Quantifying reactor safety margins

application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

Publisher: Division of Systems Research, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission in Washington, DC

Written in English
Published: Pages: 239 Downloads: 748
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Subjects:

  • Water cooled reactors -- Loss of coolant -- Computer simulation.,
  • Light water reactors -- Accidents -- Computer simulation.,
  • Light water reactors -- Safety measures.

Edition Notes

Statementprepared by technical program group, B. Boyack ... [et al.] ; contributing authors, K.R. Katsma ... [et al.].
ContributionsBoyack, B. E., Katsma, K. R., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research., Idaho National Engineering Laboratory., EG & G Idaho.
The Physical Object
FormatMicroform
Paginationxv, 114, [239] p.
Number of Pages239
ID Numbers
Open LibraryOL14695690M

Probabilistic safety assessment (PSA) of nuclear facilities on external multi-hazards has become a major issue after the Fukushima accident in However, the existing external hazard PSA methodology is for single hazard events and cannot cover the impact of multi-hazards. Therefore, this study proposes a methodology for quantifying multi-hazard risks for nuclear energy : Shinyoung Kwag, Jeong Gon Ha, Min Kyu Kim, Jung Han Kim. The applications of predictive modeling to nuclear energy includes sensitivity analysis and uncertainty quantification in reactor physics and thermal-hydraulics, reducing the uncertainties in reactor design margins, reprocessing, and waste disposal, non-proliferation and safeguards. Prof. Dan Gabriel Cacuci Guest Editors. Uncertainty is ubiquitous in modern decision-making supported by quantitative modeling. While uncertainty treatment has been initially largely developed in risk or environmental assessment, it is gaining large-spread interest in many industrial fields generating knowledge and practices going beyond the classical risk versus uncertainty or epistemic versus aleatory by: 3.

Quantifying reactor safety margins Download PDF EPUB FB2

Conference: Quantifying reactor safety margins: Application of CSAU (Code Scalability, Applicability and Uncertainty) methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codesAuthor: W.

Wulff, B.E. Boyack, R.B. Duffey, P. Griffith, K.R. Katsma, G.S. Lellouche, S. Levy, U.S. Rohatgi. Quantification of Margins and Uncertainties (QMU) is a methodology for assessing the confidence in the performance of complex systems, and has been applied to general engineering areas.

Code scaling, applicability, and uncertainty or CSAU methodology is a systematic approach proposed by the US Nuclear Regulatory Commission that can be used to identify and quantify overall nuclear reactor uncertainties, and this estimate methods for reactor safety analysis is in lieu of the earlier licensing practice that used deterministic methods with conservative.

@article{osti_, title = {Quantifying reactor safety margins: Application of CSAU (Code Scalability, Applicability and Uncertainty) methodology to LBLOCA: Part 3, Assessment and ranging of parameters for the uncertainty analysis of LBLOCA codes}, author = {Wulff, W and Boyack, B E and Duffey, R B and Griffith, P and Katsma, K R and Lellouche, G S and Levy, S.

Get this from a library. Quantifying reactor safety margins: application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident. [B E Boyack; K R Katsma; U.S.

Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Systems Research.; Idaho National Engineering Laboratory. Quantifying reactor safety margins, Part 1: An overview of the code scaling, applicability, and uncertainty evaluation methodology.

Nuclear Engineering and Design.1–Author: William L. Oberkampf, Christopher J. Roy. In a method of determining an operating margin to a given operating limit in a nuclear reactor, operational plant data from an on-line nuclear reactor plant is accessed, and reactor operation is simulated off-line using the operational plant data to generate predicted dependent variable data representative of the given operating by: 4.

In addition, the book could serve as a text for a course on fast reactor safety since many topics other than those appearing in the safety chapters relate to FBR safety.

Methodology in fast. Roache’s book on V&V is available from Hermosa Publishers and well worth the price. A discussion with recent references can also be found in "Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety Applications," by Mahaffy et al. Errors.

Reactor Safety Analysis. Such analysis is performed for all phases of the facility’s life cycle. A modified method of quantifying risk through the use of systems analysis has been adopted by the process industries.

Among these factors are the fundamental safety principles of defence-in-depth, safety margins, the principle of ALARA.

Quantifying Reactor Safety Margins—Part I: An Overview of the Code Scaling, Applicability and Uncertainty Evaluation Methodology Proceedings of the International Fast Reactor Safety Meeting, Snowbird, Utah, American Nuclear Society, La Grange Park, IL Bootstrap and Order Statistics for Quantifying Thermal-Hydraulic Code Cited by: 1.

Lellouche, G.S., Levy, S.: Quantifying reactor safety margins, part 4: quantifying reactor safety margins part 4: Uncertainty evaluation of lbloca analysis based on trac-pf1/mod 1.

Nucl. Eng. Des.67–95 () CrossRef Google ScholarAuthor: Mihály Makai, János Végh. Independent of reactor class, existing modern safety analyses are based on the twin directions of (1) assessing potential event initiators and quantifying estimates of the sequence evolution and responses using PSA/PRA and (2) Deterministic Analyses (DA) in which postulated events are analyzed largely independent of their likelihood.

A 'read' is counted each time someone views a publication summary (such as the title, abstract, and list of authors), clicks on a figure, or views or downloads the full-text. Unfortunately, this book can't be printed from the OpenBook. If you need to print pages from this book, we recommend downloading it as a PDF.

Visit to get more information about this book, to buy it in print, or to download it as a free PDF. NUREG/CR “Quantifying Reactor Safety Margins Application of Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large Break, Loss-of-Coolant Accident” provides the general methodologies to be used in the development of realistic loss of coolant safety : Andrew B.

French. The Emergency Core Coolant (ECC) is injected in the reactor cold legs to flow down the downcomer and to cool the hot core during a Large Break Loss of Coolant Accident. The ECC flows down in the downcomer before reaching the core entrance.

The delivery of ECC to the core is critical to the by: 1. Introduction. Safety analysis of nuclear reactors (both research and power) (IAEA, a, IAEA, b) is performed by constructing models in best-estimate computer codes that use the best available physics (Hernández-Solís, ), techniques, and practices to simulate steady-state, accident, and anticipated transient r, the use of best-estimate codes.

Garrick, B.J., “Research and Development in Reactor Safety,” a booklet prepared for the U.S. Atomic Energy Commission describing the U.S. Atomic Energy Commission Reactor Safety Research Program, for Distribution at the Second International Conference on the Peaceful Uses of Atomic Energy, Geneva, Switzerland,U.S.

Government Printing. The full book can be obtained from Hermosa Publishers and is worth the price. A discussion with recent references can also be found in "Best Practice Guidelines for the use of CFD in Nuclear Reactor Safety Applications," by Mahaffy et al. Verification must precede Validation.

This paper presents an as-loaded canister-specific criticality analysis approach for quantifying inherent uncredited margins in already loaded DPCs. The as-loaded analysis approach has been implemented in a new SNF management and analysis tool - The Used Nuclear Fuel-Storage, Transportation, and Disposal Analysis Resource and Data System (UNF.

good book retains its interior heat and will warm a generation yet unborn.— replicating the operation of the reactor and quantifying the current radioisotope inventory. Extensive criticality safety reviews were performed in MCNP5 for the establishment of safety margins for the TSF-SNAP dismantlement.

This work is. Safety and Reliability – Theory and Applications contains the contributions presented at the 27th European Safety and Reliability Conference (ESRELPortorož, Slovenia, June).

The book covers a wide range of topics, including: • Accident and Incident modelling • Economic Analysis in Risk ManagementAuthor: Marko Cepin.

The analysis of the mechanical integrity of Light Water Nuclear Reactors has experienced a strong evolution in the two latest decades. Until the nineteen eighties, the structural design was practically based on static loads with amplifying factors to take dynamic effects into account, (see e.

Laheyand Moody, ), and on wide safety margins. Even if an incipient methodology Cited by: 2. Safety, Reliability and Risk Analysis. Theory, Methods and Applications contains the papers presented at the joint ESREL (European Safety and Reliability) and SRA-Europe (Society for Risk Analysis Europe) Conference (Valencia, Spain, September ).

The book covers a wide range of topics, including: Accident and Incident Investigation; CrisiCited by: 3. Nuclear experts from across the Atlantic have reached out to U.S. Department of Energy engineers for help evaluating a new reactor design that could increase safety margins while reducing waste :: Read Article» at the INL website: AM: Rising Stars in Nuclear Science and Engineering Symposium.

related to reactor safety. This risk characterization method was the first of a set of methods and tools developed that became central elements of the Significance Determination Process (SDP) to determine reactor inspection finding significance consistent with the thresholds used for the risk-informed plant Performance Indicators (PIs).

Main Safety, reliability and risk analysis theory, methods and applications. Safety, reliability and risk analysis theory, methods and applications Martorell S, Soares C G, Barnett J (eds) Categories: Mathematics\\Analysis You can write a book review and share your experiences. Other readers will always be interested in your opinion of the.

PROBLEM TO BE SOLVED: To provide a method for and a device of reactor simulation. SOLUTION: In this method for reactor simulation, a user modifies one or more design inputs used for producing response surfaces (S16).

The response surface defines the relation between the design input and operation outputs in one or more by: 4. Safety and Reliability – Theory and Applications contains the contributions presented at the 27th European Safety and Reliability Conference (ESRELPortorož, Slovenia, June).

The book covers a wide range of topics, including:• Accident and Incident modelling• Economic Analy. Risk, Reliability and Safety contains papers describing innovations in theory and practice contributed to the scientific programme of the European Safety and Reliability conference (ESREL ), held at the University of Strathclyde in Glasgow, Scotland (25—29 September ).Authors include scientists, academics, practitioners, regulators and other key individuals Price: $The health and safety risks to be discussed in this chapter arise from exposures of people who travel, work, or live near transportation routes, and transportation workers themselves, to radiation (Sidebar ) from loaded spent fuel and high-level waste transportation packages.

During normal operations, such exposures can occur as the result.Sixteenth Water Reactor Safety Information Meeting Volume 1 - Plenary Session - Decontamination and Decommissioning safety margins or cause unexpected common mode failures in vital safety Book Company, 2) W.

J. Marble.