Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a Nordic type BWR

Grishchenko, Dmitry; Galushin, Sergey; Kudinov, Pavel

doi:10.1016/j.nucengdes.2018.12.013

Cited by 6 publications

(12 citation statements)

References 27 publications

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“…The jet speed can be derived through the debris mass flow rate ð _ M deb Þ, total breach area (A br Þ, and melt density (qÞ-which is currently assumed to take value equal to 8000 kg/m 3 (average corium density, based on Ref. [17])…”

Section: Postprocessing Of the Results Postprocessing Of Melcor Analmentioning

confidence: 99%

“…Ejected debris density is assumed to be equal to q ¼ 8000ðkg=m 3 Þ. This vale was assumed based on SERENA-II BWR benchmark exercise [17]. Debris ejection rate _ M deb is uniform during MELCOR plotting time-step DT P ðsÞ.…”

Section: Postprocessing Of the Results Postprocessing Of Melcor Analmentioning

confidence: 99%

“…It is important to emphasize that the current ex-vessel steam explosion surrogate model-SEIM SM (see Refs. [8] and [17] for details) is implemented to predict steam explosion loads on the containment per single melt jet, without taking into account possible interactions between multiple jets. Thus, current implementation of the MEM SM is limited to predict melt release conditions (such as jet radius, initial jet velocity) per single jet.…”

Section: Data Base Of Melcor Codementioning

confidence: 99%

“…The results show that in some cases the MEM SM slightly underestimates the tails of the distribution; however, these discrepancies can be taken into account by application of the approaches for quantification of the uncertainty due to SM approximation of FM, implemented in Refs. [16] and [17].…”

Section: Data Base Of Melcor Codementioning

confidence: 99%

“…Steam explosion impact map surrogate model is a fast running surrogate model that has been developed [16] for the assessment of the risk of containment failure due to steam explosion in Nordic BWR. Development of the SM relies on a database of solutions generated by a 1D FCI code TEXAS-V. TEXAS-V is a 1D threefield transient code with Eulerian fields for gas and liquid and a Lagrangian field for fuel particles.…”

Section: Ex-vessel Steam Explosion Surrogatementioning

confidence: 99%

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Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework

Galushin

Grishchenko

Kudinov

2020

Journal of Nuclear Engineering and Radiation Science

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Nordic Boiling Water Reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management strategy. Depending on melt release conditions from the vessel and core-melt coolant interactions, containment integrity can be threatened by formation of non-coolable debris bed, or energetic steam explosion. Melt release from the vessel affect ex-vessel phenomena and is recognized as the major source of uncertainty. The Risk Oriented Accident Analysis Methodology (ROAAM+) is used for quantification of the risk of containment failure in Nordic BWR where Melt Ejection Surrogate Model (MEM SM) provides initial conditions for the analysis of debris agglomeration and ex-vessel steam explosion which determine the respective loads on the containment. MEM SM is based on the system analysis code MELCOR. Modelling of vessel failure and melt release from the vessel in MELCOR is based on parametric models, allowing a user to select different assumptions that effectively control lower head behavior and melt release. The work addresses the effect of epistemic uncertain parameters and modelling assumptions in MEM SM on the containment loads due to ex-vessel steam explosion in Nordic BWR. Sensitivity and uncertainty analysis performed to identify the most influential parameters and uncertainty in the risk of containment failure due to ex-vessel steam explosion. The results of the analysis provide valuable insights regarding the effect of MELCOR models, modelling parameters and sensitivity coefficients on melt release conditions and predictions of ex-vessel steam explosion loads on the containment structures.

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Section: Postprocessing Of the Results Postprocessing Of Melcor Analmentioning

confidence: 99%

Section: Postprocessing Of the Results Postprocessing Of Melcor Analmentioning

confidence: 99%

Section: Data Base Of Melcor Codementioning

confidence: 99%

Section: Data Base Of Melcor Codementioning

confidence: 99%

Section: Ex-vessel Steam Explosion Surrogatementioning

confidence: 99%

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Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework

Galushin

Grishchenko

Kudinov

2020

Journal of Nuclear Engineering and Radiation Science

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Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code

Galushin

Kudinov

2020

Annals of Nuclear Energy

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Integrated deterministic and probabilistic safety assessment of a superconducting magnet cryogenic cooling circuit for nuclear fusion applications

Bellaera

Bonifetto

Maio

et al. 2020

Reliability Engineering & System Safety

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The most promising configuration of a nuclear energy fusion system is the tokamak, the largest of which, called ITER, is under construction in Cadarache, France, which uses a complex system of superconducting magnets to generate a field of several tesla (T), aimed at confining the plasma in the toroidal chamber where nuclear fusion reactions occur. For industrial development, the safety of nuclear fusion systems has to be proved and verified by a systematic analysis of operational transients and accidental conditions. Although the final aim of fusion reactors is to reach steady state operation, present-day tokamaks present complex dynamic features, as their operation is based on the transformer principle with a subset of the superconducting magnets operating in a pulsed mode, to inductively generate plasma currents of the order of several MA. We adopt the framework of Integrated Deterministic and Probabilistic Safety Assessment (IDPSA), for identifying the component failures that may cause a Loss-Of-Flow-Accident (LOFA) in the cooling circuit of a superconducting magnet for fusion applications. Post-processing of the simulated scenarios for the identification of the abnormal transients is performed in an unsupervised manner resorting to a spectral clustering approach embedding a Fuzzy-C Means (FCM) that is compared with an Extended Symbolic Aggregate approXimation (ESAX) from the literature that also resorts to the FCM for the classification.The proposed approach turns out to be more efficient than ESAX in the identification of clusters and "prototypical states" of abnormal system behavior. Results show that none of the identified scenarios (even those leading to a LOFA) are critical for the ITER central solenoid module integrity, in the mode of operation considered.

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Failure domain analysis and uncertainty quantification using surrogate models for steam explosion in a Nordic type BWR

Cited by 6 publications

References 27 publications

Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework

Analysis of the Effect of Vessel Failure and Melt Release on Risk of Containment Failure Due to Ex-Vessel Steam Explosion in Nordic Boiling Water Reactor Using ROAAM+ Framework

Sensitivity and uncertainty analysis of the vessel lower head failure mode and melt release conditions in Nordic BWR using MELCOR code

Integrated deterministic and probabilistic safety assessment of a superconducting magnet cryogenic cooling circuit for nuclear fusion applications

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