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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 Finally, the process is defined by which the existing version of baseline PSA model is superseded by a new one.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. Perspectives of Living PSA in NPP Krsko

    International Nuclear Information System (INIS)

    Vrbanic, I.; Kastelan, M.

    1996-01-01

    Nuclear power plant Krsko has completed the Level 1/Level 2 Probabilistic Safety Analysis (PSA) for internal initiating events and is in the process of completing the same for the external initiators. The analysis completed up to now has provided a valuable insight into a plant risk profile. In NPP Krsko there is a plan to use the PSA model as a permanent tool for the risk based applications and incorporate it into a decision making process. In order to achieve this there is a need to permanently maintain the PSA model in a manner that it reflects both the plan configuration/design at a time point and the operational experience up to the time point. All the activities aimed toward keeping the PSA model up-to-dated in this sense are usually referred to as a Living PSA (LPSA) program. NPP Krsko is in the process of defining and proceduralizing a LPSA program that would be plant specific and based on known world practices. Further, in order to be suitable for risk based applications the PSA model must be flexible in a sense that modifications to the base case model may be done easily and requantifications performed quickly as to evaluate various conditions imposed by real or hypothetical situations. NPP Krsko PSA model has been based on licensing type software. The requirements specified above dictate the transfer of the overall model to an application oriented software of newer generation with larger capabilities. The transfer becomes a part of a mentioned ongoing effort aimed at establishing LPSA model and concept. The paper present this effort and the perspectives of LPSA concept and risk based applications in NPP Krsko. (author)

  3. Upgrade of Common Cause Failure Modelling of NPP Krsko PSA

    International Nuclear Information System (INIS)

    Vukovic, I.; Mikulicic, V.; Vrbanic, I.

    2006-01-01

    Over the last thirty years the probabilistic safety assessments (PSA) have been increasingly applied in technical engineering practice. Various failure modes of system of concern are mathematically and explicitly modelled by means of fault tree structure. Statistical independence of basic events from which the fault tree is built is not acceptable for an event category referred to as common cause failures (CCF). Based on overview of current international status of modelling of common cause failures in PSA several steps were made related to primary technical basis for methodology and data used for CCF model upgrade project in NPP Krsko (NEK) PSA. As a primary technical basis for methodological aspects of CCF modelling in Krsko PSA the following documents were considered: NUREG/CR-5485, NUREG/CR-4780, and Westinghouse Owners Group documents (WOG) WCAP-15674 and WCAP-15167. Use of these documents is supported by the most relevant guidelines and standards in the field, such as ASME PRA Standard and NRC Regulatory Guide 1.200. WCAP documents are in compliance with NUREG/CR-5485 and NUREG/CR-4780. Additionally, they provide WOG perspective on CCF modelling, which is important to consider since NEK follows WOG practice in resolving many generic and regulatory issues. It is, therefore, desirable that NEK CCF methodology and modelling is in general accordance with recommended WOG approaches. As a primary basis for CCF data needed to estimate CCF model parameters and their uncertainty, the main used documents were: NUREG/CR-5497, NUREG/CR-6268, WCAP-15167, and WCAP-16187. Use of NUREG/CR-5497 and NUREG/CR-6268 as a source of data for CCF parameter estimating is supported by the most relevant industry and regulatory PSA guides and standards currently existing in the field, including WOG. However, the WCAP document WCAP-16187 has provided a basis for CCF parameter values specific to Westinghouse PWR plants. Many of events from NRC / INEEL database were re-classified in WCAP

  4. 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)

  5. 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

  6. 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

  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. 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)

  9. 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)

  10. 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)

  11. Start-up of NPP Krsko; Pokusno obratovanje NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Spiler, J; Aralica, J [Nuklearna elektrana Krsko, Krsko (Yugoslavia)

    1984-07-01

    The report describes a review of start-up program and its realisation. There are also described some more significant start-up results with their evaluation. The most significant operation criteria are compared between NPP Krsko and other similar plants in the world. The comparison shows that after the first contractors and operation personnel efforts have been accomplished, our first nuclear power plant is a safe and reliable source of electric power. At the end there are listed NPP Krsko start-up recommendations and experience. (author)

  12. Krsko NPP Periodic Safety Review program

    International Nuclear Information System (INIS)

    Basic, I.; Spiler, J.; Novsak, M.

    2001-01-01

    The need for conducting a Periodic Safety Review for the Krsko NPP has been clearly recognized both by the NEK and the regulator (SNSA). The PSR would be highly desirable both in the light of current trends in safety oversight practices and because of many benefits it is capable to provide. On January 11, 2001 the SNSA issued a decision requesting the Krsko NPP to prepare a program and determine a schedule for the implementation of the program for 'Periodic Safety Review of NPP Krsko'. The program, which is required to be in accordance with the IAEA safety philosophy and with the EU practice, was submitted for the approval to the SNSA by the end of March 2001. The paper summarizes Krsko NPP Periodic Safety Review Program [1] including implemented SNSA and IAEA Expert Mission comments.(author)

  13. NPP Krsko natural circulation performance evaluation

    International Nuclear Information System (INIS)

    Segon, Velimir; Bajs, Tomislav; Frogheri, Monica

    1999-01-01

    The present document deals with an evaluation of the natural circulation performance of the Krsko nuclear power plant. Two calculation have been performed using the NPP Krsko nodalization (both similar to the LOBI A2-77 natural circulation experiment) - the first with the present steam generators at NPP Krsko (Westinghouse, 18% plugged), the second with the future steam generators (Siemens, 0% plugged). The results were evaluated using the natural circulation flow map derived in /1/, and were compared to evaluate the influence of the new steam generators on the natural circulation performance. (author)

  14. 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

  15. 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)

  16. Environmental impact of the NPP Krsko

    International Nuclear Information System (INIS)

    Novosel, N.

    1996-01-01

    The Ministry of Economic Affairs has for six years now been monitoring the operation of the Krsko NPP (NEK) and its impact on the environment. A bulletin titled 'NEK - Energy and Environment' is being issued every three months. It contains information on operation of the Krsko NPP for the previous three months, a graph of duration of temperature increase of water in the Sava river (delta T) in that period, an assessment of the radiological impact of Krsko NPP on the environment through an equivalent dose cumulatively throughout the calendar year, and a short current text related to Krsko NPP. The Ministry of Economic Affairs organizes a press conference on every issue of the bulletin, as an attempt of introducing this subject to the media and to the public. This paper contains a review of information given in the NEK bulletin from 1990 to 1995 with a special emphasis on the contribution of the Krsko NPP to the artificially caused radiation on the border between the Republic of croatia and the Republic of Slovenia. (author)

  17. 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)

  18. Revision of Krsko NPP Quality Assurance Plan

    International Nuclear Information System (INIS)

    Biscan, R.; Fifnja, I.; Kavsek, D.

    2012-01-01

    International standards from nuclear power plant operation area are being frequently upgraded and revised in accordance with the continuous improvement philosophy. This philosophy applies also to the area of Quality Assurance, which has also undergone significant improvement since the early 1950s. Besides just nuclear industry, there are also other international quality standards that are being continuously developed and revised, bringing needs for upgrades also in the nuclear application. Since the beginning of Krsko NPP construction, the overall Quality Assurance program and its applicable procedures were in place to assure that all planned and systematic actions necessary to provide adequate confidence that an item or service will satisfy given requirements to quality, are in place. The overall requirements for quality as one of the major objectives for Krsko NPP operation are also set forth in the Updated Safety Analyses Report, the document that serves as a base for operating license. During more than 30 years of Krsko NPP operation, the quality requirements and related documents were revised and upgraded in several attempts. The latest revision 6 of QD-1, Quality Assurance Plan was issued during the year 2011. The bases for the revision were: Changes of the Slovenian regulatory requirements (ZVISJV, JV5, JV9?), Changes of Krsko NPP licensing documents (USAR section 13?), SNSA inspection requirements, Changes of international standards (IAEA, ISO?), Conclusions of first PSR, Implementation of ISO standards in Krsko NPP (ISO14001, ISO17025), Changes of plant procedures, etc. One of the most obvious changes was the enlargement of the QA Plan scope to cover interdisciplinary areas defined in the plant management program MD-1, such as Safety culture, Self-assessment, Human performance, Industrial Safety etc. The attachment of the QA Plan defining relationships between certain standards was also updated to provide matrix for better correlation of requirements of

  19. 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

  20. 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)

  1. 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)

  2. Transition cycle fuel management problems of NPP Krsko; Problemi gospodarenje gorivom u prijelaznim ciklusima NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B [Institut ' Rudjer Boskovic' , Zagreb (Yugoslavia); Pevec, D [Elektrotehnicki fakultet, Zagreb (Yugoslavia); Smuc, T; Urli, N [Institut ' Rudjer Boskovic' , Zagreb (Yugoslavia)

    1989-07-01

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

  3. Transition cycle fuel management problems of NPP Krsko

    International Nuclear Information System (INIS)

    Petrovic, B.; Pevec, D.; Smuc, T.; Urli, N.

    1989-01-01

    Transition cycle fuel management problems are described and illustrated using results and experience attained during core reload design of NPP Krsko. Improved version of computer code package PSU-LEOPARD/Mcrac is successfully applied to NPP Krsko loading pattern design. (author)

  4. NPP Krsko Severe Accident Management Guidelines Implementation

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.; Bilic-Zabric, T.; Spiler, J.

    2002-01-01

    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. The USA NRC has indicated that the development of a licensee plant specific accident management program will be required in order to close out the severe accident regulatory issue (Ref. SECY-88-147). Generic Letter 88-20 ties the Accident management Program to IPE for each plant. The SECY-89-012 defines those actions taken during the course of an accident by the plant operating and technical staff to: 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. The subject of this paper is to document the severe accident management activities, which resulted in a plant specific Severe Accident Management Guidelines implementation. They have been developed based on the Krsko IPE (Individual Plant Examination) insights, Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidances) and plant specific documents developed within this effort. Among the required plant specific actions the following are the most important ones: Identification and documentation of those Krsko plant specific severe accident management features (which also resulted from the IPE investigations). The development of the Krsko plant specific background documents (Severe Accident Plant Specific Strategies and SAMG Setpoint Calculation). Also, paper discusses effort done in the areas of NPP Krsko SAMG review (internal and external ), validation on Krsko Full Scope Simulator (Severe Accident sequences are simulated by MAAP4 in real time) and world 1st IAEA Review of Accident Management Programmes (RAMP). (author)

  5. NPP Krsko secondary side analysis

    International Nuclear Information System (INIS)

    Fabijan, Lj.

    1987-01-01

    The purpose of this work is to analyze secondary side thermohydraulics response on steam generator tube plugging in order to ensure nominal NPP power. We had established that the additional opening of the governing valve No. 3 and 4 can compensate pressure drop caused by steam generator tube plugging. Two main steam flows with four governing valves were simulated. Steam expansion in turbine and feed water system was modeled separately. All important process point and steam moisture changes impact on nominal NPP power were analysed. (author)

  6. Tritium liquid effluents from the Krsko NPP

    International Nuclear Information System (INIS)

    Savli, S.; Krizman, M.; Nemec, T.; Cindro, M.; Stritar, A.; Vokal Nemec, B.; Janzekovic, H.

    2007-01-01

    In the past, 12-months' fuel cycles in the Krsko NPP had not caused any problems regarding compliance with its Technical Specifications and license limits on liquid tritium releases (20 TBq/year, 8 TBq/three months). The first 18-months' fuel cycle, which was introduced in 2004, required fuel with higher enrichment, higher boron concentration in the primary coolant and more fuel rods with burnable poisons. In 2005, the NPP operated without refueling outage for the whole year and produced the highest amount of energy so far. Due to these facts and a few unplanned shutdowns and power reductions, production of tritium and releases increased strongly in 2005. As a result, the Krsko NPP hardly succeeded to stay within regulatory limits on tritium releases. However, the three-months' limit was exceeded in the first quarter of 2006. On the basis of conclusions acquired from the SNSA's study and practice of other European countries the SNSA considerably increased the annual limit of permitted liquid tritium releases (from 20 TBq to 45 TBq) and abolished the three-months' limit. At the same time, the SNSA reduced the limit of fission and activation products by halves. (author)

  7. 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

  8. 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)

  9. 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)

  10. 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)

  11. Equipment Reliability Process in Krsko NPP

    International Nuclear Information System (INIS)

    Gluhak, M.

    2016-01-01

    To ensure long-term safe and reliable plant operation, equipment operability and availability must also be ensured by setting a group of processes to be established within the nuclear power plant. Equipment reliability process represents the integration and coordination of important equipment reliability activities into one process, which enables equipment performance and condition monitoring, preventive maintenance activities development, implementation and optimization, continuous improvement of the processes and long term planning. The initiative for introducing systematic approach for equipment reliability assuring came from US nuclear industry guided by INPO (Institute of Nuclear Power Operations) and by participation of several US nuclear utilities. As a result of the initiative, first edition of INPO document AP-913, 'Equipment Reliability Process Description' was issued and it became a basic document for implementation of equipment reliability process for the whole nuclear industry. The scope of equipment reliability process in Krsko NPP consists of following programs: equipment criticality classification, preventive maintenance program, corrective action program, system health reports and long-term investment plan. By implementation, supervision and continuous improvement of those programs, guided by more than thirty years of operating experience, Krsko NPP will continue to be on a track of safe and reliable operation until the end of prolonged life time. (author).

  12. 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)

  13. 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)

  14. 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

  15. Environmental qualification program for Krsko NPP

    International Nuclear Information System (INIS)

    Cerjak, J.; Klenovsek, P.; Pavsek, J.; Freeland, K.R.; Spalj, S.

    1998-01-01

    NEK plant components, including those critical to safe plant operation, deteriorate and wear over service life due to the effects of aging and harsh environmental conditions. Since the plant environment is a source of common-cause failures, an Environmental Qualification (EQ) program is required to ensure and demonstrate the ability of safety-related equipment to perform its design safety function during a design-basis (DBE), even after aging over its service life in the plant. EQ is a requirement for plants licensed by the US NRC, in accordance with 10 CFR 50.49, Regulatory Guide 1.89, NUREG-0588, and IEEE-323. This paper presents the current EQ Program status at Krsko NPP.(author)

  16. 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)

  17. Corrective action program at Krsko NPP

    Energy Technology Data Exchange (ETDEWEB)

    Skaler, F; Divjak, G; Kavsek, D [NPP Krsko, Krsko (Slovenia)

    2004-07-01

    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)

  18. NPP Krsko Periodic Safety Review action plan

    International Nuclear Information System (INIS)

    Bilic Zabric, T.

    2006-01-01

    In the current, internationally accepted, safety philosophy Periodic Safety Reviews (PSRs) are comprehensive reviews aimed at the verification that an operating NPP remains safe when judged against current safety objectives and practices and that adequate arrangements are in place to maintain an acceptable level of safety. These reviews are complementary to the routine and special safety reviews. They are long time-scale reviews intended to deal with the cumulative effects of plant ageing, modifications, operating experience and technical developments, which are not so easily comprehended over the shorter time-scale routine of safety reviews. The review was completed in 2005 and the next period will see the implementation of the action plan including some plant upgrades. The action plan lists issues that should be implemented at NPP Krsko together with associated milestones. The milestones were assumed based on best estimate resource availability and their ends can be potentially floated. In some cases, multiple corrective measures may be postulated to provide resolution for a given safety issue. The Slovenian Nuclear Safety Administration by decree approved the first periodic safety review and the implementation plan of activities arising from it. The entire implementation plan must be carried out by 15 October 2010. Report on the second periodic safety review must be submitted by the NEK not later than 15 December 2013. (author)

  19. 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)

  20. Intranet portal at the Krsko NPP, Slovenia

    International Nuclear Information System (INIS)

    2016-01-01

    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

  1. NPP Krsko full scope simulator verification and qualification

    International Nuclear Information System (INIS)

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

    1998-01-01

    NPP Krsko (NEK) will, as a part of the Modernization plan, obtain also Krsko Full Scope Simulator (KFSS). Contract has been awarded to CAE Electronics for the design, construction and integration. KFSS will support in real time, the training for the complete range of operation, which can be performed from the main control room and some selected plant areas (remote shutdown panels, etc). Based on the lessons learned on development of NPP Krsko Basic Principle Simulator we decided for active approach. That means that NPP Krsko personnel will be heavily involved into all phases of KFSS development and testing. Since NPP Krsko is going to replace the existing steam generators, raise the nominal power and perform necessary modifications to support the power uprate, it was decided that the development of the KFSS will be conducted in two steps: 1. Development of the models as well as all the hardware interface in the MCR for the existing plant Cycle 15 and then, 2. Models and hardware will be modified, added or replaced as needed to take into account the steam generator replacement and plant uprate projects. In spite of the fact that the simulator will be used for the training of the plant operators for the uprated conditions and with new steam generators, the upper described approach was selected since we want to be sure that the models will at the beginning adequately simulate the existing plant. For the existing conditions we have available reference data for different plant conditions, as well as data for different plant transients. By verifying that simulator will be able adequately simulate the existing conditions the level of confidence for the uprated simulator will be much higher. This is of special importance since it will support initial training for modernized plant conditions. In this paper the plan for verification and qualification of KFSS as well as the amount of the work needed on NPP Krsko side to develop the test acceptance criteria will be presented.(author)

  2. Presentation of common cause failures in fault tree structure of Krsko PSA : an historical overview

    International Nuclear Information System (INIS)

    Vrbanic, I.; Kosutic, I.; Vukovic, I.; Simic, Z.

    2003-01-01

    Failure of multiple components due to a common cause represents one of the most important issues in evaluation of system reliability or unavailability. The frequency of such events has relatively low expectancy, when compared to random failures, which affect individual components. However, in many cases the consequence is a direct loss of safety system or mitigative safety function. For this reason, the modeling of a common cause failure (CCF) and its presentation in fault tree structure is of the uttermost importance in probabilistic safety analyses (PSA). During the past decade, PSA model of Krsko NPP has undergone many small changes and a couple of major ones in fulfilling its basic purpose, which was serving as a tool for providing an appropriate information on the risk associated with actual plant design and operation. All changes to Krsko PSA model were undertaken in order to make it a better tool and / or to make it represent the plant in more accurate manner. The paper provides an overview of changes in CCF modeling in the fault tree structure from the initial PSA model development till present. (author)

  3. 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)

  4. 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)

  5. NPP Krsko core calculations to improve operational safety

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  6. 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

  7. Determination of the NPP Krsko reactor core safety limits using the COBRA-III-C code

    International Nuclear Information System (INIS)

    Lajtman, S.; Feretic, D.; Debrecin, N.

    1989-01-01

    This paper presents the NPP Krsko reactor core safety limits determined by the COBRA-III-C code, along with the methodology used. The reactor core safety limits determination is a part of reactor protection limits procedure. The results obtained were compared to safety limits presented in NPP Krsko FSAR. The COBRA-III-C NPP Krsko design core steady state thermal hydraulics calculation, used as the basis for the safety limits calculation, is presented as well. (author)

  8. Axial blanket enrichment optimization of the NPP Krsko fuel

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2001-01-01

    In this paper optimal axial blanket enrichment of the NPP Krsko fuel is investigated. Since the optimization is dictated by economic categories that can significantly vary in time, two step approach is applied. In the first step simple relationship between the equivalent change in enrichment of axial blankets and central fuel region is established. The relationship is afterwards processed with economic criteria and constraints to obtain optimal axial blanket enrichment. In the analysis realistic NPP Krsko conditions are considered. Except for the fuel enrichment all other fuel characteristics are the same as in the fuel used in the few most recent cycles. A typical reload cycle after the plant power uprate is examined. Analysis has shown that the current blanket enrichment is close to the optimal. Blanket enrichment reduction results in an approximately 100 000 US$ savings per fuel cycle.(author)

  9. Equipment performance monitoring in NPP Krsko (Summarized system health report)

    International Nuclear Information System (INIS)

    Djetelic, N.; Cicvaric, D.

    2004-01-01

    Management common goal is safe, reliable, effective, acceptable to public and conservative/cautious operation of NPP Krsko. A set of programs, including Corrective Action Program, Performance Indicators, Operating Experience, Self Assessment and System Health Report, is developed to assist NPP Krsko management in fulfilling those goals. System Health Report is a tool that management can use to quickly assess how selected systems are performing, to determine where additional management attention is required and to determine if appropriate corrective actions have been established. Summarized System Health Report is developed for management's quick overview of systems status, important system malfunctions and problems as well as major changes from previous assessment period. Summarized Report contains nine sections: status difference including brief explanation, selected performance indicators, new equipment problems, functional failures, important problem analyses, action plan for systems with Potential Danger (RED) status, maintenance rule status overview and systems availability (planned and unplanned).(author)

  10. 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)

  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. Use of the deterministic safety analyses in support to the NPP Krsko modification

    International Nuclear Information System (INIS)

    Feretic, D.; Cavlina, N.; Debrecin, N.; Grgic, D.; Bajs, T.; Spalj, S.

    2004-01-01

    The ultimate goal of the safety analysis is to verify that Nuclear Power Plant (NPP) meets safety and operational requirements. To this aim it is necessary to demonstrate that plant safety has not been deteriorated in the case of the modifications to the plant Systems, Structures and Components (SSC) or changes to the plant procedures. In addition, safety analyses are needed in the case of reassessment of an existing plant. The reasons for reassessment may be different, e.g. due to the changes in the methodology and assumptions used in the original design, if the original design basis or acceptance criteria may no longer be adequate, if the safety analysis tools used may have been superseded by more sophisticated methods or if the original design basis may no longer be met. The operation of the NPP Krsko has experienced numerous changes from the original design for the majority of the reasons that have been mentioned before. On the other side, the application of the large best-estimate thermalhydraulic codes has evolved to the wide spread support in the operation of the NPP: compliance with the regulatory goals, support to the PSA studies, analysis of the operational transients, plant modifications studies, equipment qualification, training of the operators, preparation of the operating procedures, etc. This trend has been followed at the Faculty of Electrical Engineering Zagreb (FER) and applied to the on-going needs due to the modifications and changes at NPP Krsko. In this paper, an overview of the deterministic safety analyses performed at FER in the support to the NPP Krsko modifications and changes is presented.(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. NPP Krsko on-line low pressure containment tightness monitoring implementation

    International Nuclear Information System (INIS)

    Dudas, M.; Basic, I.

    2004-01-01

    Containment Integrated Leak Rate Test (CILRT) 1999 in NPP Krsko was completely performed following regulation of 10CFR50 Appendix J Option A and ANSI/ANS 56.8-1987 at a design pressure (3.15 kp/cm2). In 2001 NPP Krsko proposed to Slovenian Nuclear Safety Administration (SNSA) the Technical Specification (TS) and Updated Safety Analysis Report (USAR) changes that describe implementation of new test intervals for Type A, B and C tests according to 10CFR50, Appendix J, Option B. After the positive final independent review of proposed changes by Authorized Institution, NPP Krsko received the License Amendment requiring from NPP Krsko to define technical solution for surveillance of containment tightness between two 10-years CILRT. This paper intends to discuss proposed methods by NPP Krsko, test equipment, performed measurements in 2004, associated analyses and evaluation.(author)

  16. Containment Leakage Rate Testing Program in NPP Krsko

    International Nuclear Information System (INIS)

    Dudas, M.; Heruc, Z.

    2002-01-01

    NPP Krsko adopted new regulations for testing of the reactor building containment as stipulated by 10CFR50 (Code of Federal Regulation) Appendix J, Option B instead of the previous requirement 10CFR50 Appendix J now renamed to 10CFR50 Appendix J, Option A. In the USA a thorough analyses of nuclear power plants reactor building containment testing was conducted. As part of these analyses the test results obtained from testing of various reactor-building containments in the last ten years were reviewed. It was concluded that it would be meaningful to, based on test results historical data, reconsider possibility of redefining testing intervals. The official proposal of such approach was reviewed and approved by the NRC and published in September of 1995 in the FR Vol.60 No.186. Based on directions from 10CFR50 Appendix J, Option B, the new criteria for definition of test intervals were created. Criteria were based upon past performance during testing (Performance-based Requirements) and safety impact. At NPP Krsko, the analyses of the Reactor Building Containment. Integrity Test results was performed . This included test results of the Containment Integrated Leak Rate Testing (CILRT or Type A tests), Containment Isolation Valves Local Leak Rater Tests (Type C tests) and Mechanical and Electrical Penetrations Local Leak Rate tests (Type B tests). In accordance with instructions from NEI 94-01 and based on analyses of test results, NPP Krsko created Containment Leakage Rate Testing Program with the purpose to establish the performance-based definition of test intervals, inspection scope, trending and reporting. Equally, the program gives instructions how to evaluate test results and how to deal with the containment penetration or isolation valve repair contingency. All changes caused with transition from Option A to Option B are marginal to public safety. (author)

  17. Heat and radiation analysis of NPP Krsko irradiated fuel

    International Nuclear Information System (INIS)

    Lalovic, M.

    1986-01-01

    Radioactive and heat potential for irradiated fuel in the region 2 with burnup of 13400 MWd/tHM, and in the region 4A with burnup of 9360 MWd/tHM for NPP KRSKO, was calculated. Computer code KORIGEN (Karlsruhe Oak Ridge Isotope Generation and Depletion Code) was used. The aspects of radiation (mainly gamma and neutrons) and of heat production was considered with respect to their impact on fuel handing and waste management. Isotopic concentrations for irradiated fuel was calculated and compared with Westinghouse data. (author)

  18. 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

  19. Reracking Possibilities of the NPP Krsko Spent Fuel Pool

    International Nuclear Information System (INIS)

    Bace, M.; Pevec, D.; Smuc, T.

    1998-01-01

    Using the SCALE-4 code package reracking possibilities of the NPP Krsko spent fuel pool were analyzed. Two cases were considered: the first case assuming the 40 years lifetime of the plant, and the second case assuming the 50 years lifetime of the plant. It was shown that it is possible to design the additional racks in free space of the spent fuel pool with the sufficient total capacity to store all the spent fuel generated during the 40 years lifetime of the plant. In the case of 50 years plant lifetime, completely new racks (capacity of 1890 spent fuel assemblies), containing 4mm boral in storage cell walls, were proposed for the NPP Krsko spent fuel pool. The effective multiplication factor of the spent fuel pool fully loaded with new racks containing spent fuel assemblies of initial enrichment 4.3 w/o, burned to 40 GWd/tU and cooled 2 years is lower than the value required by standard. It showed the possibility of the safe disposal of all spent fuel accumulated during more than 50 years lifetime of the plant. (author)

  20. Reactor Coolant Pump Motor Maintenance Experience in Krsko NPP

    International Nuclear Information System (INIS)

    Vukovic, J.; Besirevic, A.; Boljat, Z.

    2016-01-01

    After thirty years of service as well as maintenance in Krsko NPP both original Reactor Coolant Pump (RCP) motors are remanufactured by original vendor Westinghouse and a new one was purchased. Design function of the RCP motor is to drive Reactor Coolant Pump and for coast-down feature during Design Basis Accident. This paper will give a view on maintenance issues of RCP motor during the thirty years of service and maintenance in Krsko NPP to be kept functionally operational. During the processes of remanufacturing inspection and disassembly it was made possible to get a deeper perspective in the motor condition and the wear or fatigue of the motor parts. Parameters like bearing & winding temperature, absolute and relative vibration greatly affect motor operation if not kept inside design margins. Rotational speed causes heat generation at the bearings which is then associated with oil temperatures and as a consequence bearing temperatures. That is why the most critical parts of the motor are the components of upper and lower bearing assembly. The condition of motor stator and rotor assembly technical characteristics shall be explained with respect to influence of demanding environmental conditions that the motor is exposed. Assessment shall be made how does the wear of critical RCP motor parts can influence reliable performance of the motor if not maintained in proper way. Information on upgrades that were done on RCP motor shall be shared: Oil Spillage Protection System (OSPS), Stator upgrades, Dynamic Port, etc. (author).

  1. 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)

  2. Croatian Capacity for Management of the NPP Krsko Radioactive Waste

    International Nuclear Information System (INIS)

    Levanat, Ivica; Lokner, Vladimir

    2014-01-01

    Together with Slovenia, Croatia is responsible for management of the NPP Krsko radioactive waste and spent fuel. So far, no firm agreement on specific solutions has been reached between the two countries. On the contrary, all activities related to revision and development of the joint Program of the NPP decommissioning and spent fuel and radioactive waste management were discontinued several years ago. Unless Slovenia and Croatia definitely agree on joint solutions in the meantime, Croatia will have to begin transfer of one half of the NPP Krsko spent fuel and radioactive waste into its territory in about nine years. Presently, however, Croatian capacities for such an undertaking are seriously inadequate in several respects, and they are not developing in any promising way. For no rational reason, this state of the Croatian capacities has been maintained (recently even deteriorating in some respects) for at least a decade, disregarding also for at least two years the explicit requirements of the EU Directive 2011/70/EURATOM aimed at establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste. In fall of 2012, Fund for financing the NPP Krsko decommissioning and spent fuel and radioactive waste disposal was appointed as the Croatian expert organization for revision and development of the program of these same activities it had originally been supposed only to finance. In 2013 the Fund expanded its activities to a low-profile attempt at revitalization of the Croatian radioactive waste repository project, although the Fund is not yet properly capacitated for either of these tasks. The above is hardly in compliance with the Directive requirements, such as to establish 'a national legislative, regulatory and organisational framework' 'that allocates responsibility... between relevant competent bodies'. The lack of competent experts in the field appears to affect the quality of Croatian legislative

  3. 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

  4. 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

  5. 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)

  6. Criticality safety analysis of the NPP Krsko storage racks

    International Nuclear Information System (INIS)

    Kromar, M.; Kurincic, B.

    2002-01-01

    NPP Krsko is going to increase the capacity of the spent fuel storage pool by replacement of the existing racks with high-density racks. This will be the second reracking campaign since 1983 when storage was increased from 180 to 828 storage locations. The pool capacity will increase from 828 to 1694 with partial reracking by the spring 2003. The installed capacity will be sufficient for the current design plant lifetime. Complete reracking of the spent fuel pool will additionally increase capacity to 2321 storage locations. The design, rack manufacturing and installation has been awarded to the Framatome ANP GmbH. Burnup credit methodology, which was approved by the Slovenian Nuclear Safety Administration in previous licensing of existing racks, will be again implemented in the licensing process with the recent methodology improvements. Specific steps of the criticality safety analysis and representative results are presented in the paper.(author)

  7. Determination of source term for Krsko NPP extended fuel cycle

    International Nuclear Information System (INIS)

    Nemec, T.; Persic, A.; Zagar, T.; Zefran, B.

    2004-01-01

    The activity and composition of the potential radioactive releases (source term) is important in the decision making about off-site emergency measures in case of a release into environment. Power uprate of Krsko NPP during modernization in 2000 as well as changing of the fuel type and the core design have influenced the source term value. In 2003 a project of 'Jozef Stefan' Institute and Slovenian nuclear safety administration determined a plantspecific source term for new conditions of fuel type and burnup for extended fuel cycle. Calculations of activity and isotopic composition of the core have been performed with ORIGEN-ARP program. Results showed that the core activity for extended 15 months fuel cycle is slightly lower than for the 12 months cycles, mainly due to larger share of fresh fuel. (author)

  8. Trending of Events - Practical Use at Krsko NPP

    International Nuclear Information System (INIS)

    Skaler, F.; Bach, B.

    2008-01-01

    Krsko NPP (NEK) in 2004 developed a new application, a corrective action program (CAP) software, to acquire all events within the same environment. In 2006 NEK started to code all acquired records of the events. In 2007 NEK developed within CAP application new tools for analyzing the events. These new tools are common cause analyses Root cause, Direct causes, Processes, and Equipment failures. The new tools are user friendly and easy to use. With these new tools the useful queries are generated easily. The results are presented in graphs, which have drill down options all the way down to the reported event itself. With these new tools many common cause analyses can be done within seconds, which in the past took hours or even days. Also the department leads can easily perform common cause analyses within their departments or processes, they can compare the different periods or projects and they can look for different type of causes. NEK next step is a learning module for the department leads. They have to learn how to use these tools and to understand the results correctly. NEK has to establish the process, where these analyses can be formalized, checked and the most effective correction actions to prevent the events can be taken. Also all the process can be formalized. The presentation will explain the theory of trending the minor events and practical implementation at Krsko NPP. The presentation will also present the new tools in CAP application with some very interesting examples of common cause analyses. The presentation will explain possibilities for the future process at NEK to fully implement these tools in order to prevent the events.(author)

  9. 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

  10. 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)

  11. NPP Krsko: LILW Repository or Long Term Storage

    International Nuclear Information System (INIS)

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

    2008-01-01

    Construction of the facilities for LILW and SF management, as planned in Decommissioning and LILW and SF management program for NPP Krsko, would be a rather expensive and challenging project for such a small nuclear program. In order to accommodate waste arising from a single nuclear power plant, one LILW repository should be constructed before the end of the NPP operation, then one SF dry storage, and finally one geological repository. This requires relatively urgent identification within Slovenian/Croatian territory of three locations that meet the criteria for establishment of such facilities and are acceptable to the local communities. There are very few such potential locations. The siting process for the first of the three facilities is well under way in Slovenia, because the country wants to have its LILW repository in operation by the year 2013. In order to facilitate public acceptance, Slovenian government has introduced financial incentives to local communities for the repository construction and operation. These 'compensations for limited land use' may significantly increase the overall costs of disposal if the repository is in operation for a long period. In the recent years, however, a possibility of long term storage (LTS) is gaining an increased attention in the waste management community, and has already been introduced e.g. in the Netherlands. It is a particularly viable option for limited waste quantities. Disposal remains the final solution, but present technologies have made possible a relatively inexpensive storage up to about hundred years, which can accommodate LILW, HLW and SF from nuclear programs as well as research reactor waste and NORM. Such storage would be a safe and simple temporary solution, encompassing all immediate and near future waste management needs. In addition, it would increase flexibility and reduce financing requirements for the final waste disposal: providing additional time for reduction of radiation emission and heat

  12. Developing Effective Corrective Action Plan in Krsko NPP

    International Nuclear Information System (INIS)

    Bach, Bruno; Cizmek, Rudi; Bozin, Bojan

    2014-01-01

    Experience shows that many events could have been prevented if lessons had been learned from previous incidents. Event reporting thus has become an increasingly important aspect of the operation and regulation of all safety-related and public health industries. Different industries such as aeronautics, chemicals, transport and of course nuclear depend on Operating Experience (OE) feedback programs to provide lessons learned about safety. The information available under an OE programme for these organizations comprises internal event reports and analysis and external operating experience including reports on low level events and near misses and other relevant operating performance information. The worldwide OE programme (such as WANO OE) in nuclear power plants provides opportunity to learn from events at other plants. In particular, it alerts plants to mistakes or events that have occurred at other nuclear power plants and enables them to take corrective actions to prevent similar occurrences at their own plant. The intent of the effective and efficient OE program is therefore to improve personnel/plant safety, reliability and commercial performance of the operating nuclear power plants. Such a programme ensures that operating experience is analysed, events important to safety are reviewed in depth, lessons learned are disseminated to the staff and to the relevant national and international organizations and corrective actions are effectively implemented. 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 its own OE program by using the most relevant INPO/WANO/IAEA guidelines as well as its own knowledge, skills an operating practice. The OE is a part of the Corrective Action Program (CAP), which is among top management programs. The purpose of this article is to present a part of the Krško NPP

  13. Revision 2 of the NPP Krsko Decommissioning Program Is Stalled

    International Nuclear Information System (INIS)

    Levanat, I.; Lokner, V.; Rapic, A.; Zeleznik, N.; Kralj, M.

    2012-01-01

    Revision 2 of the joint Slovenian-Croatian Program of NPP Krsko Decommissioning and SF andLILW Disposal was scheduled to be finished and formally approved by the end of 2009, in accordance with the bilateral Agreement on the NPP. Slightly behind the schedule, the Project team completed the entire document during spring of 2010, and in June 2010 drafted a proposal for a peer review of the Program by a dedicated IAEA expert mission. This procedure was agreed upon at the last session (May 2010) of the Intergovernmental Commission for implementation of the Agreement, when the Commission was acquainted with the five scenarios of the Revision 2 and with the estimates of their costs/financing. It was expected that the peer review would be performed soon, and that formal adoption of the Revision 2 would follow. Although in this process of approval some decisions remained to be made by the stakeholders, the Project team did select and recommend one scenario to be used for costing purposes, in order to ensure that most necessary corrections in Program financing would be timely adopted. However, the planned IAEA review was cancelled by the Advisory board, the body nominated by the Commission ''to supervise the activities and resolve the issues raised by the Project team''. By this cancellation, the process of Program revision was effectively stalled, because the Advisory board could not clearly define further course of action: differing views between the Slovenian and the Croatian part of the Advisory board appeared, in particular regarding the set of Program scenarios and regarding the appropriateness of the Revision 2 document for the IAEA review; nonetheless, the Advisory board sent to the Project team a compilation of requests to modify Revision 2 document. The Project team determined that some minor requests were easy to fulfill, but other modifications could only be carried out after changes in the boundary conditions (approved by the Commission), or changes in national

  14. 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)

  15. 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

  16. Revision 2 of the Program of NPP Krsko Decommissioning and SF and LILW Disposal

    International Nuclear Information System (INIS)

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

    2010-01-01

    First joint Slovenian-Croatian Program of NPP Krsko Decommissioning and SF and LILW Disposal (DP) was completed in 2004 and formally adopted in 2005. As bilateral agreement on the NPP requires periodic revisions at least each 5 years, revision 2 of DP was started in September 2008, with the purpose to incorporate relevant developments since the 1st revision, to improve the level of details and reliability of DP, and to propose updated and more accurate cost estimates and appropriate financing models. In the first phase of the revision, new supporting studies for DP modules were prepared. Among these studies, the most demanding was the NPP Krsko specific Preliminary Decommissioning Plan (PDP), complying with the IAEA-recommended format, which included development of the NPP decommissioning inventory database. For upgrade of SF management, new and more detailed descriptions with improved cost estimates were prepared. Update of LILW disposal concept was based on new developments and projects prepared for the Slovenian repository. In the second phase of the revision, integrated DP scenarios were formulated and analyzed. They integrate NPP decommissioning together with RW and SF management/disposal into rationally inter-related sequences. Boundary conditions for this revision required: (a) that the reference scenario from the previous revision should be re-examined, with appropriate variations or new alternatives; (b) that the option of the NPP Krsko life extension should also be included; and (c) that the possibility of diverging interests of the contracting parties should also be analyzed (i.e. waste division and separate management). Finally, scenario evaluation is intended to compare the analyzed scenarios taking into account both their feasibility and estimated costs. It should provide the basis for determining future financing of DP, namely the annuities to be paid by the NPP Krsko owners into the national decommissioning funds.(author).

  17. External Events PSA for the Paks NPP

    International Nuclear Information System (INIS)

    Bareith, Attila; Karsa, Zoltan; Siklossy, Tamas; Vida, Zoltan

    2014-01-01

    Initially, probabilistic safety assessment of external events was limited to the analysis of earthquakes for the Paks Nuclear Power Plant in Hungary. The level 1 seismic PSA was completed in 2002 showing a significant contribution of seismic failures to core damage risk. Although other external events of natural origin had previously been screened out from detailed plant PSA mostly on the basis of event frequencies, a review of recent experience on extreme weather phenomena made during the periodic safety review of the plant led to the initiation of PSA for external events other than earthquakes in 2009. In the meantime, the accident of the Fukushima Dai-ichi Nuclear Power Plant confirmed further the importance of such an analysis. The external event PSA for the Paks plant followed the commonly known steps: selection and screening of external hazards, hazard assessment for screened-in external events, analysis of plant response and fragility, PSA model development, and risk quantification and interpretation of results. As a result of event selection and screening the following weather related external hazards were subject to detailed analysis: extreme wind, extreme rainfall (precipitation), extreme snow, extremely high and extremely low temperatures, lightning, frost and ice formation. The analysis proved to be a significant challenge due to scarcity of data, lack of knowledge, as well as limitations of existing PSA methodologies. This paper presents an overview of the external events PSA performed for the Paks NPP. Important methodological aspects are summarised. Key analysis findings and unresolved issues that need further elaboration are highlighted. Development of external events PSA for the Paks NPP was completed by the end of 2012. The analysis followed the commonly known steps: selection and screening of external hazards, hazard assessment for screened-in external events, analysis of plant response and fragility, PSA model development, and risk

  18. 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).

  19. 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)

  20. Analysis of the network protection devices action connected with the Krsko NPP operation safety

    International Nuclear Information System (INIS)

    Omahen, P.; Struc, S.

    1995-01-01

    The example of failure that took place in Zagreb on 30 March, 1995, has been investigated in order to obtain appropriate answers to the unexpected electric power system (ESP) response and consecutive outage of the Krsko nuclear power plant (NPP). The analysis has been made of particular operating and stability conditions, related to the functioning of distance relays line protection. Using a consequent simulation model of the EPS (European interconnection) under specified fault condition, the research of EPS operation (relays action) and stability has been done. The appropriate simulation results have been compared to the available measured data which had been collected by the SOREL on-line data acquisition system. In the end, a practise proposal for measures to be taken with the target of achieving the foreseen and expected operation of the EPS equipment that effects operation of the Krsko NPP is given. (author)

  1. 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)

  2. Risk assessment of the Krsko NPP normal operation on the public in the Zagreb area

    International Nuclear Information System (INIS)

    Skanata, D.; Pevec, D.

    1994-01-01

    In the paper, a very short review of the ICRP recommendations related to effective dose limitation for the workers as well as for the public, is given. Particular attention is paid to the risk coefficients values. In the short manner, description of the applied methodology and achieved result related to the risk assessment of the Krsko NPP normal operation on the public in the Zagreb Area, are given too. The mentioned assessment was performed as an activity within the Case Study Zagreb Project developed under the Inter-Agency Programme on the Assessment and Management of Health and Environmental Risks from Energy and Other Complex Industrial Systems (UNEP, WHO, IAEA and UNIDO). Making comparison between assessment risk of the Krsko NPP normal operation and risks to which inhabitants of the City of Zagreb are normally exposed, it is concluded that living in the vicinity of such nuclear power plant as it is Krsko NPP (PWR, 664 MWe), is not so risky as risky are some everyday activities

  3. 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)

  4. 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)

  5. 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)

  6. Analysis of steam generator plugging on core thermohydraulic performance of NPP Krsko; Analiza vpliva cepljenja cevi v upravljaniku na termohidravliko sredice JE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Kostadinov, V; Petelin, S; Sarler, B [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1988-07-01

    Nuclear safety analysis of NPP Krsko core operating at full power with 4% steam generator tubes plugged have been performed. Influence of individual parameters on core thermohydraulic performance have been evaluated. Using COBRA-III-C computer code we have analysed a core design (evaluation) model. The DNBR change was calculated as a consequence of 4% plugging. The influence of thermohydraulic parameters change on DNBR was analysed. (author)

  7. 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

  8. Integrated use of Primavera and ORAM codes in outage 1999 at NPP Krsko

    International Nuclear Information System (INIS)

    Krajnc, J.; Skaler, F.; Basic, I.; Kocnar, R.

    1999-01-01

    The paper deals with the following postulated main goals of outage scheduling with Primavera tool at Krsko NPP: planning and controlling of resources (people, equipment, locations, sources), controlling the safety aspects of an outage and assuring defense-in-depth philosophy (through integrated safety assessment by ORAM code), diversity use of the plan during preparations period and outage progress (MCB, work leaders, management, planning Dept., subcontractors, support, etc.), allowing for optimization of outage duration. A snapshot in Primavera of what actually happened in outage 1999, lessons learned and a new work template is the scope of the next year outage.(author)

  9. The profile of quantitative risk indicators in Krsko NPP

    International Nuclear Information System (INIS)

    Vrbanic, I.; Basic, I.; Bilic-Zabric, T.; Spiler, J.

    2004-01-01

    During the past decade strong initiative was observed which was aimed at incorporating information on risk into various aspects of operation of nuclear power plants. The initiative was observable in activities carried out by regulators as well as utilities and industry. It resulted in establishing the process, or procedure, which is often referred to as integrated decision making or risk informed decision making. In this process, engineering analyses and evaluations that are usually termed traditional and that rely on considerations of safety margins and defense in depth are supplemented by quantitative indicators of risk. Throughout the process, the plant risk was most commonly expressed in terms of likelihood of events involving damage to the reactor core and events with radiological releases to the environment. These became two commonly used quantitative indicators or metrics of plant risk (or, reciprocally, plant safety). They were evaluated for their magnitude (e.g. the expected number of events per specified time interval), as well as their profile (e.g. the types of contributing events). The information for quantitative risk indicators (to be used in risk informing process) is obtained from plant's probabilistic safety analyses or analyses of hazards. It is dependable on issues such as availability of input data or quality of model or analysis. Nuclear power plant Krsko has recently performed Periodic Safety Review, which was a good opportunity to evaluate and integrate the plant specific information on quantitative plant risk indicators and their profile. The paper discusses some aspects of quantitative plant risk profile and its perception.(author)

  10. 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)

  11. NPP Krsko analysis of the loss of the RHR system during mid-loop operation

    International Nuclear Information System (INIS)

    Bencik, V.; Feretic, D.; Debrecin, N.

    2002-01-01

    In the paper the results of the NPP Krsko analysis of the loss of the Residual Heat Removal (RHR) system at mid-loop conditions using RELAP5/MOD 3.2.2 Gamma and RELAP5/MOD 3.3 Beta code are presented. Both pressurizer and Steam Generator (SG) 1 manway were open. The facility was open to the containment atmosphere and filled with air above liquid level. Loss of the RHR system when Reactor Coolant System (RCS) is open causes quick boiling in the core and loss of the inventory available for the cooling of the core through the openings. Aims of the analysis were threefold. First, the consequences of the transient for NPP Krsko, i.e., the time to core uncovery was determined. Secondly, the influence of the applied RELAP5 code version (MOD 3.2.2 Gamma and MOD 3.3 Beta) in the analysis of that particular transient with noncondensable gases was assessed. Third, the analysis with SG secondary sides under wet lay-up conditions was performed in order to assess the influence of condensation in the SG U tubes on liquid inventory in the system and core cooling capability.(author)

  12. Applicability of living PSA in NPP modernization

    International Nuclear Information System (INIS)

    Himanen, R.

    1999-01-01

    Recently the utility Teollisuuden Voima Oy (TVO) has modernized the Olkiluoto 1 and 2 nuclear units and increased the net electric power by 18 per cent. Level 2 PSA was performed during the modernization project and the living level 1 PSA was used to support the design of the plant modifications. The plant specific living PSA model was a powerful tool when evaluating modernization alternatives. Successive support of safety management with the PSA model requires, that both the utility and the Regulatory Body understand capability and limitations of the model in details. TVO has prepared an internal procedure that presents in detail the practices and responsibilities concerning living PSA. The procedure is based on general guidelines and requirements on probabilistic safety analysis of nuclear power plants in Finland, released by the Regulatory Body. Living PSA requires that also the procedure for the use of living PSA is living. The recently published USNRC Regulatory Guides on PSA will be taken into account in the next version of the TVO PSA procedure. The PSA Peer Review Certification Process is one way to evaluate the quality of PSA in general, but also to detect the weaknesses of the PSA. However, the Certification Process cover only limited scope of PSA omitting e.g. all other external events except internal floods. This paper gives an overview on the scope of living PSA for Olkiluoto 1 and 2, and presents some examples on the real use of PSA concerning the modernization of the plant. Definition of quantitative dependability requirements for renovated systems is possible, but on the other hand, proving of these targets is in some cases extremely difficult, because of lacking dependability data. The problems are mainly concerned in systems with of programmable logic control. (au)

  13. Analysis of inadvertent containment spray actuation for NPP Krsko

    International Nuclear Information System (INIS)

    Grgic, D.; Spalj, S.; Fancev, T.

    2000-01-01

    Refueling Water Storage Tank (RWST) supplies borated water to the Chemical and Volume Control System, Emergency Core Cooling System and Containment Spray System. In the analyses of the containment external pressure the spray temperature is assumed to be equal to the RWST lower temperature limit. This value ensures that the design negative containment pressure will not be exceeded in the event of inadvertent actuation of the Containment Spray. For NPP Kriko the negative containment pressure has to be kept below 0.1 kp/cm2 to avoid the loss of containment integrity. This paper pursuents the analysis of Inadvertent Containment Spray Actuation in order to check the influence of change in RWST water temperature on containment negative pressure. GOTHIC computer code was used for calculation of containment thermal hydraulic behavior during this accident. (author)

  14. RELAP5 Low Temperature Overpressurization Analysis for NPP Krsko

    International Nuclear Information System (INIS)

    Basic, I.; Krajnc, B.; Bajs, T.

    2000-01-01

    NEK has expressed interest in the acceptability of removing the autoclosure interlock (ACI) on the Residual Heat Removal System (RHRS) suction/isolation valves. This interest is in response to growing concerns about the loss of residual heat removal capability during cold shutdown and refueling operations. This article describes effort done to justify removal of the ACI in the light of low temperature overpressure protection of Reactor Coolant System (RCS) and RHRS and interfacing system LOCA potential. The intent of this article is to review the NEK RHRS relief valves sizing design basis and verify if the relief valves provide RHRS overpressure protection for the events possible at cold shutdown. Inadvertent isolation of RHRS during hot and cold shutdown (with reactor coolant system closed and temperatures below 177o Code 4 and5) presents one of the major safety concerns in this mode of operation. Detailed RELAP5 model of NPP Kriko following steam generator (SG) replacement and core uprate has been used in the frame of this analysis verification of RHRS relief valves sizing. The following limiting cases for cold shutdown with RCS solid conditions have been analyzed: - ransients that affect the system input/output mass balance, - ransients that affect the heat input/removal balance. (author)

  15. Annual and seasonal variations In the gamma activities in Sava river sediments upstream and downstream of NPP Krsko

    International Nuclear Information System (INIS)

    Stipe, Lulic

    2006-01-01

    Results of the five years monitoring of artificial and natural occurring radionuclides in the Sava river sediments are presented. Measurements were conducted as a part of the regular Krsko Nuclear Power Plant radioactivity control and the independent supervisions of the input of radionuclides into larger environment (immission). In order to estimate seasonal variations samples were taken from seven locations (one upstream and five downstream of the Krsko NPP) during four sampling period (seasonal) in each year. Selected radionuclides in the sediment fraction less than 0.5 mm were determined with gamma spectrometer equipped with BE3830 model High Purity Ge detector with 30% relative efficiency. (authors)

  16. Annual and seasonal variations In the gamma activities in Sava river sediments upstream and downstream of NPP Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Stipe, Lulic [Rudjer Boskovic Institute, Lab. for radioecology, Zagreb (Croatia)

    2006-07-01

    Results of the five years monitoring of artificial and natural occurring radionuclides in the Sava river sediments are presented. Measurements were conducted as a part of the regular Krsko Nuclear Power Plant radioactivity control and the independent supervisions of the input of radionuclides into larger environment (immission). In order to estimate seasonal variations samples were taken from seven locations (one upstream and five downstream of the Krsko NPP) during four sampling period (seasonal) in each year. Selected radionuclides in the sediment fractiess than 0.5 mm were determined with gamma spectrometer equipped with BE3830 model High Purity Ge detector with 30% relative efficiency. (authors)

  17. Decommissioning and Waste Disposal Programme of NPP Krsko - How to Proceed in the Future

    International Nuclear Information System (INIS)

    Mele, I.; Zeleznik, N.; Levanat, I.; Lokner, V.

    2006-01-01

    By the agreement between Slovenia and Croatia on the ownership and exploitation of the NPP Krsko, which is effective since March 2003, the decommissioning and the disposal of spent fuel and low and intermediate level waste from NPP Krsko is the responsibility of both countries. In article 10 the agreement requires that within a year after putting it into force both parties jointly prepare a decommissioning and waste disposal programme with more detailed elaboration of these issues. According to these requirements such a programme was prepared by the waste management organisations from both countries - APO from Croatia and ARAO from Slovenia - and in March 2004 submitted to the Intergovernmental Commission for adoption. Later in 2004 the document was accepted also by both Governments and in Croatia also by the Parliament. By the agreement it is also anticipated that the decommissioning and waste disposal programmes be revised at least every 5 years. Such an approach is quite common and practiced in many countries, and some countries prepare revisions even more frequently. The purpose of these new revisions is two folded: on one hand to improve the technical solutions for the decommissioning as well as for waste disposal by including new or better known data and new technological developments and experience, and on the other hand to update the cost calculation of these future nuclear liabilities. Having in mind that these cost estimations are made for the rather distant future it is extremely important that regular updating and adjustment of estimates be performed in order to meet the future needs. Although the Decommissioning and Waste Disposal Programme has just recently passed the adoption procedure and its implementation has not yet been fully achieved, the time of the next revision is approaching fast. To make good progress in the next revision serious preparations including some strategic decisions should start immediately. The programme from 2004 was prepared

  18. The new revision of NPP Krsko decommissioning, radioactive waste and spent fuel management program: analyses and results

    International Nuclear Information System (INIS)

    Zeleznik, Nadja; Kralj, Metka; Lokner, Vladimir; Levanat, Ivica; Rapic, Andrea; Mele, Irena

    2010-01-01

    The preparation of the new revision of the Decommissioning and Spent Fuel (SF) and Low and Intermediate level Waste (LILW) Disposal Program for the NPP Krsko (Program) started in September 2008 after the acceptance of the Term of Reference for the work by Intergovernmental Committee responsible for implementation of the Agreement between the governments of Slovenia and Croatia on the status and other legal issues related to investment, exploitation, and decommissioning of the Nuclear power plant Krsko. The responsible organizations, APO and ARAO together with NEK prepared all new technical and financial data and relevant inputs for the new revision in which several scenarios based on the accepted boundary conditions were investigated. The strategy of immediate dismantling was analyzed for planned and extended NPP life time together with linked radioactive waste and spent fuel management to calculate yearly annuity to be paid by the owners into the decommissioning funds in Slovenia and Croatia. The new Program incorporated among others new data on the LILW repository including the costs for siting, construction and operation of silos at the location Vrbina in Krsko municipality, the site specific Preliminary Decommissioning Plan for NPP Krsko which included besides dismantling and decontamination approaches also site specific activated and contaminated radioactive waste, and results from the referenced scenario for spent fuel disposal but at very early stage. Important inputs for calculations presented also new amounts of compensations to the local communities for different nuclear facilities which were taken from the supplemented Slovenian regulation and updated fiscal parameters (inflation, interest, discount factors) used in the financial model based on the current development in economical environment. From the obtained data the nominal and discounted costs for the whole nuclear program related to NPP Krsko which is jointly owned by Slovenia and Croatia have

  19. 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

  20. 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)

  1. LWRA analysis of inadvertent closing of the main steam isolation valve in NPP Krsko

    International Nuclear Information System (INIS)

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

    1996-01-01

    The paper describes the use of system code RELAP5/mod2 and analyzer code LWRA in analysis of inadvertent closing of the main steam isolation valve that happened in NPP Krsko on September, 25 1995. Three cases were calculated in order to address different aspects of the modelled transient. This preliminary calculation showed that, even though the real plant behaviour was not completely reproduced, such kind of analysis can help to better understand plant behaviour and to identify important phenomena in the plant during transient. The results calculated by RELAP5 and LWRA were similar and both codes indicated lack of better understanding of the plant systems status. The LWRA was more than 5 times faster than real time. (author)

  2. External Reactor Vessel Cooling Evaluation for Severe Accident Mitigation in NPP Krsko

    International Nuclear Information System (INIS)

    Mihalina, M.; Spalj, S.; Glaser, B.

    2016-01-01

    The In-Vessel corium Retention (IVR) through the External Reactor Vessel Cooling (ERVC) is mean for maintaining the reactor vessel integrity during a severe accident, by cooling and retaining the molten material inside the reactor vessel. By doing this, significant portion of severe accident negative phenomena connected with reactor vessel failure could be avoided. In this paper, analysis of NPP Krsko applicability for IVR strategy was performed. It includes overview of performed plant related analysis with emphasis on wet cavity modification, plant's site specific walk downs, new applicable probabilistic and deterministic analysis, evaluation of new possibilities for ERVC strategy implementation regarding plant's post-Fukushima improvements and adequacy with plant's procedures for severe accident mitigation. Conclusion is that NPP Krsko could perform in-vessel core retention by applying external reactor vessel cooling strategy with reasonable confidence in success. Per probabilistic and deterministic analysis, time window for successful ERVC strategy performance for most dominating plant damage state scenarios is 2.5 hours, when onset of core damage is observed. This action should be performed early after transition to Severe Accident Management Guidance's (SAMG). For loss of all AC power scenario, containment flooding could be initiated before onset of core damage within related emergency procedure. To perform external reactor vessel cooling, reactor water storage tank gravity drain with addition of alternate water is needed to be injected into the containment. ERVC strategy will positively interfere with other severe accident strategies. There are no negative effects due to ERVC performance. New flooding level will not threaten equipment and instrumentation needed for long term SAMGs performance and eventually diluted containment sump borated water inventory will not cause return to criticality during eventual recirculation phase due to the

  3. 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)

  4. Performance indicators at Embalse NPP: PSA and safety system indicators based on PSA models

    International Nuclear Information System (INIS)

    Fornero, D.A.

    2001-01-01

    Several indicators have been implemented at Embalse NPP. The objective was selecting some representative parameters to evaluate the performance of both the plant and the personnel activities, important for safety. A first set of indicators was defined in accordance with plant technical staff criteria. A complementary set of them was addressed later based on WANO guidance. This report presents the set of indicators used at Embalse NPP, centering the description to related to safety systems performance indicators (SSPI). Some considerations are done about the calculation methods, the need for aligning and updating their values following Embalse Probabilistic Safety Assessment (PSA) development, and some pros and cons of using the PSA model for getting systems indicators. Owing to the fact that PSA ownership by utilities is also a subject of the meeting, some characteristics of the organization of the PSA Project are described at the beginning of the report. At Embalse NPP a Level 1 PSA has been developed under the responsibility of its own plant and with an important contribution from the IAEA. PSA was developed at the site, conducting this to a study strongly interactive with the station staff. (author)

  5. 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)

  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. Quality Control Activities Related to Mechanical Maintenance of Safety Related Components at Krsko NPP

    International Nuclear Information System (INIS)

    Djakovic, D.

    2016-01-01

    For successful, safe and reliable operation of nuclear power plant, maintenance processes have to be systematically controlled and procedures for quality control of maintenance activities shall be established. This is requested by the quality assurance program, which shall provide control over activities affecting the quality of structures, systems, and components, considering their importance to safety. As a part of Quality and Nuclear Oversight Division (QNOD; SKV), the Quality Control Department (QC) provides quality control activities, which are deeply involved in maintenance processes at Krsko NPP, both on safety related and non-safety related (non-nuclear safety) components. QC activities on safety related components have to fulfil all requirements, which will enable the components to perform their intended safety functions. This paper describes quality control activities related to mechanical maintenance of safety related components at Krsko NPP and significant role of the Krsko plant QC Department in three particular maintenance cases connected with safety related components. In these three specific cases, the QC has confirmed its importance in compliance with quality assurance program and presented its significant added value in providing safe and reliable operation of the plant. The first maintenance activity was installation of nozzle check valves in the scope of a modification for improving regulation of spent fuel pit pumps. The QC Department performed receipt inspection of the valves. Using non-destructive examination methods and X-ray spectrometry, it was found out that the valve diffuser was made of improper material, which could cause progressive corrosion of the valve diffuser in borated water and consequently a loss of safety function of the valves followed by long-term consequences. The second one was the receipt inspection of containment ventilation fan coolers. The coolers were claimed and sent back to the supplier because the QC Department

  8. Krsko NPP Quality Assurance Plan Application to Nuclear Safety Upgrade Projects (PCFV System and PAR System)

    International Nuclear Information System (INIS)

    Biscan, Romeo; Fifnja, Igor

    2014-01-01

    Nuklearna Elektrarna Krsko (NEK) has undertaken Nuclear Safety Upgrade Projects as a safety improvement driven by the lessons learned from the Fukushima-Daiichi Accident. Among other projects, new modification 1008-VA-L Passive Containment Filtered Vent (PCFV) System has been installed which acts as the last barrier minimizing the release of radioactive material into the environment in case of failure of all safety systems, and to insure containment integrity during beyond design basis accidents (BDBA). In addition, modification 1002-GH-L Severe Accident Hydrogen Control System (PAR) has been implemented to prevent and mitigate the consequences of explosive gas generation (hydrogen and carbon monoxide) in case of reactor core melting. To ensure containment integrity for all design basis accidents (DBA) and BDBA conditions, NEK has eliminated existing safety-related electrical recombiners, replaced them with two safety-related passive autocatalytic recombiners (PARs) and added 20 new PARs designed for the BDBA conditions. Krsko NPP Quality Assurance Plan has been applied to Nuclear Safety Upgrade Projects (PCFV System and PAR System) through the following activities: · Internal audit of modification process was performed. · Supplier audits were performed to evaluate QA program efficiency of the main design organization and engineering organizations. · Evaluation and approval of Suppliers were performed. · QA engineer was involved in the review and approval of 1008-VA-L and 1002-GH-L modification documentation (Conceptual Design Package, Design Modification Package, Installation Package, Field Design Change Request, Problem/Deficiency Report, and Final Documentation Package). · Purchasing documentation for modifications 1008-VA-L and 1002-GH-L (technical specifications, purchase orders) has been verified and approved by QA. · QA and QC engineers were involved in oversight of production and testing of the new 1008-VA-L and 1002-GH-L plant components.

  9. 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)

  10. RELAP5/MOD3.3 assessment against MSIV closure events in Krsko NPP

    International Nuclear Information System (INIS)

    Parzer, I.

    2002-01-01

    The paper presents RELAP5/MOD3.3 analysis of two abnormal events occurred in Krsko NPP originating from sudden closure of Main Steam Isolation Valve (MSIV). Both events occurred before the SG replacement in 2000, the first one in September 1995 and the second one in January 1997. Valuable plant data were obtained from real plant transients and the RELAP5 code assessment was performed. Recently the last frozen version RELAP5/MOD3.3 has been released, before merging with another best-estimate thermalhydraulic system code TRAC into an integrated code. It is thus of utmost importance to assess models built in RELAP5 code against real plant transients before the code merger. A full twoloop plant model, developed at Jozef Stefan Institute (JSI), has been used for the analyses. The model includes old Westinghouse D4 type steam generators (SGs) with assumed 18% Utubes plugged in both steam generators. In the first case a malfunction in the MSIV in SG-1 caused inadvertent valve closure, while in the second case the valve stem has been broken in the SG-2, which also caused sudden valve closure.(author)

  11. RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Debrecin, N.; Foretic, D.

    2003-01-01

    In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

  12. 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

  13. 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)

  14. 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)

  15. 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

  16. RELAP5/MOD 3.2 Analysis of the Loss of RHR System Experiment Scaled to NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Bajs, T.; Prah, M.

    1998-01-01

    In the paper the RELAP5/MOD 3.2 analysis of the loss of Residual Heat Removal (RHR) system during midloop operation experiment performed at the Rig of Safety Assessment (ROSA)-IV/Large Scale Test Facility (LSTF) together with the analysis of the same test scenario scaled to NPP Krsko are presented. The experiment consisted in a loss of the RHR system at cold shutdown conditions along with a 5% cold leg break in the loop without pressurizer. The Safety Injection (SI) system was disable in the calculation. The aims of the work were to study the physical phenomena encountered under low power and low system pressure conditions while the upper part of the Reactor Coolant System (RCS) is filled with noncondensable. The impact of the bypass flow between upper plenum and downcomer inlet on transient responses was investigated. The transient was simulated for 6000 s. (author)

  17. Comparison and lessons learned from plant specific PSA of German NPP

    International Nuclear Information System (INIS)

    Balfanz, Hans-Peter; Berg, H.P.

    2000-01-01

    PSA are launched in frame of Periodic Safety Reviews (PSR) in Germany. The aims are to identify overall safety level and relative weak points. Some backfitting measures have been realized for older plants to remove relative weak points and to bring these plants to the state of the art. In this field PSA is well accepted today and is seen as a valuable tool supplementing the deterministic analysis. Main application of PSA within PSR is planned to become mandatory as part of the revision of the German Atomic Energy Act. According to the German PSA Guideline plant specific PSA level 1+ were performed for all 19 In comparison with international practice German PSA are very detailed. Otherwise they do not handle all external events, non-power states and accident management measures as discussed before. The New PSA guideline will cover these aspects and therefore analysts have to take them into account in further PSA. Moreover gathering of plant specific data is needed. The development in this field is driven by the utilities (for instance in frame of their so-called ZEDB project). Public discussion about quantitative risk of industrial hazards is quite limited in Germany and PSA results have only few impacts to this respect. Independent from this PSA for NPP is understood as a diverse tool in supporting the deterministic licensing and supervision process. Risk based decision making as well as informed regulation are just only of the beginning. State of PSA of NPP in Germany, comparison of PSA result of different NPP, German PSA guideline and state of discussion of further development and recommendation of further development of PSA of NPP are discussed in this paper in more detail. (S.Y.)

  18. Comparison and lessons learned from plant specific PSA of German NPP

    Energy Technology Data Exchange (ETDEWEB)

    Balfanz, Hans-Peter [TUEV Nord, Hamburg (Germany); Berg, H.P.

    2000-07-01

    PSA are launched in frame of Periodic Safety Reviews (PSR) in Germany. The aims are to identify overall safety level and relative weak points. Some backfitting measures have been realized for older plants to remove relative weak points and to bring these plants to the state of the art. In this field PSA is well accepted today and is seen as a valuable tool supplementing the deterministic analysis. Main application of PSA within PSR is planned to become mandatory as part of the revision of the German Atomic Energy Act. According to the German PSA Guideline plant specific PSA level 1+ were performed for all 19 In comparison with international practice German PSA are very detailed. Otherwise they do not handle all external events, non-power states and accident management measures as discussed before. The New PSA guideline will cover these aspects and therefore analysts have to take them into account in further PSA. Moreover gathering of plant specific data is needed. The development in this field is driven by the utilities (for instance in frame of their so-called ZEDB project). Public discussion about quantitative risk of industrial hazards is quite limited in Germany and PSA results have only few impacts to this respect. Independent from this PSA for NPP is understood as a diverse tool in supporting the deterministic licensing and supervision process. Risk based decision making as well as informed regulation are just only of the beginning. State of PSA of NPP in Germany, comparison of PSA result of different NPP, German PSA guideline and state of discussion of further development and recommendation of further development of PSA of NPP are discussed in this paper in more detail. (S.Y.)

  19. 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)

  20. 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

  1. Use of MAAP code for identification of key plant vulnerabilities for the beyond design accidents and their mitigation at NPP Krsko

    International Nuclear Information System (INIS)

    Krajnc, B.

    1995-01-01

    NPP Krsko performed according to GL 88-20, Supplement 1-4 and RUJV requirement the Individual Plant Examination analyses. For the required deterministic analyses the MAAP 3.0B code was used. It was proven that such severe accident analysis can be used for evaluation of the overall level of safety improvement that can be gained with the different modifications and alternate design. In this paper one such important outcomes from these analyses will be presented. (author)

  2. Influence of the CVCS Modelling on Results of the Loss of Offsite Power (LOOP) Safety Analysis for NPP Krsko

    International Nuclear Information System (INIS)

    Bencik, V.; Bajs, T.; Debrecin, N.

    2006-01-01

    A Loss of Offsite Power (LOOP) transient scenario is based on a complete loss of non-emergency AC power that results in the loss of all power to the plant auxiliaries, i.e., the Reactor Coolant Pumps (RCPs), condensate pumps, etc. An actual LOOP event would cause a loss of all feedwater, a loss of forced Reactor Coolant System (RCS) flow and a reactor trip within less than 2 seconds as a result of either loss of power to the rod cluster assembly gripper coils or any RCS flow trips. For safety analysis purposes the LOOP event is conservatively modelled as a Loss of Normal Feedwater (LONF) transient with a subsequent loss of offsite power as a result of a reactor trip. The reactor trip followed by RCP trip are delayed until a low-low Steam Generator (SG) level signal is reached. This is a more conservative scenario than the LOOP event because the least amount of SG secondary side water mass available for heat removal and the increased amount of the stored energy in the primary circuit at the time of the loss of RCS flow result. The standard LOOP safety analysis is aimed to demonstrate the natural circulation capability of the RCS to remove residual and decay heat from the core aided by Auxiliary Feedwater in the secondary system. In addition to this goal the presented work is aimed to resolve the potential safety issue resulting from the influence of the Chemical and Volume Control System (CVCS) operation during LOOP event for NPP Krsko. The potential safety concern for the LOOP analysis is that the loss of instrument air system may occur thus leading to the CVCS charging and letdown flow imbalance. A net RCS inventory addition may result with water solid pressurizer condition. Water discharge through the pressurizer relief and safety valves could lead to overpressurization of the Pressurizer Relief Tank (PRT) and rupture of the PRT rupture disks. Additional concern is that pressurizer relief and safety valves may fail to properly reseat when exposed to water relief

  3. Short-term and long-term strategies for NPP KRSKO radwaste management

    International Nuclear Information System (INIS)

    Fink, K.; Tankosic, D.; Feizullahi, F.

    1990-01-01

    All radioactive waste generated by the Nuklearna Elektrana Krsko (NEK) has been stored in a temporary storage building located at the site. In 1987, the plant owner embarked upon a program to develop and operate a low-level repository for permanent disposal of NEK waste by 1992. However, due to institutional and political considerations the schedule for the repository program has been substantially delayed. As a result, the plant owner has developed new strategies for the plant systems modifications as needed to cope with the shortage of the much needed storage and disposal space. This paper contains a presentation of these strategies and summarized the process which was used to develop them. 2 figs

  4. 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)

  5. Trends of degradation in steam generator tubes of Krsko NPP before the last planned inspection

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.; Androjna, F.

    1998-01-01

    Full-length inspection of all active tubes in both Krsko steam generators resulted in a huge amount of inspection records. A computerized database was developed by Reactor Engineering division to accelerate the management of about 200.000 records. The database was designed to support the development and decision related to the plugging criteria for damaged tubes and is utilized to gain as much experience concerning the degradation of SG tube as possible. In this paper, two prevailing group of data are statistically analyzed: the axial cracks in expansion transitions at the top of tube sheet (TTS) and Outside Diameter Stress Corrosion Cracking at tube support plates (TSP). Especially ODSCC caused a vast majority of repaired tubes (e.g., plugs and sleeves). The influence of plant startups involving oxidizing transient on the repair rates of tubes affected by ODSCC is analyzed in some detail. The results are promising and show excellent correlation in SG 2 and reasonable fit in SG 1. Predictions of maximum expected number of tubes repaired due to ODSCC at the last planned inspection is given as 67 in SG 1 and 400 in SG 2. (author)

  6. 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)

  7. 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

  8. Overview of Mobile Equipment Used in Case of Beyond Design Basis Accident at NPP Krsko

    International Nuclear Information System (INIS)

    Lukacevic, H.; Kopinc, D.; Ivanjko, M.

    2016-01-01

    Terrorist attack in USA in the September 11, 2001 and accident at the Fukushima - Daiichi Nuclear Power Station in the March 11, 2011 highlight the importance of mitigating strategies in responding to Beyond Design Basis Accident (BDBA), while ensuring cooling of reactor core, containment and spent fuel pool. Nuclear Power Plant Krsko (NEK) has acquired additional mobile equipment and made necessary modifications on existing systems for the connection of this equipment (fast couplers). Usage of mobile equipment is not only limited to design basis accident (DBA), but, also to prevent and mitigate the consequences in case of BDBA, when other plant systems are not available. NEK also decided to take steps for upgrade of safety measures and prepared Safety Upgrade Program (SUP), which is consistent with the nuclear industry response to the Fukushima accident and is implementing main projects and modifications related to SUP. NEK mobile equipment is not required to operate under normal reactor plant operation except for periodic surveillance testing and is incorporated into the normal training process. Equipment is dislocated from the reactor building and most of the equipment is located in the new building, able to withstand extreme natural events, including earthquakes and tornadoes. The usage of all mobile equipment is prescribed as an additional option in NEK operating procedures in following cases and enables following options: filling various tanks, filling the steam generators, filling the containment, additional compressed air source, spent fuel pool refilling and spraying, alternative power supply. This document provides an overview of NEK mobile equipment, which consists of various mobile fire protection pumps, air compressors, protective equipment, fire trucks, diesel generators. Sufficient fuel supply for the equipment is provided on site for a minimum three days of operation. (author).

  9. NPP Temelin safety analysis reports and PSA status

    International Nuclear Information System (INIS)

    Mlady, O.

    1999-01-01

    To enhance the safety level of Temelin NPP, recommendations of the international reviews were implemented into the design as well as into organization of the plant construction and preparation for operation. The safety assessment of these design changes has been integrated and reflected in the Safety Analysis Reports, which follow the internationally accepted guidelines. All safety analyses within Safety Analysis Reports were repeated carefully considering technical improvements and replacements to complement preliminary safety documentation. These analyses were performed by advanced western computer codes to the depth and in the structure required by western standards. The Temelin NPP followed a systematic approach in the functional design of the Reactor Protection System and related safety analyses. Modifications of reactor protection system increase defense in depth and facilitate demonstrating that LOCA and radiological limits are met for non-LOCA events. The rigorous safety analysis methodology provides assurance that LOCA and radiological limits are met. Established and accepted safety analysis methodology and accepted criteria were applied to Temelin NPP meeting US NRC and Czech Republic requirements. IAEA guidelines and recommendations

  10. Experience with high percent step load decrease from full power in NPP Krsko

    International Nuclear Information System (INIS)

    Vukovic, V.

    2000-01-01

    The control system of NPP Kriko, is designed to automatically control the reactor in the power range between 15 and 100 percent of rated power for the following designed transients; - 10 percent step change in load; 5 percent per minute loading and unloading; step full load decrease with the aid of automatically initiated and controlled steam dump. Because station operation below 15 percent of rated power is designed for a period of time during startup or standby conditions, automatic control below 15 percent is not provided. The steam dump accomplishes the following functional tasks: it permits the nuclear plants to accept a sudden 95 percent loss of load without incurring reactor trip; it removes stored energy and residual heat following a reactor trip and brings the plant to equilibrium no-load conditions without actuation of the steam generator safety valves; it permits control of the steam generator pressure at no-load conditions and permits a manually controlled cooldown of the plant. The first two functional tasks are controlled by Tavg. The third is controlled by steam pressure. Interlocks minimise any possibility of an inadvertent actuation of steam dump system. This paper discusses relationships between designed (described) characteristics of plant and the data which are obtained during startup and/or first ten years of operation. (author)

  11. Human reliability analysis during PSA at Trillo NPP: main characteristics and analysis of diagnostic errors

    International Nuclear Information System (INIS)

    Barquin, M.A.; Gomez, F.

    1998-01-01

    The design difference between Trillo NPP and other Spanish nuclear power plants (basic Westinghouse and General Electric designs) were made clear in the Human Reliability Analysis of the Probabilistic Safety Analysis (PSA) for Trillo NPP. The object of this paper is to describe the most significant characteristics of the Human Reliability Analysis carried out in the PSA, with special emphasis on the possible diagnostic errors and their consequences, based on the characteristics in the Emergency Operations Manual for Trillo NPP. - In the case of human errors before the initiating event (type 1), the existence of four redundancies in most of the plant safety systems, means that the impact of this type or error on the final results of the PSA is insignificant. However, in the case common cause errors, especially in certain calibration errors, some actions are significant in the final equation for core damage - The number of human actions that the operator has to carry out during the accidents (type 3) modelled, is relatively small in comparison with this value in other PSAs. This is basically due to the high level of automation at Rillo NPP - The Plant Operations Manual cannot be strictly considered to be a symptoms-based procedure. The operation Group must select the chapter from the Operations Manual to be followed, after having diagnosed the perturbing event, using for this purpose and Emergency and Anomaly Decision Tree (M.O.3.0.1) based on the different indications, alarms and symptoms present in the plant after the perturbing event. For this reason, it was decided to analyse the possible diagnosis errors. In the bibliography on diagnosis and commission errors available at the present time, there is no precise methodology for the analysis of this type of error and its incorporation into PSAs. The method used in the PSA for Trillo y NPP to evaluate this type of interaction, is to develop a Diagnosis Error Table, the object of which is to identify the situations in

  12. External hazards analysis approach to level 1 PSA of Mochovce NPP - Slovakia

    International Nuclear Information System (INIS)

    Stojka, Tibor

    2000-01-01

    Analyses of external events had been first time performed at the design stage of the Mochovce NPP showing sufficiently low contribution of external hazards to core damage frequency. But, based on IAEA document 'Safety problems of WWER-440/213 NPPs and the categorization' (IAEA-EBP-WWER-03, 1996), the need of new reassessment arose due to discrepancy of some origin recommendations in compare with present IAEA ones. Mochovce NPP Nuclear Safety Improvements Program elaborated at the same time included the IAEA recommendations and following improvements were proposed to perform in context of external events. 1. Seismic project and new locality seismic evaluation This safety improvement includes also some 'on site' technical improvements in seismic stability of structures and equipment. 2. Unit specific analyses of extreme meteorologic conditions. This safety improvement focuses on impact of feasible extreme conditions on NPP systems caused by rain, snow and hail storms, frost, winds, low and high temperatures. 3. Analyses of external hazards caused by humans. In this safety improvement were specified: feasible sources of explosions; analyses of hydrogen, gas and propane-calor gas depots; air crash risk. The results of these implemented safety improvements were considered in the PSA study. The External hazards analysis is also part of Level 1 PSA Mochovce NPP performed by PSA Department of VUJE Trnava Inc., Engineering, Design and Research Organization, Slovakia. Some partial analyses are performed in cooperation with following companies DS and S - SAIC, USA and Geophysical Institute Academy of Science, Slovakia Relko, Slovakia. Basic documents are: NUREG/CR-2300 'PRA Procedures Guide - A Guide to the Performance of Probabilistic Risk Assessments for Nuclear Power Plants' and IAEA SS No. 50-P-7 'Treatment of External Hazards in PSA for NPPs. The external hazards analysis consists of following parts: 1. Geography and plant locality; 2. Nearby industry; 3. Extreme

  13. 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)

  14. 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)

  15. 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

  16. 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

  17. Analytical tool for the periodic safety analysis of NPP according to the PSA guideline. Vol. 1

    International Nuclear Information System (INIS)

    Balfanz, H.P.; Boehme, E.; Musekamp, W.; Hussels, U.; Becker, G.; Behr, H.; Luettgert, H.

    1994-01-01

    The SAIS (Safety Analysis and Informationssystem) Programme System is based on an integrated data base, which consists of a plant-data and a PSA related data part. Using SAIS analyses can be performed by special tools, which are connected directly to the data base. Two main editors, RISA+ and DEDIT, are used for data base management. The access to the data base is done via different types of pages, which are displayed on a displayed on a computer screen. The pages are called data sheets. Sets of input and output data sheets were implemented, such as system or component data sheets, fault trees or event trees. All input information, models and results needed for updated results of PSA (Living PSA) can be stored in the SAIS. The programme system contains the editor KVIEW which guarantees consistency of the stored data, e.g. with respect to names and codes of components and events. The information contained in the data base are called in by a standardized users guide programme, called Page Editor. (Brunsbuettel on reference NPP). (orig./HP) [de

  18. 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)

  19. Significant aspects of the external event analysis methodology of the Jose Cabrera NPP PSA

    International Nuclear Information System (INIS)

    Barquin Duena, A.; Martin Martinez, A.R.; Boneham, P.S.; Ortega Prieto, P.

    1994-01-01

    This paper describes the following advances in the methodology for Analysis of External Events in the PSA of the Jose Cabrera NPP: In the Fire Analysis, a version of the COMPBRN3 CODE, modified by Empresarios Agrupados according to the guidelines of Appendix D of the NUREG/CR-5088, has been used. Generic cases were modelled and general conclusions obtained, applicable to fire propagation in closed areas. The damage times obtained were appreciably lower than those obtained with the previous version of the code. The Flood Analysis methodology is based on the construction of event trees to represent flood propagation dependent on the condition of the communication paths between areas, and trees showing propagation stages as a function of affected areas and damaged mitigation equipment. To determine temporary evolution of the flood area level, the CAINZO-EA code has been developed, adapted to specific plant characteristics. In both the Fire and Flood Analyses a quantification methodology has been adopted, which consists of analysing the damages caused at each stage of growth or propagation and identifying, in the Internal Events models, the gates, basic events or headers to which safe failure (probability 1) due to damages is assigned. (Author)

  20. NPP life management program - status report for Slovenia

    International Nuclear Information System (INIS)

    Glumac, B.

    1998-01-01

    Status report on NPP life management in Slovenia is dealing with possible life extension of NPP Krsko which comprises: replacement of steam generator; power upgrade; exchange of plant process computer; snubber reduction program, additional forced ventilation cooling system. Fuel improvements are predicted as well as the problems of storing spent fuel, low and intermediate waste if the plant is to operate through 2023 and possibly beyond that date. Related research activities are concerned with radiation damage, modelling of reactor core parameters by Monte Carlo calculations and PSA and severe accidents studies. Most of the activities are performed in cooperation with foreign organisations

  1. Performance of nuclear fuel in the Krsko reactor; Spremljanje delovanja jedrskega goriva v reaktorju NE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Jurcevic, M; Kurincic, B; Levstek, M F; Sambo, B; Vrcko, P [Nuklearna elektrana Krsko, Krsko (Yugoslavia)

    1987-07-01

    In this paper activities to follow performance of the nuclear fuel and operational status of the reactor of Nuclear Power Plant Krsko are presented. Short descriptions of the methods as well as nuclear and process instrumentation used for surveillance of the reactor performance are given. The purpose of the subject activities is to assure safe operation of the reactor in accordance with the Final safety Analysis Report of NPP Krsko. (author)

  2. Soundness of Krsko Nuclear Power Plant Performance in Terms of Energy and Finance

    International Nuclear Information System (INIS)

    Curkovic, A.; Vrankic, K.; Magdic, M.

    1998-01-01

    Compared to existing conventional thermal power plants in Croatian electric power system, as well as to alternative (potential) imported coal and gas fired thermal power plants, Krsko NPP (nuclear power plant) generates electricity with lower production costs. This cost margin in favour of the Krsko NPP represents the soundness of this nuclear power plant in terms of energy and finance. (author)

  3. 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)

  4. Krsko NPP ecological information system

    International Nuclear Information System (INIS)

    Kovac, A.; Breznik, B.

    1996-01-01

    The Ecological Information System was developed and is used for continuous data collecting from different measuring points as well as for dose calculation in case of emergency. The system collects all the data which are continuously measured in the environment or might have influence to the environment or are necessary for evaluation of impact to the environment. (author)

  5. Post-reconstruction full power and shut down level 2 PSA study for Unit 1 of Bohunice V1 NPP

    International Nuclear Information System (INIS)

    Kovacs, Z.

    2003-01-01

    The level 2 PSA model of the J. Bohunice V1 NPP was developed in the RISK SPECTRUM Professional code with the following objectives: to identify the ways in which radioactive releases from the plant can occur following the core damage; to calculate the magnitudes and frequency of the release; to provide insights into the plant behaviour during a severe accident; to provide a framework for understanding containment failure modes; the impact of the phenomena that could occur during and following core damage and have the potential to challenge the integrity of the confinement; to support the severe accident management and development of SAMGs. The magnitudes of release categories are calculated using: the MAAP4/VVER for reactor operation and shutdown mode with closed reactor vessel and the MELCOR code for shutdown mode with open reactor vessel. In this paper an overview of the Level 2 PSA methodology; description of the confinement; the interface between the level 1 and 2 PSA and accident progression analyses are presented. An evaluation of the confinement failure modes and construction of the confinement event trees as well as definition of release categories, source term analysis and sensitivity analyses are also discussed. The presented results indicate that: 1)for the full power operation - there is an 25% probability that the confinement will successfully maintain its integrity and prevent an uncontrolled fission product release; the most likely mode of release from the confinement is a confinement bypass after SGTM with conditional probability of 30%; the conditional probability for the confinement isolation failure probability without spray is 5%, for early confinement failure at the vessel failure is 4%, for other categories 1% or less; 2) for the shutdown operating modes - the shutdown risk is high for the open reactor vessel and open confinement; important severe accident sequences exists for release categories: RC5.1, RC5.2 and RC6.2

  6. Integration of the Ovation Based Distributed Turbine Control System into the Existing Process Information System and Full Scope Simulator at NPP Krsko

    International Nuclear Information System (INIS)

    Cicvaric, D.

    2010-01-01

    Programmable Digital Electro Hydraulic System (PDEH) is Turbine Control System (TCS), built on Emerson OVATION Distributed Control System (DCS) platform and installed by the Westinghouse Electric Company at the Krsko Nuclear Power Plant as the replacement for the DEH Mod II turbine control system. Core of the PDEH system consist of three pairs of redundant controllers (ETS, OA/OPC and ATC/MSR) configured for the Turbine Generator (TG) set protection, control and monitoring functions. Existing serial data link between replaced DEH Mod II and Process Information System (PIS) was removed and replaced with redundant bi-directional Ethernet TCP/IP data link via two Data Link servers in client-server architecture configuration. All hardwired signals and some of the important calculated signals are being transferred from PDEH to PIS. Main purpose of PIS data link is trending utilization at the existing PIS workstations with pre-configured trend groups and centralized data archiving. Most of the PDEH display screens (mimics) were also replicated on the PIS platform, so that TG set monitoring and operation overview can be performed over the PIS network as well as over the Process Computer Network (PCN) with PMSNT-view utility. The simulator is implemented using a stimulated Windows based Ovation platform and a SGI IRIX based Plant Model Computer (PMC) using the L-3 MAPPS simulation software platform. Two PDEH stimulated systems are installed at the Krsko Full Scope Simulator (KFSS), one for foreground and another for background simulation. Stimulated PDEH hardware is essentially identical to that installed in the plant with the exception of hardware redundancy, isolation features and interface with physical plant I/O. The Ovation control logic sheets are executed with virtual controllers hosted on a simulator specific Virtual Controller Host (VCH) Workstation. The data interface between the simulator Ovation system and the PMC is accomplished through the Ethernet

  7. 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)

  8. Impact of power plant KRSKO on the environment

    International Nuclear Information System (INIS)

    Hak, Nena; Lulic, Stipe

    1993-01-01

    The Sava river is among the largest rivers in the Republic of Croatia. It drains 95.000 square kilometers before meeting the Danube River. The Sava river and its surroundings we being exploited in agriculture, forestry, power generation (one nuclear power plant and several thermal power plants), oil transportation, gravel extraction and recreation. At last, the Sava River is the major source of fresh water for industry and population. Different authorized institutions from the Republic of Slovenia and the Republic of Croatia are included in programs of nonradiological and radiological monitoring of Nuclear Power Plant Krsko (NPPK). Quarterly, the institutions from the Republic of Croatia, NPP Krsko and the Ministry of Energy of the Republic of Croatia, submit public information in the Republic of Croatia about NPP Krsko operation and its environmental impact

  9. ATWS analyses for Krsko Full Scope Simulator verification

    Energy Technology Data Exchange (ETDEWEB)

    Cerne, G; Tiselj, I; Parzer, I [Reactor Engineering Div., Inst. Jozef Stefan, Ljubljana (Slovenia)

    2000-07-01

    The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for verification of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD2 code and the input card deck for NPP Krsko was used. The analyses for ATWS were performed to assess the influence and benefit of ATWS Mitigation System Actuation Circuitry (AMSAC). In the presented paper the most severe ATWS scenarios have been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied regarding the AMSAC availability. (author)

  10. Krsko 2 Project Preparation and Status

    International Nuclear Information System (INIS)

    Novsak, M.; Spiler, J.; Bergant, R.; Zagar, T.

    2008-01-01

    The Government of the Republic of Slovenia has adopted a document 'Resolution on National Development Projects for the period 2007 - 2023' in October 2006. Among several projects which all support sustainable development of Slovenia, an option of a new nuclear power plant was proposed. Thus, the Slovenian Government showed obvious intention and direction of Slovenian future energetic situation. The supply of electricity has been sharpening in Slovenia in the last few years. Growing of gross domestic product (GDP) and approaching to the life standard of developed EU countries has caused increased consumption of electricity. Since domestic electricity production does not follow the increased consumption, Slovenia already imports nearly a quarter of its needed electricity. Slovenia would need 400 MW of installed electrical power now for covering of its base load needs only. Based on recent electricity consumption forecasts, Slovenia will need 800 MW of new electricity generation capacities by 2015 and 1500 MW by 2025 respectively. Besides high annual growth rate and high dependence on electricity import, a problem of relatively old power generation objects and realization of new energy and climate package for Europe exists. All together dictates our options for expansion of capacity production of the proposed Nuclear power plant Krsko 2. Experience and knowledge of light water pressurized reactor can be effectively used at NPP Krsko 2. Planned NPP Krsko 2 would have between 1100 and 1600 MW of installed electrical power. If the decision is taken in 2009, construction could start in 2013 and finish in 2017. (author)

  11. ATWS thermal-hydraulic analysis for Krsko Full Scope Simulator validation

    International Nuclear Information System (INIS)

    Parzer, I.; Kljenak, I.

    2005-01-01

    The purpose of this analysis was to simulate Anticipated Transient without Scram transient for Krsko NPP. The results of these calculations were used for annual ANSI/ANS validation of reactor coolant system thermal-hydraulic response predicted by Krsko Full Scope Simulator. For the thermal-hydraulic analyses the RELAP5/MOD3.3 code and the input model for NPP Krsko, delivered by NPP Krsko, was used. In the presented paper the most severe ATWS scenario has been analyzed, starting with the loss of Main Feedwater at both steam generators. Thus, gradual loss of secondary heat sink occurred. On top of that, control rods were not supposed to scram, leaving the chain reaction to be controlled only by inherent physical properties of the fuel and moderator and eventual actions of the BOP system. The primary system response has been studied assuming AMSAC availability. (author)

  12. PSA analysis focused on Mochovce NPP safety measures evaluation from operational safety point of view

    International Nuclear Information System (INIS)

    Cillik, I.; Vrtik, L.

    2001-01-01

    Mochovce NPP consists of four reactor units of WWER 440/V213 type and it is located in the south-middle part of Slovakia. At present first unit operated and the second one under the construction finishing. As these units represent second generation of WWER reactor design, the additional safety measures (SM) were implemented to enhance operational and nuclear safety according to the recommendations of performed international audits and operational experience based on exploitation of other similar units (as Dukovany and J. Bohunice NPPs). These requirements result into a number of SMs grouped according to their purpose to reach recent international requirements on nuclear and operational safety. The paper presents the bases used for safety measures establishing including their grouping into the comprehensive tasks covering different areas of safety goals as well as structural organization of a project management of including participating companies and work performance. More, results are given regarding contribution of selected SMs to the total core damage frequency decreasing.(author)

  13. 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)

  14. SB LOCA analyses for Krsko Full Scope Simulator verification

    International Nuclear Information System (INIS)

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

    2000-01-01

    Nuclear power plant simulators are intended to be used for training and maintaining competence to ensure safe, reliable operation of nuclear power plants throughout the world. The simulator shall be specified to a reference unit and its performance validation testing shall be provided. In this study a small-break loss-of-coolant accident (SB LOCA) response of Krsko nuclear power plant (NPP) was calculated for full scope simulator verification. The investigation included five cases with varying the break size in the cold leg of reactor coolant system. The plant specific and verified RELAP5/MOD2 model of Krsko nuclear power plant (NPP), developed in the past for 1882 MWt power, was adapted for 2000 MWt power (cycle 17) including the model for replacement steam generators. The results showed that the plant system response to breaks with small break area was slower compared to breaks with larger break area. The core heatup occurred in most of the cases analyzed. The acceptance criteria for emergency core cooling system were also met. The predicted results of the SB LOCA analysis for Krsko NPP suggest that they may be used for verification of the Krsko Full Scope Simulator performance. (author)

  15. Krsko Nuclear Power Plant's Environmental Management System in Compliance with ISO 14001:2004

    International Nuclear Information System (INIS)

    Kusar, A.; Kavsek, D.

    2010-01-01

    Krsko Nuclear Power Plant (NPP) pays special attention to environmental protection and practices environmental safety in all plant processes and management. In 2008, Krsko NPP introduced the Environmental Management System in compliance with ISO 14001:2004 standard. The plant management announced the Environmental policy which is a part of the business strategy of Krsko NPP which is an eco-friendly company. The Policy is a commitment of the plant management and all staff to act in compliance with requirements of ISO 14001:2004. The standard served as a basis for developing some new documentation such as Environmental Management System Quality Manual, Environmental planning procedures identifying legal and other requirements, Register of environmental aspects, Register of legal and other requirements etc. When establishing the Register of environmental aspects, all possible environmental impacts of the plant were carefully reviewed and estimated. Following the introduction and certification audit in October and December 2008 of Bureau Veritas Certification, Krsko NPP was awarded certificate ISO 14001:2004 attesting conformity of its Environmental Management System with this standard. The Environmental Certificate means that Krsko NPP will promote a positive environmental culture and maintain a safe, healthy and environmentally-sound workplace for all its employees, contractors and visitors.(author).

  16. Prioritization of design changes based on PSA

    International Nuclear Information System (INIS)

    Krajnc, B.; Mavko, B.

    1996-01-01

    Effective use of Probabilistic Safety Analyses (PSA) in the day to day plant operation is subject of intensive discussions among plant operators and regulators. There are several possible applications in which the PSA can be used, among those also to use the PSA approach for the quantification of influence of different proposed design changes to nuclear safety - influence on public safety - health. NPP Krsko is one of those plants that successfully completed its PSA project, with Level 1 and Level 2 analyses and effective know-how transfer. It also faces a number of regulatory and internally generated requirements for different design changes, mainly due to the fact that the plant is committed to continuous augmentation of nuclear safety. It is considered that the available tools and knowledge should be used and therefore applicable methodology should be developed for effective prioritization of proposed design changes by performing cost-benefit analyses for all major modifications - focusing on their influence on nuclear safety. Based on the above a new method for prioritization of design changes is proposed. The method uses Level 1 results (in the sense of plant damage states and their frequencies) directly as an input for further processing - first decision step to decide whether the proposed modification has or has no influence on nuclear safety. In Level 2 analyses the combination of probabilistic and deterministic approach was adopted. In fact the results of the deterministic analyses of severe accidents are treated in probabilistic manner due to large uncertainty of results. Finally to be able to perform plant specific cost benefit analyses so called partial Level 3 was defined. The proposed methods was preliminary tested and it gave favorable results. (author)

  17. 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

  18. 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

  19. Performance assessment for Nuclear Power Plant Krsko intermediate and low-level rad-waste repository

    International Nuclear Information System (INIS)

    Jelavic, V.; Skanata, D.; Plecjas, I.

    1990-01-01

    Performance safety Assessment for NPP Krsko radwaste repository was performed (LLW/ILW). Shallow land and tunnel type concept were analyzed. Because it was based on two unknown referent sites, one for the shallow land concept and the other for the tunnel type, analysis was generic in nature. Scenario selecting process and consequence analysis were performed by using deterministic approach. Results for both concepts of disposal suggests that proposed NPP Krsko radwaste repository reference site and disposal technology will fully meet radiation limits imposed by the Yugoslav regulations and ICRP guidelines. (author)

  20. Development and verification of a leningrad NPP unit 1 living PSA model in the INL SAPHIRE code format for prompt operational safety level monitoring

    International Nuclear Information System (INIS)

    Bronislav, Vinnikov

    2007-01-01

    The first part of the paper presents results of the work, that was carried out in complete conformity with the Technical Assignment, which was developed by the Leningrad Nuclear Power Plant. The initial scientific and technical information, contained into the In-Depth Safety Assessment Reports, was given to the author of the work. This information included graphical Fault Trees of Safety Systems and Auxiliary Technical Systems, Event Trees for the necessary number of Initial Events, and also information about failure probabilities of basic components of the nuclear unit. On the basis of this information and fueling it to the Usa Idaho National Laboratory (INL) SAPHIRE code, we have developed an electronic version of the Data Base for failure probabilities of the components of technical systems. Then, we have developed both the electronic versions of the necessary Fault Trees, and an electronic versions of the necessary Event Trees. And at last, we have carried out the linkage of the Event Trees. This work has resulted in the Living PSA (LPSA - Living Probabilistic Safety Assessment) Model of the Leningrad NPP Unit 1. The LPSA-model is completely adapted to be consistent with the USA INL SAPHIRE Risk Monitor. The second part of the paper results in analysis of fire consequences in various places of Leningrad NPP Unit 1. The computations were carried out with the help of the LPSA-model, developed in SAPHIRE code format. On the basis of the computations the order of priority of implementation of fire prevention measures was established. (author)

  1. Integrated approach to fire safety at the Krsko nuclear power plant - fire protection action plan

    International Nuclear Information System (INIS)

    Lambright, J.A.; Cerjak, J.; Spiler, J.; Ioannidi, J.

    1998-01-01

    Nuclear Power Plant Krsko (NPP Krsko) is a Westinghouse design, single-unit, 1882 Megawatt thermal (MWt), two-loop, pressurized water nuclear power plant. The fire protection program at NPP Krsko has been reviewed and reports issued recommending changes and modifications to the program, plant systems and structures. Three reports were issued, the NPP Krsko Fire Hazard Analysis (Safe Shout down Separation Analysis Report), the ICISA Analysis of Core Damage Frequency Due to Fire at the NPP Krsko and IPEEE (Individual Plant External Event Examination) related to fire risk. The Fire Hazard Analysis Report utilizes a compliance - based deterministic approach to identification of fire area hazards. This report focuses on strict compliance from the perspective of US Nuclear Regulatory Commission (USNRC), standards, guidelines and acceptance criteria and does not consider variations to comply with the intent of the regulations. The probabilistic analysis methide used in the ICISA and IPEEE report utilizes a risk based nad intent based approach in determining critical at-risk fire areas. NPP Krsko has already completed the following suggestions/recommendations from the above and OSART reports in order to comply with Appendix R: Installation of smoke detectors in the Control Room; Installation of Emergency Lighting in some plant areas and of Remote Shout down panels; Extension of Sound Power Communication System; Installation of Fire Annunciator Panel at the On-site Fire Brigade Station; Installation of Smoke Detection System in the (a) Main Control Room Panels, (b) Essential Service Water Building. (c) Component Cooling Building pump area, chiller area and HVAC area, (d) Auxiliary Building Safety pump rooms, (e) Fuel Handling room, (f) Intermediate Building AFFW area and compressor room, and (g) Tadwaste building; inclusion of Auxiliary operators in the Fire Brigade; training of Fire Brigade Members in Plant Operation (9 week course); Development of Fire Door Inspection and

  2. Implementation of procedures to NPP Krsko INTRANEK

    International Nuclear Information System (INIS)

    Gradisar, D.; Krajnc, M.; Kocnar, R.; Spiler, J.

    1999-01-01

    Part of NEK documentation has already been presented on NEK Intranet such as USAR, Technical Specifications, QA Plan as well as some frequently used series of drawings. At the time being the process of presentation of all procedures (thereinafter INTRANEK procedures) is in progress. The purpose of this project is the presentation of 1600 procedures with average size of 30 pages what is more than 48000 pages altogether. ADOBE PDF (Portable Document Format) has been chosen as the most suitable format for the presentation of procedures on INTRANEK. PDF format meets the following criteria: the outlook of a document page is always the same as an original one and cannot be changed without control. In addition to this, full text search is available as well as easy jump from procedure to procedure. Some changes of working process on internal procedures have to be made before the project start, which determine the responsibility of individual users in the process. The work flow, which enables easy daily maintenance, has been prepared, the rules of both procedure numbering as well as folder contents/name have been set and the server selected. The project was managed and implemented with the extensive use of compute-aided management, document distribution and control, databases, electronics mail and Intranet tools. The results of practical implementation of NEK procedures and our experience with INTRANEK are presented in this paper.(author)

  3. 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

  4. State of the art of probabilistic safety analysis (PSA) in the FRG, and principles of a PSA-guideline

    International Nuclear Information System (INIS)

    Balfanz, H.P.

    1987-01-01

    Contents of the articles: Survey of PSA performed during licensing procedures of an NPP; German Nuclear Standards' requirements on the reliability of safety systems; PSA-guideline for NPP: Principles and suggestions; Motivation and tasks of PSA; Aspects of the methodology of safety analyses; Structure of event tree and fault tree analyses; Extent of safety analyses; Performance and limits of PSA. (orig./HSCH)

  5. Performance assessment for Nuclear Power Plant Krsko intermediate and low-level rad-waste repository; Provjera sigurnosnih performansi odlagalishta radioaktivnog otpada niske i srednje aktivnosti nuklearne elektrane Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Jelavic, V [Inst. za Elektroprivredu, Zagreb. (Yugoslavia).; Tankosic, Dj [Bechtel Inc., San Francisco (United States); Skanata, D [Nuklearna elektrana Krshko, Krshko (Yugoslavia); Plecjas, I [Institut za Nuklearne Nauke Boris Kidric, Belgrade (Yugoslavia)

    1990-07-01

    Performance safety Assessment for NPP Krsko radwaste repository was performed (LLW/ILW). Shallow land and tunnel type concept were analyzed. Because it was based on two unknown referent sites, one for the shallow land concept and the other for the tunnel type, analysis was generic in nature. Scenario selecting process and consequence analysis were performed by using deterministic approach. Results for both concepts of disposal suggests that proposed NPP Krsko radwaste repository reference site and disposal technology will fully meet radiation limits imposed by the Yugoslav regulations and ICRP guidelines. (author)

  6. Potential need for re-definition of the highest priority recovery action in the Krsko SAG-1

    International Nuclear Information System (INIS)

    Bilic Zabric, T.; Basic, I.

    2005-01-01

    Replacement of old SG (Steam Generators) [7] and the characteristic of new ones throws the question of proper accident management strategy, which leans on philosophy that repair and recovery actions have first priority. In the current NPP Krsko SAMGs (Severe Accident Management Guidelines), water supply to the SG has priority over re-injection water into the core. NPP Krsko reconsidered the highest priority of SAG-1 (inject water to the SG), against the WOG (Westinghouse Owners Group) generic approach (inject water into the core) and potential revision of Severe Accident Phenomenology Evaluations using MAAP (Modular accident Analysis Program) 4.0.5 code. (author)

  7. Human reliability analysis approach to level 1 PSA - shutdown and low power operation of Mochovce NPP, Unit 1, Slovakia

    International Nuclear Information System (INIS)

    Stojka, Tibor; Holy, Jaroslav

    2003-01-01

    The paper presents general approach, used methods and form of documentation of the results as have been applied within the Human Reliability Analysis (HRA) task of the shutdown and low power PSA (SPSA) study for Mochovce nuclear power plant, Unit 1, Slovakia. The paper describes main goals of the HRA task within the SPSA project, applied methods and data sources. Basic steps of the HRA task and human errors (HE) classification are also specified in its first part. The main part of the paper deals with pre-initiator human errors, human-induced initiators and response to initiator human errors. Since the expert judgment method (SLIM) was used for the last type of human errors probability assessment, also related activities are described including preparation works (performance shaping factors (PSFs) selection, development of PSF classification tables, preparation of aid tools for interview with plant experts), qualitative analysis (sources of information and basic steps) and quantitative analysis itself (human errors classification for final quantification including criteria used for the classification, description of structure of the spreadsheet used for quantification and treatment with dependencies). The last part of the paper describes form of documentation of the final results and provides some findings. (author)

  8. 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

  9. Tritium in organic matter around Krsko Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kristof, Romana; Zorko, Benjamin; Kozar Logar, Jasmina; Kosenina, Suzana

    2017-01-01

    The aim of the research was to obtain first results of tritium in the organic matter of environmental samples in the vicinity of Krsko NPP. The emphasis was on the layout of suitable sampling network of crops and fruits in nearby agricultural area. Method for determination of tritium in organic matter in the form of Tissue Free Water Tritium (TFWT) and Organically Bound Tritium (OBT) has been implemented. Capabilities of the methods were tested on real environmental samples and its findings were compared to modeled activities of tritium from atmospheric releases and literature based results of TFWT and OBT. (author)

  10. NPP Prevlaka - Preparation of construction

    International Nuclear Information System (INIS)

    Bojic, K.

    1984-01-01

    On the basis of study 'Optimal electricity generation structure till the year 2000' production of 3 x 500 MWe in nuclear power plants has been anticipated. Second Croatian-Slovenian NPP project will be based on the same principles the first one (NPP Krsko) was based on. Preconstruction investigation studies are performed at site Prevlaka on river Sava downstream of Zagreb. Licensing procedure has started with republic Urban countryside planning activities. Preconstruction activities are planned to be finished by the end of 1986. while the construction is expected to start during 1987. Parallel to investigation studies for NPP Prevlaka, evaluation of nuclear technology and reactor type is planned to be made. (author)

  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. 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

  14. Ideal scaling of BETHSY 9.1.B test results to NPP

    International Nuclear Information System (INIS)

    Petelin, S.; Guntel, I.

    1995-01-01

    The transient scenario standard problem 27 (ISP-27) was implemented in the RELAP5 analysis of small break loss of coolant accident for Krsko nuclear power plant (Krsko NPP). The objective was to evaluate the effectiveness of ISP-27 proposed accident management procedure for real NPP and to compare the physical phenomena known from experimental background with the phenomena predicted by RELAP5 simulation of real plant transient. The analyses showed that, if relevant break, power scaling criteria, primary and secondary pressure are fulfilled the RELAP5 model of Krsko NPP cannot completely ensure the simulation of typical thermal-hydraulic phenomena observed in BETHSY facility during ISP-27. Much better satisfaction is observed on ideal scaled up model. (author)

  15. Flooding risk reduction for the ASCO NPP PSA; Reduccion del riesgo por inundaciones en el APS de C.N. ASCO

    Energy Technology Data Exchange (ETDEWEB)

    Nos Llorens, V [Empresarios Agrupados, A.I.E., Madrid (Spain); Faig Sureda, J [Asociacion Nuclear ASCO, Barcelona (Spain)

    1993-12-15

    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)

  16. Involvement of Union Fenosa skills in the thermohydraulic area of the Jose Cabrera NPP PSA. Applications of the RELAPS5/MOD2 Code; Implicacion de las capacidades de union fenosa dentro del area de termohidraulica en el APS de la C.N. Jose Cabrera. Aplicaciones del codigo RELAP5/MOD2

    Energy Technology Data Exchange (ETDEWEB)

    Martin, L; Saenz Tejada, P [Empresarios Agrupados, A.I.E., Madrid (Spain)

    1993-12-15

    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)

  17. Evaluation of implementation 18-month cycle in NPP Krsko

    International Nuclear Information System (INIS)

    Novsak, M.; Bilic Zabric, T.

    2003-01-01

    The feasibility study, which evaluates the impacts of the transition to the 18-month cycle on plant design bases, operational and maintenance requirements, long-term economics of the operation, is performed. Activities, which have to be implemented for adopting 18- month cycle, are presented. The study shows that implementation of 18-month cycle is possible with some investments through transition 15-month cycle. (author)

  18. 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

  19. 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)

  20. PSA methodology

    Energy Technology Data Exchange (ETDEWEB)

    Magne, L

    1997-12-31

    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{sup 1} and EPS 900{sup 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.

  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. Radioactivity monitoring programme of Krsko nuclear power plant

    International Nuclear Information System (INIS)

    Miklavzic, U.; Martincic, R.; Kanduc, M.; Lulic, S.; Kovac, J.; Breznik, B.

    1996-01-01

    As a successor to the preoperational surveillance programme, the regular offsite radioactivity monitoring programme (RMP) of the Krsko Nuclear Power Plant (NPP) was implemented in 1982, when the power plant formally commenced operating. Observations collected during the first years of its operation were later also the basis for setting up the official 'Regulatory guide on monitoring of nuclear installations', issued not earlier than in 1986. The basic criterion which governed the selection of measuring methods, sampling techniques and locations, was the extent to which the data obtained could serve for the realistic assessment of the committed dose to a member of the population, and later on to members of a representative (critical) group. To be able to differentiate the radioactivity released through the liquid and gaseous effluents of the NPP from other radiation sources (natural radioactivity, global contamination), and especially because of the varying radiotoxicity of different radionuclides, in principle monitoring in the environment, as at the source, had to provide activity data for each individual radionuclide appearing in the effluents. Therefore, as early as 1982 the programme attributed the main weight to high resolution gamma spectrometry, combined with specific radiochemical analytical methods (e.g. 90 Sr/ 89 Sr, 3 H, 14 C, alpha spectrometry of Pu isotopes) which together made feasible determination of individual specific activities of the most significant man-made and natural radionuclides. By weighting the specific activities of the radionuclides identified and measured in the media surveyed by dose factors for intake, the quantity 'B' - the so-called 'radiological burden', was calculated and introduced in the yearly-summary tables. Expressed in relative units, from which the committed dose could be readily calculated, the burden B very lucidly disclosed the relative importance of different artificial pollutants and natural radioactivity present in

  3. 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

  4. PSA applications

    Energy Technology Data Exchange (ETDEWEB)

    Dubreuil Chambardel, A

    1997-12-31

    The IAEA now defines three types of PSA applications: Validation of design and of operation procedures; optimization of plant operation; and regulatory applications. The applications of PSA are manifold: only a few are dealt with here (precursor analysis is dealt with in session 3, topic 4). For each of them, we will do the utmost to demonstrate the main difficulties encountered, EDF`s viewpoint on the matter, and the points remaining to be solved. In what follows, unless explicitly stated otherwise, we have made every effort to represent the different applications as they are practiced by all concerned in the international community, and to describe the inherent difficulties the international community has encountered with these applications with all objectivity. It goes without saying that the comments below are simply those of the ESF department, and are submitted here for discussion by the experts. 13 refs.

  5. PSA applications

    International Nuclear Information System (INIS)

    Dubreuil Chambardel, A.

    1996-01-01

    The IAEA now defines three types of PSA applications: Validation of design and of operation procedures; optimization of plant operation; and regulatory applications. The applications of PSA are manifold: only a few are dealt with here (precursor analysis is dealt with in session 3, topic 4). For each of them, we will do the utmost to demonstrate the main difficulties encountered, EDF's viewpoint on the matter, and the points remaining to be solved. In what follows, unless explicitly stated otherwise, we have made every effort to represent the different applications as they are practiced by all concerned in the international community, and to describe the inherent difficulties the international community has encountered with these applications with all objectivity. It goes without saying that the comments below are simply those of the ESF department, and are submitted here for discussion by the experts. 13 refs

  6. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Holz, R; Stach, U [Siemens AG, Erlangen, Offenbach (Germany); Gloaguen, C [Framatome, Paris (France)

    2000-07-01

    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{sup th} of April 2000 and the plant was back on line on 15{sup 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)

  8. 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)

  9. 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)

  10. External flood probabilistic safety analysis of a coastal NPP

    International Nuclear Information System (INIS)

    Pisharady, Ajai S.; Chakraborty, M.K.; Acharya, Sourav; Roshan, A.D.; Bishnoi, L.R.

    2015-01-01

    External events pose a definitive challenge to safety of NPP, solely due to their ability to induce common cause failures. Flooding incidents at Le Blayais NPP, France, Fort Calhoun NPP, USA and Fukushima Daiichi have pointed to the importance of external flooding as an important contributor to NPP risk. A methodology developed for external flood PSA of a coastal NPP vulnerable to flooding due to tsunami, cyclonic storm and intense local precipitation is presented in this paper. Different tasks for EFPSA has been identified along with general approach for completing each task

  11. 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)

  12. PSA Review Handbook

    International Nuclear Information System (INIS)

    Hallman, Anders; Nyman, Ralph; Knochenhauer, Michael

    2004-05-01

    The Swedish Nuclear Power Inspectorate (SKI) expresses requirements on the performance of PSAs as well as on PSA activities in general in the the regulatory document 'Regulations Concerning Safety in Certain Nuclear Facilities', SKlFS 1998:1. The follow-up of these activities is part of the inspection tasks of the SKI. In view or this, there is a need for documented guidelines on now to perform these inspections and reviews. The SKI PSA Review Handbook is intended to be a support in the SKI inspection and control of the PSA activities or the licensees. These PSA activities include both the organisation and working procedures of the licensee, the layout and contents of the PSA, and its areas of application. Using the regulation SKIFS 1998:1 as a starting point, the review handbook presents important aspects to be considered when judging whether a licensee fulfils the requirements on PSA activities, including the performance of PSA:s or PSA applications. The handbook shall also be a guidance for the review of PSA:s. However, the intention of the PSA Review Handbook is not to be a handbook for how a PSA is performed. The PSA Review Handbook is applicable to all types or initiating events and all operating conditions, and has been structured in a way, which stresses the integrated characteristics of PSA in the creation of the risk picture of a plant. The PSA Review Handbook has been based on the requirements for PSA of nuclear power plants, as this is the most extensive application. However, the relevant parts of it are also applicable when analysing other nuclear installations. The PSA Review Handbook is published as a research report as its contents are judged to be of general interest, and the SKI welcomes comments to the handbook. An update or the PSA Review Handbook may be required as experience with the use of the handbook is acquired and if general PSA requirements change

  13. 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

  14. 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)

  15. 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

  16. External hazards considered for Paks NPP

    International Nuclear Information System (INIS)

    Kiss, Tibor

    2000-01-01

    PAKS NPP was built according to Soviet construction standards which took into account meteorological aspects but no documents for other external hazards were available. Main activities concerning earthquakes cover reevaluation of the plant site, seismic safety technological concept, improving the seismic resistance, installation of seismic monitoring and protection system, and seismic PSA

  17. 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)

  18. 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)

  19. Spent fuel pool risk analysis for the Dukovany NPP

    Energy Technology Data Exchange (ETDEWEB)

    Hust' ak, S.; Jaros, M.; Kubicek, J. [UJV Rez, a.s., Husinec-Rez (Czech Republic)

    2013-07-01

    UJV Rez, a.s. maintains a Living Probabilistic Safety Assessment (Living PSA) program for Dukovany Nuclear Power Plant (NPP) in the Czech Republic. This project has been established as a framework for activities related to risk assessment and to support for risk-informed decision making at this plant. The most extensively used PSA application at Dukovany NPP is risk monitoring of instantaneous (point-in-time) risk during plant operation, especially for the purpose of configuration risk management during plant scheduled outages to avoid risk significant configurations. The scope of PSA for Dukovany NPP includes also determination of a risk contribution from spent fuel pool (SFP) operation to provide recommendations for the prevention and mitigation of SFP accidents and to be applicable for configuration risk management. This paper describes the analysis of internal initiating events (IEs) in PSA for Dukovany NPP, which can contribute to the risk from SFP operation. The analysis of those IEs was done more thoroughly in the PSA for Dukovany NPP in order to be used in instantaneous risk monitoring. (orig.)

  20. 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...

  1. 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

  2. A Tsunami PSA for Nuclear Power Plants in Korea

    International Nuclear Information System (INIS)

    Kim, Min Kyu; Choi, In Kil; Park, Jin Hee; Seo, Kyung Suk; Seo, Jeong Moon; Yang, Joon Eon

    2010-06-01

    For the evaluation of safety of NPP caused by Tsunami event, probabilistic safety assessment (PSA) method was applied in this study. At first, an empirical tsunami hazard analysis performed for an evaluation of tsunami return period. A procedure for tsunami fragility methodology was established, and target equipment and structures for investigation of Tsunami Hazard assessment were selected. A several fragility calculations were performed for equipment in Nuclear Power Plant and finally accident scenario of tsunami event in NPP was presented. Finally, a system analysis performed in the case of tsunami event for an evaluation of a CDF of Ulchin 56 NPP site. For the evaluation of safety of NPP caused by Tsunami event, probabilistic safety assessment (PSA) method was applied. A procedure for tsunami fragility methodology was established, and target equipment and structures for investigation of Tsunami Hazard assessment were selected. A several fragility calculations were performed for equipment in Nuclear Power Plant and finally accident scenario of tsunami event in NPP was presented. As a result, in the case of tsunami event, functional failure is mostly governed total failure probability of facilities in NPP site

  3. 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...

  4. Krsko: A case study for western nuclear power issues

    International Nuclear Information System (INIS)

    Bizal, M.R.

    1996-01-01

    This article is a review of issues surrounding the operation of the Krsko reactor in the former Yugoslav state of Slovenia. This reactor, a Westinghouse-supplied 632-MWe PWR, supplies 25% of Slovenia's electrical energy and is therefore a major element in the country's energy mix. There are environmental issues and political issues noted, and the project is currently evaluating the need for costly steam generator replacement

  5. Belene NPP

    International Nuclear Information System (INIS)

    Tsvetanov, P.

    1990-01-01

    The book presents the main results of the studies of the Bulgarian Academy of Sciences (BAS) on the construction of a new nuclear power plant at Belene on the Danube river. The programme of the studies comprises five areas: the socio-economic and energy development and the necessity of the commissioning; a technical project and design level of the equipment (safety, radioactivity control, waste disposal and economic efficiency of the power plant); the seismic properties of the construction site; the corresponding risk and design features of the plant; the ecological impacts of the NPP and public opinion. The studies in the different areas have been carried out by independent teams, fully responsible for the formulated topical conclusions. The general opinion of the BAS voiced in the book is that the construction of Belene NPP is not sufficiently substantiated and is considered unacceptable. 94 refs., 53 fig., 56 tabs. (R.Ts.)

  6. Development and application of probabilistic safety assessment PSA in Daya Bay Nuclear Power Station

    International Nuclear Information System (INIS)

    Huang Weigang; Chen Jiefei; Guo Jianbing; Zhen Wei

    2005-01-01

    This paper discusses the development and application of Level 1 PSA used for safety review, risk monitoring and on line maintenance of the nuclear power plant. PSA development includes the analysis of event tree, fault tree, FMEA, PSA quantification and the equipment reliability database. We have collected and processed the reliability data of external power source, the equipment reliability data and the initial event frequency. The thermal-hydraulics analysis of some important events and accidents, human factor analysis, and the calculation of human error probability have been made. During the development of event trees and fault trees, the effect of some support systems such as compressed air distribution system, ventilation system and electrical system have been taken into account. The system manuals, operation procedures and emergency operating procedures of Daya Bay NPP are referred to in this project. The operators of the NPP were involved in the establishment of all event trees and fault trees analysis. Furthermore, we have accepted the suggestion of IAEA experts, completed the logic chart of initial events to the initial events analysis of Daya Bay NPP, and optimized the code system of PSA model again. Together with the development of the reliability database, by absorbing the advanced experience of EDF, we have gained the reports about equipment's classifying, function and experience feedback information of Daya Bay NPP. According to the quantitative calculation of the latest Level 1 PSA Model of Daya Bay NPP, the results of Core Damage Frequency (CDF) is: CDF = 2.13E-5/reactor . year. The latest PSA Model of Daya Bay NPP includes: 1) 12 sorts of initial events, 67 sub-initial events, 70 fault trees; 2) 25 nuclear safety related systems were developed by fault trees and FMEA ; 3) 2609 fault tree logic gates; 4) 2146 basic events; 5) 680 core damage accident sequences. (authors)

  7. Whooping Cough PSA (:30)

    Centers for Disease Control (CDC) Podcasts

    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.

  8. Review of UCN 5,6 Fire PSA Model based on ANS Fire PRA Standard

    International Nuclear Information System (INIS)

    Yang, Joon Eon; Lee, Yoon Hwan

    2006-12-01

    Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). This approach uses the fire risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In fire risk informed/performance-based decision/operation, fire PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of fire PSA. ANS (American Nuclear Society) has developed a guidance called 'ANS Fire PRA Methodology Standard'. However, in Korea, there have been no attempts to evaluate the quality of fire PSA model itself. Therefore, we cannot be sure about the quality of fire PSA whether or not the present fire PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of fire PSA model quality is the basis for the fire risk informed/performance-based decision/operation. In this report, we have evaluated the quality of fire PSA model for Ulchin 5 and 6 units based on the ANS Fire PRA Standard. We, also, have derived what items are to be improved to upgrade the quality of fire PSA model and how it can be improved. This report can be used as the base of the fire risk informed/performance-based decision/operation work in Korea

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

    International Nuclear Information System (INIS)

    Tankosic, Djurica; Fink, Kresimir

    1991-01-01

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

  10. 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)

  11. 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

  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. 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

  14. Motor-operated Valve Program at NPP Krsko (NEK) - Status and Overview

    International Nuclear Information System (INIS)

    Nikolic, M; Jagodar, N.; Cerjak, J.; Butkovic, V.

    2002-01-01

    On the basics of US NRC Generic Letter 89-10 Safety-related Motor-operated Valve Testing and Surveillance and subsequent generic letters, Motor-operated Valve (MOV) Program at NEK has been developing. Namely, the holders of nuclear power plant operating licenses has to verify the design basis capability of safety-related e.g. important-to-safety MOVs, as well as to ensure the same for the life of the plant. In light of that, each plant should establish a program to address stressed issues for each program MOV (124 at NEK). Such comprehensive task requires significant effort in many aspects, and basically multidisciplinary skills. NEK MOV Program represents a blend of engineering and in-plant testing, comprised of three phases: Phase I Engineering, Phase II Field Implementation and Phase III Trending. Currently, the program is about the end of Phase I and II, as well as in development of engineering basis for launching Phase III. Overview of the major programmatic issues will be given in this paper along with ongoing activities: testing process, gear-ratio modification, pressure locking/thermal binding susceptibility screening and preventive maintenance. (author)

  15. 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)

  16. 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)

  17. ASAMPSA-E guidance for level 2 PSA Volume 3. Verification and improvement of SAM strategies with L2 PSA

    International Nuclear Information System (INIS)

    Rahni, N.; Raimond, E.; Jan, P.; Lopez, J.; Loeffler, H.; Mildenberger, O.; Kubicek, J.; Vitazkova, J.; Ivanov, I.; Groudev, P.; Lajtha, G.; Serrano, C.; Zhabin, O.; Prosek, Andrej; Dirksen, G.; Yu, S.; Oury, L.; Hultqvist, G.

    2016-01-01

    For each NPP, severe accident management (SAM) strategies shall make use of components or systems and human resources to limit as far as possible the consequences of any severe accident on-site and off-site. L2 PSA is one of the tools that can be used to verify and improve these strategies. The present report (deliverable D40.5 of the project ASAMPSA-E) provides an opportunity for a comparison of objectives in the different countries in terms of SAM strategies verification and improvement. The report summarizes also experience of each partner (including potential deficiencies) involved in this activity, in order to derive some good practices and required progress, addressing: - SAM modeling in L2 PSA, - Positive and negative aspects in current SAM practices, - Discussion on possible criteria related to L2 PSA for verification and improvement: risk reduction (in relation with WP30 activities on risk metrics), reduction of uncertainties on the severe accident progression paths until NPP stabilization, reduction of human failure conditional probabilities (depending on the SAM strategy, the environmental conditions...), - Review with a perspective of verification and improvement of the main SAM strategies (corium cooling, RCS depressurization, control of flammable gases, reactivity control, containment function, containment pressure control, limitation of radioactive releases,...), - SAM strategies to be considered in the context of an extended L2 PSA (as far possible, depending on existing experience), taking into account all operating modes, accidents also occurring in the SFPs and long term and multi-unit accidents. The deliverable D40.5 is developed from the partners' experience. Many of the topics described here are beyond the common practices of L2 PSA applications: in some countries, L2 PSA application is limited to the calculations of frequencies of release categories with no formal requirement for SAM verification and improvement. (authors)

  18. Seismic Level 2 PSA

    International Nuclear Information System (INIS)

    Dirksen, Gerben; Pellissetti, Manuel; Duncan-Whiteman, Paul

    2014-01-01

    For most external events, the calculation of the core damage frequency (CDF) in Level 1 PSA is sufficient to be able to show that the contribution of the event to the plant risk is negligible. However, it is not sufficient to compare the CDF due to the external event to the total plant CDF; instead the Level 1 PSA result for the event should be compared to the large early release frequency (LERF), or alternatively arguments should be given why the CDF from the external event will not contribute mostly to LERF. For seismic events in particular, it can often not be easily excluded that sequences leading to core damage would not also result in LERF. Since the confinement function is one of the most essential functions for Level 2 PSA, special care must be taken of the containment penetrations. For example systems with containment penetrations that are normally closed during operation or are designed to withstand more than the maximum containment pressure are normally screened out in the Level 2 PSA for the containment isolation function, however the possibility of LOCA in such systems due to an earthquake may nevertheless lead to containment bypass. Additionally, the functionality of passive features may be compromised in case of a beyond design earthquake. In the present paper, we present crucial ingredients of a methodology for a Level 2 seismic PSA. This methodology consists of the following steps: Extension of the seismic equipment list (SEL) to include Level 2 PSA relevant systems (e.g. containment isolation system, features for core melt stabilization, hydrogen mitigation systems), Determination of the systems within the existing SEL with increased demands in case of severe accidents, Determination of essential components for which a dedicated fragility analysis needs to be performed. (author)

  19. Re-racking the spent fuel pit in nuclear power plant Krsko

    International Nuclear Information System (INIS)

    Volaric, B.; Krajnc, B.

    2003-01-01

    Krsko NPP was designed to temporary store a limited number of spent fuel assemblies (SFA). They were planned to be either removed for reprocessing or permanently stored. By the design the plant would run out of capacity for temporary storing of SFAs in 2003. This means that the plant could not operate further without additional changes since valid regulations require the Spent Fuel Pit (SFP) free storage capacity for the whole emergency core unloading (ECU). The purpose of the SFP Reracking Project is to assure a safe storage of all SFAs in the existing SFP during the plant lifetime. Design solutions and related analysis did not only consider the quantity of spent fuel for the plant lifetime, i.e. up to the year 2023, but also a possibility of the SFP extension for eventually extended plant lifetime of 20 years. According to the project, racks were designed to provide up to 1694 cells for storage of FAs in Phase I which is sufficient for normal NPP operation up to its lifetime, i.e. the year 2023. The extension of temporary storage capacity of SFAs in the SFP was made by a combination of existing and new high-density racks. Three modules from the existing 12 were eliminated. Thus leaving 621 cells. There are nine new racks added with up to 1073 supercompacted cells installed in the empty part of the pool. The spacing between new cells is smaller than the one between existing cells because of special plates, made of borated stainless steel. The design of the racks fulfils all the applicable requirements to ensure sub-criticality of all stored SFAs, enriched up to 5 % U 235 with peak pellet burnup above 40 GWD/MTU. Besides static and dynamic seismic loading of both, the racks and the fuel handling building, sufficient cooling of FAs is provided as well as accident conditions precluded. Greater cooling capacity is achieved by installing the third heat exchanger, connected in parallel to existing two exchangers. The design and manufacturing of the new heat exchanger

  20. 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)

  1. 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

  2. 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

  3. 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)

  4. Tools for PSA reviews

    International Nuclear Information System (INIS)

    Linden, J. von

    1998-01-01

    It is desirable to have a uniform and competent procedure for the review of PSAs which are performed within the framework of the Periodic Safety Review of German Nuclear Power Plants. Guidelines for the review process should therefore be evaluated within task A. 1 of project SR 2096. The basis for this work is the experience and knowledge within GRS derived from PSA-related work and from several review projects as well as the German PSA Guide with its appendices. Furthermore, the review processes in the USA, Switzerland and Sweden and the Guidelines for the International Peer Review Service (IPERS Guidelines) were utilized. As a result, recommendations are given for the review process, with individual recommendations concerning the organization of the review, task allocation between the reviewers, interface problems, assessment criteria, the scope and depth of the review as well as the supporting documents. An additional result are checklists for the technical elements of the PSA, which are listed to facilicate the review work. It is not the intention of this report to work out complete review guidelines. Its aims is rather more to give recommondations and support for the review in addition to what can be derived from the existing documents that should be used for the review. The recommendations reflect the view of GRS and go beyond the statements given in the German PSA Guide (Leitfaden Probabilistische Sicherheitsanalyse /PSUe97/) in some points. (orig.) [de

  5. Binge Drinking PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    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.

  6. 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

  7. 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

  8. 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.

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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

  14. Risk-informed decision making during Bohunice NPP safety upgrading

    International Nuclear Information System (INIS)

    Lipar, M.; Muzikova, E.; Kubanyi, J.

    2001-01-01

    The paper summarizes some facts of risk-informed regulation developments within UJD regulatory environment. Based on national as well as international operating experience and indications resulted from PSA, Nuclear Regulatory Authority of the Slovak Republic (UJD) since its constituting in 1993 has devoted an effort to use PSA technology to support the regulatory policy in Slovakia. The PSA is considered a complement, not a substitute, to the deterministic approach. Suchlike integrated approach is used in decision making processes and the final decision on scope and priorities is based on it. The paper outlines risk insights used in the decision making process concerning Bohunice NPP safety upgrading and focuses on the role of PSA results in Gradual Reconstruction of Bohunice VI NPP. Besides, two other examples of the PSA results application to the decision making process are provided: the assessment of proposal of modifications to the main power supply diagram (incorporation of generator switches) and the assessment of licensee request for motor generator AOT (Allowable Outage Time) extension. As an example of improving support of Bohunice V-2 risk-informed operations, concept of AOT calculations and Bohunice V-2 Risk Monitor Project are briefly described. (author)

  15. 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.

  16. 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.

  17. 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

  18. 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

  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. 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)

  1. 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

  2. Implementation guidelines for seismic PSA

    International Nuclear Information System (INIS)

    Coman, Ovidiu; Samaddar, Sujit; Hibino, Kenta; )

    2014-01-01

    The presentation was devoted to development of guidelines for implementation of a seismic PSA. If successful, these guidelines can close an important gap. ASME/ANS PRA standards and the related IAEA Safety Guide (IAEA NS-G-2.13) describe capability requirements for seismic PSA in order to support risk-informed applications. However, practical guidance on how to meet these requirements is limited. Such guidelines could significantly contribute to improving risk-informed safety demonstration, safety management and decision making. Extensions of this effort to further PSA areas, particularly to PSA for other external hazards, can enhance risk-informed applications

  3. Lesson learned from PSA in the design improvement program of KSNP+

    International Nuclear Information System (INIS)

    Kim, M.R.; Lim, H.K.; Kang, S.K.; Park, K.S.

    2001-01-01

    KOPEC (Korea Power Engineering Co.) in conjunction with the client KEPCO (Korea Electric Power Corp.) has been developing a highly competitive Improved Korean Standard Nuclear Power Plant named KSNP+. From the beginning of Design Improvement Program, PSA was carried out to assure that the safety level of KSNP+ is maintained or improved in comparison with that of KSNP, the Korean Standard NPP. To achieve the safety goal of KSNP+, PSA team reviewed all design changes that might affect the plant safety. Design vulnerabilities were identified from the PSA results and safety improvement items were recommended to the system designers. Through the Design Improvement Program, KSNP+ became more reliable, safer and economically competitive than KSNP. This was achieved by systematic approach for design optimization and effective use of PSA technology based on past experience and expertise of nuclear power plant. (author)

  4. 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)

  5. Review of KSNP LPSD PSA model based of ANS LPSD PRA standard, rev.0

    International Nuclear Information System (INIS)

    Jang, S. C.; Park, J. H.; Kim, T. W.; Lim, H. G.; Yang, J. E.; Ha, J. J.

    2004-02-01

    Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-informed In-service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, we cannot be sure about the quality of PSA whether or not the present PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of PSA model quality is the basis for the RIPBO. In this report, we have evaluated the quality of PSA model at Low power and Shutdown operation model for Yongkwang 5 and 6 units based on the ANS LPSD PRA Standard. We, also, have derived what items are to be improved to upgrade the quality of LPSD PSA model and how it can be improved. This report can be used as the base of RIPBO work in Korea

  6. Prostate-Specific Antigen (PSA) Test

    Science.gov (United States)

    ... Cancer Prostate Cancer Screening Research Prostate-Specific Antigen (PSA) Test On This Page What is the PSA ... parts of the body before being detected. The PSA test may give false-positive or false-negative ...

  7. Development of PSA workstation KIRAP

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. 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

  9. 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...

  10. 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)

  11. 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.)

  12. 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

  13. 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.

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

    International Nuclear Information System (INIS)

    Lee, Chan Young; Seong, Poong Hyun

    2016-01-01

    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

  15. NPP life management (abstracts)

    International Nuclear Information System (INIS)

    Litvinskij, L.L.; Barbashev, S.V.

    2002-01-01

    Abstracts of the papers presented at the International conference of the Ukrainian Nuclear Society 'NPP Life Management'. The following problems are considered: modernization of the NPP; NPP life management; waste and spent nuclear fuel management; decommissioning issues; control systems (including radiation and ecological control systems); information and control systems; legal and regulatory framework. State nuclear regulatory control; PR in nuclear power; training of personnel; economics of nuclear power engineering

  16. Determining if pretreatment PSA doubling time predicts PSA trajectories after radiation therapy for localized prostate cancer

    International Nuclear Information System (INIS)

    Soto, Daniel E.; Andridge, Rebecca R.; Pan, Charlie C.; Williams, Scott G.; Taylor, Jeremy M.G.; Sandler, Howard M.

    2009-01-01

    Introduction: To determine if pretreatment PSA doubling time (PSA-DT) can predict post-radiation therapy (RT) PSA trajectories for localized prostate cancer. Materials and methods: Three hundred and seventy-five prostate cancer patients treated with external beam RT without androgen deprivation therapy (ADT) were identified with an adequate number of PSA values. We utilized a linear mixed model (LMM) analysis to model longitudinal PSA data sets after definitive treatment. Post-treatment PSA trajectories were allowed to depend on the pre-RT PSA-DT, pre-RT PSA (iPSA), Gleason score (GS), and T-stage. Results: Pre-RT PSA-DT had a borderline impact on predicting the rate of PSA rise after nadir (p = 0.08). For a typical low risk patient (T1, GS ≤ 6, iPSA 10), the predicted PSA-DT post-nadir was 21% shorter for pre-RT PSA-DT 24 month (19 month vs. 24 month). Additional significant predictors of post-RT PSA rate of rise included GS (p < 0.0001), iPSA (p < 0.0001), and T-stage (p = 0.02). Conclusions: We observed a trend between rapidly rising pre-RT PSA and the post-RT post-nadir PSA rise. This effect appeared to be independent of iPSA, GS, or T-stage. The results presented suggest that pretreatment PSA-DT may help predict post-RT PSA trajectories

  17. 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

  18. Outlet from the condenser of nuclear power plant Krsko into Sava river

    International Nuclear Information System (INIS)

    Rek, Z.

    1990-01-01

    Paper deals with hot water outflow from condenser of the Nuclear power plant Krsko into river Sava. We are interested in temperature and velocity field along the river. Boundary-domain integral method is used to solve a system of conservative equations. As a result, the time development of the velocity and temperature field at nodes of discrete model is obtained. (author)

  19. 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.)

  20. Blackpool: More evidence on PSA

    International Nuclear Information System (INIS)

    Cullingford, M.

    1985-01-01

    PSA is spreading widely throughout the world, with 30 IAEA Member States having active programmes in this area. The main reason for its popularity is that it offers insights critical in the safety decision-making process available from no other method. It allows power plant designers, regulators, and operators to discriminate between issues important to safety and those which are trivial. Although it is beneficial to perform a PSA to utilize the potential products available from such a study, it is especially important to realize that the PSA process itself is a valuable experience. The main points such as potential users of PSA, safety evaluations, treatment of uncertainties as well as future trends are briefly discussed in this paper

  1. 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

  2. Human behaviour in PSA

    International Nuclear Information System (INIS)

    Schott, H.

    1999-01-01

    Based on the current international state of the art of methodology for evaluation of human errors for PSA, many research projects have been initiated by the competent departments of the BMU and the BfS (Federal Min. of the Environment and Reactor Safety, Federal Radiation Protection Office). Three major areas of the research activities are discussed: Database: - Specific investigations into the applicability of generic data (THERP) to other than the original cases, possibly elaboration of approaches for application-specific modification, further evaluation of operating results; - general enhancement of insight into human performance and errors, e.g. with respect to causes of error and application areas (influence of organisation, cognitive performance); interviews with experts as a supplementary approach for data verification and database enhancement. Sensitivity analysis: - Identification of information describing human errors essentially contributing to frequency of occurrence of incidents and system non-availability; - establishment of relevance rating system, methodology for uncertainty analysis. Further development of methodology: - Modelling of repair activities and knowledge-based behaviour. (orig./CB) [de

  3. 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

  4. 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

  5. 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

  6. Survey of Dynamic PSA Methodologies

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hansul; Kim, Hyeonmin; Heo, Gyunyoung [Kyung Hee University, Yongin (Korea, Republic of); Kim, Taewan [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-05-15

    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. NRC regulatory uses of PSA

    International Nuclear Information System (INIS)

    Murley, T.E.

    1991-01-01

    The publication in 1975 of WASH-1400, with its new probabilistic safety assessment (PSA) methodology, had the effect of presenting a pair of eyeglasses to a man with poor eyesight. Suddenly, it gave us a view of nuclear safety with a new clarity, and it allowed us to sort out the important safety issues from the unimportant. In the intervening years, PSA insights have permeated the fabric of nearly all our safety judgments. This acceptance can be seen from the following list of broad areas where the Nuclear Regulatory Commission (NRC) staff uses PSA insights and methodology: evaluating the safety significance of operating events and recommending safety improvements where warranted; requesting licensees to systematically look for design vulnerabilities in each operating reactor; evaluating the safety significance of design weaknesses or non-compliances when judging the time frame for necessary improvements; conducting sensitivity analyses to judge where safety improvements are most effective; assessing the relative safety benefits of design features for future reactors. In judging where PSA methodology can be improved to give better safety insights, it is believed that the following areas need more attention: better modeling of cognitive errors; more comprehensive modeling of accident sequences initiated from conditions other than full power; more comprehensive modeling of inter-system loss of coolant accident (ISLOCA) sequences. Although PSA is widely used in the staff's regulatory activities, the NRC deliberately chooses not to include probabilistic prescriptions in regulations or guidance documents. The staff finds the bottom line risk estimates to be one of the least reliable products of a PSA. The reason for this view is that PSA cannot adequately address cognitive errors nor assess the effects of a pervasive poor safety attitude

  8. 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

  9. Review of UCN 3,4 PSA model based on NEI PRA peer review process guidance, rev.0

    International Nuclear Information System (INIS)

    Yang, Joon Eon; Kang, D. I.; Kim, K. Y.; Lee, Y. H.; Jang, S. C.; Ha, J. J.; Han, S. H.; Han, S. J.; Hwang, M. J.

    2003-05-01

    Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-Based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-Informed In-Service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, we cannot be sure about the quality of PSA whether or not the present PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of PSA model quality is the basis for the RIPBO. In this report, we have evaluated the quality of PSA model for Ulchin 3 and 4 units based on the NEI guidance. We, also, have derived what items are to be improved to upgrade the quality of PSA model and how it can be improved. This report can be used as the base of RIPBO work in Korea. The review result based on ASME Standard is published as the separated technical report of KAERI

  10. 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)

  11. Level 1 shutdown and low power operation of Mochovce NPP, Unit 1, Slovakia

    International Nuclear Information System (INIS)

    Halada, P.; Cillik, I.; Stojka, T.; Kuzma, M.; Prochaska, J.; Vrtik, L.

    2004-01-01

    The paper presents general approach, used methods and form of documentation of the results that have been applied within the shutdown and low power PSA (SPSA) study for Mochovce NPP, Unit 1, Slovakia. The SPSA project was realized by VUJE Trnava Inc., Slovakia in 2001-2002 years. The Level 1 SPSA study for Mochovce NPP Unit 1 covers internal events as well as internal (fires, floods and heavy load drop) and external (aircraft crash, extreme meteorological conditions, seismic event and influence of surrounding industry) hazards. Mochovce NPP consists of two operating units equipped with VVER 440/V213 reactors safety upgraded before construction finishing and operation start. 87 safety measures based on VVER 440 operational experience and international mission insights were implemented to enhance its operational and nuclear safety. The SPSA relates to full power PSA (FPSA) as a continuation of the effort to create a harmonized level 1 PSA model for all operational modes of the plant with the goal to use it for further purposes as follows: Real Time Risk Monitor, Maintenance Optimization, Technical Specifications Optimization, Living PSA. (author)

  12. 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

  13. 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

  14. Use of CEDB for PSA

    International Nuclear Information System (INIS)

    Balesteri, S.; Besi, A.; Carlesso, S.; Colombo, A.G.; Jaarsma, R.J.

    1987-01-01

    The Component Event Data Bank (CEDB) is a centralized bank collecting, at the European level, data describing the operational behaviour of components of Nuclear Power Plants (NPP's) operating in various European countries. It is one of the three event data banks of the European Reliability Data System (ERDS). The CEDB stores information on the operational history (operational times and/or number of demands of intervention in a year, failure-events reports) of components of NPP's well identified by their engineering and operation characteristics. The CEDB (as well as the whole of the ERDS) was conceived as a support to the analyst in his safety assessments for the design of a new NPP or the backfitting of an old one. (orig./HSCH)

  15. The application of model with lumped parameters for transient condition analyses of NPP

    International Nuclear Information System (INIS)

    Stankovic, B.; Stevanovic, V.

    1985-01-01

    The transient behaviour of NPP Krsko during the accident of pressurizer spray valve stuck open has been simulated y lumped parameters model of the PWR coolant system components, developed at the faculty of Mechanical Engineering, University of Belgrade. The elementary volumes which are characterised by the process and state parameters, and by junctions which are characterised by the geometrical and flow parameters are basic structure of physical model. The process parameters obtained by the model RESI, show qualitative agreement with the measured valves, in a degree in which the actions of reactor safety engineered system and emergency core cooling system are adequately modelled; in spite of the elementary physical model structure and only the modelling of thermal process in reactor core and equilibrium conditions of pressurizer and steam generator. The pressurizer pressure and liquid level predicted by the non-equilibrium pressurizer model SOP show good agreement until the HIPS (high pressure pumps) is activated. (author)

  16. 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)

  17. 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.)

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. Methods and use of PSA

    International Nuclear Information System (INIS)

    Vaurio, J.

    2005-01-01

    The objective is to introduce possible methods and methodological issues for application of PSA for decision making at nuclear power plants based on risk and cost, to present a number of application examples, and to describe recommendations and regulatory guides for risk-informed applications. (orig.)

  3. 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)

  4. 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

  5. State of living PSA and further development

    International Nuclear Information System (INIS)

    1999-01-01

    In October 1985 OECD-Principal Working Group (PWG 5) - Risk Assessment has initiated the Task Force 7 'Use of PSA in Nuclear Power Plant Management' to explore and report on the principles, characteristics, requirements and status of PSA oriented safety management. During this study, it became apparent that the utilisation of PSA techniques in nuclear plant safety management requires the development of supporting programmes to ensure that PSA models are being updated to reflect plant changes, and to direct their use towards the evaluation and determination of plant changes. These requirements also influence the software and hardware characteristics necessary to support the programme. This overall process is known as Living PSA. In this context OECD-PWG 5 has arranged international workshops on Living PSA application to support this development, to facilitate exchange of international experience and to summarise the state-of-the-art of L-PSA methodology. These activities were accompanied by following Task Groups of OECD-PWG 5 and the work results were published in state-of-the-art reports. According to the increasing development of Living PSA in the international field and its capacity to support plant safety management in a broad sense, OECD PWG 5 continues its work in setting up the Task Group 96-1 'State of Living PSA and Further Development' to clarify specific aspects of Living PSA. This report summarises the state of Living PSA in the international field based on the four Living PSA Workshops from 1988 to 1994 (Chapter 2) and the state of Reliability Data Collection based on the results of Task Group 12 'Reliability Data Collection and Analysis to Support PSA' and the two Data-Workshops from 1995 and 1998 (Chapter 3). The specific items of further development of Living PSA application as mentioned above are treated in Chapter 4. Chapter 5 gives a summary of the current state of Living PSA as well as outlook and recommendations of further development

  6. 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.

  7. 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)

  8. Methodology for seismic PSA of NPPs

    International Nuclear Information System (INIS)

    Jirsa, P.

    1999-09-01

    A general methodology is outlined for seismic PSA (probabilistic safety assessment). The main objectives of seismic PSA include: description of the course of an event; understanding the most probable failure sequences; gaining insight into the overall probability of reactor core damage; identification of the main seismic risk contributors; identification of the range of peak ground accelerations contributing significantly to the plant risk; and comparison of the seismic risk with risks from other events. The results of seismic PSA are typically compared with those of internal PSA and of PSA of other external events. If the results of internal and external PSA are available, sensitivity studies and cost benefit analyses are performed prior to any decision regarding corrective actions. If the seismic PSA involves analysis of the containment, useful information can be gained regarding potential seismic damage of the containment. (P.A.)

  9. 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.

  10. 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)

  11. PSA based vulnerability and protectability analysis for NPPs

    International Nuclear Information System (INIS)

    Gopika, V.; Sanyasi Rao, V.V.S.; Ghosh, A.K.; Kushwaha, H.S.

    2012-01-01

    Highlights: ► The paper describes the generation of location sets and protection sets. ► Vulnerability and protectability used to rank location sets and protection sets. ► Ranking helps in adequacy of protection measures employed in various locations. ► The procedure for PSA based vital area identification is demonstrated. ► This method has found practical applicability for Indian NPP. -- Abstract: Identification of vital areas in a facility involves assessing the facility and the locations, whose sabotage can result in undesirable (radiological) consequences. Probabilistic Safety Assessment (PSA) technique can find the component failures leading to core damage (a surrogate for radiological consequence) in a systematic manner, which can be extended to identification of vital areas. This paper describes the procedure for the generation of location sets (set of locations whose sabotage can lead to possible core damage) and protection sets (set of locations that must be protected to prevent possible core damage). In addition, measures such as vulnerability and protectability have been introduced, which can be used to rank location sets and protection sets.

  12. Too Much Sodium PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    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.

  13. 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.)

  14. 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

  15. Dukovany NPP maintenance management

    International Nuclear Information System (INIS)

    Siegel, F.

    2005-01-01

    Maintenance planning and management for the technological equipment of the Dukovany NPP are described. A Maintenance Control and Evaluation System has been developed and is in use to help manage the complex maintenance issue. Practical examples of outputs of the System, with a comprehensive use of the cost, reliability and safety related data, are presented. (author)

  16. 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 ...

  17. Two managerial grids in NPP

    International Nuclear Information System (INIS)

    Zhao Hui

    2012-01-01

    Today, the nuclear power corporation (NPC) enjoys the profit of LCEP (the low carbon economic policy). at the same time, they also enduring more and more pressure. For example, the partner competition or the NPP potential occupational risk . The efficient counterplot of risk is the self-ability cultivation. It is essential to research the NPP managerial flow. The nuclear power plant (NPP) unit is a carrier of the NPC enterprise management system, and has taken on a new look 'pull one portion then the whole moving'. The NPP has three systematical characters, the security responsibility center, the man-machine system and the input-output system. The manufacturing system and the enterprise management system are the great constituents of the NPP managerial flows. Means of systems analysis, we can find out the truth of the NPP running interface. In CHINA, there are many operating experiences near 20 years. It indicates that the NPP manufacturing system and the enterprise system are the roots of the nuclear power corporation, the core of the all NPP systems must be based on it. So the ability cultivation is the work core to NPP. It is reliably to ensure the NPP to be up against problems, for instance, the security duty, the costing control and the man-machine system running harmoniously. This paper introduces the NPP managerial flow and the present state of QNPC, also come up with a proposal to refer for the NPC development actions of collective measure, specialization, standardization, fine. (author)

  18. 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

  19. Use of PSA in a regulatory framework

    International Nuclear Information System (INIS)

    Ross, P.J.

    1994-01-01

    The paper will briefly describe the use of PSA in the licensing process for the Sizewell 'B' PWR Power Station currently under construction in the U.K. There are two distinct phases in the licensing process - (i) A PSA has been performed to support the application to construct Sizewell 'B'. At that stage the PSA was used as a design tool (along with deterministic design requirements) for Sizewell 'B' and as such lead to a number of significant design changes in the early design process. (ii) A PSA is currently being performed to support the application to operate Sizewell 'B'. The PSA is required to support the claim that the design has included all reasonably practical measures to prevent and mitigate accidents. The comprehensive PSA being produced for the second phase of the licensing process will be described. The way the regulators/designer/analysts have interacted over the years has affected the scope, complexity, detail and bias of the comprehensive PSA. The paper will discuss these issues and highlight some of the more significant ones. The benefits and drawbacks of providing a PSA in a regulatory framework will be discussed. One of the conclusions of the paper is that the use of true ''best-estimates'' in the PSA is difficult to achieve in a regulatory framework where persistent bias to the conservative side is apparent in the designers, analysts and regulators judgements. The usefulness of the PSA is therefore, potentially, compromised by giving misleading outputs or diverting resources to unnecessary areas. (author)

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

    International Nuclear Information System (INIS)

    Han, Sang Hoon; Lim, Hogon; Ahn, Kwang Il

    2014-01-01

    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

  1. 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...

  2. 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.

  3. 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

  4. PSA, subjective probability and decision making

    International Nuclear Information System (INIS)

    Clarotti, C.A.

    1989-01-01

    PSA is the natural way to making decisions in face of uncertainty relative to potentially dangerous plants; subjective probability, subjective utility and Bayes statistics are the ideal tools for carrying out a PSA. This paper reports that in order to support this statement the various stages of the PSA procedure are examined in detail and step by step the superiority of Bayes techniques with respect to sampling theory machinery is proven

  5. Probabilistic safety assessment for Balakovo 1000 MW NPP

    International Nuclear Information System (INIS)

    Foden, R.W.

    1995-01-01

    In July 1993 the Commission of the European Communities (CEC) placed a contract with NNC Ltd (National Nuclear Corporation) for performing a Probabilistic Safety Assessment (PSA) for a 1000 MW NPP in the Russian Federation. The contract is part (Project 3.1) of the 1991 TACIS (Technical Assistance to the CIS) programme. This paper describes the objectives and scope of the Project and provides a description of the progress that has been made. For this Project, NNC is the leader of a Consortium of Western European companies that has been formed to undertake this Project and other Projects in the TACIS 91 programme. NNC therefore has overall responsibility for the coordination and management of the complete PSA Project. Other members of the Consortium involved in this Project are Empresarios Agrupados from Spain, Belgatom from Belgium and AEA-Technology from the UK. The analytical work for the Project is performed by the Russian Company Atomenergoproekt in Moscow, under contract to NNC. The official recipient institution for the results of the Project is the Russian Utility, Rosenergatom. The NPP chosen to be the subject of the Project is the Balakovo Unit 4 VVER 1000. (author)

  6. Status and use of PSA in Sweden

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. 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

  9. 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)

  10. 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.

  11. Outages 1999 and 2000, investments in safety and long-term operation of NE Krsko

    International Nuclear Information System (INIS)

    Sirola, P.; Krajnc, J.; Androjna, F.

    1999-01-01

    Plant outage is an important part of nuclear power plant operation. During that time the conditions are established for the performance of specific activities, such as refueling, tests, inspections, preventive and corrective maintenance and modifications, that are intended to confirm proper condition and availability of safety and other important components and improve overall plant safety and reliability. It is well know that in Nuclear Power Plant Krsko (Nuklearna elektrarna Krsko NEK) during Outage 2000 new Steam Generators (SGs) will be placed in service, while Outage '99 was used for preparatory works. But the importance of those two outages is even greater, because they are implementing a broad number of improvements and establishing a basis for long-term plant operation. Outage '99 required very detailed planning to assure a good control over the outage activities and operational plant systems necessary for safe shutdown. Numerous activities took place in a relatively narrow space in the Reactor Building. Some of these activities will have a big significance for the future. The article treats the status update and summarizes the specifics and importance of the mentioned activities to long-term plant safe and reliable operation.(author)

  12. Transition from 12 months to 18 months cycles at Krsko in a core physics perspective

    International Nuclear Information System (INIS)

    Jensen-Tornehed, J.

    2004-01-01

    Krsko has historically been operating in 12 months cycles with an annual outage for refuelling and maintenance work. Krsko is now in a transition from 12 months to 18 months cycles. Cycle 19, June 2002 - May 2003, was the last 12 month cycle. Cycle 20, June 2003 - September 2004 is a 15 month transition cycle and cycle 21, September 2004 - April 2006 will be the first 18 month cycle. This paper will describe the effects of the transition in a core physics perspective. There are big differences in how to design an 18 month cycle in comparison with a 12 month cycle. The required number of feed assemblies increases, as well as the content of burnable absorbers in the fuel. The strategy of the loading pattern has to be changed with the increased number of fresh fuel assemblies. The most limiting margins can be different for different cycle lengths which also affect the fresh assembly design and loading pattern during the transition. During the core design for cycle 21 the Moderator Temperature Coefficient was the main issue, which caused the need for extra amount of burnable absorbers. (author)

  13. 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.

  14. 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

  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. PSA Level 2:Scope And Method Of PSA Level 2 For Nuclear Power Plant

    International Nuclear Information System (INIS)

    Widodo, Surip; Antariksawan, Anhar R.

    2001-01-01

    A study of scope and method of PSA Level 2 had been conducted. The background of the study is the need to gain the capability to well perform PSA Level 2 for nuclear facilities. This study is a literature survey. The scope of PSA Level 2 consists of generating plant damage states, accident progression analysis, and grouping source terms. Concerning accident progression analysis, several methods are used, among others event tree method, named accident progression event tree (APET) or containment event tree (CET), and fault tree method. The end result of PSA Level 2 is release end states which is grouped into release bins. The results will be used for PSA Level 3

  17. [PSA interest and prostatitis: literature review].

    Science.gov (United States)

    Bruyère, F; Amine Lakmichi, M

    2013-12-01

    Prostatitis is easily diagnosed but sometimes associated with PSA measurement. An increased PSA in an asymptomatic patient may be associated with antibiotic use to eliminate the inflammatory part and to confirm prostate biopsy. It seems interesting to confirm or infirm these attitudes with a systematic review of the literature We performed a literature review using the words [prostatitis], [acute prostatitis], [prostate specific antigen], [PSA], in the MEDLINE, Pubmed and AMBASE database searching for articles in French or English published in the past 20 years. PSA is not always increased during an acute prostatitis episode. An increased PSA in an asymptomatic man does not seem to be systematically correlated to prostate inflammation. Analyzing the studies, it seems inaccurate to measure PSA value during a febrile urinary infection episode in men. Systematic use of antibiotic to decrease PSA and not performing prostate biopsy is not relevant and may induce resistance to antibiotic and doesn't induce a reduction risk of having prostate biopsy. PSA is unnecessary in case of febrile urinary tract infection in men. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

  18. 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)

  19. Enhancing NPP Safety Through an Effective Dependability Management

    Energy Technology Data Exchange (ETDEWEB)

    Vieru, G., E-mail: g_vieru@yahoo.com [AREN, Bucharest (Romania)

    2014-10-15

    Taking into account the importance of the continuous improvement of the performance and reliability of a NPP and practical measures to strengthen nuclear safety and security, it is to be noted that a good management for a nuclear power reactor involves a ''good dependability management'' of the activities, such as: Reliability, Availability, Maintainability (RAM) and maintenance support. In order to evaluate certain safety assessment criteria intended to be applied at the level of the nuclear reactor unit management, equipment dependability indicators and their impact over the availability and reactor safety have to be evaluated. Reactor equipment dependability indicators provide a quantitative indication of equipment RAM performances (Reliability, Availability and Maintenance). One of the important benefits of maintenance and failure data gathering is that it can be used as a support of probabilistic safety assessment (PSA). Also, a good dependability management implementation may be used to complement reactor level unit performance indicators in the field of safe operation, maintenance and improving operating parameters, as well as for Strengthening Safety and Improving Reliability of a NPP. This paper underlines the importance of nuclear safety and security as prerequisites for nuclear power. In addition, it demonstrates how different technical aspects, through implementation of a good dependability management, contribute to a strengthened safety and an improvement of availability of the NPP through dependability indicators determination and evaluation. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sook Kwan; Park, Seong Gyu; Han, Sang Koo [ACT Co., Daejeon (Korea, Republic of)

    2016-10-15

    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.

  1. 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)

  2. 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.)

  3. 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

  4. 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

  5. Preparation for Ignalina NPP decommissioning

    International Nuclear Information System (INIS)

    Medeliene, D.

    2004-01-01

    Latest developments of atomic energy in Lithuania, works done to prepare Ignalina NPP for final shutdown and decommissioning are described. Information on decommissioning program for Ignalina NPP unit 1, decommissioning method, stages and funding is presented. Other topics: radiation protection, radioactive waste management and disposal. Key facts related to nuclear energy in Lithuania are listed

  6. 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)

  7. 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)

  8. Radiometric assays for the measurement of PSA

    International Nuclear Information System (INIS)

    Venkatesh, M.

    1997-01-01

    Prostate Specific Antigen, a serine protease enzyme, of M.W. ∼ 26-33 kDa, is widely considered to be a very useful marker for prostate cancer. It satisfies nearly all the requirements of an ideal 'Tumour Marker' and has hence attracted a lot of attention in the past decade. PSA is present in multiple forms in serum, with an appreciable fraction bound to the protease inhibitor α-1-antichymotrypsin (ACT) and to a small extent to other proteins such as α-2-macroglobulin (AMG) leaving the rest in the free form. The total PSA levels have been reported to have 80% sensitivity and 60% specificity towards the detection of prostate cancer. The lack of specificity occurs mainly due to the high levels of t-PSA in benign prostatic hypertrophy(BPH) apart from the cancer. The concept of free PSA has been introduced in the recent past and the ratio of free/total PSA levels have been shown to be advantageous in the differential diagnosis of BPH from prostate cancer. The f/t ratio is considered to be particularly useful in the grey zones of decision making (t-PSA levels 4-20 ng/mL). The need for the development of assays for total and free PSA is felt due to: a. the high incidence of prostate cancers being detected currently; b. the high cost of tests (higher for free PSA assay, and the cost becomes an important parameter when a patient has to be regularly monitored after therapy) that is not affordable for many patients; c. the potential for research in the area of prostate cancer management where the PSA (total and free) assays will be of great help

  9. 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

  10. PSA as a tool for decision making

    International Nuclear Information System (INIS)

    Niehaus, F.; Lederman, L.

    1986-01-01

    The question on ''How safe is safe enough'' is being responded presently by deterministic criteria. Probabilistic criteria in support to more rational and less emotional decisions in regulatory and licensing issues, rationalization of resource allocation and research prioritization, among others, have a potential which is only marginally being explored. This paper discussed PSA limitations and proposes three areas for the use of PSA in decision making, namely: preventing accidents, mitigating accidents, and defining regulatory requirements. Current activities of the International Atomic Energy Agency in these areas are mentioned. PSA studies depict clearly the uncertainties and this is viewed as a positive aspect, which is unique to the use of probabilistic methods. (orig.)

  11. PSA as a tool for decision making

    Energy Technology Data Exchange (ETDEWEB)

    Niehaus, F; Lederman, L

    1986-05-01

    The question on ''How safe is safe enough'' is being responded presently by deterministic criteria. Probabilistic criteria in support to more rational and less emotional decisions in regulatory and licensing issues, rationalization of resource allocation and research prioritization, among others, have a potential which is only marginally being explored. This paper discussed PSA limitations and proposes three areas for the use of PSA in decision making, namely: preventing accidents, mitigating accidents, and defining regulatory requirements. Current activities of the International Atomic Energy Agency in these areas are mentioned. PSA studies depict clearly the uncertainties and this is viewed as a positive aspect, which is unique to the use of probabilistic methods.

  12. 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)

  13. 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)

  14. 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)

  15. 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)

  16. 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)

  17. 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)

  18. Issues reporting PSA in prostate cancer

    International Nuclear Information System (INIS)

    Lange, Paul H.

    1996-01-01

    The National Cancer Institute Prostate; Lung; Colon; Ovarian Cancer Screening (PLCO) project is a multi-center trial developed to investigate the effectiveness of DRE and PSA testing in the early detection and outcome of patients with prostate cancer. Accordingly, the Prostate Cancer Intervention versus Observation Trial (PIVOT) has been launched and is a randomized trial comparing radical prostatectomy versus expectant management for ALCaP. PSA: Initially PSA was thought to be of little value for diagnosis because 20% of men undergoing radical prostatectomy have 'normal' PSA and patients with apparently only symptomatic BPH have 'elevated' levels as follows: 4-10 ng/ml (Tandem-R) - 20%, >10 ng/ml -3%. Yet, PSA has looked attractive as a diagnostic tool in many studies; for example, when PSA was used in a screening approach as the first test which then drove further evaluation (Catalona, Brawer). It was shown that the positive predictive value for PSA's between 4 and 10 is approximately 20% and > 10 approximately 55%. The value of serial PSA's (velocity) is unknown but is under intense study: one major issue is determination of what represents a significant rise (details to be presented). Studies have also revealed that a DRE and PSA are important for optimal results. About 18% of clinically detectable cancers are only DRE positive while about 25 - 30% are only PSA positive. When both a DRE and PSA are used together, very few clinically apparent cancers are missed (3-5%). Recent ROC curves suggest that 4 ng/ml is reasonable. Recently, PSA values for men without apparent cancer were stratified by age, and taking the 2SD, age specific reference values were generated as follows: age 40-49 (0-2.5 ng/ml), 50-59 (0-3.5), 60-69 (0-4.5), 70-70 (0-6.5). Finally, there is the issue about different PSA assays regarding the compatabilities/reliability of the upper limit of normal and serial values. Much of the confusion is because there is no international PSA standard and

  19. 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

  20. 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

  1. More Adults Are Walking PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    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.

  2. Methodology for fire PSA during design process

    International Nuclear Information System (INIS)

    Kollasko, Heiko; Blombach, Joerg

    2009-01-01

    Fire PSA is an essential part of a full scope level 1 PSA. Cable fires play an important role in fire PSA. Usually, cable routing is therefore modeled in detail. During the design of new nuclear power plants the information on cable routing is not yet available. However, for the use of probabilistic safety insights during the design and for licensing purposes a fire PSA may be requested. Therefore a methodology has been developed which makes use of the strictly divisional separation of redundancies in the design of modern nuclear power plants: cable routing is not needed within one division but replaced by the conservative assumption that all equipment fails due to a fire in the concerned division; critical fire areas are defined where components belonging to different divisions may be affected by a fire. For the determination of fire frequencies a component based approach is proposed. The resulting core damage frequencies due to fire are conservative. (orig.)

  3. A preliminary investigation of PSA validation methods

    Energy Technology Data Exchange (ETDEWEB)

    Unwin, S D [Science Applications International Corp., (United States)

    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.

  4. 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

  5. PDS4 Challenges in the PSA

    Science.gov (United States)

    Saiz, J.; Barbarisi, I.; Docasal, R.; Rios, C.; Montero, A.; Macfarlane, A.; Laantee, C.; Besse, S.; Vallat, C.; Marcos, J.; Arenas, J.; Osinde, J.; Arviset, C.

    2018-04-01

    The Planetary Science Archive (PSA) stores products from all planetary ESA missions. Adopting PDS4 as the standard for new missions, while being compatible with existing PDS3 products, has driven a design with several difficulties to overcome.

  6. 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)

  7. 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)

  8. 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

  9. Technology for NPP decantate treatment realized at Kola NPP

    International Nuclear Information System (INIS)

    Stakhiv, Michael; Avezniyazov, Slava; Savkin, Alexander; Fedorov, Denis; Dmitriev, Sergei; Kornev, Vladimir

    2007-01-01

    At Moscow SIA 'Radon' jointly with JSC 'Alliance Gamma', the technology for NPP Decantate Treatment was developed, tested and realized at Kola NPP. This technology consists of dissolving the salt residue and subsequent treatment by ozonization, separation of the deposits formed from ozonization and selective cleaning by ferro-cyanide sorbents. The nonactive salt solution goes to an industrial waste disposal site or a repository specially developed at NPP sites for 'exempt waste' products by IAEA classification. This technology was realized at Kola NPP in December 2006 year. At this time more than 1000 m 3 of decantates log time stored are treated. It allows solving very old problem to empty decantates' tanks at NPPs in environmentally safe manner and with high volume reduction factor. (authors)

  10. NPOESS Preparatory Project (NPP) Science Overview

    Science.gov (United States)

    Butler, James J.

    2011-01-01

    NPP Instruments are: (1) well understood thanks to instrument comprehensive test, characterization and calibration programs. (2) Government team ready for October 25 launch followed by instrument activation and Intensive Calibration/Validation (ICV). NPP Data Products preliminary work includes: (1) JPSS Center for Satellite Applications and Research (STAR) team ready to support NPP ICV and operational data products. (2) NASA NPP science team ready to support NPP ICV and EOS data continuity.

  11. Review of APR+ Level 2 PSA

    International Nuclear Information System (INIS)

    Lehner, J.R.; Mubayi, V.; Pratt, W.T.

    2012-01-01

    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 and Nuclear Power Co., Ltd (KHNP) and KEPCO Engineering and Construction Co., Inc. (KEPCO-E and 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 and 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 and 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.

  12. 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)

  13. 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

  14. 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

  15. Development of seismic PSA methodology at JAERI

    International Nuclear Information System (INIS)

    Muramatsu, K.; Ebisawa, K.; Matsumoto, K.; Oikawa, T.; Kondo, M.

    1995-01-01

    The Japan Atomic Energy Research Institute (JAERI) is developing a methodology for seismic probabilistic safety assessment (PSA) of nuclear power plants, aiming at providing a set of procedures, computer codes and data suitable for performing seismic PSA in Japan. In order to demonstrate the usefulness of JAERI's methodology and to obtain better understanding on the controlling factors of the results of seismic PSAs, a seismic PSA for a BWR is in progress. In the course of this PSA, various improvements were made on the methodology. In the area of the hazard analysis, the application of the current method to the model plant site is being carried out. In the area of response analysis, the response factor method was modified to consider the non-linear response effect of the building. As for the capacity evaluation of components, since capacity data for PSA in Japan are very scarce, capacities of selected components used in Japan were evaluated. In the systems analysis, the improvement of the SECOM2 code was made to perform importance analysis and sensitivity analysis for the effect of correlation of responses and correlation of capacities. This paper summarizes the recent progress of the seismic PSA research at JAERI with emphasis on the evaluation of component capacity and the methodology improvement of systems reliability analysis. (author)

  16. 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.)

  17. Shock absorber in Ignalina NPP

    International Nuclear Information System (INIS)

    Bulavas, A.; Muralis, J.

    1996-09-01

    Theoretical calculation and experimental analysis of models of shock absorber in Ignalina NPP is presented. The results obtained from the investigation with model of shock absorber coincide with the theoretical calculation. (author). 2 figs., 3 refs

  18. Dukovany NPP - Safely 16 TERA

    International Nuclear Information System (INIS)

    Vlcek, J.

    2008-01-01

    In this presentation increasing of power output of the Dukovany NPP is reviewed. To operate all Dukovany Units safely with the perspective of long-term operation (LTO) of 50 - 60 years it is proposed.

  19. 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

  20. 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

  1. Assessment of core damage frequency owing to possible fires at NPP with RBMK type reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vinnikov, B. [National Research Centre Kurchatov Inst., 1, Kurchatov Square, Moscow, 123 182 (Russian Federation); NRC Kurchatov Inst. (Russian Federation)

    2012-07-01

    According to Scientific and Technical Cooperation between the USA and Russia in the field of nuclear engineering the Idaho National Laboratory has transferred to the possession of the National Research Center ' Kurchatov Inst. ' the SAPHIRE software without any fee. With the help of the software Kurchatov Inst. developed a Pilot Living PSA- Model of Leningrad NPP Unit 1. Computations of core damage frequencies were carried out for additional Initiating Events. In the submitted paper such additional Initiating Events are fires in various compartments of the NPP. During the computations of each fire, structure of the PSA - Model was not changed, but Fault Trees for the appropriate systems, which are removed from service during the fire, were changed. It follows from the computations, that for ten fires Core Damaged Frequencies (CDF) are not changed. Other six fires will cause additional core damage. On the basis of the calculated results it is possible to determine a degree of importance of these fires and to establish sequence of performance of fire-prevention measures in various places of the NPP. (authors)

  2. Assessment of core damage frequency owing to possible fires at NPP with RBMK type reactors

    International Nuclear Information System (INIS)

    Vinnikov, B.

    2012-01-01

    According to Scientific and Technical Cooperation between the USA and Russia in the field of nuclear engineering the Idaho National Laboratory has transferred to the possession of the National Research Center ' Kurchatov Inst. ' the SAPHIRE software without any fee. With the help of the software Kurchatov Inst. developed a Pilot Living PSA- Model of Leningrad NPP Unit 1. Computations of core damage frequencies were carried out for additional Initiating Events. In the submitted paper such additional Initiating Events are fires in various compartments of the NPP. During the computations of each fire, structure of the PSA - Model was not changed, but Fault Trees for the appropriate systems, which are removed from service during the fire, were changed. It follows from the computations, that for ten fires Core Damaged Frequencies (CDF) are not changed. Other six fires will cause additional core damage. On the basis of the calculated results it is possible to determine a degree of importance of these fires and to establish sequence of performance of fire-prevention measures in various places of the NPP. (authors)

  3. 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)

  4. 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

  5. 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.

  6. 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

  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. Current status of low power/shutdown PSA and accident sequence analysis for loss of RHR during mid-loop operation

    International Nuclear Information System (INIS)

    Park, Chang Kyu; Choi, Young; Kim, Tae Woon; Jin, Young Ho

    1994-07-01

    Probabilistic safety assessment (PSA) has been applied to only full-power operation of nuclear power plant (NPP), but some events which were recently occurred could reach severe plant damage state. Thus, various countries around the world have focused their interests on the evaluation for low power/shutdown (LP/S) operation. This report covers the main stream of LP/S PSA methodology, current status of LP/S PSA practices and results, and accident sequence analysis for loss of RHR during mid-loop operation. Therefore this report would be helpful for us to practice LP/S PSA for YGN 5,6 NPP which will be built in the near future. Also the results of accident sequence analysis show that operator's mis-diagnosis and failure of recovery action would initiate core damage during LP/S operation. In summary, overall environmental improvements (equipments, procedures, Tech Spec, etc, ...) and operating support system will be very useful to reduce risk during LP/S operation. (Author) 5 figs., 9 tabs

  9. Correlative study of SPECT bone scan, serum tPSA and fPSA/tPSA ratio and the pathological grade of prostate cancer with bone metastasis

    International Nuclear Information System (INIS)

    Xu Haiqing; Duan Jun

    2011-01-01

    Objective: To study the rules and characteristics of SPECT bone scan, serum TPSA, fPSA/tPSA ratio and the pathological grade of prostate cancer with bone metastasis. Methods: Nuclear medicine SPECT bone scan as the gold standard, retrospective analysis of the in vitro radioimmunoassay in 107 patients with prostate cancer serum PSA (prostate specific antigen) levels, serum fPSA/tPSA ratio and whole body bone imaging studies and pathological classification. Results: 107 patients with prostate cancer : 49 patients had bone metastases, accounting for 45.8% (49/107), in which groups of different pathological comparison between the incidence of bone metastasis significantly, the lower the degree of differentiation, the more the incidence of bone metastases high; with elevated levels of tPSA, the incidence of bone metastasis increased significantly; serum tPSA 4 - 40 ng/ml, the use of fPSA/tPSA ratio may improve the diagnostic specificity of prostate cancer. Conclusion: Patients with bone metastases of prostate cancer incidence and degree of differentiation of prostate cancer, serum PSA levels and fPSA/tPSA ratio of a certain relationship. The lower degree of differentiation,the higher the incidence of bone metastasis. (authors)

  10. 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.

  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. 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.

  13. 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

  14. The application of model with lumped parameters for transient condition analyses of NPP; Primena modela sa koncentrisanim parametrima za analize pelaznih stanja nukleane elektrane

    Energy Technology Data Exchange (ETDEWEB)

    Stankovic, B [Institut GOSA, Beograd (Yugoslavia); Stevanovic, V [Masinski fakultet, Beograd (Yugoslavia)

    1985-07-01

    The transient behaviour of NPP Krsko during the accident of pressurizer spray valve stuck open has been simulated y lumped parameters model of the PWR coolant system components, developed at the faculty of Mechanical Engineering, University of Belgrade. The elementary volumes which are characterised by the process and state parameters, and by junctions which are characterised by the geometrical and flow parameters are basic structure of physical model. The process parameters obtained by the model RESI, show qualitative agreement with the measured valves, in a degree in which the actions of reactor safety engineered system and emergency core cooling system are adequately modelled; in spite of the elementary physical model structure and only the modelling of thermal process in reactor core and equilibrium conditions of pressurizer and steam generator. The pressurizer pressure and liquid level predicted by the non-equilibrium pressurizer model SOP show good agreement until the HIPS (high pressure pumps) is activated. (author)

  15. 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

  16. 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.

  17. 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

  18. Optimization of maintenance periodicity of complex of NPP safety systems

    International Nuclear Information System (INIS)

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

    2006-01-01

    The analysis of the positive and negative aspects connected to maintenance of the safety systems equipment which basically is in a standby state is executed. Tests of systems provide elimination of the latent failures and raise their reliability. Poor quality of carrying out the tests can be a source of the subsequent failures. Therefore excess frequency of tests can result in reducing reliability of safety systems. The method of optimization of maintenance periodicity of the equipment taking into account factors of its reliability and restoration procedures quality is submitted. The unavailability factor is used as a criterion of optimization of maintenance periodicity. It is offered to use parameters of reliability of the equipment and each of safety systems of NPPs received at developing PSA. And it is offered to carry out the concordance of maintenance periodicity of systems within the NPP maintenance program taking into account a significance factor of the system received on the basis of the contribution of system in CDF. Basing on the submitted method the small computer code is developed. This code allows to calculate reliability factors of a separate safety system and to determine optimum maintenance periodicity of its equipment. Optimization of maintenance periodicity of a complex of safety systems is stipulated also. As an example results of optimization of maintenance periodicity at Zaporizhzhya NPP are presented. (author)

  19. 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)

  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. Remaining life prediction of I and C cables for reliability assessment of NPP systems

    International Nuclear Information System (INIS)

    Santhosh, T.V.; Ghosh, A.K.; Fernandes, B.G.

    2012-01-01

    Highlights: ► A framework for time dependent reliability prediction of I and C cables for use in PSA of NPP has been developed using stress–strength interference theory. ► The proposed methodology has been illustrated with the accelerated thermal aging data on a typical XLPE cable. ► The behavior of insulation resistance when the degradation process is linear or exponential has also been modeled. ► The reliability index or probability of failure obtained from this approach can be used in system reliability evaluation to account for cable aging for PSA of NPP. - Abstract: Instrumentation and control (I and C) cables are one of the most important components in nuclear power plants (NPPs) because they provide power to safety-related equipment and also to transmit signals to and from various controllers to perform safety operations. I and C cables in NPP are subjected to a variety of aging and degradation stressors that can produce immediate degradation or aging-related mechanisms causing the degradation of cable components over time. Although, there exits several life estimation techniques, currently there is no any standard methodology or an approach toward estimating the time dependent reliability of I and C cables that can be directly used in probabilistic safety assessment (PSA) applications. Hence, the objective of this study is to develop an approach to estimate and confirm the continued acceptable margin in cable insulation life over time subjected to aging. This paper presents a framework based on the structural reliability theory to quantify the life time of I and C cable subjecting to thermal aging. Since cross-linked polyethylene (XLPE) cables are extensively being used in Indian NPPs, the remaining life time evaluations have been carried out for a typical XLPE cable. However, the methodology can be extended to other cables such as polyvinyl chloride (PVC), ethylene propylene rubber (EPR), etc.

  2. Probabilistic safety analysis for fire events for the NPP Isar 2

    International Nuclear Information System (INIS)

    Schmaltz, H.; Hristodulidis, A.

    2007-01-01

    The 'Probabilistic Safety Analysis for Fire Events' (Fire-PSA KKI2) for the NPP Isar 2 was performed in addition to the PSA for full power operation and considers all possible events which can be initiated due to a fire. The aim of the plant specific Fire-PSA was to perform a quantitative assessment of fire events during full power operation, which is state of the art. Based on simplistic assumptions referring to the fire induced failures, the influence of system- and component-failures on the frequency of the core damage states was analysed. The Fire-PSA considers events, which will result due to fire-induced failures of equipment on the one hand in a SCRAM and on the other hand in events, which will not have direct operational effects but because of the fire-induced failure of safety related installations the plant will be shut down as a precautionary measure. These events are considered because they may have a not negligible influence on the frequency of core damage states in case of failures during the plant shut down because of the reduced redundancy of safety related systems. (orig.)

  3. 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)

  4. 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

  5. A Preliminary Fire PSA on PGSFR

    International Nuclear Information System (INIS)

    Kim, Kilyoo; Han, Sanghoon; Lee, KwiLim

    2017-01-01

    A Prototype Generation IV Sodium Fast Reactor (PGSFR) is under design with defense in depth concept with active, passive, and inherent safety features to acquire a design approval for PGSFR from Korean regulatory authority by around 2017. A preliminary fire PSA on PGSFR is done in 2016 and a final fire PSA of PGSFR will be done in 2017. The characteristics of the preliminary fire PSA on PGSFR are described in this paper. Since PGSFR is very safe reactor, it is not bad approach to use a conservative assumption in the preliminary PSA. In addition, several drawings including cable routing are not yet issued, a conservative calculation for CDF is performed. As shown in Table 2, the CDF caused by the fire in the control room takes 89% portion of total CDF. Thus, a detailed fire modeling for control room is necessary for the final fire PSA on PGSFR. Also, the increased ignition frequency due to sodium leak would be derived by considering the sodium piping complexity in the final fire PSA on PGSFR. The 4th column of Table 2 is derived the 3rd column by multiplying the factor (592/1177). The 5th column is the ignition frequency caused by the sodium leak. The 6th column is derived by summing the 4th column and the 5th column. The 7th column is the CDF portion of each fire area. The control room (fire area F-A404A) is the most important area since the control room fire takes 89% portion of total CDF.

  6. Regulatory approach to NPP ageing in Bulgaria

    International Nuclear Information System (INIS)

    Vassilev, D.

    2000-01-01

    In this contribution summary information of Kozloduy NPP units is presented. The nuclear legislation, regulatory approach for managing safety aspects on NPP ageing, short term programme, complex programme PRG'97 ant other aspects of ageing management are discussed

  7. Some problems of NPP personnel training

    International Nuclear Information System (INIS)

    Vajshnis, P.P.; Kumkov, L.P.; Omel'chuk, V.V.

    1984-01-01

    Shortcomings of NPP personnel training are discussed. Development of full-scale training systems is necessary for qualitative training operative personnel. Primary problems that should be necessarily solved for ensuring effective training NPP personnel are considered

  8. 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

  9. Individual protection of NPP personnel

    International Nuclear Information System (INIS)

    Koshcheev, V.S.; Gol'dshtejn, D.S.; Chetverikova, Z.S.

    1983-01-01

    Specific features of NPP personnel individual protection are considered, mainly with respect to maintenance and repair works on various type reactors. The major concern is given to the selection and application reglamentations of the individual protection system (IPS), employment of sanitary locks, the organization of individual protection under the conditions of a heating microclimate. The ways are specified to the development and introduction of the most effective IPS and improvement of the entire NPP personnel individual protection system with respect to providing the necessary protection effect for maintaining high working capability of the personnel and minimizing the IPS impact on human organism functional systems. The accumulated experience in the personnel individual protection can be applied during construction and operation of NPP's in CMEA member-countries [ru

  10. 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)

  11. 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....

  12. 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

  13. Use of PSA to support accident management at NPPs

    International Nuclear Information System (INIS)

    Gomez Cobo, A.

    1997-01-01

    The presentation discusses the following: Overview of PSA level 2; Introduction: Framework; Accident Progression Phenomena in the Confinement/containment; Severe Accident Sequences; Examples; Results and Insights. Accident Management: Concepts; Process; Use of PSA to support Accident; Management

  14. 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

  15. 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.

  16. 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

  17. 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%

  18. Neutron doze distribution in capsules for surveillance of radiation embrittlement of pressure vessel in Krsko nuclear power plant; Porazdelitev nevtronske doze v sondah za kontrolo povecanja krhkosri tlacne posode JE Krsko

    Energy Technology Data Exchange (ETDEWEB)

    Najzer, M; Remec, I; Kodeli, I [Institut Jozef Stefan, Ljubljana (Yugoslavia)

    1984-07-01

    Calculation of neutron fluence and spectrum distribution in the capsule with samples for radiation embrittlement of PWR pressure vessel surveillance program of Krsko nuclear power plant is presented. Two dimensional computer code DOT 3 has been used and neutron cross sections were taken from DLC-2D library. result is that fluence magnitude in the capsules changes for up to 70%, so when evaluating results of mechanical tests of samples it is necessary to take into account actual position of samples within the capsule. (author)

  19. 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

  20. PSA methodology development and application in Japan

    International Nuclear Information System (INIS)

    Kazuo Sato; Toshiaki Tobioka; Kiyoharu Abe

    1987-01-01

    The outlines of Japanese activities on development and application of probabilistic safety assessment (PSA) methodologies are described. First the activities on methodology development are described for system reliability analysis, operational data analysis, core melt accident analysis, environmental consequence analysis and seismic risk analysis. Then the methodoligy application examples by the regulatory side and the industry side are described. (author)

  1. 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.

  2. 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.

  3. French 900 MWe PWR PSA preliminary results

    International Nuclear Information System (INIS)

    Lanore, J.M.; Brisbois, J.

    1988-10-01

    A PSA is performed by the Safety Assessment Department of CEA for a 900 MWe standardized plant. The paper presents the objectives, the scope of the study and the relative preliminary results. Some general insights are drawn, especially the benefit related to the implementation of emergency procedures

  4. Deepwater Horizon Oil Spill PSA (:24)

    Centers for Disease Control (CDC) Podcasts

    This 24 second PSA from Dr. Regina Benjamin, U.S. Surgeon General, encourages people to get help if they are feeling sad, angry, depressed, or stressed as a result of the Gulf oil spill. It was recorded so that localities can add their own "For more information" at the end, as appropriate.

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

    Centers for Disease Control (CDC) Podcasts

    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).

  6. Communication Can Save Lives PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    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.

  7. Safer Food Saves Lives PSA (:60)

    Centers for Disease Control (CDC) Podcasts

    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.

  8. 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.

  9. 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

  10. 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.

  11. 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.

  12. 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.

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

    Centers for Disease Control (CDC) Podcasts

    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.

  14. 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.

  15. 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...

  16. The V-1 NPP and V-2 NPP upgrading

    International Nuclear Information System (INIS)

    1998-01-01

    A brief account of activities in the V-1 NPP and V-2 NPP upgrading as well as maintenance carried out by the Nuclear power plants Jaslovske Bohunice in 1997 is presented. The V-1 NPP applied the so called 'Small Backfitting Programme'covering 81 points of the Czechoslovak Atomic energy Commission Decree No 5/91. Continual upgrading continued after the Backfitting Programme completion with the Safety Report and following Nuclear Regulatory Authority of Slovak Republic (NRA SR) Decrees No 1/94 and 110/94 setting spheres and procedure for adopting and implementation of measures enabling the units to operate further on. Results of expert missions, analyses and assessments of components identified by Basic Engineering became the basis for the development of the Gradual Reconstruction Programme. The Programme outputs underwent economic and probabilistic assessing their contribution to nuclear safety. This process resulted in finalizing the Gradual Reconstruction Programme which started to be implemented in 1996 and will be completed in 1999. It is implemented by the REKON consortium and covers 17 areas including Instrumentation and Control, self-consumption emergency supply, leakage monitoring, emergency core cooling system, seismic reinforcement and radioactivity localisation. Both units will reach internationally acceptable safety standards for the remaining life-time period. The V-2 NPP Upgrading and Safety Enhancement Programme includes results of activities performed in the course of last years to define all important activities leading to enhancement of nuclear safety and performance reliability and effectiveness within the plant life-time period and to establish conditions for extending the life-time of these units for 40 years. The V-2 NPP Upgrading and Safety Enhancement Programme aims to assure safe operation with a probability of the core damages less than 10 -4 /reactor · year

  17. Management and organisational factors in PSA; Organisations- und Management-Faktoren in der PSA

    Energy Technology Data Exchange (ETDEWEB)

    Balfanz, H.P. [TUEV Nord e.V., Hamburg (Germany)

    1999-04-01

    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) [German] Mit zunehmender Anwendung der PSA (Probabilistische Sicherheitsanalyse) im Sicherheitsmanagement von KKW (vergl. US-NRC, Einfuehrung des Konzepts 'Risk Informed Regulation' NRC-1) gewinnt die Beachtung der Grenzen der PSA zusaetzliche Bedeutung. International ist eine intensive Diskussion ueber die Moeglichkeiten einer Einbindung der Einflussgroesse von Organisation und Management in der PSA zu verzeichnen und wird belegt durch vielfaeltige Forschungs- und Entwicklungsarbeiten (vergl. PSAM 4). Dieser Beitrag setzt sich in erster Linie mit diesem Entwicklungsstand auseinander und diskutiert seinen Anwendungsstand fuer die PSA. Die hierzu vom TUeV Nord durchgefuehrten Arbeiten basieren auf dem BMU/BfS-Vorhaben SR 2260, 'Weiterentwicklung probabilistischer Methoden zur Sicherheitsbeurteilung von KKW'. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  19. 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)

  20. 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

  1. 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

  2. Clinical performance of serum [-2]proPSA derivatives, %p2PSA and PHI, in the detection and management of prostate cancer.

    Science.gov (United States)

    Huang, Ya-Qiang; Sun, Tong; Zhong, Wei-De; Wu, Chin-Lee

    2014-01-01

    Prostate-specific antigen (PSA) has been widely used as a serum marker for prostate cancer (PCa) screening or progression monitoring, which dramatically increased rate of early detection while significantly reduced PCa-specific mortality. However, a number of limitations of PSA have been noticed. Low specificity of PSA may lead to overtreatment in men who presenting with a total PSA (tPSA) level of PHI) and %p2PSA, which were defined as [(p2PSA/fPSA) × √ tPSA] and [(p2PSA/fPSA) × 100] respectively, have been suggested to be increased in PCa and can better distinguish PCa from benign prostatic diseases than tPSA or fPSA. We performed a systematic review of the available scientific evidences to evaluate the potentials of %p2PSA and PHI in clinical application. Mounting evidences suggested that both %p2PSA and PHI possess higher area under the ROC curve (AUC) and better specificity at a high sensitivity for PCa detection when compare with tPSA and %fPSA. It indicated that measurements of %p2PSA and PHI significantly improved the accuracy of PCa detection and diminished unnecessary biopsies. Furthermore, elevations of %p2PSA and PHI are related to more aggressive diseases. %p2PSA and PHI might be helpful in reducing overtreatment on indolent cases or assessing the progression of PCa in men who undergo active surveillance. Further studies are needed before being applied in routine clinical practice.

  3. An approach to develop a PSA workstation in KAERI

    International Nuclear Information System (INIS)

    Kim, T. W.; Han, S. H.; Park, C. K.

    1995-01-01

    This paper describes three kinds of efforts for the development of PSA workstation in KAERI; Development of a PSA tool, KIRAP, Reliability Database Development, Living PSA tool development. Korea has 9 nuclear power plants (NPPs) in operation and 9 NPPs under design or construction. For the NPPs recently constructed or designed, the probabilistic safety assessments (PSAs) have been performed by the Government requirements. For these PSAs, the MSDOS version of KIRAP has been used. For the consistent data management and the easiness of information handling needed in PSA, APSA workstation, KIRAP-Win is under development under Windows environment. For the reliability database on component failure rate, human error rate, and common cause failure rate, data used in international PSA or reliability data handbook are collected and processed to use in Korean new plants' PSAs. Finally, an effort for the development of a living PSA tool in KAERI based on dynamic PSA concept is described

  4. 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

  5. A PSA study for the SMART basic design

    International Nuclear Information System (INIS)

    Han, Sang Hoon; Kim, H. C.; Yang, S. H.; Lee, D. J.

    2002-03-01

    SMART (System-Integrated Modular Advanced Reactor) is under development that is an advanced integral type small and medium category nuclear power reactor with the rated thermal power of 330 MW. A Probabilistic Safety Analysis (PSA) for the SMART basic design has been performed to evaluate the safety and optimize the design. Currently, the basic design is done and the detailed design is not available for the SMART, we made several assumptions about the system design before performing the PSA. The scope of the PSA was limited to the Level-1 internal full power PSA. The level-2 and 3 PSA, the external PSA, and the low power/shutdown PSA will be performed in the final design stage

  6. 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)

  7. Experience from the comparison of two PSA-studies

    International Nuclear Information System (INIS)

    Holmberg, J.; Pulkkinen, U.

    2001-03-01

    Two probabilistic safety assessments (PSA) made for nearly identical reactors units (Forsmark 3 and Oskarshamn 3) have been compared. Two different analysis teams made the PSAs, and the analyses became quite different. The goal of the study is to identify, clarify and explain differences between PSA-studies. The purpose is to understand limitations and uncertainties in PSA, to explain reasons for differences between PSA-studies, and to give recommendations for comparison of PSA-studies and for improving the PSA-methodology. The reviews have been made by reading PSA-documentation, using the computer model and interviewing persons involved in the projects. The method and findings have been discussed within the project group. Both the PSA-project and various parts in the PSA-model have been reviewed. A major finding was that the two projects had different purpose and thus had different resources, scope and even methods in their study. The study shows that comparison of PSA results from different plants is normally not meaningful. It takes a very deep knowledge of the PSA studies to make a comparison of the results and usually one has to ensure that the compared studies have the same scope and are based on the same analysis methods. Harmonisation of the PSA-methodology is recommended in the presentation of results, presentation of methods, scope main limitation and assumption, and definitions for end states, initiating events and common cause failures. This would facilitate the comparison of the studies. Methods for validation of PSA for different application areas should be developed. The developed PSA review standards can be applied for a general validation of a study. The most important way to evaluate the real feasibility of PSA can take place only with practical applications. The PSA-documentation and models can be developed to facilitate the communication between PSA-experts and users. In any application consultation with the PSA-expert is however needed. Many

  8. 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)

  9. 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

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. 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)

  16. 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.

  17. The organizational factor in PSA framework

    International Nuclear Information System (INIS)

    Farcasiu, Mita; Nitoi, Mirela

    2015-01-01

    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.

  18. Identification of initiating events using a master logic diagram in low-power and shutdown PSA for nuclear power plant

    International Nuclear Information System (INIS)

    Han, S. J.; Park, J. H.; Kim, T. W.; Ha, J. J.

    2003-01-01

    It is necessary to apply a formal technique instead of an empirical technique in the identification of initiating events for Low Power and ShutDown (LPSD) Probabilistic Safety Assessment (PSA) of Nuclear Power Plant (NPP). The present study focuses on the examination of Master Logic Diagram (MLD) technique as a formal technique in the identification of initiating events. The MLD technique is a deductive tool using top-down approach for the formal and logical indentification of initiating events. The present study modified the MLD used in the full power PSA considering the characteristics of LPSD operation. The modified MLD introduced a systematic formation in decomposition process of which the MLD for full power PSA lacked. The modified MLD was able to identify initiating events systematic and logical. However, the formal techniques including the MLD have a limitation for precisely identifying all of the initiating events. In order to overcome this limitation, it is necessary to combine it with an empirical technique. We expect that the modified MLD can be used in an upgrade of the current LPSD PSAs

  19. Building Newcomer Competence for NPP Safety Assessment through Learning by Doing: Development of Level 1 Probabilistic Safety Assessment for Research Reactors

    International Nuclear Information System (INIS)

    Kuzmina, Irina

    2014-01-01

    Final remarks: • COMPASS-M project is a very fruitful study. 1. State-of-the-art competence for PSA technique in Malaysia (applicable to nuclear installations, incl. RR and NPP). 2. PSA model and report for the operating research reactor in Malaysia. → Risk estimate of core damage and ranking contributors to the risk; → Basis for further safety improvement of RR as appropriate. 3. Input for IAEA’s publications on PSA for research reactors. • The results will be available to interested Member States (security considerations be addressed); → Completion in mid-2014, paper to be published in PSAM-12; ► Managerial support is instrumental for success of learning-by-doing projects

  20. 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

  1. 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

  2. ASSET experience at Paks NPP

    Energy Technology Data Exchange (ETDEWEB)

    Szabo, I [Operational Safety Dept., Paks NPP, Paks (Hungary)

    1997-10-01

    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.

  3. Nuclear Fuel in Cofrentes NPP

    International Nuclear Information System (INIS)

    2002-01-01

    Fuel is an essential in the nuclear power generating business because of its direct implications on safety, generating costs and the operating conditions and limitations of the facility. Fuel management in Cofrentes NPP has been targeted at optimized operation, enhanced reliability and the search for an in-depth knowledge of the design and licensing processes that will provide Iberdrola,as the responsible operator, with access to independent control of safety aspects related to fuel and free access to manufacturing markets. (Author)

  4. Welding facilities for NPP assembling

    International Nuclear Information System (INIS)

    Rojtenberg, S.S.

    1987-01-01

    Recommendations concerning the choice of equipment for welding in pre-assembling work shops, in the enlarging assembling shops and at the assembling site, are given. Advanced production automatic welders and semiautomatic machines, applied during the NPP equipment assembling as well as automatic machines specially produced for welding the main reactor components and pipelines are described. Automatic and semiautomatic machine and manual welding post supply sources are considered

  5. Clinical performance of serum [-2]proPSA derivatives, %p2PSA and PHI, in the detection and management of prostate cancer

    OpenAIRE

    Huang, Ya-Qiang; Sun, Tong; Zhong, Wei-De; Wu, Chin-Lee

    2014-01-01

    Prostate-specific antigen (PSA) has been widely used as a serum marker for prostate cancer (PCa) screening or progression monitoring, which dramatically increased rate of early detection while significantly reduced PCa-specific mortality. However, a number of limitations of PSA have been noticed. Low specificity of PSA may lead to overtreatment in men who presenting with a total PSA (tPSA) level of < 10 ng/mL. As a type of free PSA (fPSA), [-2]proPSA is differentially expressed in peripheral ...

  6. New approach of second Romanian NPP siting

    International Nuclear Information System (INIS)

    Mauna, Traian

    2010-01-01

    The NPP sitting studies in Romania began before 1975. The first Romanian NPP CANDU 6 type reactor gone to erection in 1980 on Cernavoda site planned to have 5 units. Gained the experience from Cernavoda NPP sitting, the first mission of new multi-branch of specialists team was to choose new NPP sites adapting the NPP Cernavoda project to the new parameters of close water cooling circuit and hard less and no rock foundation strata. The studies were carrying out in different stages on the inner rivers Olt, Mures, Somes in Transylvania historical region. This paper tries to reconsider shortly the old analysis according to the last IAEA Safety Standards, taking into account the new NPP generation requirement. Paper is focused on geological aspects and other local sites characteristics. (authors)

  7. 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

  8. On the NPP structural reliability

    International Nuclear Information System (INIS)

    Klemin, A.I.; Polyakov, E.F.

    1980-01-01

    Reviewed are the main statements peculiarities and possibilities of the first branch guiding technical material (GTM) ''The methods of calculation of structural reliability of NPP and its systems at the stage of projecting''. It is stated, that in GTM presented are recomendations on the calculation of reliability of such specific systems, as the system of the reactor control and protection the system of measuring instruments and automatics and safe systems. GTM are based on analytical methods of modern theory of realibility with the Use of metodology of minimal cross sections of complex systems. It is stressed, that the calculations on the proposed methods permit to calculate a wide complex of reliability parameters, reflecting separately or together prorerties of NPP dependability and maintainability. For NPP, operating by a variable schedule of leading, aditionally considered are parameters, characterizing reliability with account of the proposed regime of power change, i.e. taking into account failures, caused by decrease of the obtained power lower, than the reguired or increase of the required power higher, than the obtained

  9. Upgrade of internal events PSA model using the AESJ level-1 PSA standard for operating state

    International Nuclear Information System (INIS)

    Sato, Teruyoshi; Yoneyama, Mitsuru; Hirokawa, Naoki; Sato, Chikahiro; Sato, Eisuke; Tomizawa, Shigeatsu

    2009-01-01

    In 2003, the Atomic Energy Society of Japan (AESJ) started to develop the Level-1 Probabilistic Safety Assessment (PSA) standard of internal events for operating state (AESJ standard). The AESJ standard has been finished to be asked for public comment. Using the AESJ standard (draft version), the authors have upgraded the PSA model for Tokyo Electric Power Company (TEPCO) BWR-5 plant not only to reflect latest knowledge but also to ensure high quality of PSA model (not yet peer-reviewed) for the purpose of better operation and maintenance management of TEPCO BWR plants. For example, the categorization of structures, systems and components (SSCs) will be performed to improve nuclear reactor safety using information of risk importance. (author)

  10. 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)

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. Experiences of Uncertainty in Men With an Elevated PSA.

    Science.gov (United States)

    Biddle, Caitlin; Brasel, Alicia; Underwood, Willie; Orom, Heather

    2015-05-15

    A significant proportion of men, ages 50 to 70 years, have, and continue to receive prostate specific antigen (PSA) tests to screen for prostate cancer (PCa). Approximately 70% of men with an elevated PSA level will not subsequently be diagnosed with PCa. Semistructured interviews were conducted with 13 men with an elevated PSA level who had not been diagnosed with PCa. Uncertainty was prominent in men's reactions to the PSA results, stemming from unanswered questions about the PSA test, PCa risk, and confusion about their management plan. Uncertainty was exacerbated or reduced depending on whether health care providers communicated in lay and empathetic ways, and provided opportunities for question asking. To manage uncertainty, men engaged in information and health care seeking, self-monitoring, and defensive cognition. Results inform strategies for meeting informational needs of men with an elevated PSA and confirm the primary importance of physician communication behavior for open information exchange and uncertainty reduction. © The Author(s) 2015.

  17. 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)

  18. Evaluating the Phoenix definition of biochemical failure after (125)I prostate brachytherapy: Can PSA kinetics distinguish PSA failures from PSA bounces?

    Science.gov (United States)

    Thompson, Anna; Keyes, Mira; Pickles, Tom; Palma, David; Moravan, Veronika; Spadinger, Ingrid; Lapointe, Vincent; Morris, W James

    2010-10-01

    To evaluate the prostate-specific antigen (PSA) kinetics of PSA failure (PSAf) and PSA bounce (PSAb) after permanent (125)I prostate brachytherapy (PB). The study included 1,006 consecutive low and "low tier" intermediate-risk patients treated with (125)I PB, with a potential minimum follow-up of 4 years. Patients who met the Phoenix definition of biochemical failure (nadir + 2 ng/mL(-1)) were identified. If the PSA subsequently fell to ≤0.5 ng/mL(-1)without intervention, this was considered a PSAb. All others were scored as true PSAf. Patient, tumor and dosimetric characteristics were compared between groups using the chi-square test and analysis of variance to evaluate factors associated with PSAf or PSAb. Median follow-up was 54 months. Of the 1,006 men, 57 patients triggered the Phoenix definition of PSA failure, 32 (56%) were true PSAf, and 25 PSAb (44%). The median time to trigger nadir + 2 was 20.6 months (range, 6-36) vs. 49 mo (range, 12-83) for PSAb vs. PSAf groups (p < 0.001). The PSAb patients were significantly younger (p < 0.0001), had shorter time to reach the nadir (median 6 vs. 11.5 months, p = 0.001) and had a shorter PSA doubling time (p = 0.05). Men younger than age 70 who trigger nadir +2 PSA failure within 38 months of implant have an 80% likelihood of having PSAb and 20% chance of PSAf. With adequate follow-up, 44% of PSA failures by the Phoenix definition in our cohort were found to be benign PSA bounces. Our study reinforces the need for adequate follow-up when reporting PB PSA outcomes, to ensure accurate estimates of treatment efficacy and to avoid unnecessary secondary interventions. 2010. Published by Elsevier Inc. All rights reserved.

  19. Evaluating the Phoenix Definition of Biochemical Failure After 125I Prostate Brachytherapy: Can PSA Kinetics Distinguish PSA Failures From PSA Bounces?

    International Nuclear Information System (INIS)

    Thompson, Anna; Keyes, Mira; Pickles, Tom

    2010-01-01

    Purpose: To evaluate the prostate-specific antigen (PSA) kinetics of PSA failure (PSAf) and PSA bounce (PSAb) after permanent 125 I prostate brachytherapy (PB). Methods and Materials: The study included 1,006 consecutive low and 'low tier' intermediate-risk patients treated with 125 I PB, with a potential minimum follow-up of 4 years. Patients who met the Phoenix definition of biochemical failure (nadir + 2 ng/mL -1 ) were identified. If the PSA subsequently fell to ≤0.5 ng/mL -1 without intervention, this was considered a PSAb. All others were scored as true PSAf. Patient, tumor and dosimetric characteristics were compared between groups using the chi-square test and analysis of variance to evaluate factors associated with PSAf or PSAb. Results: Median follow-up was 54 months. Of the 1,006 men, 57 patients triggered the Phoenix definition of PSA failure, 32 (56%) were true PSAf, and 25 PSAb (44%). The median time to trigger nadir + 2 was 20.6 months (range, 6-36) vs. 49 mo (range, 12-83) for PSAb vs. PSAf groups (p < 0.001). The PSAb patients were significantly younger (p < 0.0001), had shorter time to reach the nadir (median 6 vs. 11.5 months, p = 0.001) and had a shorter PSA doubling time (p = 0.05). Men younger than age 70 who trigger nadir +2 PSA failure within 38 months of implant have an 80% likelihood of having PSAb and 20% chance of PSAf. Conclusions: With adequate follow-up, 44% of PSA failures by the Phoenix definition in our cohort were found to be benign PSA bounces. Our study reinforces the need for adequate follow-up when reporting PB PSA outcomes, to ensure accurate estimates of treatment efficacy and to avoid unnecessary secondary interventions.

  20. 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)

  1. Low Power Shutdown PSA for CANDU Type Plants

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yeon Kyoung; Kim, Myung Su [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    KHNP also have concentrated on full power PSA. Some recently constructed OPR1000 type plants and APR1400 type plants have performed the low power and shutdown (LPSD) PSA. The purpose of LPSD PSA is to identify the main contributors on the accident sequences of core damage and to find the measure of safety improvement. After the Fukushima accident, Korean regulatory agency required the shutdown severe accident management guidelines (SSAMG) development for safety enhancement. For the reliability of SSAMG, KHNP should develop the LPSD PSA. Especially, the LPSD PSA for CANDU type plant had developed for the first time in Korea. This paper illustrates how the LPSD PSA for CANDU type developed and the core damage frequency (CDF) is different with that of full power PSA. KHNP performed LPSD PSA to develop the SSAMG after the Fukushima accidents. The results show that risk at the specific operation mode during outage is higher than that of full power operation. Also, the results indicated that recovery failure of class 4 power at the POS 5A, 5B contribute dominantly to the total CDF from importances analysis. LPSD PSA results such as CDF with initiating events and POSs, risk results with plant damage state, and containment failure probability and frequency with POSs can be used by inputs for developing the SSAMG.

  2. PSA data base, comparison of the German and French approach

    International Nuclear Information System (INIS)

    Kreuser, A.; Tirira, J.

    2001-01-01

    The results of probabilistic safety assessments (PSA) of nuclear power plants strongly depend on the reliability data used. This report describes coarsely the general process to generate reliability data for components and resumes the differences between the German and French approaches. As has been shown in former studies which compared international PSA data, PSA data are closely related to the model definitions of the PSA. Therefore single PSA data cannot be compared directly without regard e.g. to the corresponding fault trees. These findings are confirmed by this study. The comparison of German and French methods shows a lot of differences concerning various details of the data generation process. Some differences between single reliability data should be eliminated when taking into account the complete fault tree analysis. But there are some other differences which have a direct impact on the obtained results of a PSA. In view of the all differences between both approaches concerning the definition of data and the data collection process, it is not possible to compare directly German and French PSA data. However, the database differences give no indication on the influence on the PSA results. Therefore, it is a need to perform a common IPSN/GRS assessment on how the different databases impact the PSA results. (orig.)

  3. 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

  4. DIAGNOSTIC AND PROGNOSTIC UTILITY OF SERUM PSA IN BREAST CANCER

    Institute of Scientific and Technical Information of China (English)

    张淑群; 强水云; 李妙羡; 纪宗正

    2004-01-01

    Objective To investigate the diagnostic and prognostic value of total and free prostate-specific antigen (PSA) in breast cancer women. Methods Using the microparticle enzyme immunoassay system, we measured the concentrations of these markers in the sera of 85 women with breast cancer and in 30 healthy women.Results Free PSA levels were significantly higher in women with breast cancer than healthy women (P <0. 05 ).The percentage of free PSA predominant subjects was 37. 6% in breast cancer patients and 3. 3% in healthy women.In women with breast cancer,total PSA positivity was 23.5% and free PSA positivity was 27. 1%. When compared to negatives,total PSA positive patients had a higher percentage of lymph node involvement tamours ( P >0. 05).However, patients with predominant free PSA had a higher percentage of early stage than patients with predominant PSA-ACT. Conclusion This study indicate clinical significance of preoperative measurement of serum total and free PSA in diagnosis and prognosis of women with breast cancer. The expression of KLKs is correlated with carcinogenesis of breast cancer.

  5. 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.

  6. Institute of nuclear power operations perspectives on PSA applications

    International Nuclear Information System (INIS)

    Webster, W.E.; Miller, W.J. Jr.

    1996-01-01

    The investment to develop a PSA is very substantial, and therefore, there is motivation to recover this investment through further use of the techniques used to develop it. It is not surprising that nuclear power plant staff are beginning to use PSA to make operational decisions. The Institute of Nuclear Power Operations is interested in those factors that impact the conduct of plant operations and therefore is actively monitoring the increased usage of PSA techniques. The purpose of this paper is to provide some thoughts and perspectives on the use of PSA as a factor in operational decision making, including decision making in activities performed by engineering, maintenance and operation personnel. (author)

  7. 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

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

    Centers for Disease Control (CDC) Podcasts

    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.

  9. 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

  10. 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

  11. Display systems for NPP control

    International Nuclear Information System (INIS)

    Rozov, S.S.

    1988-01-01

    Main trends in development of display systems used as the means for image displaying in NPP control systems are considered. It is shown that colour display devices appear to be the most universal means for concentrated data presentation. Along with digital means the display systems provide for high-speed response, sufficient for operative control of executive mechanisms. A conclusion is drawn that further development of display systems will move towards creation of large colour fields (on reflection base or with multicolour gas-discharge elements)

  12. 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

  13. Safety culture in Ignalina NPP, regulatory view

    Energy Technology Data Exchange (ETDEWEB)

    Maksimovas, G [VATESI (Lithuania)

    1997-09-01

    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.

  14. 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

  15. Constructive approaches to the space NPP designing

    International Nuclear Information System (INIS)

    Eremin, A.G.; Korobkov, L.S.; Matveev, A.V.; Trukhanov, Yu.L.; Pyshko, A.P.

    2000-01-01

    An example of designing a space NPP intended for power supply of telecommunication satellite is considered. It is shown that the designing approach based on the introduction of a leading criterion and dividing the design problems in two independent groups (reactor with radiation shield and equipment module) permits to develop the optimal design of a space NPP [ru

  16. 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)

  17. NPP Bohunice experience with ASSET services

    Energy Technology Data Exchange (ETDEWEB)

    Klimo, J [Bohunice NPP (Slovakia)

    1997-12-31

    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).

  18. Bohunice V1 NPP upgrading programme

    International Nuclear Information System (INIS)

    Kerak, J.

    2001-01-01

    The paper describes whole process of Bohunice V1 NPP nuclear safety and operational reliability level increase which has been performed since units commissioning (1. unit in 1978, 2. unit in 1980), continued Small Reconstruction (1991 -1993) and finished Gradual Upgrading(1994 -2000). The main purpose is to last stage -Gradual upgrading of Bohunice V1 NPP. (author)

  19. 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

  20. Prostate-Specific Antigen (PSA) Test: MedlinePlus Lab Test Information

    Science.gov (United States)

    ... medlineplus.gov/labtests/prostatespecificantigenpsatest.html Prostate-Specific Antigen (PSA) Test To use the sharing features on this ... enable JavaScript. What is a prostate-specific antigen (PSA) test? A prostate-specific antigen (PSA) test measures ...

  1. PSA - A utility perspective from the United Kingdom

    International Nuclear Information System (INIS)

    Ross, P.J.

    2004-01-01

    This paper provides an outline of the three stages of PSA used for the Sizewell B PWR. These stages cover the use of PSA during design, licensing and (in future) operation of Sizewell B. The paper discusses each of these stages in some detail, highlighting the differences in approach and lessons learnt at each stage. (author)

  2. 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

  3. 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)

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

    Centers for Disease Control (CDC) Podcasts

    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.

  5. 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

  6. PSA in design of passive/active safety reactors

    International Nuclear Information System (INIS)

    Sato, T.; Tanabe, A.; Kondo, S.

    1995-01-01

    PSAs in the design of advanced reactors are applied mainly in level 1 PSA areas. However, even in level 1 PSA, there are certain areas where special care must be taken depending on plant design concepts. This paper identifies these areas both for passive and active safety reactor concepts. For example, 'long-term PSA' and shutdown PSA are very important for a passive safety reactor concept from the standpoint of effectiveness of a grace period and passive safety systems. External events are also important for an active safety reactor concept. These kinds of special PSAs are difficult to conduct precisely in a conceptual design stage. This paper shows methods of conducting these kinds of special PSAs simply and conveniently and the use of acquired insights for the design of advanced reactors. This paper also clarifies the meaning or definition of a grace period from the standpoint of PSA

  7. 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...

  8. 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

  9. 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)

  10. 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.)

  11. 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.

  12. 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.

  13. 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.

  14. 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

  15. 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

  16. 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

  17. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2009-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

  18. ESTE EMO and ESTE EBO - emergency response system for NPP Mochovce and NPP Bohunice V-2

    International Nuclear Information System (INIS)

    Caeny, P.; Chyly, M.; Suchon, D.; Smejkalova, E.; Fabova, V.; Mancikova, M.; Muller, P.

    2008-01-01

    Programs ESTE EMO and ESTE EBO are emergency response systems that help the crisis staff of the NPP in assessing the source term (predicted possible release of radionuclides to the atmosphere ), in assessing the urgent protective measures and sectors under threat, in assessing real release (symptoms of release really detected and observed), in calculating radiological impacts of real release, averted or avertable doses, potential doses and doses during transport or evacuation on specified routes. Both systems serve as instruments in case of severe accident (DBA or BDBA) at NPP Mochovce or NPP Bohunice, accidents with threat of release of radioactivity to the atmosphere. Systems are implemented at emergency centre of Mochovce NPP and Bohunice NPP and connected online to the sources of technological and radiological data from the reactor, primary circuit, confinement, secondary circuit, ventilation stack, from the area of NPP (TDS 1) and from the emergency planning zone (TDS 11). Systems are connected online to the sources of meteorological data, too. (authors)

  19. Study on human factor at NPP

    International Nuclear Information System (INIS)

    Nopp, I.

    1984-01-01

    Factors affecting the reliabilty of the reactor control by an NPP operator are considered on the base of the Czechoslovakia NPP operating experience. The reliability level of NPP operators depends on objective factors (conditions and regime of labour) determining the labour productivity and on subjective ones (psychological morale, physical and mental abilities and occupational level of personnel). Problems of the effect of physical and mental abilities and professional level on the reliability of personnel are considered to be the most important ones. The effect of individual abilities and specific features of the human body on changes in his occupational abilities can be estimated only to a certain degree

  20. 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