| 1. |
- Galusin, Sergey, 1985-, et al.
(författare)
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An approach to joint application of integrated deterministic-probabilistic safety analysis and PSA Level 2 to severe accident issues in Nordic BWRs
- 2016
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Ingår i: PSAM 2016 - 13th International Conference on Probabilistic Safety Assessment and Management2017. - : International Association for Probabilistic Safety Assessment and Management (IAPSAM).
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Konferensbidrag (refereegranskat)abstract
- In this paper we outline a conceptual approach for combined use of Probabilistic Safety Assessment (PSA) and Integrated Deterministic-Probabilistic Safety Assessment (IDPSA), considering Nordic Boiling Water Reactor (BWR) severe accident issues (specifically ex-vessel steam explosion and debris bed coolability) for illustration.We describe a conceptual approach based on post processing of the results generated by IDPSA to update the “static” Boolean structures in the standard PSA representation. The challenge in the evaluation is to retain the failure combinations from the PSA to allow for component importance evaluation, to be able to perform the calculations in a reasonable time frame and to use all relevant information from the IDPSA results.We discuss the approaches for determination of the event space (for IDPSA analysis) which is consistent with PSA damage states from PSA-L1 and L2. We also discuss application of post processing approach for analysis of huge amount of data generated in the process of uncertainty space exploration, which is difficult to use directly in decision making process including incorporation of such data into PSA framework, to update structures of “static” Boolean structures in standard PSA. Using data post processing approach we can significantly reduce amount of information which represents results from IDPSA analysis.
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| 2. |
- Bandini, G., et al.
(författare)
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Assessment of systems codes and their coupling with CFD codes in thermal-hydraulic applications to innovative reactors
- 2015
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 281, s. 22-38
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Tidskriftsartikel (refereegranskat)abstract
- The THINS project of the 7th Framework EU Program on nuclear fission safety is devoted to the investigation of crosscutting thermal hydraulic issues for innovative nuclear systems. A significant effort in the project has been dedicated to the qualification and validation of system codes currently employed in thermal hydraulic transient analysis for nuclear reactors. This assessment is based either on already available experimental data, or on the data provided by test campaigns carried out in the frame of THINS project activities. Data provided by TALL and CIRCE facilities were used in the assessment of system codes for HLM reactors, while the PHENIX ultimate natural circulation test was used as reference for a benchmark exercise among system codes for sodium-cooled reactor applications. In addition, a promising grid-free pool model based on proper orthogonal decomposition is proposed to overcome the limits shown by the thermal hydraulic system codes in the simulation of pool-type systems. Furthermore, multi-scale system-CFD solutions have been developed and validated for innovative nuclear system applications. For this purpose, data from the PHENIX experiments have been used, and data are provided by the tests conducted with new configuration of the TALL-3D facility, which accommodates a 3D test section within the primary circuit. The TALL-3D measurements are currently used for the validation of the coupling between system and CFD codes.
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| 3. |
- Bandini, G., et al.
(författare)
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Safety Analysis Results of Representative DEC Accidental Transients for the ALFRED Reactor
- 2013
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Konferensbidrag (refereegranskat)abstract
- The conceptual design of the Advanced Lead Fast Reactor European Demonstrator (ALFRED) is under development within the LEADER project to meet the safety objectives of Gen IV nuclear energy systems. This paper presents the main results of the safety analysis for beyond design basis conditions, namely design extension conditions (DEC), which include the failure of prevention and mitigation systems, like the reactor scram in the so called unprotected transients. The main objective of this analysis is to evaluate the impact of the core and plant design features on the intrinsic safety behaviour of the ALFRED reactor. Several computer codes: SIM LFR, RELAP5, CATHARE, SPECTRA and TRACE are applied to evaluate the consequences of representative unprotected accident scenarios such as Loss of Flow, Loss of Heat Sink and Reactivity initiated accidents. Additionally, the consequences of steam generator tube rupture and partial sub assembly flow blockage events are assessed by means of appropriate fluid dynamic codes.
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| 4. |
- Basso, Simone, et al.
(författare)
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Development of scalable empirical closures for self-leveling of particulate debris bed
- 2014
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Ingår i: Proceedings of ICAPP 201, Paper 14330. - : American Nuclear Society. - 9780894484605 ; , s. 14330-
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Konferensbidrag (refereegranskat)abstract
- Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel is employed as a severe accident mitigation strategy in several designs of light water reactors. Geometrical configuration of the debris bed is one of the factors which define if the decay heat can be removed from the debris bed by natural circulation. A bed can be coolable if spread uniformly, while the same debris forming a tall mound-shape debris bed can be non-coolable. Two-phase flow inside the bed serves as a source of mechanical energy which can move debris, thus flatten and gradually reduce the height of the debris bed. There is a competition between the time scales for (i) reaching a coolable configuration of the bed by such “self-leveling” phenomenon, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local (i) gas velocity, and (ii) slope angle of the bed. The goal of this work is to obtain a dependency of particle motion rate on local slope angle and gas velocity expressed in non-dimensional variables, universal for particles of different shapes, sizes and materials. Such scaling approach is proposed in this work and validated against experimental data.
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| 5. |
- Basso, Simone, et al.
(författare)
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Effectiveness of the debris bed self-leveling under severe accident conditions
- 2016
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 95, s. 75-85
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Tidskriftsartikel (refereegranskat)abstract
- Melt fragmentation, quenching and long term coolability in a deep pool of water under the reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. The success of such strategy is contingent upon the natural circulation effectiveness in removing the decay heat generated in the porous debris bed. The maximum height of the bed is one of the important factors which affect the debris coolability. The two-phase flow within the bed generates mechanical energy which can change the geometry of the debris bed by the "self-leveling" phenomenon. In this work.we developed an approach to modeling of the self-leveling phenomenon. Sensitivity analysis was carried out to rank the importance of the model uncertainties and uncertain input parameters i.e. the conditions of the accident scenario and the debris bed properties. The results provided some useful insights for further improvement of the model and reduction of the output uncertainties through separate-effect experimental studies. Finally, we assessed the self-leveling effectiveness, quantified its uncertainties in prototypic severe accident conditions and demonstrated that the effect of self-leveling phenomenon is robust with respect to the considered input uncertainties.
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| 6. |
- Basso, Simone, et al.
(författare)
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Empirical closures for particulate debris bed spreading induced by gas-liquid flow
- 2016
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 297, s. 19-25
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Tidskriftsartikel (refereegranskat)abstract
- Efficient removal of decay heat from the nuclear reactor core debris is paramount for termination of severe accident progression. One of the strategies is based on melt fragmentation, quenching and cooling in a deep pool of water under the reactor vessel. Geometrical configuration of the debris bed is among the important factors which determine possibility of removing the decay heat from the debris bed by natural circulation of the coolant. For instance, a tall mound-shape debris bed can be non-coolable, while the same debris can be coolable if spread uniformly. Decay heat generates a significant amount of thermal energy which goes to production of steam inside the debris bed. Two-phase flow escaping through the top layer of the bed becomes a source of mechanical energy which can move the particulate debris along the slope of the bed. The motion of the debris will lead to flattening of the bed. Such process is often called "self-leveling" phenomenon. Spreading of the debris bed by the self-leveling process can take significant time, depending on the initial debris bed configuration and other parameters. There is a competition between the time scales for reaching (i) a coolable configuration of the bed, and (ii) onset of dryout and re-melting of the debris. In the previous work we have demonstrated that the rate of particulate debris spreading is determined by local gas velocity and local slope angle of the bed. In this work we develop a scaling approach and a closure for prediction of debris spreading rate based on generalization of available experimental data. We demonstrate that introduced scaling criteria are universal for particles of different shapes and size distributions.
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| 7. |
- Basso, Simone
(författare)
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Particulate Debris Spreading and Coolability
- 2017
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- In Nordic design of boiling water reactors, a deep water pool under the reactor vessel is employed for the core melt fragmentation and the long term cooling of decay heated corium debris in case of a severe accident. To assess the effectiveness of such accident management strategy the Risk-Oriented Accident Analysis Methodology has been proposed. The present work contributes to the further development of the methodology and is focused on the issue of ex-vessel debris coolability.The height and shape of the porous debris bed are among the most important factors that determine if the debris can be cooled by natural circulation of water. The bed geometry is formed in the process of melt release, fragmentation, sedimentation and packing of the debris in the pool. Bed shape is affected by the coolant flow that induces movement of particles in the pool and after settling on top of the bed. The later one is called debris bed self-leveling phenomenon.In this study, the self-leveling was investigated experimentally and analytically. Experiments were carried out in order to collect data necessary for the development of a numerical model with an empirical closure. The self-leveling model was coupled to a model for prediction of the debris bed dryout. Such coupled code allows to calculate the time necessary to have a coolable configuration of the bed. The influence of input parameters was assessed through sensitivity analysis in order to screen out the less influential parameters.Results of the risk analysis are reported as complementary cumulative distribution functions of the conditional containment failure probability (CCFP).Sensitivity analyses identified: effective particle diameter and debris bed porosity as the parameters that provide the largest contribution to the CCFP uncertainty. It is found that the effect of the initial maximum height of the bed on the CCFP is reduced by the self-leveling.
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| 8. |
- Basso, Simone, et al.
(författare)
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Preliminary Risk assessment of ex-vessle debris bed coolability for a Nordic BWR
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Ingår i: Nuclear Engineering and Design. - 0029-5493 .- 1872-759X.
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Tidskriftsartikel (refereegranskat)abstract
- In Nordic design of boiling water reactors (BWRs) a deep water pool under the reactor vessel is employed as a severe accident management strategy for the core melt fragmentation and the long term cooling of corium debris. The height and shape of the debris bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry is formed as a result of melt release, fragmentation, sedimentation and settlement on the containment basemat. After settlement, the shape can change with time due to movement of particles promoted by the coolant flow (debris bed self-leveling process). Both aleatory (accident scenario, stochastic) and epistemic (modeling, lack of knowledge) uncertainties are important for assessing the risks. The present work describes a preliminary risk analysis of debris bed coolability for Nordic BWRs under severe accident conditions. It was assumed that once debris remelting starts containment failure becomes imminent. Such assumption allows to estimate the containment failure probability by calculating the probability that the time necessary for the spreading debris bed to achieve a coolable configuration will be shorter than the onset time of debris bed re-melting. An artificial neural network was employed as a surrogate model (SM) for the mechanistic full model (FM) of the debris spreading in order to achieve computationally efficient propagation of uncertainties. The effect of uncertainty in the ranges and probability density functions (PDFs) of the input parameters was addressed. Parameters defining shapes of the PDFs were varied for three different distribution families (beta, truncated normal and triangular). The results of the risk analysis were reported as complementary cumulative distribution functions (CCDFs) of the conditional containment failure probability (CCFP). It is demonstrated that CCFP can vary in wide ranges depending on the randomly selected combinations of the PDFs of the input parameters. Given the selected ranges of the input parameters, sensitivity analyses identified: the effective particle diameter and the debris bed porosity as the largest contributors to the CCFP uncertainty. It was shown that the self-leveling phenomenon reduces sensitivity of debris coolability to the initial shape of the bed. However, the initial shape remains an important uncertainty factor for the most likely values of the particle size and porosity. Importance of the initial shape increases when the effectiveness of the self-leveling is small (e.g. in case of high initial temperature or heat up rate of the debris). Findings of this work in combination with consideration of the necessary efforts can be used for prioritization of the future research on obtaining new information on the uncertain parameters.
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| 9. |
- Basso, Simone, et al.
(författare)
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Sensitivity and uncertainty analysis for predication of particulate debris bed self-leveling in prototypic Severe Accident (SA) conditions
- 2014
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Ingår i: Proceedings of ICAPP 2014. - : American Nuclear Society. ; , s. 14329-
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Konferensbidrag (refereegranskat)abstract
- Melt fragmentation, quenching and long term coolability in a deep pool of water under reactor vessel are employed as a severe accident mitigation strategy in several designs of light water reactors. Success of the strategy is contingent upon effectiveness of natural circulation in removing the decay heat generated by the porous debris bed. Geometrical configuration of the bed is one of the factors which affect coolability of the bed. Boiling and two-phase flow inside the bed serve as a source of mechanical energy which can change the geometry of the debris bed by so called “self-leveling” phenomenon. The goals of this work are (i) to further develop self-leveling modeling approach and validate it against data produced in a new series of PDS-C (Particulate Debris Spreading Closures) experiments, and (ii) to carry out sensitivity-uncertainty analysis for the debris bed spreading for the selected cases of prototypic severe accident conditions. The model has been extended to predict spreading in both planar and axisymmetric geometries. The performed sensitivity analysis ranks the importance of different uncertain input parameters such as accident conditions, debris bed properties, modeling parameters and closures. The knowledge about the most influential parameters is important for further improvement of the model and for efficient reduction of output uncertainties through focused, separate-effect experimental studies. Finally, we report results for particulate debris spreading in prototypic severe accident scenarios with assessment of uncertainties.
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| 10. |
- Basso, Simone, et al.
(författare)
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The effect of self-leveling on debris bed coolability under severe accident conditions
- 2016
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Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 305, s. 246-259
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Tidskriftsartikel (refereegranskat)abstract
- Nordic-type boiling water reactors employ melt fragmentation, quenching, and long term cooling of the debris bed in a deep pool of water under the reactor vessel as a severe accident (SA) mitigation strategy. The height and shape of the bed are among the most important factors that determine if decay heat can be removed from the porous debris bed by natural circulation of water. The debris bed geometry depends on its formation process (melt release, fragmentation, sedimentation and settlement on the containment basemat), but it also changes with time afterwards, due to particle redistribution promoted by coolant flow (self-leveling). The ultimate goal of this work is to develop an approach to the assessment of the probability that debris in such a variable-shape bed can reach re-melting (which means failure of SA mitigation strategy), i.e. the time necessary for the slumping debris bed to reach a coolable configuration is larger than the time necessary for the debris to reach the re-melting temperature. For this purpose, previously developed models for particulate debris spreading by self-leveling and debris bed dryout are combined to assess the time necessary to reach a coolable state and evaluate its uncertainty. Sensitivity analysis was performed to screen out less important input parameters, after which Monte Carlo simulation was carried out in order to collect statistical characteristics of the coolability time. The obtained results suggest that, given the parameters ranges typical of Nordic BWR5, only a small fraction of debris beds configurations exhibits the occurrence of dryout. Of the initially non-coolable configurations, a significant portion becomes coolable due to debris bed self-leveling.
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| 11. |
- Basso, Simone, et al.
(författare)
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Validation of DECOSIM code against experiments on particle spreading by two-phase flows in water pool
- 2016
-
Ingår i: Proceedings of the 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety, NUTHOS-11.
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Konferensbidrag (refereegranskat)abstract
- Validation simulations by DECOSIM code are performed against recent PDS-P experiments on particle spreading in a planar vertical water pool with bottom air injection. The model implemented in the code considers two-fluid formulation (water, air), turbulence effects in liquid phase are taken into account by k-epsilon model with additional generation terms accounting for two-phase effects. Particles are described by Lagrangian model, with turbulent dispersion modeled by random-walk model. Simulations are performed in conditions corresponding to experimental setup, the test section was a plane rectangular tank of variable length (0.9 and 1.5 m) and pool depth (0.5, 0.7, and 0.9 m), the superficial gas injection velocity ranged between 0.12 and 0.69 m/s. Sedimentation of spherical stainless steel (1.5 and 3 mm) and glass (3 mm) particles was calculated and compared with experiments with respect to the mean spreading distance and lateral distributions of mass fraction of particles. Reasonable agreement between the results obtained and experimental measurements is achieved for all pool geometries, gas injection rates, and particle types, confirming adequacy of the modeling approach and suitability of DECOSIM code for severe accident analysis related to debris bed formation. Possible ways to further reduction of uncertainty in model validation are discussed.
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| 12. |
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| 13. |
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| 14. |
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| 15. |
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| 16. |
- Cadinu, Francesco, et al.
(författare)
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Study of algorithmic requirements for a system-to-CFD coupling strategy
- 2008
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Ingår i: Experiments and CFD Code Application to Nuclear Reactor Safety (XCFD4NRS).
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Konferensbidrag (refereegranskat)abstract
- Over the last decades, the analysis of transients and accidents in nuclear power plants has beenperformed by system codes. Though they will remain the analyst’s tool of choice for the foreseeablefuture, their limitations are also well known. It has been suggested that an improvement in thesimulation technology can be obtained by “coupling” system codes with Computational FluidDynamics (CFD) calculations. This is usually attempted in a domain decomposition fashion: the CFDsimulation is only performed in a selected subdomain and its solution is “matched” with the systemcode solution at the interface. However, another coupling strategy can be envisioned. Namely, CFDsimulations can be used to provide closures to a system code.This strategy is based on the following two assumptions. The first assumption is that there aretransients which cannot be simulated by system codes because of the lack of adequate closures. Thesecond assumption is that appropriate closures can be provided by CFD simulations. In this paper,such a coupling strategy, inspired by the Heterogeneous Multiscale Method (HMM), is presented. Thephilosophy underlying this strategy is discussed with the help of a computational example.
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| 17. |
- Carlson, A., et al.
(författare)
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A Multiscale Approach for Thin-Film Slug Flow
- 2009
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Ingår i: Proceedings of the 7<sup>th</sup> World Conference on Experimental Heat Transfer, Fluid Mechanics and Thermodynamics, ExHFT-7. - 9788374642354
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Konferensbidrag (refereegranskat)abstract
- A multiscale modeling approach is presented for multiphase flow phenomena featuring a thin-film bounding two phases. A Micro Scale Solver predicts the thin film dynamics, influenced by an antagonistic Van der Waals force and a stabilizing repulsive force, which is mapped onto a Macro Scale Solver through a multiscale coupling. Numerical experiments of thin-film slug flow in a micropipe demonstrate that the key to capture multiscale phenomena lies in the accurate modelling of the microscale parameters. Faitful results are obtained with the multiscale treatment for the modelling of slug flow with a 10.4 nm thin-film, where pure computational multi-fluid dynamics is deficient.
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| 18. |
- Carlson, Andreas, et al.
(författare)
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Prediction of Two-Phase Flow in Small Tubes : A Systematic Comparison of State-of-The-Art CMFD Codes
- 2008
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Ingår i: 5th European Thermal-Sciences Conference (EUROTHERM).
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Konferensbidrag (refereegranskat)abstract
- Multiphase dynamics and its characteristics for two-phase gas-liquid flow have been investigatedby means of advanced numerical simulations. Although important in many engineering applications, methods for robust and accurate simulations for high density and viscosity ratios remainelusive. A comprehensive comparison of two state-of-the-art Computational Multi–Fluid Dynamics (CMFD) codes, Fluent and TransAT, have been performed. The two commonly usedmethods for two–phase flow simulations, namely Volume of Fluid implemented in Fluent andLevel Set implemented in TransAT, could be compared as a result. Significant differences wereobserved between the two flow topologies predicted by the two codes. For the bubbly flow case,a recirculating flow was predicted inside the bubbles by TransAT, meanwhile no significantrecirculation was observed in the solution with Fluent. For the slug flow case a significantdeviation was observed between the results from Fluent and TransAT on the slug formationand frequency. Periodic slug formation was observed with TransAT, in agreement with theexperimental result of Chen et al. [4]. A periodic slug formation was not obtained with Fluent.
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| 19. |
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| 20. |
- Dombrovsky, L.A., et al.
(författare)
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Thermal radiation modeling in numerical simulation of melt-coolant interaction
- 2008
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Ingår i: Proc. Int. Symp. Adv. Comput. Heat Transfer (CHT-08).
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Konferensbidrag (refereegranskat)abstract
- This paper is concerned with radiation heat transfer modeling in multiphase disperse systems, which are formed in high-temperature melt-coolant interactions. This problem is important for complex interaction of the core melt with water in the case of a hypothetical severe accident in light-water nuclear reactors. The nonlocal effects of thermal radiation due to the semitransparency of water in the visible and near-infrared spectral ranges are taken into account by use of the recently developed large-cell radiation model (LCRM) based on the spectral radiation energy balance for single computational cells. In contrast to the local approach for radiative heating of water by particles (OMM—opaque medium model), the LCRM includes radiative heat transfer between the particles of different temperatures. The regular integrated code VAPEX-P, intended to model the premixing stage of FCI, was employed for verification of the LCRM in a realistic range of the problem parameters. A comparison with the OMM and the more accurate P1 approximation showed that the LCRM can be recommended for the engineering problem under consideration. The effects of the temperature difference in solidifying particles are analyzed by use of the recently suggested approximation of transient temperature profile in the particles. It is shown that the effect of the temperature difference on heat transfer from corium particles to ambient water is considerable and should not be ignored in the calculations. An advanced computational model based on the LCRM for the radiation source function and subsequent integration of radiative transfer equation along the rays is also discussed.
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| 21. |
- Dombrovsky, L.A., et al.
(författare)
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Thermal radiation modeling in numerical simulation of melt-coolant interaction
- 2009
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Ingår i: Computational Thermal Sciences. - 1940-2503. ; 1:1, s. 1-35
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Tidskriftsartikel (refereegranskat)abstract
- This paper is concerned with radiation heat transfer modeling in multiphase disperse systems, which are formed in high-temperaturemelt-coolant interactions. This problem is important for complex interaction of the core melt with water in the case of a hypothetical severe accident in light-water nuclear reactors. The nonlocal effects of thermal radiation due to the semitransparency of water in the visible and near-infrared spectral ranges are taken into account by use of the recently developed large-cell radiation model (LCRM) based on the spectralradiation energy balance for single computational cells. In contrast to the local approach for radiative heating of water by particles (OMMopaque medium model), the LCRM includes radiative heat transfer between the particles of different temperatures. The regular integrated code VAPEX-P, intended to model the premixing stage of FCI, was employed for verification of the LCRM in a realistic range of the problem parameters. A comparison with the OMM and the more accurate P1 approximation showed that the LCRM can be recommended for the engineering problem under consideration. The effects of the temperature difference in solidifying particles are analyzed by use of the recently suggested approximation of transient temperature profile in the particles. It is shown that the effect of the temperature difference on heat transfer from corium particles to ambient water is considerable and should not be ignored in the calculations. An advanced computational model based on the LCRM for theradiation source function and subsequent integration of radiative transfer equation along the rays is also discussed.
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| 22. |
- Estévez-Albuja, S., et al.
(författare)
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Modelling of a Nordic BWR containment and suppression pool behavior during a LOCA with GOTHIC 8.1
- 2020
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 136
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Tidskriftsartikel (refereegranskat)abstract
- Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher water surface temperatures than in mixed conditions and thus, to higher containment pressures. The Effective Heat (EHS) and Momentum (EMS) Source models were previously introduced to predict the effect of small-scale direct contact condensation phenomena on the large-scale pool water circulation. In this paper, the EHS/EMS models are extended by adding the effect of non-condensable gases on the chugging regime. The EHS/EMS models are implemented in the GOTHIC code to model a full-scale Nordic BWR containment under different LOCA scenarios. The results show that thermal stratification can be developed in the PSP.
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| 23. |
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| 24. |
- Gallego-Marcos, Ignacio, et al.
(författare)
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Effective momentum induced by steam condensation in the oscillatory bubble regime
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Annan publikation (övrigt vetenskapligt/konstnärligt)abstract
- The spargers used in Boiling Water Reactors (BWR) discharge steam from the primary coolant system into a pool of water. Direct steam condensation in subcooled water creates sources of heat and momentum determined by the condensation regimes, called “effective sources” in this work. Competition between the effective sources can result in thermally stratification or mixing of the pool. Thermal stratification is a safety concern in BWRs since it reduces the steam condensation and pressure suppression capacity of the pool. In this work, we present semi-empirical correlations to predict the effective momentum induced by steam condensation in the oscillatory bubble regime, relevant for the operation of spargers in BWRs. A Separate Effect Facility (SEF) was designed and built at LUT, Finland, in order to provide the necessary data. An empirical correlation for the effective momentum as a function of the Jakob number is proposed. The Kelvin Impulse theory was also applied to estimate the effective momentum based on information about the bubble dynamics. To do this, new correlations for the bubble collapse frequencies, maximum bubble radius, velocities, pressure gradient and heat transfer coefficient are proposed and compared to available data from the literature. The effective momentum induced by sonic steam jets appears to be constant in a wide range of studied Jakob number. However, further experimental data is necessary at larger Jakob numbers and steam mass fluxes.
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| 25. |
- Gallego-Marcos, Ignacio, et al.
(författare)
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Effective momentum induced by steam condensation in the oscillatory bubble regime
- 2019
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Ingår i: International Journal of Multiphase Flow. - : Elsevier BV. - 0301-9322 .- 1879-3533. ; 350, s. 259-274
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Tidskriftsartikel (refereegranskat)abstract
- The spargers used in Boiling Water Reactors (BWR) discharge steam from the primary coolant system into a pool of water. Direct steam condensation in subcooled water creates sources of heat and momentum determined by the condensation regimes, called “effective sources” in this work. Competition between the effective sources can result in thermally stratification or mixing of the pool. Thermal stratification is a safety concern in BWRs since it reduces the steam condensation and pressure suppression capacity of the pool. In this work, we present semi-empirical correlations to predict the effective momentum induced by steam condensation in the oscillatory bubble regime, relevant for the operation of spargers in BWRs. A Separate Effect Facility (SEF) was designed and built at LUT, Finland, in order to provide the necessary data. An empirical correlation for the effective momentum as a function of the Jakob number is proposed. The Kelvin Impulse theory was also applied to estimate the effective momentum based on information about the bubble dynamics. To do this, new correlations for the bubble collapse frequencies, maximum bubble radius, velocities, pressure gradient and heat transfer coefficient are proposed and compared to available data from the literature. The effective momentum induced by sonic steam jets appears to be constant in a wide range of studied Jakob number. However, further experimental data is necessary at larger Jakob numbers and steam mass fluxes.
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| 26. |
- Gallego-Marcos, Ignacio, et al.
(författare)
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Modeling of Thermal Stratification and Mixing in a Pressure Suppression Pool Using GOTHIC
- 2016
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Konferensbidrag (refereegranskat)abstract
- The development of thermal stratification in the pressure suppression pool of a BWR is a safety issue since it can lead to higher containment pressures than in completely mixed conditions. The thermal hydraulic code of GOTHIC offers a very suitable platform to simulate the pool and containment behavior during a long term accident. However, for a computationally efficient code such as GOTHIC, direct contact condensation cannot be resolved accurately enough to obtain a good estimation of the momentum induced by the condensing steam, and thus, to predict the pool behaviour. In this paper, we present how to implement the previously validated Effective Heat Source (EHS) and Effective Momentum Source (EMS) models, developed for pool analysis during a steam injection, in GOTHIC. The implementation was done using control variables and Dynamically Linked Libraries (DLL). A time averaging model to minimize the effect of the numerical oscillations appearing in GOTHIC when steam is injected into the pool is also proposed.
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| 27. |
- Gallego-Marcos, Ignacio, et al.
(författare)
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Modeling of Thermal Stratification and Mixing Induced by Steam Injection Through Spargers Into a Large Water Pool
- 2016
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Konferensbidrag (refereegranskat)abstract
- The pressure suppression pool of a Boiling Water Reactor (BWR) is designed to protect the containment from over pressure by condensing steam. Under certain steam injection conditions, thermal stratification can develop in the pool and significantly reduce its pressure suppression capacity. In this work, we propose a model to simulate the pool behavior during a steam injection through spargers, which are multi-hole injection pipes connecting the main steam lines to the wetwell pool. The aim of the model is to predict the global pool behavior. Effective Heat and Momentum Sources (EHS/EMS) approach is used to model time averaged effects of small scale direct contact condensation phenomena on the large scale pool circulation. The model was implemented in Fluent 16.2 and validated against experimental data obtained in PANDA facility at PSI (Switzerland). The scaling of the experiments was done to address the most important physical phenomena that can occur in plant scale. The results show that the global pool behavior can be predicted using the Standard Gradient Diffusion Hypothesis (SGDH) in k-Omega turbulence model.
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| 28. |
- Gallego Marcos, Ignacio, et al.
(författare)
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Modelling of pool stratification and mixing induced by steam injectionthrough blowdown pipes
- 2018
-
Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 112, s. 624-639
-
Tidskriftsartikel (refereegranskat)abstract
- Containment overpressure is prevented in a Boiling Water Reactor (BWR) by condensing steam into thepressure suppression pool. Steam condensation is a source of heat and momentum. Competition betweenthese sources results in thermal stratification or mixing of the pool. The interplay between the sources isdetermined by the condensation regime, steam mass flow rate and pool dimensions. Thermal stratificationis a safety issue since it limits the condensing capacity of the pool and leads to higher containmentpressures in comparison to a completely mixed pool with the same average temperature. The EffectiveHeat Source (EHS) and Effective Momentum Source (EMS) models were previously developed for predictingthe macroscopic effect of steam injection and direct contact condensation phenomena on the developmentof stratification and mixing in the pool. The models provide the effective heat and momentumsources, depending on the condensation regimes. In this work we present further development of theEHS/EMS models and their implementation in the GOTHIC code for the analysis of steam injection intocontainment drywell and venting into the wetwell through the blowdown pipes. Based on thePPOOLEX experiments performed in Lappeenranta University of Technology (LUT), correlations arederived to estimate the steam condensation regime and effective heat and momentum sources as functionsof the pool and steam injection conditions. The focus is on the low steam mass flux regimes withcomplete condensation inside the blowdown pipe or chugging. Validation of the developed methodswas carried out against the PPOOLEX MIX-04 and MIX-06 tests, which showed a very good agreementbetween experimental and simulation data on the pool temperature distribution and containmentpressure.
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| 29. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Pool stratification and mixing during a steam injection through spargers: analysis of the PPOOLEX and PANDA experiments
- 2018
-
Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 337, s. 300-316
-
Tidskriftsartikel (refereegranskat)abstract
- Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Advanced Pressurized (AP) reactors to condense steam in large water pools. A steam injection induces heat, momentum and mass sources that depend on the steam injection conditions and can result in thermal stratification or mixing of the pool. Thermal stratification reduces the steam condensation capacity of the pool, increases the pool surface temperature and thus the containment pressure. Development of models with predictive capabilities requires the understanding of basic phenomena that govern the behavior of the complex multi-scale system. The goals of this work are (i) to analyze and interpret the experiments on steam injection into a pool through spargers performed in the large-scale facilities of PPOOLEX and PANDA, and (ii) to discuss possible modelling approaches for the observed phenomena. A scaling approach was developed to address the most important physical phenomena and regimes relevant to prototypic plant conditions. The focus of the tests was on the low steam mass flux and oscillatory bubble condensation regimes, which are expected during a long-term steam injection transient, e.g. in the case of a Station Black Out (SBO). Exploratory tests were also done for chugging and stable jet conditions. The results showed a similar behavior in PPOOLEX and PANDA in terms of jet induced by steam condensation, pool stratification, and development of hot layer and erosion of the cold one. A correlation using the Richardson number is proposed to model the erosion rate of the cold layer as a function of the pool dimensions and steam injection conditions.
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| 30. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Pool stratification and mixing induced by steam injection through spargers : CFD modelling of the PPOOLEX and PANDA experiments
- 2019
-
Ingår i: Nuclear Engineering and Design. - : ELSEVIER SCIENCE SA. - 0029-5493 .- 1872-759X. ; 347, s. 67-85
-
Tidskriftsartikel (refereegranskat)abstract
- Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Generation III/III+ Pressurized Water Reactors (PWR) to condense steam in large water pools. During the steam injection, high pool surface temperatures induced by thermal stratification can lead to higher containment pressures compared with completely mixed pool conditions, the former posing a threat for plant safety. The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models were previously developed and validated for the modelling of a steam injection through blowdown pipes. The goal of this paper is to extend the EHS/EMS model capabilities towards steam injection through multi-hole spargers. The models are implemented in ANSYS Fluent 17.0 Computational Fluid Dynamics (CFD) code and calibrated against the spargers experiments performed in the PPOOLEX and PANDA facilities, analysed by the authors in Gallego-Marcos et al. (2018b). CFD modelling guidelines are established for the adequate simulation of the pool behaviour. A new correlation is proposed to model the turbulent production and dissipation caused by buoyancy. Sensitivity studies addressing the effect of different assumptions on the effective momentum magnitude, profile, angle and turbulence are presented. Calibration of the effective momentum showed an inverse proportionality to the sub-cooling. Differences between the effective momentum calibrated for PPOOLEX and PANDA are discussed. Analysis of the calculated flow above the cold stratified layer showed that the erosion of the layer is induced by the action of turbulence rather than mean shear flow.
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| 31. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Pool Stratification and Mixing Induced by Steam Injection through Spargers: CFD modelling of the PPOOLEX and PANDA experiments
-
Annan publikation (övrigt vetenskapligt/konstnärligt)abstract
- Spargers are multi-hole injection pipes used in Boiling Water Reactors (BWR) and Advanced Pressurized (AP) reactors to condense steam in large water pools. During the steam injection, high pool surface temperatures induced by thermal stratification can lead to higher containment pressures compared with completely mixed pool conditions, the former posing a threat for plant safety. The Effective Heat Source (EHS) and Effective Momentum Source (EMS) models were previously developed and validated for the modelling of a steam injection through blowdown pipes. The goal of this paper is to extend the EHS/EMS model capabilities towards steam injection through multi-hole spargers. The models were implemented in the CFD code of ANSYS Fluent 17.0 and calibrated against the PPOOLEX and PANDA experiments with spargers analysed by the authors in [1] (Gallego-Marcos, I., et al., 2018). Modelling guidelines are established for the adequate simulation of the pool behaviour. A new correlation is proposed to model the turbulent production and dissipation caused by buoyancy. Sensitivity studies addressing the effect of different assumptions on the effective momentum magnitude, profile, angle and turbulence are presented. Calibration of the momentum magnitude showed that it varies between 0.2 to 1.2 times the steam momentum at the injection holes. Differences of this fraction between the PPOOLEX and PANDA simulations are discussed. Analysis of the calculated flow above the cold stratified layer shows that the erosion of the layer is induced by the action of turbulence rather than mean shear flow.
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| 32. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Possibility of Air Ingress into a BWR Containment during a LOCA in case of Activation of Containment Venting System
- 2014
-
Konferensbidrag (refereegranskat)abstract
- The pressure relief systems installed in BWRs protect the containment from overpressure in case of a Loss of Coolant Accident (LOCA). This paper analyzes the possibility of air ingress, which can cause hydrogen burn, through the rupture disks of the filtered and non-filtered venting systems. Two scenarios were considered: a LOCA without SBO (Station Blackout) and a LOCA with SBO. The thermal-hydraulic code GOTHIC® was used with 3D models of the drywell and wetwell of a Nordic-type BWR. In the LOCA event, we found no activation of the rupture disks within the considered transient simulation. Moreover, the containment spray ensured a low pressure in the drywell and induced a continuous mixing of the wetwell pool. In the LOCA with SBO event, the development of thermal stratification in the wetwell pool accelerated the pressure increase in the drywell, which led to activation of the rupture disk of the filtered venting system. However, no air ingress through the vent was found during the depressurization of the containment, and hence no risk of hydrogen burn under the given assumptions.
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| 33. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Scaling and CFD Modelling of the Pool Experiments with Spargers Performed in the PANDA Facility
- 2016
-
Konferensbidrag (refereegranskat)abstract
- The development of thermal stratification in the pressure suppression pool of a BWR is a safety issue since it reduces its cooling capability and leads to higher containment pressures than in completely mixed conditions. In this work, we propose a model to simulate the pool behavior during a steam injection through spargers. The model provides the time averaged heat and momentum transferred from the steam condensation to the large scale pool circulation. Small scale phenomena such as direct contact condensation is not resolved, only its effect on the pool behaviour. The model was implemented in Fluent 16.2 and validated against experimental data obtained in PANDA facility at PSI (Switzerland). The scaling of the experiments, done to preserve the most important physical phenomena occurring in plant scale is also presented in the paper. The results show that the model is able to predict well the global pool behavior. However, flow instabilities were observed to induce a sudden mixing of the upper part of the stratified layer during the transition from the stratification to the mixing phases. This led to a faster erosion of the layer than in the experiment. Simulations done with 2D and 3D meshes and scale adaptive turbulence models were performed to clarify this issue and are presented in the paper.
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| 34. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Scaling of the Erosion of a Thermally Stratified Layer in a Large Water Pool during a Steam Injection Through Spargers
- 2016
-
Konferensbidrag (refereegranskat)abstract
- The development of thermal stratification in the pressure suppression pool of a BWR is a safety issue since it reduces its cooling capability and leads to higher containment pressures than in completely mixed conditions. In this work, we analyzed the experiments on pool thermal stratification and mixing performed with spargers in the PPOOLEX facility at Lappeenranta University of Technology (LUT), and in the PANDA facility at the Paul Scherrer Institut (PSI). It was observed that at high momentum injections the mean flow was able to penetrate into the stratified layer and induce complete mixing of the pool. At lower momentum injections, the stratified layer was slowly eroded at the interface by the action of small scale turbulence and the breakup of waves. We demonstrated that the erosion mechanism and erosion velocity can be obtained as a function of a single non dimensional number, the Richardson number. Based on this finding, a preliminary coarse resolution model was implemented in the thermal-hydraulic code of GOTHIC to simulate pool behavior during a steam injection through spargers.
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| 35. |
- Gallego-Marcos, Ignacio, 1990-
(författare)
-
Steam condensation in a water pool and its effect on thermal stratification and mixing
- 2018
-
Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- The Pressure Suppression Pool (PSP) of a Boiling Water Reactor (BWR) is a large heat sink designed to limit the containment pressure by condensing steam released from the primary coolant system. The development of thermal stratification is a safety concern since it leads to higher containment pressures than in completely mixed conditions, and can affect the performance of systems such as the emergency core cooling and containment spray, which the use PSP as a source of water.The goal of this thesis is to develop and validate models for the prediction of the PSP behavior during a steam injection in a Nordic BWR. The framework of the Effective Heat Source and Effective Momentum Source (EHS/EMS) models is used to provide the integral heat and momentum sources induced by the steam condensation. The EHS/EMS can be implemented in a containment thermal-hydraulic or a CFD code, where the pool is modelled with a single-phase liquid solver.EHS/EMS models are developed for the low steam mass flux regimes appearing in (i) large diameter blowdown pipes connecting the drywell to the wetwell pool; and (ii) multi-hole sparger pipes connecting the primary system to the pool.Empirical correlations are developed to predict the effective momentum induced by chugging in the blowdown pipes. The correlations are implemented in GOTHIC, where a containment model is proposed to enable capturing the feedback between pool conditions and drywell pressure. Validation is performed against the PPOOLEX experiments.Conceptual designs are proposed for a set of large-scale pool experiments with spargers in the PPOOLEX and PANDA facilities. Correlations are proposed for the erosion velocity of a cold layer, and ranges are estimated for the angle, profile and turbulence of the momentum sources created by steam injection. CFD simulations of the experiments is done to calibrate the momentum sources in the oscillatory bubble regimes. A concept of the Separate Effect Facility (SEF) is proposed to provide a measurements of the effective momentum. Empirical correlations for the bubble radius, velocity, heat transfer coefficient, etc. are also developed and compared to available data from the literature.Application of the developed CFD and EHS/EMS models to full-scale containment behavior shows that thermal stratification can occur during prototypic steam injection conditions. Recommendations are given on how to avoid this.
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| 36. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression Pool
-
Annan publikation (övrigt vetenskapligt/konstnärligt)abstract
- The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 oC pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached ~7 h after the beginning of the blowdown.
|
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| 37. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Thermal Stratification and Mixing in a Nordic BWR Pressure Suppression Pool
-
Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
-
Tidskriftsartikel (refereegranskat)abstract
- The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 oC pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached ~7 h after the beginning of the blowdown.
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| 38. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Thermal stratification and mixing in a Nordic BWR pressure suppression pool
- 2019
-
Ingår i: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 132, s. 442-450
-
Tidskriftsartikel (refereegranskat)abstract
- The pressure suppression pool of a Nordic Boiling Water Reactor (BWR) serves as a heat sink to condense steam from the primary coolant system in normal operation and accident conditions. Thermal stratification can develop in the pool when buoyancy forces overcome the momentum created by the steam injection. In this case, hot condensate forms a hot layer at the top of the pool, reducing the pool cooling and condensation capacity compared to mixed conditions. The Effective Heat Source and Effective Momentum Source (EHS/EMS) models were previously proposed to model the large-scale pool behavior during a steam injection. In this work, we use CFD code of ANSYS Fluent with the EHS/EMS models to simulate the transient behavior of a Nordic BWR pool during a steam injection through spargers. First, a validation against a Nordic BWR pool test with complete mixing is presented. Prediction of the pool behavior for other possible injection scenarios show that stratification can occur at prototypic steam injection conditions, and that the hot layer temperature above the injection point can be non-uniform. In cases with significant steam condensation inside the sparger pipes, the 95 degrees C pool temperature limit for the Emergency Core Cooling System (ECCS) pumps was reached similar to 7 h after the beginning of the blowdown.
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| 39. |
- Gallego-Marcos, Ignacio, et al.
(författare)
-
Water Distribution in a Nordic BWR Containment During a LOCA
- 2016
-
Ingår i: 2016 International Congress on Advances in Nuclear Power Plants, ICAPP 2016.
-
Konferensbidrag (refereegranskat)abstract
- During a main steam line break in a Boiling Water Reactor (BWR) the pressure suppression pool is used as a water source for the Emergency Core Cooling System (ECCS) and the Containment Spray (CS). These systems drain water from the pool through strainers, which are long perforated plates or cylinders submerged to a certain depth. Proper functioning of the ECCS and the CS must be ensured to maintain the water inventory in the vessel and to limit the containment pressure. However, if the liquid level in the suppression pool goes below the level of the strainers intake, the operators would be forced to stop their pumps. The liquid level in the suppression pool can be reduced when a significant fraction of ECCS and CS flow is relocated to the lower drywell. In this work, we use the thermal-hydraulic code GOTHIC to simulate the containment evolution during a main steam line break inside the biological shield. The containment volumes and their connections were modeled with 2D and 3D volumes. With this model, scenarios considering different operational conditions were assessed: (i) full capacity of all the safety systems, (ii) half capacity of all the safety systems, (iii) ECCS stops injecting water after a certain liquid level is restored in the vessel, and (iv) the pipes used to drain water from the suppression pool and flood the lower drywell are partially or totally clogged in different directions. The results showed that there is a risk of an early shut down of the ECCS and CS systems in the case of main steam line break inside the biological shield. It was observed that when the ECCS provided a continuous water injection into the vessel, the water spilled through the break into the biological shield flowed downwards driven by gravity and went directly into the lower drywell. This caused a fast decrease in the liquid level of the suppression pool, which led to an uncovery of the ECCS and CS strainers about 2000 s after the break. The activation at 1800 s of the flooding of the lower drywell led to a backward flow, from the lower drywell to the suppression pool, since at that time the liquid level in the suppression pool was lower than in the lower drywell. However, this backward flow was not enough to maintain the liquid level in the suppression pool, which continued to decrease. In the case where the pipes used for the flooding were clogged in the direction of the suppression pool, uncovery of the strainers was observed even earlier.
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| 40. |
- Galushin, Sergey, et al.
(författare)
-
Analysis of core degradation and relocation phenomena and scenarios in a Nordic-type BWR
- 2016
-
Ingår i: NUCLEAR ENGINEERING AND DESIGN. - : Elsevier. - 0029-5493. ; 310, s. 125-141
-
Tidskriftsartikel (refereegranskat)abstract
- Severe Accident Management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel cooling of core melt debris. The melt is released from the failed vessel and poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by a natural circulation and evaporation of water. Success of the strategy is contingent upon melt release conditions from the vessel and melt-coolant interaction that determine (i) properties of the debris bed and its coolability (ii) potential for energetic melt-coolant interactions (steam explosions). Risk Oriented Accident Analysis Methodology (ROAAM+) framework is currently under development for quantification of the risks associated with formation of non-coolable debris bed and occurrence of steam explosions, both presenting a credible threats to containment integrity. The ROAAM+ framework consist of loosely coupled models that describe each stage of the accident progression. Core relocation analysis framework provides initial conditions for melt vessel interaction, vessel failure and melt release frameworks. The properties of relocated debris and melt release conditions, including in-vessel and ex-vessel pressure, lower drywell pool depth and temperature, are sensitive to the accident scenarios and timing of safety systems recovery and operator actions. This paper illustrates a methodological approach and relevant data for establishing a connection between core relocation and vessel failure analysis in ROAAM+ approach. MELCOR code is used for analysis of core degradation and relocation phenomena. Properties of relocated debris are obtained as functions of the accident scenario parameters. Pattern analysis is employed in order to characterize typical behavior of core relocation transients. Clustering analysis is employed for grouping of different accident scenarios, which result in similar core relocation behavior and properties of the debris.
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| 41. |
- Galushin, Sergey, et al.
(författare)
-
Analysis of the Effect of MELCOR Modelling Parameters on In-Vessel Accident Progression in Nordic BWR
- 2019
-
Ingår i: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 350, s. 243-258
-
Tidskriftsartikel (refereegranskat)abstract
- Nordic Boiling Water Reactors (BWRs) rely on the flooding of the lower drywell (LDW) as a severe accident management (SAM) strategy. The termination of a SA is achieved by fragmenting and quenching of the melt released from the vessel. Success of SAM strategy depends on melt release and water pool conditions. The characteristics of the melt release are the major source of uncertainty in quantification of the risk of SAM failure. Vessel failure and melt release modes are subject to aleatory and epistemic uncertainties at the in-vessel accident progression stage. In this work we focus on predicting the properties of debris relocated to the lower plenum using MELCOR code. We address the effect of epistemic uncertainty in modeling parameters and models in the MELCOR code in different severe accident scenarios on main characteristics of the in-vessel accident progression in Nordic BWRs. Sensitivity analysis is performed to rank the importance of MELCOR modelling parameters and the effect of different MELCOR models is addressed by using different versions of the code. The results provide valuable insights regarding the effect of MELCOR models, modelling parameters and sensitivity coefficients on code predictions.
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| 42. |
- Galushin, Sergey, et al.
(författare)
-
Analysis of the Effect of Severe Accident Scenario on Debris Properties in Lower Plenum of Nordic BWR Using Different Versions of MELCOR Code
- 2019
-
Ingår i: Science and Technology of Nuclear Installations. - : Hindawi Publishing Corporation. - 1687-6075 .- 1687-6083.
-
Tidskriftsartikel (refereegranskat)abstract
- Nordic Boiling Water Reactors (BWRs) employ ex-vessel debris coolability as a severe accident management strategy (SAM). Core melt is released into a deep pool of water where formation of noncoolable debris bed and ex-vessel steam explosion can pose credible threats to containment integrity. Success of the strategy depends on the scenario of melt release from the vessel that determines the melt-coolant interaction phenomena. The melt release conditions are determined by the in-vessel phase of severe accident progression. Specifically, properties of debris relocated into the lower plenum have influence on the vessel failure and melt release mode. In this work we use MELCOR code for prediction of the relocated debris. Over the years, many code modifications have been made to improve prediction of severe accident progression in light-water reactors. The main objective of this work is to evaluate the effect of models and best practices in different versions of MELCOR code on the in-vessel phase of different accident progression scenarios in Nordic BWR. The results of the analysis show that the MELCOR code versions 1.86 and 2.1 generate qualitatively similar results. Significant discrepancy in the timing of the core support failure and relocated debris mass in the MELCOR 2.2 compared to the MELCOR 1.86 and 2.1 has been found for a domain of scenarios with delayed time of depressurization. The discrepancies in the results can be explained by the changes in the modeling of degradation of the core components and changes in the Lipinski dryout model in MELCOR 2.2.
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| 43. |
|
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| 44. |
- Galushin, Sergey, et al.
(författare)
-
Analysis of the effect of severe accident scenario on the vessel lower head failure in Nordic BWR using MELCOR code
- 2018
-
Ingår i: PSAM 2018 - Probabilistic Safety Assessment and Management. - : International Association for Probablistic Safety Assessment and Management (IAPSAM).
-
Konferensbidrag (refereegranskat)abstract
- Severe accident management (SAM) in Nordic boiling water reactors (BWR) relies on ex-vessel core debris coolability. In case of core melt and vessel failure, melt is poured into a deep pool of water located under the reactor. The melt is expected to fragment, quench, and form a debris bed, coolable by natural circulation of water. Success of the strategy is contingent upon melt release conditions from the vessel which determine (i) properties and thus coolability of the debris bed, and (ii) potential for energetic steam explosion. Both non-coolable debris bed and steam explosion are credible threats to containment integrity. Melt release conditions are the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs using ROAAM+ Framework. The melt release conditions, including in-vessel\ex-vessel pressure, lower drywell pool depth and temperature, are affected by aleatory (severe accident scenario) and epistemic (modeling) uncertainties. In this work we use MELCOR code to perform the analysis of the effects of Severe accident scenarios and modelling options in MELCOR on the properties of debris relocated to the lower head, the time and the mode of vessel lower head failure. We identify the most influential uncertain factors and discuss the needs for improvements in the modeling approaches.
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| 45. |
- Galushin, Sergey, et al.
(författare)
-
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 roaam1 framework
- 2020
-
Ingår i: Journal of Nuclear Engineering and Radiation Science. - : ASME International. - 2332-8983 .- 2332-8975. ; 6:4
-
Tidskriftsartikel (refereegranskat)abstract
- Nordic boiling water reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core-melt coolant interactions, containment integrity can be threatened by: (i) formation of noncoolable debris bed or (ii) energetic steam explosion. Melt is released 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 mode 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. Melt ejection SM is based on the system analysis code methods for estimation of leakages and consequences of releases (computer code) (MELCOR). Modeling 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 (LH) behavior and melt release. The work addresses the effect of epistemic uncertain parameters and modeling 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, modeling parameters, and sensitivity coefficients on melt release conditions and predictions of ex-vessel steam explosion loads on the containment structures.
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| 46. |
- Galushin, Sergey, et al.
(författare)
-
Comparison of melcor code versions predictions of the properties of relocated debris in lower plenum of nordic BWR
- 2016
-
Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM).
-
Konferensbidrag (refereegranskat)abstract
- Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the vessel failure and melt release mode from the vessel, which determine conditions for (i) the formation of debris bed and its coolability, and (ii) steam explosion. Non-coolable debris and strong explosions present credible threats to containment integrity. A risk oriented accident analysis framework (ROAAM+) is under development for assessment of the effectiveness of the severe accident management strategy. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine conditions for the corium interactions with the vessel structures, vessel failure and melt release. In this work we perform comparison of predictions of different MELCOR code versions used for the analysis of the effect severe accident scenario and uncertainties on the processes of core degradation and relocation, and resulting properties of relocated debris in Nordic BWR lower plenum. Properties of relocated debris are obtained as functions of the accident scenario parameters, such as timing of activation of different safety systems. We perform the analysis of the codes predictions and discuss possible reasons for the discrepancies in observations. The main goal of this work is to provide insights regarding the effect of code uncertainty, sensitivity coefficients and user effect on the code predictions, which is of importance for the analysis of in-vessel debris coolability and vessel failure mode in the ROAAM+ framework.
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| 47. |
- Galushin, Sergey, et al.
(författare)
-
Comparison of vessel failure mode and melt release conditions in unmitigated and mitigated station blackout scenarios in nordic BWR using melcor code
- 2019
-
Ingår i: International Conference on Nuclear Engineering, Proceedings, ICONE. - : American Society of Mechanical Engineers (ASME).
-
Konferensbidrag (refereegranskat)abstract
- Nordic boiling water reactors (BWR) employ ex-vessel debris coolability as a severe accident management (SAM) strategy. Effectiveness of this strategy depends on melt release conditions from the vessel, that recognized as the major source of uncertainty in quantification of the risk of containment failure in Nordic BWRs in ROAAM+ Framework. In this work we perform the analysis of the effect of water recovery in station blackout accident in Nordic BWR on the timing and mode of vessel failure and melt release conditions using MELCOR code. The analysis is performed using Morris method in order to evaluate the sensitivity of the vessel failure mode and melt release conditions and associated uncertainty due to user-defined modelling parameters and modelling options in MELCOR code.
|
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| 48. |
- Galushin, Sergey
(författare)
-
Development of Risk Oriented Accident Analysis Methodology for Assessment of Effectiveness of Severe Accident Management Strategy in Nordic BWR
- 2019
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Nordic Boiling Water Reactor (BWR) design employs ex-vessel debris coolability as a severe accident management strategy (SAM). In case of a severe accident, the debris ejected from the vessel are expected to fragment, quench and form a debris bed, which is coolable by a natural circulation of water. Success of the existing SAM strategy depends on melt release conditions from the vessel which determine (i) properties of ejected debris and, thus, ex-vessel debris bed coolability, and (ii) potential for energetic melt-coolant interactions (steam explosion). The strategy involves complex interactions between physical phenomena (deterministic) and transient accident scenarios (probabilistic).The aim of this work is further extension, implementation and application of the Risk-Oriented Accident Analysis Methodology (ROAAM) to assessment of the severe accident management strategy effectiveness. ROAAM was originally developed for rare, high-consequence hazards, where both aleatory (stochastic) and epistemic (modeling) uncertainties play a significant role in the risk assessment. The main purpose of ROAAM is to provide the input material to an underlying decision making regarding current safety design acceptance, procedures and possible design modifications.This work reports results of (i) development and implementation of probabilistic framework (ROAAM+) for streamlining sensitivity analysis, uncertainty quantification and risk analysis; (ii) analysis of in-vessel phase of accident progression and melt release conditions in Nordic BWR reactor design with MELCOR code; (iii) analysis of the effect of melt release conditions predicted by MELCOR code on the risk of ex-vessel steam explosion.In ROAAM+, “full models”, such as MELCOR code, are used to develop computationally efficient “surrogate models” to enable extensive uncertainty quantification and failure domain analysis. ROAAM+ analysis identified specific assumptions in MELCOR models, which are currently the major contributors to the uncertainty in the assessment of the SAM effectiveness.
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| 49. |
- Galushin, Sergey, et al.
(författare)
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Effect of severe accident scenario and modeling options in melcor on the properties of relocated debris in nordic BWR lower plenum
- 2016
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Ingår i: 17th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH 2017. - : Association for Computing Machinery (ACM).
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Konferensbidrag (refereegranskat)abstract
- Severe accident management (SAM) in Nordic Boiling Water Reactors (BWR) employs ex-vessel core debris coolability. Core melt is poured into a deep pool of water and is expected to fragment, quench, and form a coolable debris bed. Success of the strategy is contingent upon the vessel failure and melt release mode from the vessel, which determine conditions for (i) the formation of debris bed and its coolability, and (ii) steam explosion. Non-coolable debris and strong explosions present credible threats to containment integrity. A risk oriented accident analysis framework (ROAAM+) is under development for assessment of the effectiveness of the severe accident management strategy. The characteristics of melt release are determined by the in-vessel accident scenarios and phenomena subject to aleatory and epistemic uncertainties respectively. Specifically, properties of the debris relocated into the lower head determine conditions for the corium interactions with the vessel structures, vessel failure and melt release. This work is focused on the evaluation of uncertainty in core degradation progression and its effect on the resultant properties of relocated debris in lower plenum of Nordic BWR. We use MELCOR code for prediction of the accident progression. The main goal of this paper is to characterize the range of possible debris properties in lower plenum and its sensitivity towards different modelling parameters, which is of paramount importance for the analysis of in-vessel debris coolability and vessel failure mode in the ROAAM+ framework.
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| 50. |
- Galushin, Sergey, et al.
(författare)
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Implementation of framework for assessment of severe accident management effectiveness in Nordic BWR
- 2020
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Ingår i: Reliability Engineering & System Safety. - : Elsevier BV. - 0951-8320 .- 1879-0836. ; 203
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Tidskriftsartikel (refereegranskat)abstract
- Nordic Boiling Water Reactor (BWR) design employs ex-vessel debris coolability in a deep pool of water as a severe accident management (SAM) strategy. Depending on melt release conditions from the vessel and core-melt coolant interactions, containment integrity can be threatened by (i) formation of non-coolable debris bed, or (ii) energetic steam explosion. In order to assess the effectiveness of SAM the Risk Oriented Accident Analysis Methodology framework (ROAAM +) has been developed. The framework is further extension of the approach originally developed and applied by Prof. Theofanous and co-workers. This paper presents the implementation of ROAAM + probabilistic framework for uncertainty quantification and risk analysis. We further apply ROAAM + to the analysis of steam explosion risk in Nordic BWR assuming different state-of knowledge situations and different containment fragilities. We employ an iterative processes of knowledge refinement using risk analysis as a guiding tool in identification of the major sources of uncertainty. We estimate failure domains and discuss ROAAM + results in terms of possibility vs necessity of containment failure.
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