NACOM code for analysis of postulated sodium spray fires in LMFBRs

Tsai, S.S.

doi:10.2172/5120107

Cited by 29 publications

(7 citation statements)

References 6 publications

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“…Historically, a variety of sodium fire codes have been utilized in the U.S. for the assessment of room temperatures and pressures associated with sodium leaks or spills. These include SOFIRE-II [323] and NACOM [324], which are no longer supported. The models within these codes were later integrated in CONTAIN-LMR 1.0 [325] and have now also been added to MELCOR 2.2 [83].…”

Section: Sodium Firesmentioning

confidence: 99%

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

Clark¹,

Luxat²,

Higgins³

et al. 2020

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Section: Sodium Firesmentioning

confidence: 99%

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

Clark¹,

Luxat²,

Higgins³

et al. 2020

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“…Sodium spray fire analysis. [34] Not currently maintained. The most recent available report was published in 1980 by Brookhaven National Laboratory as NUREG/CR-1405.…”

Section: Nacommentioning

confidence: 99%

Fast Reactor Methods Development Plan

Fanning

Smith

Brunett

2018

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Fast reactor design and safety analysis methods (computer codes) have been under development for approximately sixty years. During the first forty years, code development was driven by the immediate needs of operational facilities such as Experimental Breeder Reactors I and II and the Fast Flux Test Facility. Development was further driven by programmatic needs to support advanced reactor programs such as the Clinch River Breeder Reactor Project and the Advanced Liquid Metal Reactor Program. By the early 1990s, design and safety analysis codes had been well established to addressed the unique characteristics of fast reactors.During the subsequent twenty years, however, development was nearly stagnant as a result of U.S. Government policy changes that abruptly canceled nearly all of the fast reactor research in the United States. As a result, the state of fast reactor design and safety analysis methods has deteriorated.This report defines a Fast Reactor Methods Development Plan that describes the scope of maintenance and development activities that are required to ensure the existing fast reactor design and safety analysis codes are available to adequately support a fast reactor authorization basis or license application. Of highest priority is the need to establish a software development team that will implement and execute quality software engineering practices to preserve the tremendous amount of knowledge and experience gained during the preceding six decades.Design and safety analysis codes for fast reactors provide unique capabilities not available in codes designed for light-water reactors. By 1980, the number of codes available for fast reactor design and safety analysis was around 130 in the U.S. alone. By 2010, the number had dwindled to approximately 60 worldwide, with less than eighteen potentially available in the U.S. Of these, seven are identified in this report as being critical for design and safety analysis:

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“…Sodium droplet combustion model of SPHINCS and AQUA-SF is based on a theory of so-called D 2 -law. While more detailed modeling basis and assumption are described in the other papers (30), (13) , the combustion rate m f of sodium droplet is calculated with the following equation.…”

Section: Computational Model Of Droplet Combustionmentioning

confidence: 99%

Development of PIRT and Assessment Matrix for Verification and Validation of Sodium Fire Analysis Codes

Ohno

Ohshima

Tajima³

et al. 2012

JPES

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Thermodynamic consequence in liquid sodium leak and fire accident is one of the important issues to be evaluated when considering the safety aspect of fast reactor plant building. The authors are therefore initiating systematic verification and validation (V&V) activity to assure and demonstrate the reliability of numerical simulation tool for sodium fire analysis. The V&V activity is in progress with the main focuses on already developed sodium fire analysis codes SPHINCS and AQUA-SF. The events to be evaluated are hypothetical sodium spray, pool, or combined fire accidents followed by thermodynamic behaviors postulated in a plant building. The present paper describes that the 'Phenomena Identification and Ranking Table (PIRT)' is developed at first for clarifying the important validation points in the sodium fire analysis codes, and that an 'Assessment Matrix' is proposed which summarizes both separate effect tests and integral effect tests for validating the computational models or whole code for important phenomena. Furthermore, the paper shows a practical validation with a separate effect test in which the spray droplet combustion model of SPHINCS and AQUA-SF predicts the burned amount of a falling sodium droplet with the error mostly less than 30%.

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NACOM code for analysis of postulated sodium spray fires in LMFBRs

Cited by 29 publications

References 6 publications

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

Fast Reactor Methods Development Plan

Development of PIRT and Assessment Matrix for Verification and Validation of Sodium Fire Analysis Codes

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