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| | výběr: všechny záznamy| nalezeno záznamů: 556 | |
| [1] | Periodic Safety Review of a Nuclear Power Plant| R. Dumitru | | AMEC NUCLEAR RO, Bucharest, Romania | | In the past, routine safety reviews and special safety reviews were the primary means of the normal safety oversight process. These reviews were often tied to license requirements: periodic license renewals (2, 3 years) and justifications for restart after normal scheduled plant outages. Now, in most countries a PSR is either required by law or strongly encouraged by the Regulator Authority, in order to deliver a relicense for a NPP after a minimum 10 years of operation. The PSR is a systematic reassessment of a nuclear power plant safety carried out at regular intervals to deal with the cumulative effects of ageing, modifications, operating experience, technical developments and siting aspects, and is aimed at ensuring a high level of safety throughout the operating lifetime of the plant. The PSR concept was developed to review the design and current plant status against: significant developments in Safety Standards, Design Standards, Operating practices, Technology, Analytical methods and Scientific and technical knowledge the need to review cumulative effects of plant modifications and ageing mechanisms and possibility of significant staffing changes of both NPP and regulatory organization (CNCAN) in Romania. | kategorie:ENYGFpublikováno:ENYGF 2011, 17.-22. května 2011, Masarykova kolej, Praha - No.01_03 | | umístění:| zde | | | | | | | |
| [2] | Complex Emergency Management in Paks Nuclear Power Plant| A. I. Herman | | Nuclear Power Plant Ltd, Paks, Hungary | | Like in many countries in Europe nuclear energy plays a determinative role in Hungarian electricity production. The only one nuclear power plant in Paks -taking into account a fact: provides 40% of national energy production- is elemental asset for Hungary, that’s why could be signed an Achilles’ heel as well. Following international directives and realizing general tendencies in EU efforts to establish a common Critical Infrastructure Protection (CIP) have binding conditions. | kategorie:ENYGFpublikováno:ENYGF 2011, 17.-22. května 2011, Masarykova kolej, Praha - No.01_04 | | umístění:| zde | | | | | | | |
| [3] | Ageing within PSA: Development of an Analytical Unavailability Model and Its Application| D. Kančev, M. Čepin | | Reactor Engineering Division, Jozef Stefan Institute, Ljubljana, Slovenia | | This paper briefly summarizes the focal points of the performed work by the authors associated with the analyses of the effects of nuclear industry equipment ageing and its integration within the probabilistic safety assessment (PSA). The paper consists of three general sections. The basic, traditional approaches for transforming the PSA into an age-dependent PSA (APSA) are addressed into the first section. The successive stepwise constant failure rates approach, the substitution of ageing models into PSA and the risk importance approach for evaluating ageing effects are briefly discussed herein. The second section addresses the development of a new analytical unavailability model and its application as a basis for trade-off analysis between risk and cost is initially addressed. Uncertainty and sensitivity analyses, associated with the analytical model and its application, are additionally conducted herein. Subsequently, the development of a multiobjective optimization method, based on the developed analytical model, utilizing the genetic algorithm technique as an optimization tool, is described in the third section. The obtained results generally indicate the fact that riskinformed surveillance requirements differ from existing ones in technical specifications as well as show the importance of considering ageing data uncertainties in component ageing modelling. | kategorie:ENYGFpublikováno:ENYGF 2011, 17.-22. května 2011, Masarykova kolej, Praha - No.01_05 | | umístění:| zde | | | | | | | |
| [4] | Analysis of the Presence of Vapor in Residual Heat Removal System in Modes 3/4 Loss-of-Coolant Accident Conditions using RELAP5| K. Mathy | | Westinghouse Electric Belgium S.A., Belgium | | The Westinghouse Nuclear Safety Advisory Letter NSAL-09-8 investigated the possibility of presence of vapor in Residual Heat Removal (RHR) System in Modes 3/4 Loss-of-Coolant Accident (LOCA) Conditions. This concerns the Westinghouse standard 3-loops plant for which the RHR is the low pressure part of the Safety Injection (SI). In some cases one or both RHR trains may become inoperable for Safety Injection (SI) function. As a response to this letter, Westinghouse Electric Belgium is providing RELAP5 analyzes for Westinghouse Nuclear Steam Supply System (NSSS) European plants to assess the thermal hydraulic behavior of the RHR suction piping system for Emergency Core Cooling System (ECCS) initiation events postulated to occur during startup/shutdown operations. Several concerns including condensation induced water hammer and voiding at the RHR pump have been investigated. As a conclusion, the analysis allowed to define the bounding Hot Leg temperature conditions under which both RHR Trains remain safely operable. These bounding conditions are then implemented by the customer in their Operating Procedures (OPs) to achieve safe operations and successful accident management. | kategorie:ENYGFpublikováno:ENYGF 2011, 17.-22. května 2011, Masarykova kolej, Praha - No.01_06 | | umístění:| zde | | | | | | | |
| [5] | Influence of Large LOCA by Seismic Event| K. Demjančuková | | University of West Bohemia, Department of Power System Engineering, Pilsen, Czech Republic | | The limiting condition for the emergency core cooling system requirements is provided by the double-ended-guillotine break criterion of the largest primary piping system in the nuclear power plant (NPP). United Stated Nuclear Regulatory Commision (US NRC) reached the conclusion that the rupture of primary circuit pipeline with a 850 mm diameter is improbable. Expert elicitation method determined the transition break size (TBS) for pressurized-water reactors (PWR) within the range of 305 – 356 mm (for the pipelines connected to the main circulation loop) and for boiling water reactors (BWR) of 500 mm, controlled primarily by the surge line and is expected to have a frequency less then 10-5 per year. The purpose of this paper is to present the main idea and basic assumptions for the assessment of the range of transition break size influenced by a seismic event. | kategorie:ENYGFpublikováno:ENYGF 2011, 17.-22. května 2011, Masarykova kolej, Praha - No.01_07 | | umístění:| zde | | | | | | | |
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