Progress in Methodologies for the Assessment of Passive Safety System Reliability in Advanced Reactors PDF Download
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Author: International Atomic Energy Agency
Publisher: IAEA Tecdoc
ISBN: 9789201086143
Category : Science
Languages : en
Pages : 223
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Book Description
This publication summarizes the results of an IAEA coordinated research project on the development of advanced methodologies for the assessment of passive safety system performance in advanced reactors. This includes discussions on various methodologies to assess the performance of passive engineered safety features in innovative small reactors, including the Indian AHWR 300 LEU and the Argentinian CAREM25. The publication focuses on the different reliability assessment approaches, methodologies, analysis and evaluation of the results and technical challenges. It provides the insights resulting from the analysis on the technical issues associated with assessing the reliability of passive systems in the context of nuclear safety and probabilistic safety analysis. A viable path towards the implementation of the research efforts in the related areas is also delineated.
Author: International Atomic Energy Agency
Publisher: IAEA Tecdoc
ISBN: 9789201086143
Category : Science
Languages : en
Pages : 223
Get Book Here
Book Description
This publication summarizes the results of an IAEA coordinated research project on the development of advanced methodologies for the assessment of passive safety system performance in advanced reactors. This includes discussions on various methodologies to assess the performance of passive engineered safety features in innovative small reactors, including the Indian AHWR 300 LEU and the Argentinian CAREM25. The publication focuses on the different reliability assessment approaches, methodologies, analysis and evaluation of the results and technical challenges. It provides the insights resulting from the analysis on the technical issues associated with assessing the reliability of passive systems in the context of nuclear safety and probabilistic safety analysis. A viable path towards the implementation of the research efforts in the related areas is also delineated.
Author:
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 0
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Book Description
"This publication summarizes the results of an IAEA coordinated research project on the development of advanced methodologies for the assessment of passive safety system performance in advanced reactors. This includes discussions on various methodologies to assess the performance of passive engineered safety features in innovative small reactors, including the Indian AHWR 300 LEU and the Argentinian CAREM25. The publication focuses on the different reliability assessment approaches, methodologies, analysis and evaluation of the results and technical challenges. It provides the insights resulting from the analysis on the technical issues associated with assessing the reliability of passive systems in the context of nuclear safety and probabilistic safety analysis. A viable path towards the implementation of the research efforts in the related areas is also delineated."--Publisher's description.
Author:
Publisher:
ISBN:
Category :
Languages : en
Pages : 17
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Book Description
Many advanced light water reactor designs incorporate passive rather than active safety features for front-line accident response. A method for evaluating the reliability of these passive systems in the context of probabilistic risk assessment has been developed at Sandia National Laboratories. This method addresses both the component (e.g. valve) failure aspect of passive system failure, and uncertainties in system success criteria arising from uncertainties in the system's underlying physical processes. These processes provide the system's driving force; examples are natural circulation and gravity-induced injection. This paper describes the method, and provides some preliminary results of application of the approach to the Westinghouse AP600 design.
Author: Tania M. Hake
Publisher:
ISBN:
Category : Light water reactors
Languages : en
Pages : 184
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Book Description
Author:
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ISBN:
Category :
Languages : en
Pages : 6
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Book Description
The advanced nuclear power plants must achieve higher levels of safety than the first generation of plants. Showing that this is indeed true provides new challenges to reliability and risk assessment methods in the analysis of the designs employing passive and semi-passive protection. Reliability assurance of the advanced reactor systems is important for determining the safety of the design and for determining the plant operability. Safety is the primary concern, but operability is considered indicative of good and safe operation. This paper discusses several concerns for reliability assurance of the advanced design encompassing reliability determination, level of detail required in advanced reactor submittals, data for reliability assurance, systems interactions and common cause effects, passive component reliability, PRA-based configuration control system, and inspection, training, maintenance and test requirements. Suggested approaches are provided for addressing each of these topics.
Author: C. Guedes Soares
Publisher: Elsevier
ISBN: 0080552153
Category : Technology & Engineering
Languages : en
Pages : 791
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Book Description
These three volumes comprise the papers presented at the ESREL '97 International Conference on Safety and Reliability held in Lisbon, Portugal, 17-20 June 1997. The purpose of the annual ESREL conferences is to provide a forum for the presentation of technical and scientific papers covering both methods and applications of safety and reliability to a wide range of industrial sectors and technical disciplines and, in so doing, to enhance cross-fertilization between them. A broad view is taken of safety and reliability which includes probabilistically-based methods, or, more generally, methods that deal with the quantification of the uncertainty in the knowledge of the real world and with decision-making under this uncertainty. The areas covered include: design and product liability; availability, reliability and maintainability; assessment and management of risks to technical systems; health and the environment; and mathematical methods of reliability and statistical analysis of data. The organization of the book closely follows the sessions of the conference with each of the three volumes containing papers from two parallel sessions, comprising a total of 270 papers by authors from 35 countries.
Author:
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Category :
Languages : en
Pages :
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Book Description
In this paper, the details of the development of a demonstration problem to assess the reliability of a passive safety system are presented. An advanced small modular reactor (advSMR) design, which is a pool-type sodium fast reactor (SFR) coupled with a passive reactor cavity cooling system (RCCS) is described. The RELAP5-3D models of the advSMR and RCCS that will be used to simulate a long-term station blackout (SBO) accident scenario are presented. Proposed benchmarking techniques for both the reactor and the RCCS are discussed, which includes utilization of experimental results from the Natural convection Shutdown heat removal Test Facility (NSTF) at the Argonne National Laboratory. Details of how mechanistic methods, specifically the Reliability Method for Passive Systems (RMPS) approach, will be utilized to determine passive system reliability are presented. The results of this mechanistic analysis will ultimately be compared to results from dynamic methods in future work. This work is part of an ongoing project at Argonne to demonstrate methodologies for assessing passive system reliability.
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Category :
Languages : en
Pages :
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Book Description
Advanced small modular reactor designs include many advantageous design features such as passively driven safety systems that are arguably more reliable and cost effective relative to conventional active systems. Despite their attractiveness, a reliability assessment of passive systems can be difficult using conventional reliability methods due to the nature of passive systems. Simple deviations in boundary conditions can induce functional failures in a passive system, and intermediate or unexpected operating modes can also occur. As part of an ongoing project, Argonne National Laboratory is investigating various methodologies to address passive system reliability. The Reliability Method for Passive Systems (RMPS), a systematic approach for examining reliability, is one technique chosen for this analysis. This methodology is combined with the Risk-Informed Safety Margin Characterization (RISMC) approach to assess the reliability of a passive system and the impact of its associated uncertainties. For this demonstration problem, an integrated plant model of an advanced small modular pool-type sodium fast reactor with a passive reactor cavity cooling system is subjected to a station blackout using RELAP5-3D. This paper discusses important aspects of the reliability assessment, including deployment of the methodology, the uncertainty identification and quantification process, and identification of key risk metrics.
Author: John Joseph Vandenkieboom
Publisher:
ISBN:
Category :
Languages : en
Pages : 428
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Book Description
Author: Nicolas Zweibaum
Publisher:
ISBN:
Category :
Languages : en
Pages : 231
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Book Description
The development of advanced nuclear reactor technology requires understanding of complex, integrated systems that exhibit novel phenomenology under normal and accident conditions. The advent of passive safety systems and enhanced modular construction methods requires the development and use of new frameworks to predict the behavior of advanced nuclear reactors, both from a safety standpoint and from an environmental impact perspective. This dissertation introduces such frameworks for scaling of integral effects tests for natural circulation in fluoride-salt-cooled, high-temperature reactors (FHRs) to validate evaluation models (EMs) for system behavior; subsequent reliability assessment of passive, natural- circulation-driven decay heat removal systems, using these validated models; evaluation of life cycle carbon dioxide emissions as a key environmental impact metric; and recommendations for further work to apply these frameworks in the development and optimization of advanced nuclear reactor designs. In this study, the developed frameworks are applied to the analysis of the Mark 1 pebble-bed FHR (Mk1 PB-FHR) under current investigation at the University of California, Berkeley (UCB). The capability to validate integral transient response models is a key issue for licensing new reactor designs. This dissertation presents the scaling strategy, design and fabrication aspects, and startup testing results from the Compact Integral Effects Test (CIET) facility at UCB, which reproduces the thermal hydraulic response of an FHR under forced and natural circulation operation. CIET provides validation data to confirm the performance of the direct reactor auxiliary cooling system (DRACS) in an FHR, used for natural-circulation-driven decay heat removal, under a set of reference licensing basis events, as predicted by best-estimate codes such as RELAP5-3D. CIET uses a simulant fluid, Dowtherm A oil, which at relatively low temperatures (50-120°C) matches the Prandtl, Reynolds, Froude and Grashof numbers of the major liquid salts simultaneously, at approximately 50% geometric scale and heater power under 2% of prototypical conditions. The studies reported here include isothermal pressure drop tests performed during startup testing of CIET, with extensive pressure data collection to determine friction losses in the system, as well as subsequent heated tests, from parasitic heat loss tests to more complex feedback control tests and natural circulation experiments. For initial code validation, coupled steady-state single-phase natural circulation loops and simple forced cooling transients were conducted in CIET. For various heat input levels and temperature boundary conditions, fluid mass flow rates and temperatures were compared between RELAP5- 3D results, analytical solutions when available, and experimental data. This study shows that RELAP5-3D provides excellent predictions of steady-state natural circulation and simple transient forced cooling in CIET. The code predicts natural circulation mass flow rates within 8%, and steady-state and transient fluid temperatures, under both natural and forced circulation, within 2°C of experimental data, suggesting that RELAP5-3D is a good EM to use to design and license FHRs. A key element in design and licensing of new reactor technology lies in the analysis of the plant response to a variety of potential transients. When applicable, this involves understanding of passive safety system behavior. This dissertation develops a framework to assess reliability and propose design optimization and risk mitigation strategies associated with passive decay heat removal systems, applied to the Mk1 PB-FHR DRACS. This investigation builds upon previous detailed design work for Mk1 components and the use of RELAP5-3D models validated for FHR natural circulation phenomenology. For risk assessment, reliability of the point design of the passive safety system for the Mk1 PB-FHR, which depends on the ability of various structures to fulfill their safety functions, is studied. Whereas traditional probabilistic risk assessment (PRA) methods are based on event and fault trees for components of the system that perform in a binary way - operating or not operating -, this study is mostly based on probability distributions of heat load compared to the capacity of the system to remove heat, as recommended by the reliability methods for passive safety functions (RMPS) that are used here. To reduce computational time, the use of response surfaces to describe the system in a simplified manner, in the context of RMPS, is also demonstrated. The design optimization and risk mitigation part proposes a framework to study the elements of the design of the reactor, and more specifically its passive safety cooling system, which can contribute to enhanced reliability of heat removal under accident conditions. Risk mitigation measures based on design, startup testing, in-service inspection and online monitoring are proposed to narrow probability distributions of key parameters of the system and increase reliability and safety. Another major aspect in the development of novel energy systems is the assessment of their impacts on the environment compared to current technologies. While most existing life cycle assessment (LCA) studies have been applied to conventional nuclear power plants, this dissertation proposes a framework to extend such studies to advanced reactor designs, using the example of the Mk1 PB-FHR. The Mk1 uses a nuclear air-Brayton combined cycle designed to produce 100 MWe of base-load electricity when operated with only nuclear heat, and 242 MWe using natural gas co-firing for peaking power. The Mk1 design provides a basis for quantities and costs of major classes of materials involved in building the reactor and fabricating fuel, and operation parameters. Existing data and economic input-output LCA models are used to calculate greenhouse gas emissions per kWh of electricity produced over the life cycle of the reactor. Baseline life cycle emissions from the Mk1 PB-FHR in base-load configuration are 26% lower than average Generation II light water reactors in the U.S., 98% lower than average U.S. coal plants and 96% lower than average U.S. natural gas combined cycle plants using the same turbine technology. In peaking configuration, due to its nuclear component and higher thermal efficiency, the Mk1 plant only produces 32% of the emissions of average U.S. gas turbine simple cycle peaking plants. One key contribution to life cycle emissions results from the amount and type of concrete used for reactor construction. This is an incentive to develop innovative construction methods using optimized steel-concrete composite wall modules and new concrete mixes to reduce life cycle emissions from the Mk1 and other advanced reactor designs.
