Download or read book Safety of Sodium Cooled Fast Reactors written by Songbai Cheng and published by Springer Nature. This book was released on 2021-09-27 with total page 313 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book highlights the advances and trends in the safety analysis of sodium-cooled fast reactors, especially from the perspective of particle bed-related phenomena during core disruptive accidents. A sodium-cooled fast reactor (SFR) is an optimized candidate of the next-generation nuclear reactor systems. Its safety is a critical issue during its R&D process. The book elaborates on research progresses in particle bed-related phenomena in terms of the molten-pool mobility, the molten-pool sloshing motion, the debris bed formation behavior, and the debris bed self-leveling behavior. The book serves as a good reference for researchers, professionals, and postgraduate students interested in sodium-cooled fast reactors. Knowledge provided is also useful for those who are engaging in severe accident analysis for lead-cooled fast reactors and light water reactors.

Download or read book Analysis of Accidents in Sodium cooled Fast Reactors written by Whitney Ann Wutzler and published by . This book was released on 2011 with total page 70 pages. Available in PDF, EPUB and Kindle. Book excerpt: Abstract: The research presented in this paper is part of a project to help develop a risk-informed approach to the design optimization and licensing of a sodium-cooled fast reactor (SFR). The objectives of this research include analyzing accident scenarios that occur in SFRs and assessing their offsite consequences.

Download or read book Safety of Sodium Cooled Fast Reactors Molten pool Mobility Chapter 3 Molten pool Sloshing Motion Chapter 4 Debris Bed Formation Behavior written by Songbai Cheng and published by . This book was released on 2021 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book highlights the advances and trends in the safety analysis of sodium-cooled fast reactors, especially from the perspective of particle bed-related phenomena during core disruptive accidents. A sodium-cooled fast reactor (SFR) is an optimized candidate of the next-generation nuclear reactor systems. Its safety is a critical issue during its R&D process. The book elaborates on research progresses in particle bed-related phenomena in terms of the molten-pool mobility, the molten-pool sloshing motion, the debris bed formation behavior, and the debris bed self-leveling behavior. The book serves as a good reference for researchers, professionals, and postgraduate students interested in sodium-cooled fast reactors. Knowledge provided is also useful for those who are engaging in severe accident analysis for lead-cooled fast reactors and light water reactors.

Download or read book A Methodology for Analyzing the Consequences of Accidents in Sodium cooled Fast Reactors written by Acacia Joann Brunett and published by . This book was released on 2010 with total page 118 pages. Available in PDF, EPUB and Kindle. Book excerpt: Several accident scenarios are examined in the radionuclide release and transport analysis: varying sizes of sodium spills inside containment, arrested melt scenarios, an energetic recriticality event, core uncovery and a case of sustained sodium vaporization. The scenarios will be analyzed for both containment intact and containment failed, where the containment has a design basis leak rate of 1.0 v/o per day. Resulting offsite consequences are then compared with the NRC'S frequency-consequence (F-C) curve in the Technology Neutral Framework (TNF) for compliance. Calculations are also performed to examine the implications of satisfying the F-C curve and its relationship to satisfaction of the NRC's Quantitative Health Objectives.

Download or read book Containment Source Terms for Sodium cooled Fast Reactor Accidents written by Marissa Umbel and published by . This book was released on 2011 with total page 65 pages. Available in PDF, EPUB and Kindle. Book excerpt: Abstract: The objective of this research was to develop a methodology to determine containment source terms for sodium fast reactors as well as to develop a code to be used for modeling severe accident progressions in SFRs. An integrated analysis tool comparable to MELCOR does not exist for SFR accident scenario analysis that is capable of predicting radionuclide release and transport, so it was necessary to develop the RCS code. The RCS code developed in this project assesses transient RCS thermal-hydraulic conditions, release of radionuclides from fuel, transport of gases, vapors, and aerosols via bubbles to the cover gas region, and vaporization of sodium from the pool and partitioning of radionuclides in solution in the pool to the cover gas region. The RCS code also performs a mass and energy balance of a two-region pool and calculates reactor kinetics. In order to support the demonstration of risk-informed approach to the design optimization of an SFR in a DOE-funded research program, it was necessary to make realistic estimates of the consequences of severe accident scenarios. This thesis describes the assumptions used to estimate the magnitude of representative containment source terms for characteristic accident scenarios, which are used as surrogates for licensing basis events in the DOE study . The results of this paper were used in the DOE study as input to the MELCOR code to determine radionuclide release from containment and the magnitude of offsite doses. The results were also compared with NUREG-1465 which describes the NRC's light water reactor containment source terms.

Download or read book Effects of Fuel Type on the Safety Characteristics of a Sodium Cooled Fast Reactor written by Tyler Sumner and published by . This book was released on 2010 with total page pages. Available in PDF, EPUB and Kindle. Book excerpt: A series of accident simulations were performed using INL's thermal hydraulics code RELAP5-3D to analyze steady-state and transient behavior of a sodium cooled fast reactor. The reactor chosen for this study was General Electric's S-PRISM, which is a 1,000 MWt pool-type sodium-cooled fast reactor, designed for either an Oxide or Metal fueled core. Once key core characteristics including power profiles, reactivity feedback coefficients and delayed neutron parameters were calculated, S-PRISM was redesigned for a Nitride fueled core to take advantage of the Nitride fuel's high thermal conductivity and melting temperature. Loss of flow, loss of heat sink, loss of power and inadvertent control rod withdrawal accidents were simulated for each core at beginning, middle and end of cycle to determine if one fuel type provides significant safety advantages over the others.

Download or read book Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety written by and published by . This book was released on 2011 with total page 54 pages. Available in PDF, EPUB and Kindle. Book excerpt: This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the experienced user-base and the experimental validation base was decaying away quickly.

Download or read book Feasibility Study of a 1000 Mwe Sodium cooled Fast Reactor Research and development program written by North American Aviation. Atomics International Division and published by . This book was released on 1966 with total page 82 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Small Liquid Metal Cooled Reactor Safety Study written by N. Brown and published by . This book was released on 2005 with total page 63 pages. Available in PDF, EPUB and Kindle. Book excerpt: The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout the core life, which leads to large reactivity worth in the control systems. The conclusions from the evaluations support the high level of safety that can be achieved with small liquid metal cooled reactors using either approach.

Download or read book Analysis of clad motion during a loss of flow LOF accident in a fast sodium cooled reactor written by P. Henkel and published by . This book was released on 1985 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Advanced Sodium Fast Reactor Accident Source Terms written by and published by . This book was released on 2010 with total page 100 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Hypothetical Core Disassembly Accident Analysis for the Gas cooled Fast Reactor and Liquid Metal Fast Breeder Reactor written by David C. Losey and published by . This book was released on 1979 with total page 144 pages. Available in PDF, EPUB and Kindle. Book excerpt:

Download or read book Reactor Protection System Design Alternatives for Sodium Fast Reactors written by Jacob Dominic DeWitte and published by . This book was released on 2011 with total page 179 pages. Available in PDF, EPUB and Kindle. Book excerpt: Historically, unprotected transients have been viewed as design basis events that can significantly challenge sodium-cooled fast reactors. The perceived potential consequences of a severe unprotected transient in a sodium-cooled fast reactor include an energetic core disruptive accident, vessel failure, and a large early release. These consequences can be avoided if unprotected transients are properly defended against, potentially improving the economics of sodium fast reactors. One way to defend against such accidents is to include a highly reliable reactor protection system. The perceived undesirability of the consequences arising from an unprotected transient has led some sodium fast reactor designers to consider incorporating several design modifications to the reactor protection system, including: self-actuated shutdown systems, articulated control rods, and seismic anticipatory scram systems. This study investigates the performance of these systems in sodium fast reactors. To analyze the impact of these proposed design alternatives, a model to analyze plant performance that incorporates uncertainty analysis is developed using RELAP5-3D and the ABR-1000 as the reference design. The performance of the proposed alternatives is analyzed during unprotected loss of flow and unprotected transient overpower scenarios, each exacerbated by a loss of heat sink. The recently developed Technology Neutral Framework is used to contextually rate performance of the proposed alternatives. Ultimately, this thesis offers a methodology for a designer to analyze reactor protection system design efficacy. The principle results of this thesis suggest that when using the Technology Neutral Framework as a licensing framework for a sodium-cooled fast reactor, the two independent scram systems of the ABR- 1000's reactor protection system perform well enough to screen unprotected transients from the design basis. While a regulator may still require consideration of accidents involving the failure of the reactor protection system, these events will not drive the design of the system. However, self-actuated shutdown systems may be called for to diversify the reactor protection system. Of these, the Curie point latch marginally reduces the conditional cladding damage probability for metal cores because of their rapid inherent feedback effects, but is more effective for the more sluggish oxide cores given reasonably long pump coastdown times. Flow levitated absorbers are highly effective at mitigating unprotected loss of flow events for both fuel types, but are limited in response during unprotected transient overpower events. When considered from a risk-informed perspective, a clear rationale and objective is needed to justify the inclusion of an additional feature such as self-actuated shutdown systems. The use of articulated safety rods as one of the diverse means of reactivity insertion and the implementation of an anticipatory seismic scram system may be the most cost-effective alternatives to provide defense in depth in light of the sodium fast reactor's susceptibility to seismic events.

Download or read book Status of Fast Reactor Research and Technology Development written by International Atomic Energy Agency and published by . This book was released on 2012 with total page 0 pages. Available in PDF, EPUB and Kindle. Book excerpt: "Based on a recommendation from the Technical Working Group on Fast Reactors, this publication is a regular update of previous publications on fast reactor technology. The publication provides comprehensive and detailed information on the technology of fast neutron reactors. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors. The main issues of discussion are experience in design, construction, operation and decommissioning, various areas of research and development, engineering, safety and national strategies, and public acceptance of fast reactors. In the summary the reader will find national strategies, international initiatives on innovative (i.e. Generation IV) systems and an assessment of public acceptance as related to fast reactors."--Résumé de l'éditeur.

Download or read book Dynamic Simulation of Sodium Cooled Fast Reactors written by G Vaidyanathan and published by CRC Press. This book was released on 2022-11-18 with total page 289 pages. Available in PDF, EPUB and Kindle. Book excerpt: This book provides the basis of simulating a nuclear plant, in understanding the knowledge of how such simulations help in assuring the safety of the plants, thereby protecting the public from accidents. It provides the reader with an in-depth knowledge about modeling the thermal and flow processes in a fast reactor and gives an idea about the different numerical solution methods. The text highlights the application of the simulation to typical sodium-cooled fast reactor. The book • Discusses mathematical modeling of the heat transfer process in a fast reactor cooled by sodium. • Compares different numerical techniques and brings out the best one for the solution of the models. • Provides a methodology of validation based on experiments. • Examines modeling and simulation aspects necessary for the safe design of a fast reactor. • Emphasizes plant dynamics aspects, which is important for relating the interaction between the components in the heat transport systems. • Discusses the application of the models to the design of a sodium-cooled fast reactor It will serve as an ideal reference text for senior undergraduate, graduate students, and academic researchers in the fields of nuclear engineering, mechanical engineering, and power cycle engineering.

Download or read book Impact of Reducing Sodium Void Worth on the Severe Accident Response of Metallic fueled Sodium cooled Reactors written by and published by . This book was released on 1994 with total page 9 pages. Available in PDF, EPUB and Kindle. Book excerpt: Analyses have performed on the severe accident response of four 90 MWth reactor cores, all designed using the metallic fuel of the Integrated Fast Reactor (IFR) concept. The four core designs have different sodium void worth, in the range of -3$ to 5$. The purpose of the investigation is to determine the improvement in safety, as measured by the severe accident consequences, that can be achieved from a reduction in the sodium void worth for reactor cores designed using the IFR concept.

Download or read book Accident Analysis SNR fast Sodium Reactor According to Probability Theory written by Juergen Wehling and published by . This book was released on 1977 with total page 165 pages. Available in PDF, EPUB and Kindle. Book excerpt: