Journal ArticleDOI
A Newly Developed Suppression Pool Model Based on the ISAA Code
TLDR
A newly developed suppression pool model based on the self-developed severe accident analysis code Integrated Severe Accident Analysis (ISAA), which combines the advantages of the dedicated vent flow model and the SPARC-90 model to analyze the suppression pool’s thermal-hydraulic behavior is presented.Abstract:
The suppression pool is an important component in a boiling water reactor nuclear power plant. Under design-basis, loss-of-coolant accident conditions, pressure in the containment increases. Gas fl...read more
Citations
More filters
Station Blackout at Browns Ferry Unit One - accident sequence analysis. Volume 1
TL;DR: In this article, the authors describe the predicted response of Unit 1 at the Browns Ferry Nuclear Plant to Station Blackout, defined as a loss of offsite power combined with failure of all onsite emergency diesel-generators to start and load.
Journal ArticleDOI
Performance Evaluation of Accident Tolerant Fuel under Station Blackout Accident in PWR Nuclear Power Plant by Improved ISAA code
TL;DR: In this article , an improved version of ISAA, referred to as ISAA-ATF, is introduced to analyze the station blackout accident of PWR using ATF cladding.
Journal ArticleDOI
Study on the effect of flow blockage due to rod deformation in QUENCH experiment
TL;DR: In this paper , the authors integrated the developed core Fuel Rod Thermal-Mechanical Behavior Analysis (FRTMB) module into the self-developed severe accident analysis code ISAA to make it possible to simulate the change of flow distribution due to fuel rod deformation.
ReportDOI
PACER -- A fast running computer code for the calculation of short-term containment/confinement loads following coolant boundary failure. Volume 1: Code models and correlations
TL;DR: PACER as discussed by the authors was originally developed to calculate containment/confinement pressure and temperature time histories for loss-of-coolant accidents in Soviet-designed VVER reactors and is relevant to the activities of the US International Nuclear Safety Center.
Journal ArticleDOI
Development and Validation of Thermal-Mechanical Creep Failure Module for Reactor Pressure Vessel Lower Head
TL;DR: In this article , a thermal-mechanical creep failure (LHTCF) module is developed based on the theory of plate and shell and Norton-type constructive creep laws, and seven failure criteria are used to evaluate the integrity of the lower head.
References
More filters
ReportDOI
MELCOR computer code manuals
R.M. Summers,R.K. Cole,R.C. Smith,D.S. Stuart,S.L. Thompson,S. A. Hodge,C. R. Hyman,R. L. Sanders +7 more
TL;DR: MELCOR as discussed by the authors is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants and is developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package.
ReportDOI
Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis
K.K. Murata,D.C. Williams,R.O. Griffith,R.G. Gido,E.L. Tadios,F.J. Davis,G.M. Martinez,K.E. Washington,J. Tills +8 more
TL;DR: The CONTAIN 2.0 as discussed by the authors is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident.
ReportDOI
Fukushima Daiichi accident study : status as of April 2012.
Randall O. Gauntt,Donald A. Kalinich,Jeffrey N Cardoni,Jesse Phillips,Andrew Scott Goldmann,Susan Y. Pickering,Matthew W. Francis,Kevin R Robb,Larry J. Ott,Dean Wang,Curtis Smith,Shawn W. St. Germain,David Schwieder,Cherie Phelan +13 more
TL;DR: In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission (NRC) and Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing severe accident modeling capability of the MELCOR code as mentioned in this paper.
Journal ArticleDOI
Pressure suppression pool mixing in passive advanced BWR plants
TL;DR: In this paper, the authors look at the various phases and phenomena present during the blowdown event and identify those that are important to thermal stratification, and the scaling necessary to model them in reduced size tests.
ReportDOI
SPARC-90: A code for calculating fission product capture in suppression pools
P.C. Owczarski,K.W. Burk +1 more
TL;DR: Owczarski, Postma, and Schreck as mentioned in this paper described the technical bases and use of two updated versions of a computer code initially developed to serve as a tool for calculating aerosol particle retention in boiling water reactor (BWR) pressure suppression pools during severe accidents.
Related Papers (5)
Experimental investigation of BWR Suppression Pool stratification during RCIC system operation
Evaluation of pool swell velocity during large break loss of coolant accident in boiling water reactor Mark III containment design
Jin Yan,Francis Bolger +1 more