Simplified modeling of a PWR reactor pressure vessel lower head failure in the case of a severe accident

Koundy, V.; Durin, M.; Nicolas, L.; Combescure, Alain

doi:10.1016/j.nucengdes.2004.11.012

Cited by 21 publications

(2 citation statements)

References 2 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…(2), was used for the creep analysis. In the failure evaluation method, the creep damage variable D was calculated for each element using creep damage models based on "considère," strain, LMP, or Kachanov damage criteria (NEA, 2003), (Koundy, 2005). Damage variable D corresponds to ω for the K-R model.…”

Section: Creep Analysis and Creep Damage Evaluation 31 Details Of Thmentioning

confidence: 99%

Development of failure evaluation method for BWR Lower head in severe accident; high temperature creep test and creep damage model

Yamaguchi

Katsuyama

Nemoto

et al. 2017

Mechanical Engineering Journal

View full text Add to dashboard Cite

After the Fukushima Daiichi nuclear power plant accident due to the pacific coast of Tohoku earthquake and Tsunami, we have been developing an analysis method considering creep damage mechanisms based on threedimensional analysis in order to estimate condition of the fuel debris and failure behavior of the reactor due to severe accidents for the early completion of the decommissioning of nuclear power plants. We have also been obtaining material properties that are not provided in existing databases or literature for the analysis. In this study, we measure the tensile and creep properties of low alloy steel, Ni-based alloy, and stainless steel at high temperatures near the melting points. Using experimental data, some parameters for the creep constitutive law and creep failure evaluation method are determined. It is confirmed that the estimated rupture time by the failure evaluation method agree well with the experimental one.

show abstract

Section: Creep Analysis and Creep Damage Evaluation 31 Details Of Thmentioning

confidence: 99%

Development of failure evaluation method for BWR Lower head in severe accident; high temperature creep test and creep damage model

Yamaguchi

Katsuyama

Nemoto

et al. 2017

Mechanical Engineering Journal

View full text Add to dashboard Cite

show abstract

“…Since early failure tends to have more severe consequences it is usually emphasized in APET analyses. Many mechanisms are present that can lead to early containment failure such as rapid overpressurization from severe accident phenomena such as direct containment heating that can result from reactor pressure vessel lower head failure with the primary system at high pressure [66,67] or explosive increase in pressure due to a steam explosion [68,69] or from hydrogen deflagration or detonation [70,71,72]. The main mechanism of late containment failure is through slow pressure buildup because of failure of AC-powered containment heat removal systems.…”

Section: Containment Overpressure Failurementioning

confidence: 99%

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

Kunsman¹,

Aldemir

Rutt

et al. 2008

View full text Add to dashboard Cite

This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors-analyses which are very resource intensive-more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine.Initially, the objective of the work was to fuse the accident progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code-the type on which APET construction and analysis relies, albeit indirectly-and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment.As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties.This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc.Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully.This research and development effort has met its original goals and then exceeded them.4

show abstract

Study on the lower head failure in severe accident Part I: Model development and validation

Gao

Zhang

Yang

et al. 2023

Annals of Nuclear Energy

View full text Add to dashboard Cite

Simplified modeling of a PWR reactor pressure vessel lower head failure in the case of a severe accident

Cited by 21 publications

References 2 publications

Development of failure evaluation method for BWR Lower head in severe accident; high temperature creep test and creep damage model

Development of failure evaluation method for BWR Lower head in severe accident; high temperature creep test and creep damage model

Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

Study on the lower head failure in severe accident Part I: Model development and validation

Contact Info

Product

Resources

About