level of the plants. IRSN proposed an external events PSA development program. One of the first steps of the program is the development of methods to model in the PSA the long term accident sequences, based on the experience gained. At short term IRSN intends to enhance the modeling of the 'long term' accident sequences induced by the loss of the heat sink or/and the loss of external power supply. The experience gained by IRSN and EDF from the development of several probabilistic studies treating long term accident sequences shows that the simple extension of the mission time of the mitigation systems from 24 hours to longer times is not sufficient to realistically quantify the risk and to obtain a correct ranking of the risk contributions and that treatment of recoveries is also necessary. IRSN intends to develop a generic study which can be used as a general methodology for the assessment of the long term accident sequences, mainly generated by external hazards and their combinations. This first attempt to develop this generic study allowed identifying some aspects, which may be hazard (or combinations of hazards) or related to initial boundary conditions, which should be taken into account for further developments. (authors)

  11. The relationship between serum PSA, six sex hormones and the benign or malignant prostate diseases

    International Nuclear Information System (INIS)

    Xu Yancun

    2008-01-01

    In order to study clinical significance of serum prostate-specific antigen (PSA), free prostate specific antigen (PSA), f/tPSA and six sex hormones in prostate diseases, the serum levels of PSA, fPSA, f/tPSA, T, P, E 2 , PRL, LH and FSH in 72 cases of hyperplasia of prostate patients and 40 patients with prostate cancer were determined by RIA. The results showed that the serum levels of T, E 2 , PRL, LH, FSH in the BPH Group were significantly lower than those of in Pca group, the serum level of P in Pca group were significantly lower than those in BPH group; the levels of fPSA and f/tPSA ratio in BPH Group were significantly higher than those in Pca group. The results suggest that benign and malignant prostate disease (BPH and Pca) was related with the hormone imbalance. The serum total PSA and fPSA can be regarded as important indicators in the diagnosis of BPH and Pea. The combined determination of PSA, fPSA and f/tPSA may improve the diagnostic accuracy of Pca. (authors)

  12. Project No. 10 - Partial restoration of Ignalina NPP territory

    International Nuclear Information System (INIS)

    2000-01-01

    At present Ignalina NPP territory makes a total of 2544 ha of land. Due to termination of construction activity development and due to the decision taken to shutdown unit 1 the need in such a territory fell off. For normal and safe operation of Ignalina NPP 1440 ha is enough, including 1237 ha for of Ignalina NPP administrative area and 203 ha for auxiliary objects. Ignalina NPP will have to rearrange territory, forestry that was damaged during the construction activities of the plant and to restore the damaged farmlands and to pass the rearranged forestry that belonged to the Ignalina NPP to the Ministry of Forestry. The total estimated cost of the project is about 1.042 M EURO

  13. Repeat prostate-specific antigen (PSA) test before prostate biopsy: a 20% decrease in PSA values is associated with a reduced risk of cancer and particularly of high-grade cancer.

    Science.gov (United States)

    De Nunzio, Cosimo; Lombardo, Riccardo; Nacchia, Antonio; Tema, Giorgia; Tubaro, Andrea

    2018-03-13

    To analyse the impact of repeating a prostate-specific antigen (PSA) level assessment on prostate biopsy decision in a cohort of men undergoing prostate biopsy. From 2015 onwards, we consecutively enrolled, at a single institution in Italy, men undergoing 12-core transrectal ultrasonography-guided prostate needle biopsy. Indication for prostate biopsy was a PSA level of ≥4 ng/mL. Demographic, clinical, and histopathological data were collected. The PSA level was tested at enrolment (PSA 1 ) and 4 weeks later on the day before biopsy (PSA 2 ). Variations in PSA level were defined as: stable PSA 2 within a 10% variation, stable PSA 2 within a 20% variation, PSA 2 decreased by ≥10%, PSA 2 decreased by ≥20%, PSA 2 increased by ≥10%, PSA 2 increased by ≥20%, and PSA 2 PSA within 20% variation had a higher risk of prostate cancer (odds ratio [OR] 1.80, P PSA2 decreased by ≥20% had a lower risk of prostate cancer (OR 0.37, P PSA2 increased by ≥10% had an increased risk of high-grade prostate cancer (OR 1.93, P PSA returned to normal values (PSA levels significantly reduced the risk of high-grade prostate cancer. Further multicentre studies should validate our present results. © 2018 The Authors BJU International © 2018 BJU International Published by John Wiley & Sons Ltd.

  14. Monitors for the surveillance of NPP components

    International Nuclear Information System (INIS)

    Giera, H.D.; Grabner, A.; Hessel, G.; Koeppen, H.E.; Liewers, P.; Schumann, P.; Weiss, F.P.; Kunze, U.; Pfeiffer, G.

    1985-01-01

    Noise diagnostics have reached a level where it is possible and efficient to integrate this method as far as possible into the control and safety system of the NPP. The communication between the noise diagnostic system and the plant operator is the main problem of integration. It is necessary to refine the diagnostic results in such a manner that the operator can use them without being skilled in noise analysis respectively without contacting a noise specialist. Moreover, in this way the noise specialist can be released from routine surveillance. For selected processes which have already intensively been investigated because of their inherent risk this can be achieved by means of autonomously working monitors. Independently the monitors perform signal processing and diagnosis. In general this means that they classify the technical condition of the monitored component into one of the two categories: ''normal'' or ''anomalous''. The result will be annunciated to the plant operator who will in the first step of the development contact the noise specialist only if anomalies have occurred in order to clarify the cause. At the NPP ''Bruno Leuschner'' Greifswald, three hardware monitors for loose parts detection, control rod surveillance and main coolant pump diagnosis are being tested. Additionally a so-called software monitor for diagnosing the pressure vessel vibrations is in preparation. The techniques and the hardware used for the monitors as well as planned further improvements of the integration of noise diagnostics into the control and safety system are discussed in this paper. (author)

  15. Cernavoda NPP - Management of internal tritium exposures

    International Nuclear Information System (INIS)

    Chitu, Catalina; Popescu, Ion; Samson, Liliana; Simionov, Vasile

    2010-01-01

    Full text: During normal operation of a CANDU nuclear power plant significant tritium quantities are generated. Through design solutions that have been implemented we manage to control the tritium losses from the reactor systems and keep them as low as possible. Special dryers are designed and are used to remove moisture from different ventilation systems of a CANDU reactor in order to maintain tritium in air concentration and gaseous tritium emissions below the limits established by the national authorities. Vapor Recovery System is designed to control tritium in air concentration and to recover heavy water loss from PHT and Moderator Systems and to control the air circulation, providing atmosphere separation between different areas of the Reactor Building. Cernavoda NPP developed a special strategy in order to control workers' internal exposures to tritium and dedicated programs are running to implement this strategy: improvement of radiation protection procedures; increasing equipment performances; leakages prevention through maintenance program; finalization of the de-tritiation facility. This paper presents the evolution of workers tritium exposure and emphasizes the results of the ALARA policy promoted by Cernavoda NPP management. (authors)

  16. Fiddle at financing of Mochovce NPP

    International Nuclear Information System (INIS)

    Beer, G.

    2003-01-01

    Slovak police found out and documented racket in financing of third and fourth block of Mochovce NPP. Damage should be 144 millions Slovak crowns. Investigator consequently accused twenty-three Slovaks from foundation, trump up and supporting of criminal group and deception. Two persons are suggested to be taken to custody. Skoda Praha, which was general supplier of construction for 1,5 billions Slovak crowns, in 2001 allowed to create a connection to let the finances between Slovenske elektrarne, a.s., Bratislava as investor and Skoda as subcontractor to be paid through three mediator companies. Companies got authorization to balance the relationship among all interested subjects. Confused network of 118 treaties was created. They invoiced reward from money current based on these treaties. Reward represented up to 70 per cent of transferred resources in some cases. According to Minister of Domestic Affairs Vladimir Palko it will be necessary to find out where is actually the money. For the time being nobody from NPP is among accused. (Author)

  17. Radiographic progression with nonrising PSA in metastatic castration-resistant prostate cancer

    DEFF Research Database (Denmark)

    Bryce, A H; Alumkal, J J; Armstrong, A

    2017-01-01

    BACKGROUND: Advanced prostate cancer is a phenotypically diverse disease that evolves through multiple clinical courses. PSA level is the most widely used parameter for disease monitoring, but it has well-recognized limitations. Unlike in clinical trials, in practice, clinicians may rely on PSA...... in the PREVAIL study were analyzed post hoc for rising versus nonrising PSA (empirically defined as >1.05 vs ⩽1.05 times the PSA level from 3 months earlier) at the time of radiographic progression. Clinical characteristics and disease outcomes were compared between the rising and nonrising PSA groups. RESULTS......: Of 265 PREVAIL patients with radiographic progression and evaluable PSA levels on the enzalutamide arm, nearly one-quarter had a nonrising PSA. Median progression-free survival in this cohort was 8.3 months versus 11.1 months in the rising PSA cohort (hazard ratio 1.68; 95% confidence interval 1...

  18. Report on Fukushima Daiichi NPP precursor events

    International Nuclear Information System (INIS)

    2014-01-01

    The main questions to be answered by this report were: The Fukushima Daiichi NPP accident, could it have been prevented? If there is a next severe accident, may it be prevented? To answer the first question, the report addressed several aspects. First, the report investigated whether precursors to the Fukushima Daiichi NPP accident existed in the operating experience; second, the reasons why these precursors did not evolve into a severe accident. Third, whether lessons learned from these precursor events were adequately considered by member countries; and finally, if the operating experience feedback system needs to be improved, based on the previous analysis. To address the second question which is much more challenging, the report considered precursor events identified through a search and analysis of the IRS database and also precursors events based on risk significance. Both methods can point out areas where further work may be needed, even if it depends heavily on design and site-specific factors. From the operating experience side, more efforts are needed to ensure timely and full implementation of lessons learnt from precursor events. Concerning risk considerations, a combined use of risk precursors and operating experience may drive to effective changes to plants to reduce risk. The report also contains a short description and evaluation of selected precursors that are related to the course of the Fukushima Daiichi NPP accident. The report addresses the question whether operating experience feedback can be effectively used to identify plant vulnerabilities and minimize potential for severe core damage accidents. Based on several of the precursor events national or international in-depth evaluations were started. The vulnerability of NPPs due to external and internal flooding has clearly been addressed. In addition to the IRS based investigation, the WGRISK was asked to identify important precursor events based on risk significance. These precursors have

  19. Automated personal dosimetry monitoring system for NPP

    International Nuclear Information System (INIS)

    Chanyshev, E.; Chechyotkin, N.; Kondratev, A.; Plyshevskaya, D.

    2006-01-01

    Full text: Radiation safety of personnel at nuclear power plants (NPP) is a priority aim. Degree of radiation exposure of personnel is defined by many factors: NPP design, operation of equipment, organizational management of radiation hazardous works and, certainly, safety culture of every employee. Automated Personal Dosimetry Monitoring System (A.P.D.M.S.) is applied at all nuclear power plants nowadays in Russia to eliminate the possibility of occupational radiation exposure beyond regulated level under different modes of NPP operation. A.P.D.M.S. provides individual radiation dose registration. In the paper the efforts of Design Bureau 'Promengineering' in construction of software and hardware complex of A.P.D.M.S. (S.H.W. A.P.D.M.S.) for NPP with PWR are presented. The developed complex is intended to automatize activities of radiation safety department when caring out individual dosimetry control. The complex covers all main processes concerning individual monitoring of external and internal radiation exposure as well as dose recording, management, and planning. S.H.W. A.P.D.M.S. is a multi-purpose system which software was designed on the modular approach. This approach presumes modification and extension of software using new components (modules) without changes in other components. Such structure makes the system flexible and allows modifying it in case of implementation a new radiation safety requirements and extending the scope of dosimetry monitoring. That gives the possibility to include with time new kinds of dosimetry control for Russian NPP in compliance with IAEA recommendations, for instance, control of the equivalent dose rate to the skin and the equivalent dose rate to the lens of the eye S.H.W. A.P.D.M.S. provides dosimetry control as follows: Current monitoring of external radiation exposure: - Gamma radiation dose measurement using radio-photoluminescent personal dosimeters. - Neutron radiation dose measurement using thermoluminescent

  20. Clinical diagnostic value of combined determination of serum tPSA, cPSA and IGF-I levels in patients with prostatic disorders

    International Nuclear Information System (INIS)

    Zhang Bashan; Zhang Zigang; Lai Fudi

    2008-01-01

    Objective: To investigate the diagnostic value of combined determination of serum total prostatic specific antigen (tPSA), complex prostatic specific antigen (cPSA) and IGF-I levels in patients with prostatic disorders. Methods: Serum tPSA, cPSA (with CLIA) and IGF-I (with IRMA) levels were determined in 41 patients with prostatic carcinoma, 60 patients with benign prosta- tic hypertrophy (BPH) and 55 controls. Results: The serum tPSA, cPSA and IGF-I levels in patients with prostatic cancer were significantly higher than those in patients with BPH and controls (P<0.01). Taking the cut-off values of 4ng/ml, 3.6ng/ml and 150 for tPSA, cPSA and IGF-I respectively, the combined determination of these three items would yield a sensitivity of 88.6%, specificity of 84.9%, positive predicative value of 83% and negative predicative value of 90.0% for diagnosis of prostatic cancer. Conclusion: Combined determination of tPSA, cPSA and IGF-I would yield better sensitive and accurate diagnostic rate in patients with prostatic cancer, especially in those with laboratory values within the 'grey zone'. (authors)

  1. Human model simulation of plant anomaly diagnosis (HUMOS-PAD) to estimate time cognitive reliability curve for HRA/PSA practice

    International Nuclear Information System (INIS)

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

    2000-01-01

    A computer-simulation-based approach has been proposed to estimate fundamental human error probability (HEP) parameters, which are required for Human Reliability Analysis (HRA) as related with Probabilistic Safety Assessment (PSA) for nuclear power plant (NPP). The target HEP parameters are the ones normally represented by time versus reliability correlation (TRC), which is called as 'Time versus Cognitive Reliability' (TCR) curves. In the developed simulation system, 'human model adjustment factors' were proposed to model the inherent variety and diversity in human cognitive behaviors, which are considered as the primary factors that generate different TRC. Inter-comparisons were also made between the TCR curves obtained from a laboratory experimental data and the ones obtained by the developed simulation system. It turns out that both TCR curves agree well with each other and it suggests that the computer simulation approach would be a promising technique for estimating HEP. (author)

  2. Time from first detectable PSA following radical prostatectomy to biochemical recurrence: A competing risk analysis

    Science.gov (United States)

    de Boo, Leonora; Pintilie, Melania; Yip, Paul; Baniel, Jack; Fleshner, Neil; Margel, David

    2015-01-01

    Introduction: In this study, we estimated the time from first detectable prostate-specific antigen (PSA) following radical prostatectomy (RP) to commonly used definitions of biochemical recurrence (BCR). We also identified the predictors of time to BCR. Methods: We identified subjects who underwent a RP and had an undetectable PSA after surgery followed by at least 1 detectable PSA between 2000 and 2011. The primary outcome was time to BCR (PSA ≥0.2 and successive PSA ≥0.2) and prediction of the rate of PSA rise. Outcomes were calculated using a competing risk analysis, with univariable and multivariable Fine and Grey models. We employed a mixed effect model to test clinical predictors that are associated with the rate of PSA rise. Results: The cohort included 376 patients. The median follow-up from surgery was 60.5 months (interquartile range [IQR] 40.8–91.5) and from detectable PSA was 18 months (IQR 11–32). Only 45.74% (n = 172) had PSA values ≥0.2 ng/mL, while 15.16% (n = 57) reached the PSA level of ≥0.4 ng/mL and rising. On multivariable analysis, the values of the first detectable PSA and pathologic Gleason grade 8 or higher were consistently independent predictors of time to BCR. In the mixed effect model rate, the PSA rise was associated with time from surgery to first detectable PSA, Gleason score, and prostate volume. The main limitation of this study is the large proportion of patients that received treatment without reaching BCR. It is plausible that shorter estimated median times would occur at a centre that does not use salvage therapy at such an early state. Conclusion: The time from first detectable PSA to BCR may be lengthy. Our analyses of the predictors of the rate of PSA rise can help determine a personalized approach for patients with a detectable PSA after surgery. PMID:25624961

  3. The correlation of PSA nadir and biochemical freedom from cancer after external beam treatment: effects of stage, grade and pretreatment PSA groupings

    International Nuclear Information System (INIS)

    Pinover, W.H.; Hanlon, A.L.; Lee, W.R.; Hanks, G.E.

    1996-01-01

    Purpose: This study demonstrates the correlation of various post-irradiation PSA nadirs with long term biochemical freedom from disease (bNED) survival in patients treated mainly with conformal external beam radiation therapy. It also shows the effects of various groupings of pretreatment (prerx) PSA level, stage, and Gleason score on the rate of achieving a favorable PSA nadir. Materials and Methods: Three hundred forty patients with known pretreatment PSA, >2 years followup treated with radiation alone (278 conformal, 62 conventional) are reported. The median followup is 41 months (range 24 to 96 mos.). Patient grouping by pretreatment PSA levels are <10 ng/ml (143 patients), 10-19.9 ng/ml (108 patients), ≥20 ng/ml (89 patients); by palpation stage are T1C,2AB (240 patients) and T2C,3,4 (100 patients); and by differentiation are Gleason 2-4 (108 patients), Gleason 5-7 (221 patients), Gleason 8-10 (11 patients). The PSA nadir response is given for all patients, and for each of the above prerx groupings. The 5 year actuarial bNED survival is determined for all patients by PSA nadir. Biochemical failure is a PSA ≥1.5 ng/ml and rising on two consecutive measures. Multivariate analysis (MVA) is performed to determine factors predictive of favorable PSA nadir response and predictive of bNED survival. Results: The PSA nadir responses and 5 year bNED survival rates are shown in the table for all patients according to PSA nadir. 66% of patients achieved a favorable nadir (<1.0 ng/ml) which was associated with a 75%-87% 5 year bNED rate, while 34% achieved an unfavorable nadir associated with an 18-32% bNED survival rate at 5 years. The figure illustrates the dramatic separation in outcome associated with the nadir response. The table also illustrates the fraction of patients that achieve various nadir levels subdivided by prerx PSA level, palpation stage and Gleason score. A favorable PSA nadir is obtained in 90%, 63%, and 31% of patients with a prerx PSA <10, 10

  4. Rapid elimination kinetics of free PSA or human kallikrein-related peptidase 2 after initiation of gonadotropin-releasing hormone-antagonist treatment of prostate cancer

    DEFF Research Database (Denmark)

    Ulmert, David; Vickers, Andrew J; Scher, Howard I

    2012-01-01

    The utility of conventional prostate-specific antigen (PSA) measurements in blood for monitoring rapid responses to treatment for prostate cancer is limited because of its slow elimination rate. Prior studies have shown that free PSA (fPSA), intact PSA (iPSA) and human kallikrein-related peptidase...... of tPSA, fPSA, iPSA and hK2 after rapid induction of castration with degarelix (Firmagon(®)), a novel GnRH antagonist....

  5. Hand Hygiene in Healthcare Settings 2 PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2010-08-19

    This 30 second PSA encourages people to wash their hands often while in the hospital or visiting someone in the hospital. It also encourages them to remind their healthcare providers to wash their hands, too.  Created: 8/19/2010 by Centers for Disease Control and Prevention (CDC).   Date Released: 8/19/2010.

  6. Hand Hygiene in Healthcare Settings 1 PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2010-08-19

    This 30 second PSA encourages people to wash their hands often while in the hospital or visiting someone in the hospital. It also encourages them to remind their healthcare providers to wash their hands, too.  Created: 8/19/2010 by Centers for Disease Control and Prevention (CDC).   Date Released: 8/19/2010.

  7. Asian and Pacific Islander HIV/AIDS Awareness PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    2010-05-12

    In this PSA, Asians and Pacific Islanders are encouraged to talk about HIV and get tested for HIV.  Created: 5/12/2010 by Centers for Disease Control and Prevention (CDC).   Date Released: 5/12/2010.

  8. The PSA assessment of Defense in Depth Memorandum and proposals

    International Nuclear Information System (INIS)

    Fiorini, Gian-Luigi; La Rovere, Stefano

    2016-01-01

    This report concerns the peculiar roles of the Defence-in-Depth (DiD) concept and the Probabilistic Safety Assessment (PSA) approach for the optimization of the safety performances of the nuclear installation. It proposes a conceptual framework and related process for the assessment of the 'safety architecture' implementing DiD, which is articulated in four main steps devoted to (1) the formulation of the safety objectives, (2) the identification of loads and environmental conditions, (3) the representation of the safety architecture and (4) the evaluation of the physical performance and reliability of the levels of DiD. A final additional step achieves the practical assessment of the safety architecture and the corresponding DiD with the support of the PSA. The comprehensive safety assessment of the implemented architecture needs its multi-dimensional representation, i.e. for given initiating event, sequence of possible failures, affected safety function and level of DiD. The risk space (frequency/probability of occurrence, versus consequences) is the framework for the integration between the DiD concept and the PSA approach. Additional qualitative key-notions are introduced in order to address the compliance of the safety architecture with a number of international safety requirements. In this context, the role of the PSA is no longer limited to the verification of the fulfilment of probabilistic targets but includes different contributions to the assessment of the DiD identified in this report. (authors)

  9. Use of living PSA in regulatory decision making in Finland

    International Nuclear Information System (INIS)

    Virolainen, R.

    1999-01-01

    Consideration of severe accidents beyond the traditional design basis, including full core melt accidents, has become an important ingredient of regulatory process in Finland. Increasingly, decisions are being based, at least in part, on results of plant-specific Probabilistic Safety Assessments (PSA) studies. Plant-specific level-1 and level 2 PSA studies, including internal initiators, fires, flooding and harsh weather conditions are required by STUK. These studies are used in a living fashion both at the utilities and a STUK. PSA has got an important role in the safety management at Loviisa and Olkiluoto (OL) plants and in the regulatory process of STUK. While PSA is recognised as an effective tool and review method for many different regulatory and safety management purposes, we have to acknowledge its limitations, which are often related to the level of modelling or methods, not yet mature enough for some specific applications. Hence, in context of the aforementioned activities, STUK has to use both deterministic and probabilistic reviews in parallel while controlling and regulating the issues. (au)

  10. Safety and efficacy of procedural sedation and analgesia (PSA ...

    African Journals Online (AJOL)

    Safety and efficacy of procedural sedation and analgesia (PSA) conducted by medical officers in a level 1 hospital in Cape Town. ... Respiratory complications were treated with simple airway manoeuvres; no patient required intubation or experienced respiratory problems after waking up. There was no significant difference ...

  11. PyPSA: Python for Power System Analysis

    Directory of Open Access Journals (Sweden)

    Thomas Brown

    2018-01-01

    Full Text Available Python for Power System Analysis (PyPSA is a free software toolbox for simulating and optimising modern electrical power systems over multiple periods. PyPSA includes models for conventional generators with unit commitment, variable renewable generation, storage units, coupling to other energy sectors, and mixed alternating and direct current networks. It is designed to be easily extensible and to scale well with large networks and long time series. In this paper the basic functionality of PyPSA is described, including the formulation of the full power flow equations and the multi-period optimisation of operation and investment with linear power flow equations. PyPSA is positioned in the existing free software landscape as a bridge between traditional power flow analysis tools for steady-state analysis and full multi-period energy system models. The functionality is demonstrated on two open datasets of the transmission system in Germany (based on SciGRID and Europe (based on GridKit.   Funding statement: This research was conducted as part of the CoNDyNet project, which is supported by the German Federal Ministry of Education and Research under grant no. 03SF0472C. The responsibility for the contents lies solely with the authors

  12. GIS Technologies for the Planetary Science Archive (PSA)

    Science.gov (United States)

    Docasal, R.

    2017-09-01

    In my abstract I try to show how a GIS and 3D visual tools architecture could handle the different approaches for visualizing the spatial info, depending on the nature and shape of the object (planet, satellite, comet...etc) to be mapped in a multi-mission website such as the new PSA.

  13. Prognostic Significance of Prostate-Specific Antigen (PSA) Rate of ...

    African Journals Online (AJOL)

    Aim: To investigate the prognostic significance of Prostate-Specific Antigen (PSA) rate of change in patients with advanced prostate cancer . Patients and Methods: A total of forty-nine male patients aged between 42 and 84 years with advanced prostate cancer receiving therapy of maximum androgen bloackade were ...

  14. Reducing the Risk of Methadone Overdose PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    2012-07-03

    This 60 second PSA is based on the July 2012 CDC Vital Signs report. Approximately 14 people die every day of overdoses related to methadone. Listen to learn how to reduce your risk of an overdose.  Created: 7/3/2012 by Centers for Disease Control and Prevention (CDC).   Date Released: 7/3/2012.

  15. Use of PSA for design modifications and backfitting

    International Nuclear Information System (INIS)

    Evans, M.G.K.

    1997-01-01

    The aim of this presentation is to gain an understanding of how the living PSA can be used to evaluate proposed design changes and backfitting modifications. The topics are: Process for evaluation of design change. Application to various types of change - Design changes; Procedure changes; Temporary plant modifications; Design deviations. 1 fig

  16. Association of sociodemographic factors and prostate-specific antigen (PSA) testing.

    Science.gov (United States)

    Gorday, William; Sadrzadeh, Hossein; de Koning, Lawrence; Naugler, Christopher

    2014-11-01

    There are conflicting recommendations regarding the use of prostate specific antigen (PSA) as a screening test. Integral to this debate is an understanding of who is currently being tested. The purpose of this study was to provide a detailed account of PSA testing practices in a major Canadian city (Calgary, Alberta) and to identify variables that may affect access to the PSA test. PSA test counts were retrieved from Calgary Laboratory Services' Laboratory Information System from January 1, 2011 to December 31, 2011. A total of 75,914 individual PSA tests were included in our analysis. The frequency of PSA testing was plotted onto a dissemination area map of Calgary using ArcGIS software. Associations with sociodemographic variables were tested using Poisson regression. The median PSA value was 0.93 μg/L and the median age at collection was 58 years. Forty-three percent of men aged 60-69 received a PSA test. Visible minority status 'Black' (P=0.0002) and Métis status (P=0.0075) were associated with lower PSA testing frequencies, while median household income (P=PSA testing frequencies. There are areas in Calgary which are significantly over or under tested relative to the mean. The amount of PSA testing in men PSA testing guidelines. Copyright © 2014 The Canadian Society of Clinical Chemists. Published by Elsevier Inc. All rights reserved.

  17. Understanding PSA and its derivatives in prediction of tumor volume: Addressing health disparities in prostate cancer risk stratification.

    Science.gov (United States)

    Chinea, Felix M; Lyapichev, Kirill; Epstein, Jonathan I; Kwon, Deukwoo; Smith, Paul Taylor; Pollack, Alan; Cote, Richard J; Kryvenko, Oleksandr N

    2017-03-28

    To address health disparities in risk stratification of U.S. Hispanic/Latino men by characterizing influences of prostate weight, body mass index, and race/ethnicity on the correlation of PSA derivatives with Gleason score 6 (Grade Group 1) tumor volume in a diverse cohort. Using published PSA density and PSA mass density cutoff values, men with higher body mass indices and prostate weights were less likely to have a tumor volume PSA derivatives when predicting for tumor volume. In receiver operator characteristic analysis, area under the curve values for all PSA derivatives varied across race/ethnicity with lower optimal cutoff values for Hispanic/Latino (PSA=2.79, PSA density=0.06, PSA mass=0.37, PSA mass density=0.011) and Non-Hispanic Black (PSA=3.75, PSA density=0.07, PSA mass=0.46, PSA mass density=0.008) compared to Non-Hispanic White men (PSA=4.20, PSA density=0.11 PSA mass=0.53, PSA mass density=0.014). We retrospectively analyzed 589 patients with low-risk prostate cancer at radical prostatectomy. Pre-operative PSA, patient height, body weight, and prostate weight were used to calculate all PSA derivatives. Receiver operating characteristic curves were constructed for each PSA derivative per racial/ethnic group to establish optimal cutoff values predicting for tumor volume ≥0.5 cm3. Increasing prostate weight and body mass index negatively influence PSA derivatives for predicting tumor volume. PSA derivatives' ability to predict tumor volume varies significantly across race/ethnicity. Hispanic/Latino and Non-Hispanic Black men have lower optimal cutoff values for all PSA derivatives, which may impact risk assessment for prostate cancer.

  18. Case study on the use of PSA methods: Backfitting decisions

    International Nuclear Information System (INIS)

    1991-04-01

    This case study illustrates the process of using probabilistic risk assessment (PRA) method to evaluate proposed backfits of nuclear power plants (NPP), which are intended to enhance the plant safety by improving equipment operability. Some examples of situations in which PRA techniques have been used to address backfit issues at operating NPPs are summarized. 2 refs, 5 figs, 4 tabs

  19. ATUCHA I NPP - Emergency drill practice

    International Nuclear Information System (INIS)

    Sanda, Alejandro; Rosales, Gabriel

    2008-01-01

    Full text: Atucha I NPP performs an Emergency Drill Practice once a year. Its main goals are: -) Fulfill the requirements of the Argentine Nuclear Regulatory Authority (ARN) regarding Atucha I NPP's Operating License; -) Fulfill the commitment with the community regarding the safe and reliable operation Atucha I NPP; -) Verify the response of the Civil Organizations, Security Forces, and Armed Forces, as well as the correct application of the Emergency Plan; -) Perform the 'General Alarm Drill' periodic control; -) Perform a re-training of the members of the Security Advisor Internal Committee (CIAS) on the Internal and External Aspects of the Emergency Plan and on the related procedures; -) Test the Emergency Communications System. New goals are added every year, considering the Drill's scope. This drill comprises two different kinds of practices: Internal practices (practices in the station, with our personnel) and external practices (practices outside the station with governmental organizations). Internal practices comprise: -) Internal and external communications practices; -) Acoustic alarms; -) Personnel gathering in the Meeting Points; -) Safety of selected Meeting Points; -) Personnel count, selective evacuation; -) Iodide Potassium pills distribution; -) CICE (Internal Group for Emergency Control) Coordination. External practices comprise: -) Nuclear Regulatory Authority; -) Argentine Navy, Comando Area Naval Fluvial, Base Naval Zarate; -) Lima firemen; -) Zarate firemen; -) Municipal Civil Defense (Zarate and Lima); -) National Guard, Escuadron Atucha; -) Zarate Regional Hospital; -) Lima Police Department; -) Zarate Police Department; -) Argentine Coast Guard, Zarate; -) Local radios: Radio FM Libre, FM El Sitio; -) First Aid clinic. The following activities are performed together with the aforementioned organizations: -) Formation of an 'Operative committee'; -) Evacuation of citizens in a 3 km radio; -) Control of every access to Lima; -) Control of

  20. Decommissioning of NPP A-1 Phase I, Jaslovske Bohunice. Documentation for application for permission to Phase II of decommissioning of NPP A-1. Schedule stage II of decommissioning of NPP A-1

    International Nuclear Information System (INIS)

    2007-04-01

    In this study documentation for application for permission to Phase II of decommissioning of NPP A-1 and the schedule stage II of decommissioning of NPP A-1 are presented. This study consists of ten appendixes.

  1. GIS Technologies For The New Planetary Science Archive (PSA)

    Science.gov (United States)

    Docasal, R.; Barbarisi, I.; Rios, C.; Macfarlane, A. J.; Gonzalez, J.; Arviset, C.; De Marchi, G.; Martinez, S.; Grotheer, E.; Lim, T.; Besse, S.; Heather, D.; Fraga, D.; Barthelemy, M.

    2015-12-01

    Geographical information system (GIS) is becoming increasingly used for planetary science. GIS are computerised systems for the storage, retrieval, manipulation, analysis, and display of geographically referenced data. Some data stored in the Planetary Science Archive (PSA), for instance, a set of Mars Express/Venus Express data, have spatial metadata associated to them. To facilitate users in handling and visualising spatial data in GIS applications, the new PSA should support interoperability with interfaces implementing the standards approved by the Open Geospatial Consortium (OGC). These standards are followed in order to develop open interfaces and encodings that allow data to be exchanged with GIS Client Applications, well-known examples of which are Google Earth and NASA World Wind as well as open source tools such as Openlayers. The technology already exists within PostgreSQL databases to store searchable geometrical data in the form of the PostGIS extension. An existing open source maps server is GeoServer, an instance of which has been deployed for the new PSA, uses the OGC standards to allow, among others, the sharing, processing and editing of data and spatial data through the Web Feature Service (WFS) standard as well as serving georeferenced map images through the Web Map Service (WMS). The final goal of the new PSA, being developed by the European Space Astronomy Centre (ESAC) Science Data Centre (ESDC), is to create an archive which enables science exploitation of ESA's planetary missions datasets. This can be facilitated through the GIS framework, offering interfaces (both web GUI and scriptable APIs) that can be used more easily and scientifically by the community, and that will also enable the community to build added value services on top of the PSA.

  2. Living PSA issues in France on pressurized water reactors

    International Nuclear Information System (INIS)

    Dewailly, J.; Deriot, S.; Dubreuil Chambardel, A.; Francois, P.; Magne, L.

    1993-09-01

    Two Probabilistic Safety Assessments (PSAs) carried out in France on 900 and 1300 MWe Pressurized Water Reactor units ended in 1990. These PSAs determined the core damage frequency for all plant operating conditions ranging from cold shutdown for refuelling to full power operation. Since 1990, these PSAs have been used increasingly as tools for applications such as accident precursor analysis, risk-based Technical Specifications, and maintenance optimization. In turn, these applications are used to enhance the initial PSAs. The notion of a ''living'' PSA which can be used and updated is slowly taking form. The accident precursor analysis consists in applying PSA event trees to obtain quick information on the potential consequences of a precursor event and on the corresponding probabilities of occurrence. A feedback on PSAs is provided by comparing them with actual operating incidents. The computation of the allowed outage time during power operation, based on the computerized models of Probabilistic Safety Assessments, requires adjustments: calculation of hourly risk of core damage under different reactor conditions without equipment unavailabilities. The proposed method also turns out to be an aid in determining the safe shutdown condition and procedure. Furthermore, when introducing a sufficient level of detail, PSA reliability models make it possible to compute contributions and to perform sensitivity studies in order to highlight those components for which a maintenance effort should be made. From the experience acquired up to now, there was felt to be a strong need to create guidelines for using PSAs that would simplify their implementation by the experts in charge of determining Technical Specifications, of maintenance programs, etc. who are not generally specialists in PSAs. For this purpose, it is necessary to improve the intelligibility of the models made in order for them to be used and to offer user's guides adapted to each application. Documents

  3. [Relationship between tumor volume and PSA recurrence after radical prostatectomy].

    Science.gov (United States)

    Hashimoto, Yasuhiro; Momose, Akishi; Okamoto, Akiko; Yamamoto, Hayato; Hatakeyama, Shingo; Iwabuchi, Ikuya; Yoneyama, Takahiro; Koie, Takuya; Kamimura, Noritaka; Ohyama, Chikara

    2010-02-01

    We examined whether the tumor volume (TV) is a good predictor of PSA recurrence after radical prostatectomy. Data were collected for 158 patients with clinically localized prostate cancer undergoing radical prostatectomy without neoadjuvant hormonal therapy in our hospital since April 2005 to September 2007. Along with the routine pathological assessment, TV was assessed in all prostatectomy specimens. PSA recurrence was defined as PSA levels of greater than 0.2 ng/ml. The TVs were 1.81+/-1.66 ml (mean +/-SD) ranging from 0.02 to 8.20 ml. The TV in cT1c was 1.77+/-1.64, and 1.89+/-1.72 ml in cT2 (not significant). Significant differences were observed between TV and pT. The TVs in pT2a, pT2b and pT3/4 were 0.54+/-0.54, 1.63+/-1.47 and 2.67+/-1.80 ml, respectively. The median follow-up period was 32.3 months (range from 15 to 45) after radical prostatectomy, and PSA recurrence was observed in 32 cases. Patients with smaller TV (TV TV (TV > or = 1.3 ml, 66.7%) with a significant difference atp TV, pT, Gleason Score (GS), and surgical margins. Significant differences were observed for GS, and surgical margins, but not for TV. Clinically organ-confined disease in Japanese patients with prostate cancer included various cancers from clinically insignificant to locally advanced ones. In our series, TV was not regarded as a predictor of PSA recurrence after radical prostatectomy.

  4. IAEA activities on NPP personnel training and qualification

    International Nuclear Information System (INIS)

    Kossilov, A.

    1998-01-01

    Activities of IAEA concerning training and qualification of NPP personnel consider the availability of sufficient number of competent personnel which is one of the most critical requirements for safe and reliable NPP operation and maintenance. Competence of personnel is essential for reducing the frequency of events connected to human errors and equipment failures. The IAEA Guidebook on Nuclear Power Plant Personnel Training and its Evaluation incorporates the experience gained worldwide and provides recommendations on the use of SAT being the best practice for attaining and maintaining the qualification and competence of NPP personnel and for quality assurance of training

  5. A reliability evaluation method for NPP safety DCS application software

    International Nuclear Information System (INIS)

    Li Yunjian; Zhang Lei; Liu Yuan

    2014-01-01

    In the field of nuclear power plant (NPP) digital i and c application, reliability evaluation for safety DCS application software is a key obstacle to be removed. In order to quantitatively evaluate reliability of NPP safety DCS application software, this paper propose a reliability evaluating method based on software development life cycle every stage's v and v defects density characteristics, by which the operating reliability level of the software can be predicted before its delivery, and helps to improve the reliability of NPP safety important software. (authors)

  6. Construction of large-thickness sand cushions for NPP foundations

    International Nuclear Information System (INIS)

    Krantsfel'd, Ya.L.; Losievskaya, I.K.; Kovalenko, R.P.; Mutalipov, A.

    1982-01-01

    A study is made on some technological peculiarities of preparation of NPP foundations and control methods of foundation density. As an example the experience of cement-sand foundation construction for two 900 MW power-units at the Koeberg South Africa NPP is briefly described. The experience of artificial foundation construction at this NPP indicates both the possibility of obtaining the required quality of cement-sand cushions and commercial construction of large volume of such cushions by acceptable rates and the necessity of unification of work quality characteristics

  7. Is Systemic Inflammation Associated With Elevated PSA Serum Levels In Patients Submitted Chronic Hemodialysis?

    Directory of Open Access Journals (Sweden)

    Gilmar Pereira Silva

    2017-09-01

    Full Text Available Backgroundː whereas that systemic inflammation (SI affects 40–60% of patients on hemodialysis (HD is characterized by serum C-reactive protein (CRP level elevation or proinflammatory interleukin production or both. We evaluated the association between SI and total (tPSA and free PSA (fPSA in patients on HD with tPSA 6 months. Patients were excluded if they had local infections or SI. Hs-CRP was measured using turbidimetry, and tPSA and fPSA levels using immunochemoluminescence. Overall, 27 patients had inflammation (hs-CRP >5 mg/L and 33 had no inflammation (hs-CRP was ≤5 mg/L. In the control group, hs-CRP was ≤ 1 mg/L. Resultsː there was no significant difference in mean levels among groups 3 and 4 for age (p=0,058, tPSA (p=0,74 and fPSA (p=0,30. The SI did not promote differences between groups 1, 2 and 4 for the levels of tPSA (0,71 ± 0,18  vs   0,67 ± 0,15  vs  0,67 ± 0,11; p=0,69 and fPSA (0,34  ±  0,01  vs  0,34  ±  0,01  vs   0,35  ±  0,01, p= 0,59 . As well as maintained no correlation with tPSA and fPSA (p>0,05. Conclusionː The systemic inflammation in hemodialytic patients without clinically detectable cancer (PSA<4ng/ml is no associated with changes fractions of tPSA and fPSA.

  8. Prostate-specific antigen (PSA) density in the diagnostic algorithm of prostate cancer.

    Science.gov (United States)

    Nordström, Tobias; Akre, Olof; Aly, Markus; Grönberg, Henrik; Eklund, Martin

    2018-04-01

    Screening for prostate cancer using prostate-specific antigen (PSA) alone leads to un-necessary biopsying and overdiagnosis. PSA density is easily accessible, but early evidence on its use for biopsy decisions was conflicting and use of PSA density is not commonly recommended in guidelines. We analyzed biopsy outcomes in 5291 men in the population-based STHLM3 study with PSA ≥ 3 ng/ml and ultrasound-guided prostate volume measurements by using percentages and regression models. PSA density was calculated as total PSA (ng/ml) divided by prostate volume (ml). Main endpoint was clinically significant cancer (csPCa) defined as Gleason Score ≥ 7. The median PSA-density was 0.10 ng/ml 2 (IQR 0.075-0.14). PSA-density was associated with the risk of finding csPCa both with and without adjusting for the additional clinical information age, family history, previous biopsies, total PSA and free/total PSA (OR 1.06; 95% CI:1.05-1.07 and OR 1.07, 95% CI 1.06-1.08). Discrimination for csPCa was better when PSA density was added to a model with additional clinical information (AUC 0.75 vs. 0.73, P PSA-density. Omitting prostate biopsy for men with PSA-density ≤0.07 ng/ml 2 would save 19.7% of biopsy procedures, while missing 6.9% of csPCa. PSA-density cutoffs of 0.10 ng/ml 2 and 0.15 ng/ml 2 resulted in detection of 77% (729/947) and 49% (461/947) of Gleason Score ≥7 tumors. PSA-density might inform biopsy decisions, and spare some men from the morbidity associated with a prostate biopsy and diagnosis of low-grade prostate cancer.

  9. NPP Cernavoda Unit 2 Financing Completion Works

    International Nuclear Information System (INIS)

    Chirica, T.; Stefanescu, A.; Constantin, C.; Dobrin, M.

    2002-01-01

    NPP Cernavoda Unit 2 completion is the highest priority of the Romanian power sector strategy. The nuclear energy represents, through its technological features of adopted solution (a CANDU nuclear power plant) and also through technological and economical performance indicators, the best solution to fulfill the demands concerning the sustainable development and the electricity request. The guidelines of energy strategy regarding the nuclear sector development in Romania are framing in the general policy for energy system development at least costs and they are responding to requests concerning the environment and people protection. The paper presents the financing alternatives for Unit 2 completion works taking into consideration the financing market conditions. The paper presents the impact of the financing conditions on the project efficiency, as well as the facilities offered by the Romanian Government in order to support this project. (author)

  10. Fuel consumption organization at the Kola NPP

    International Nuclear Information System (INIS)

    Matveev, A.A.; Ignatenko, E.I.; Volkov, A.P.; Trofimov, B.A.

    1981-01-01

    Problems of using NPPs in the power systems including hydroelectric power plants and NPPs are considered on the example of the Kola power system. The methods of the WWER-440 reactor fuel loading formation, reactor power forcing, optimization of volumes and time of the NPP main equipment planned maintenance are discussed. It is concluded that the optimal methods for the WWER-440 reactor fuel loading formation are the following: reactor make-up with the lesser number of fuel assemblies with maximum designed enrichment; for the case of decreased loading energy capacity displacement of make-up fuel with 2.4% enrichment by the fuel with 3.6% enrichment when preserving the designed number of make-up fuel assemblies [ru

  11. Feature article. Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    Ekarinai, Masashi; Ake, Yutaka; Narabayashi, Tadashi

    2011-01-01

    This special feature article consisted of five reports and the minutes of emergency discussion meeting on Fukushima Daiichi Nuclear Power Plant (NPP) accident. Effects of the accident on future electricity supply of electric utilities and also on business development of nuclear industries were discussed. Activities of senior network team of atomic energy society of Japan (AESJ) to conduct severe accident analysis and early restoration from the accident were introduced. Circulating injection reactor cooling system and zeolite decontamination system of accumulated contaminated water was proposed. Effects of the accident on overseas reaction on nuclear development were also reported as well as personal experience of the professor in the US west coast on communications. (T. Tanaka)

  12. N-16 monitors: Almaraz NPP experience

    International Nuclear Information System (INIS)

    Adrada, J.

    1997-01-01

    Almaraz Nuclear Power Plant has installed N-16 monitors - one per steam generator - to control the leakage rate through the steam generator tubes after the application of leak before break (LBB) criteria for the top tube sheet (TTS). After several years of operation with the N-16 monitors, Almaraz NPP experience may be summarized as follows: N-16 monitors are very useful to follow the steam generator leak rate trend and to detect an incipient tube rupture; but they do not provide an exact absolute leak rate value, mainly when there are small leaks. The evolution of the measured N-16 leak rates varies along the fuel cycle, with the same trend for the 3 steam generators. This behaviour is associated with the primary water chemistry evolution along the cycle

  13. Students education and training for Slovak NPP

    International Nuclear Information System (INIS)

    Slugen, V.; Lipka, J.; Hascik, J.; Miglierini, M.

    2005-01-01

    Slovak University of Technology is the largest and also the oldest university of technology in Slovakia. It is certain that more than 50% of the highly-educated technicians who are currently working in the nuclear industry have graduated from this university. The Department of Nuclear Physics and Technology of the Faculty of Electrical Engineering and Information Technology as one of the seven faculties of this University feels the responsibility to impart proper engineering education and training for Slovak NPP operating staff. The education process is realised via undergraduate (BSc), graduate (MSc) and postgraduate (PhD) study as well as via specialised training courses within the framework of a continuous education system. (author)

  14. Features of the Kozloduy NPP management system

    International Nuclear Information System (INIS)

    2016-01-01

    The Kozloduy NPP management system was built taking into account the specifics of the organizational structure and management of the Company, actual processes and practices, and is oriented towards future development, with the participation of all staff. Additional requirements integrated in the system that distinguish it from general industrial requirements of management systems are: priority of nuclear safety; safety culture; knowledge management including extraction and storage of 'hidden knowledge'; periodic self-assessments; use of graded response to the products and activities; use of 'conservative approach' in decision making;; possibilities for self learning and creating of a vision of 'leaders' and 'professional workers in nuclear energy

  15. Concreting organization during Chernobylsk NPP construction

    International Nuclear Information System (INIS)

    Lysyuk, R.I.; Kareva, A.P.

    1984-01-01

    Conreting organization during the Chernobylsk NPP construction is described. Processes of extra heavy concrete production and placement, which specific mass constitutes 4t/m 3 at the age of 28 days wiath metallic aggregates and 3.3-3.5 t/m 3 at the same age without aggregates, are considered in short. Basic characteristics of this concrete are presented. At the 4th power unit labour contents for construction works were a 1.5 times lower as compared to the 3rd power unit erection. This progress was achieved by round-the-clock operation of the concrete plant with the 800 m 3 /day output and also by utilization of special equipment for mechanized concrete placement: concrete pumps, automatic concrete mixer, manipulators and concrete pipelines

  16. Construction of Belene NPP in Bulgaria

    International Nuclear Information System (INIS)

    JSC Atomstroyexport

    2010-01-01

    Presentations concluding remarks: ASE has performed its scope of responsibilities under the Agreement of 29.11.2006 and has achieved great results regarding both the Technical part of the Project and its organization; Though there is a number of unsettled issues under the Project, in particular, the issue related to financing, ASE is willing to continue the Project and works on its development; The Russian Party believes that in case the activities under the Project are continued, Belene NPP will be constructed with high quality and within the time limits prescribed in the Agreement of 29.11.2006: 59 months before Unit 1 take-over into operation and 71 month before Unit 2 take-over into operation, starting from concreting of foundation slab of Unit 1 Reactor building

  17. Electrohydraulic system to control NPP turbines

    International Nuclear Information System (INIS)

    Kosyak, Yu.F.; Virchenko, M.A.; Rozhanskij, V.E.; Rokhlenko, V.Yu.; Gapunin, A.Ya.; Zhornitskaya, T.Ya.; Rasskazov, I.Eh.; Butsenko, V.N.; Brajnin, L.S.; Makarenko, N.I.

    1985-01-01

    Operation regimes of electrohydraulic regulation system (EHRS) of NPP turbines, designed to control the turbine in start-up and working conditions, have been decribed. In start-up regimes EHRS ensures the testing of control valves of the turbine, the turn of the turbine from zero to the nominal rotation frequency (automatic, semiautomatic and manual regulation), turbine acceleration to test safety automatic systems, gradual change in rotation frequency during generator synchronization with circuit. Under working conditions EHRS ensures the maintenance of frequency, power and vapour pressure before the turbine. A block diagram of EHRS is presented. Sensors and electronic part of EHRS are supplied with triple reservation, which ensures a high relaibility of the system

  18. Contribution of allelic variability in prostate specific antigen (PSA) & androgen receptor (AR) genes to serum PSA levels in men with prostate cancer.

    Science.gov (United States)

    Chavan, Sushant V; Maitra, Anurupa; Roy, Nobhojit; Chavan, Padma R

    2014-03-01

    Wide variability in serum prostate specific antigen (PSA) levels exists in malignant conditions of the prostate. PSA is expressed in normal range in 20 to 25 per cent of prostate cancer cases even in presence of high grade Gleason score. This study was aimed to assess the influence of genetic variants exhibited by PSA and androgen receptor (AR) genes towards the variable expression of PSA in prostate cancer. Pre-treatment serum PSA levels from 101 prostate cancer cases were retrieved from medical record. PSA genotype analysis in promoter region and AR gene microsatellite Cytosine/Adenine/Guanine (CAG) repeat analysis in exon 1 region was performed using DNA sequencing and fragment analysis techniques. A total of seven single nucleotide polymorphisms (SNPs) in the PSA promoter region were noted. Only two SNPs viz., 158G/A (PPSA levels. The carriers of homozygous GG genotype (PPSA whereas homozygous AA genotype (PPSA levels. The combination effect of PSA genotypes along with stratified AR CAG repeats lengths (long, intermediate and short) was also studied. The homozygous GG genotype along with AR long CAG repeats and homozygous AA genotype along with AR short CAG repeats at position -3845 and -158 showed strong interaction and thus influenced serum PSA levels. The genetic variants exhibited by PSA gene at positions -3845G/A and -158G/A may be accountable towards wide variability of serum PSA levels in prostate cancer. Also the preferential binding of G and A alleles at these polymorphic sites along with AR long and short CAG repeats may contribute towards PSA expression.

  19. Detection of primary coolant leaks in NPP

    International Nuclear Information System (INIS)

    Slavov, S.; Bakalov, I.; Vassilev, H.

    2001-01-01

    The thermo-hydraulic analyses of the SG box behaviour of Kozloduy NPP units 3 and 4 in case of small primary circuit leaks and during normal operation of the existing ventilation systems in order to determine the detectable leakages from the primary circuit by analysing different parameters used for the purposes of 'Leak before break' concept, performed by ENPRO Consult Ltd. are presented. The following methods for leak detection: measurement of relative air humidity in SG box (can be used for detection of leaks with flow rate 3.78 l/min within one hour at ambient parameters - temperature 40 0 - 60 0 C and relative humidity form 30% to 60%); measurement of water level in SG box sumps (can not be used for reliable detection of small primary circuit leakages with flow rate about 3.78 l/min); measurement of gaseous radioactivity in SG box( can be used as a general global indication for detection of small leakages from the primary circuit); measurement of condensate flow after the air coolers of P-1 venting system (can be used for primary circuit leak detection) are considered. For determination of the confinement behaviour, a model used with computer code MELCOR has been developed by ENPRO Consult Ltd. A brief summary based on the capabilities of the different methods of leak detection, from the point of view of the applicability of a particular method is given. For both Units 3 and 4 of Kozloduy NPP a qualified complex system for small leak detection is planned to be constructed. Such a system has to unite the following systems: acoustic system for leak detection 'ALUS'; system for control of the tightness of the main primary circuit pipelines by monitoring the local humidity; system for primary circuit leakage detection by measuring condensate run-off in collecting tank after ventilation system P-1 air coolers

  20. Indicators to monitor NPP operational safety performance

    International Nuclear Information System (INIS)

    Gomez-Cobo, Ana

    2002-01-01

    Since December 1995 the IAEA activities on safety performance indicators focused on the elaboration of a framework for the establishment of an operational safety performance indicator programme. The development of this framework began with the consideration of the concept of NPP operational safety performance and the identification of operational safety attributes. For each operational safety attribute, overall indicators, envisioned as providing an overall evaluation of relevant aspects of safety performance, were established. Associated with each overall indicator is a level of strategic indicators intended to provide a bridge from overall to specific indicators. Finally each strategic indicator was supported by a set of specific indicators, which represent quantifiable measures of performance. The programme development was enhanced by pilot plant studies, conducted over a 15 month period from January 1998 to March 1999. The result of all this work is compiled in the IAEA-TECDOC-1141, to be published shortly. This paper presents a summary of this IAEA TECDOC. It describes the operational safety performance indicator framework proposed and discusses the results of and lessons learned from the pilot studies. Despite the efforts described, it is clear that additional research is still necessary in areas such as plant-specific adaptation of proposed frameworks in order to suit individual data collection systems and plant characteristics, indicator selection, indicator definition, goal setting, action thresholds, analysis of trends, indicator display systems, analysis of overall safety performance (i.e., aggregation or combination of indicators), safety culture indicators, qualitative indicators, and use of additional indicators to address issues such as industrial safety attitude and performance, staff welfare, and environmental compliance. This is the rationale for a new IAEA Coordinated Research Project on 'Development and application of indicators to monitor NPP