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Chen, Y., Zhang, H., Villanueva, W., Ma, W. & Bechta, S. (2019). A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor. Nuclear Engineering and Design, 343, 22-37
Open this publication in new window or tab >>A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor
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2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 343, p. 22-37Article in journal (Refereed) Published
Abstract [en]

This paper presents a sensitivity study of MELCOR nodalization for simulation of postulated severe accidents in a Nordic boiling water reactor, with the objective to address the nodal effect on in-vessel accident progression, including thermal-hydraulic response, core degradation and relocation, hydrogen generation, source term release, melt behavior and heat transfer in the lower head, etc. For this purpose, three meshing schemes (coarse, medium and fine) of the COR package of MELCOR are chosen to analyze two severe accident scenarios: station blackout (SBO) accident and large break loss-of-coolant accident (LOCA) combined with station blackout. The comparative results of the MELCOR simulations show that the meshing schemes mainly affect the core degradation and relocation to the lower head of the reactor pressure vessel: the fine mesh leads to a delayed leveling process of a heap-like debris bed in the lower head, and a later breach of the vessel. The simulations with fine mesh also provide more detailed distributions of corium mass and temperature, as well as heat flux which is an important parameter in qualification assessment of the In-Vessel Melt Retention (IVR) strategy.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Boiling water reactor, Reactor safety, Severe accident, MELCOR simulation, Mesh sensitivity analysis
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-244082 (URN)10.1016/j.nucengdes.2018.12.011 (DOI)000456923500003 ()2-s2.0-85059233155 (Scopus ID)
Note

QC 20190219

Available from: 2019-02-19 Created: 2019-02-19 Last updated: 2019-04-29Bibliographically approved
Gallego-Marcos, I., Kudinov, P., Villanueva, W., Kapulla, R., Paranjape, S., Paladino, D., . . . Kotro, E. (2019). Pool stratification and mixing induced by steam injection through spargers: CFD modelling of the PPOOLEX and PANDA experiments. Nuclear Engineering and Design, 347, 67-85
Open this publication in new window or tab >>Pool stratification and mixing induced by steam injection through spargers: CFD modelling of the PPOOLEX and PANDA experiments
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2019 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 347, p. 67-85Article in journal (Refereed) Published
Abstract [en]

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.

Place, publisher, year, edition, pages
ELSEVIER SCIENCE SA, 2019
Keywords
Thermocline, Turbulence production buoyancy, Richardson, C-3e coefficient, Oscillatory bubble regime
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-251271 (URN)10.1016/j.nucengdes.2019.03.011 (DOI)000465217900008 ()2-s2.0-85063478019 (Scopus ID)
Note

QC 20190514

Available from: 2019-05-14 Created: 2019-05-14 Last updated: 2019-05-29Bibliographically approved
Yu, P., Ma, W., Villanueva, W., Karbojian, A. & Bechta, S. (2019). Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment. Annals of Nuclear Energy, 133, 637-648
Open this publication in new window or tab >>Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment
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2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 133, p. 637-648Article in journal (Refereed) Published
Abstract [en]

The failure of reactor pressure vessel (RPV) during a severe accident of light water reactors is a thermal fluid-structure interaction (FSI) problem which involves melt pool heat transfer and creep deformation of the RPV. The present study is intended to explore a reliable coupling approach of thermo-fluid-structure analyses which will not only be able to reflect the transient thermal FSI feature, but also apply the advanced models and computational platforms to melt pool convection and structural mechanics, so as to improve simulation fidelity. For this purpose, the multi-physics platform of ANSYS encompassing Fluent and Structural capabilities was employed to simulate the fluid dynamics and structural mechanics in a coupled manner. In particular, the FOREVER-EC2 experiment was chosen to validate the coupling approach. The natural convection in melt pool was modeled with the SST turbulence model with a well-resolved boundary layer, while the creep deformation for the vessel made of 16MND5 steel was analyzed with a new three-stage creep model (modified theta projection model). A utility tool was introduced to transfer the transient thermal loads from Fluent to Structural which minimizes the user effort in performing the coupled analysis. The validation work demonstrated the well-posed capability of the coupling approach for prediction of the key parameters of interest, including temperature profile, total displacement of vessel bottom point and the evolution of wall thickness profile in the experiment. Ltd. All rights reserved.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Reactor pressure vessel, Creep failure, Thermal fluid-structure interaction, Computational fluid dynamics, Computational structural mechanics, Coupled analysis
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-260983 (URN)10.1016/j.anucene.2019.06.067 (DOI)000484649800061 ()2-s2.0-85068784394 (Scopus ID)
Note

QC 20191010

Available from: 2019-10-10 Created: 2019-10-10 Last updated: 2019-10-16Bibliographically approved
Fichot, F., Carenini, L., Villanueva, W. & Bechta, S. (2018). A revised methodology to assess in-vessel retention strategy for high-power reactors. In: PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 18, VOL 7: . Paper presented at 2018 26th International Conference on Nuclear Engineering, ICONE 2018; London; United Kingdom; 22 July 2018 through 26 July 2018. The American Society of Mechanical Engineers, 7
Open this publication in new window or tab >>A revised methodology to assess in-vessel retention strategy for high-power reactors
2018 (English)In: PROCEEDINGS OF THE 26TH INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, 18, VOL 7, The American Society of Mechanical Engineers , 2018, Vol. 7Conference paper, Published paper (Refereed)
Abstract [en]

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the design and SAM guidances (SAMGs) of several operating small and medium capacity LWRs (reactors below 500 MWe, e.g. VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power such as the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the "3-layers" configuration, where the "focusing effect" may cause higher heat fluxes than in steady-state (due to transient "thin" metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 W/m(2)) whereas the first type provides the lowest heat fluxes (around 500 kW/m(2)) but this model is not realistic due to the immiscibility of molten steel with oxide melt. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes used for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes can reach, in many cases, values which are well above 1 MW/m(2). This could reduce the residual thickness of the vessel considerably and threaten its strength and integrity. Therefore, it is clear that the safety demonstration of IVR in high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking the focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. Both situations are illustrated in this paper. The demonstration also requires an accurate thermo-mechanical analysis of the ablated vessel. The standard approach based on "yield stress" (plastic behaviour) is compared with more detailed calculations made on realistic profiles of ablated vessels. The validity of the standard approach is discussed. The current approach followed by many experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena, e.g. associated with molten pool transient behaviour, and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Some elements that might help to reach such harmonization are proposed in this paper, with a preliminary revision of the methodology that could be used to address the IVR issue. In the proposed revised methodology, the safety criterion is not based on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in the current approaches but on the minimum vessel thickness reached after ablation and the maximum pressure load that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion is more straightforward to be used in a deterministic approach.

Place, publisher, year, edition, pages
The American Society of Mechanical Engineers, 2018
Series
International Conference on Nuclear Engineering, Proceedings, ICONE ; 7
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-248375 (URN)10.1115/ICONE26-82248 (DOI)000461413000039 ()2-s2.0-85062420243 (Scopus ID)978-0-7918-5151-7 (ISBN)
Conference
2018 26th International Conference on Nuclear Engineering, ICONE 2018; London; United Kingdom; 22 July 2018 through 26 July 2018
Note

QC 20190409

Available from: 2019-04-09 Created: 2019-04-09 Last updated: 2019-04-09Bibliographically approved
Yu, P., Villanueva, W., Galushin, S., Ma, W. & Bechta, S. (2018). Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR. In: : . Paper presented at 12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12).
Open this publication in new window or tab >>Coupled Thermo-Mechanical Creep Analysis For A Nordic BWR Lower Head Using Non-Homogeneous Debris Bed Configuration From MELCOR
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2018 (English)Conference paper, Published paper (Refereed)
Abstract [en]

We present a coupled thermo-mechanical creep analysis for a Nordic BWR lower head with a non-homogeneous debris bed configuration generated with MELCOR code. A one-way coupling approach was adopted which uses the Phase-Change Effective Convectivity Model implemented in Fluent to simulate the convective heat transfer in the melt pool and the ANSYS Mechanical to simulate the vessel wall deformation induced by the thermal and mechanical load from the debris. An initial non-homogeneity of debris bed was estimated using MELCOR core relocation simulation results specifying the mass of each component (UO2/Zr/ZrO2/SS/SSOX) and temperature in each MELCOR cell of the lower head. A mapping scheme was designed to transfer this non-homogeneities debris bed to Fluent through User Defined Functions. All components were locally treated in Fluent as one ideal phase by averaging the weights of element-specific mass fractions inside each cell. Material properties (density, heat capacity, etc.) and volumetric heat in the debris were both spatial- and temperature-dependent. Meanwhile, additional simulations using homogeneous debris bed configuration but with the same amount of mass compositions were run for comparison. Results including temperature escalation, vessel failure timing and location were analyzed and compared.

Keywords
Boiling Water Reactor, Severe Accident, Vessel Failure, Thermo-Mechanical Analysis
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-259590 (URN)
Conference
12th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, Operation and Safety (NUTHOS-12)
Note

QCR 20191015

Available from: 2019-09-18 Created: 2019-09-18 Last updated: 2019-10-15Bibliographically approved
Jeltsov, M., Villanueva, W. & Kudinov, P. (2018). Seismic sloshing effects in lead-cooled fast reactors. Nuclear Engineering and Design, 332, 99-110
Open this publication in new window or tab >>Seismic sloshing effects in lead-cooled fast reactors
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 332, p. 99-110Article in journal (Refereed) Published
Abstract [en]

Pool-type primary system can improve the economy of lead-cooled fast reactors. However, partially filled pool of heavy liquid metal poses safety concerns related to seismic loads. Violent sloshing during earthquake-initiated fluid-structure interaction can lead to structural failures, gas entrapment and potential core voiding. Seismic isolation systems can be used to reduce the structural stresses, but its effect on sloshing is not straightforward. This paper presents a numerical study of seismic sloshing in ELSY reactor. The purpose is to evaluate the effects of seismic isolation system on sloshing at different levels of earthquake. Sloshing is modeled using computational fluid dynamics with a volume of fluid free surface capturing model. Earthquake is simulated using synthetic seismic data produced in SILER project as a boundary condition. Simultaneous verification and validation of the numerical model using a dam break experiment is presented. The adverse resonance effect of seismic isolation system is demonstrated in terms of sloshing-induced hydrodynamic loads and gas entrapment. Effectiveness of seismic isolation system is discussed separately for design and beyond design seismic levels. Partitioning baffles are proposed as a potential mitigation measure in the design and their effect is analyzed.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
CFD, FSI, Gas entrapment, LFR, Seismic isolation, Seismic sloshing
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-227559 (URN)10.1016/j.nucengdes.2018.03.020 (DOI)000430395700010 ()2-s2.0-85044166706 (Scopus ID)
Funder
EU, FP7, Seventh Framework Programme, 295485
Note

QC 20180509

Available from: 2018-05-09 Created: 2018-05-09 Last updated: 2018-05-23Bibliographically approved
Jeltsov, M., Villanueva, W. & Kudinov, P. (2018). Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core. Nuclear Engineering and Design, 328, 255-265
Open this publication in new window or tab >>Steam generator leakage in lead cooled fast reactors: Modeling of void transport to the core
2018 (English)In: Nuclear Engineering and Design, ISSN 0029-5493, E-ISSN 1872-759X, Vol. 328, p. 255-265Article in journal (Refereed) Published
Abstract [en]

Steam generator tube leakage and/or rupture (SGTL/R) is one of the safety issues for pool type liquid metal cooled fast reactors. During SGTL/R, water is injected from high-pressure secondary side to low-pressure primary side. The possible consequences of such an event include void transport to the core that has adverse effects on the reactor performance including heat transfer deterioration and reactivity insertion. This paper addresses the potential transport of steam bubbles to the core and subsequent void accumulation in the primary system in ELSY conceptual reactor. A CFD model of the primary coolant system for nominal operation is developed and verified. Bubble motion is simulated using Lagrangian tracking of steam bubbles in Eulerian flow field. The effects of uncertainties in the bubble size distribution and bubble drag are addressed. A probabilistic methodology to estimate the core and primary system voiding rates is proposed and demonstrated. A family of drag correlations by Tomiyama et al. (1998) provide the best agreement with the available experimental data. Primary system and core voiding analysis demonstrate that the smallest (sub-millimeter) bubbles have the highest probability to be entrained and remain in the coolant flow. It is found that leaks at the bottom region of the SG result in larger rates of void accumulation.

Place, publisher, year, edition, pages
Elsevier, 2018
Keywords
Bubble transport, CFD, LFR, Steam generator tube leakage/rupture
National Category
Other Physics Topics
Identifiers
urn:nbn:se:kth:diva-221684 (URN)10.1016/j.nucengdes.2018.01.006 (DOI)000427432300023 ()2-s2.0-85040467440 (Scopus ID)
Funder
EU, FP7, Seventh Framework Programme, 249668
Note

QC 20180122

Available from: 2018-01-22 Created: 2018-01-22 Last updated: 2018-05-23Bibliographically approved
Li, H., Villanueva, W., Puustinen, M., Laine, J. & Kudinov, P. (2018). Thermal stratification and mixing in a suppression pool induced by direct steam injection. Annals of Nuclear Energy, 111, 487-498
Open this publication in new window or tab >>Thermal stratification and mixing in a suppression pool induced by direct steam injection
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2018 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 111, p. 487-498Article in journal (Refereed) Published
Abstract [en]

An experimental and numerical investigation of thermal stratification and mixing in a suppression pool is presented. Steam injected into a drywell flows through a blowdown pipe and then down to the pressure suppression pool where direct contact condensation occurs. The steam venting and condensation is a source of heat and momentum. A complex interplay between the two leads either to thermal stratification or mixing of the pool. The experiments are conducted in a scaled down PPOOLEX facility at Lappeenranta University of Technology (LUT). The corresponding numerical simulations are performed using GOTHIC with the Effective Heat Source (EHS) and Effective Momentum Source (EMS) models. The EHS/EMS models, that have been previously proposed, predict the development of thermal stratification and mixing during a steam injection into a large pool of water. The experiments exhibit the development of thermal stratification in the pool at relatively low mass flow rates and then pool mixing when the mass flow rates are increased but later thermal stratification can re-develop even at the same relatively high mass flow rates, which is due to the increasing pool temperature that shifts the condensation to a different regime. The numerical simulations quantitatively capture this complex transient pool behavior and are in excellent agreement with the transient averaged pool temperature and water level in the pool. 

Place, publisher, year, edition, pages
Elsevier Ltd, 2018
Keywords
Direct steam injection, Pressure suppression pool, Thermal stratification and mixing, Condensation, Mass transfer, Mixing, Numerical models, Steam, Steam condensers, Thermal stratification, Water injection, Water levels, Direct contact condensation, Heat sources, Low mass flow rates, Mass flow rate, Momentum sources, Numerical investigations, Pool mixing, Steam injection, Lakes
National Category
Physical Sciences
Identifiers
urn:nbn:se:kth:diva-216805 (URN)10.1016/j.anucene.2017.09.014 (DOI)000413877800044 ()2-s2.0-85029704434 (Scopus ID)
Note

Export Date: 24 October 2017; Article; CODEN: ANEND; Correspondence Address: Villanueva, W.; Division of Nuclear Power Safety, Royal Institute of Technology (KTH), Roslagstullsbacken 21, Sweden; email: walterv@kth.se. QC 20171205

Available from: 2017-12-05 Created: 2017-12-05 Last updated: 2017-12-05Bibliographically approved
Yu, P., Komlev, A. A., Villanueva, W., Li, Y., Ma, W. & Bechta, S. (2016). Pre-Test Simulations of SIMECO-2 Experiments on Stratified Melt Pool Heat Transfer. In: : . Paper presented at The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11). , Article ID N11P0585.
Open this publication in new window or tab >>Pre-Test Simulations of SIMECO-2 Experiments on Stratified Melt Pool Heat Transfer
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2016 (English)Conference paper, Published paper (Refereed)
Abstract [en]

Severe accident progression in light water reactors can lead to the formation of a melt pool in the lower head that can impose thermo-mechanical loads on the pressure vessel, and subsequently can lead to vessel failure. For quantification of the thermal load which is important to in-vessel corium coolability and retention, various experiments have been carried out to investigate the heat transfer characteristics of melt pools, including the SIMECO experiment accomplished at KTH (Sehgal et al., 1998), which used low melting-point materials as the simulant of corium. In order to reduce the gaps in temperature and scale between experimental and prototypical conditions, a new test facility named SIMECO-2 is being designed at KTH (supported by the EU project IVMR), which features higher temperature (up to 900 ℃℃) and larger scale (1 meter in diameter), aiming to investigate the natural convection heat transfer of a stratified melt pool and the effects of different parameters/factors such as temperature of melt, thickness of boundary crust, thickness of top layer, top layer cooling. The present study is to provide pre-test calculations using the PECM method (Tran and Dinh, 2009), with the objectives to provide insights and analytical support to the design of the SIMECO-2 facility, including determination of required input power, as well as estimate of the temperature and heat flux distributions in the layers and time to reach steady state mode. A calculation was first performed for a reference base case with one-layer pool for which a CFD simulation was also conducted as benchmark. The calculations were then carried on to investigate the influences of different boundary conditions and internal heat sources on heat transfer. Finally the thermal behavior of a two-layer melt pool configuration was addressed in detail, and suggestions for the experimental conditions were provided.

National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-259588 (URN)
Conference
The 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11)
Note

QC 20191021

Available from: 2019-09-18 Created: 2019-09-18 Last updated: 2019-10-21Bibliographically approved
Dietrich, P., Kretzschmar, F., Miassoedov, A., Class, A., Villanueva, W. & Bechta, S. (2015). Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum. In: International Conference on Nuclear Engineering, Proceedings, ICONE: . Paper presented at 23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015, 17 May 2015 through 21 May 2015. JSME
Open this publication in new window or tab >>Coupling of melcor with the pecm for improved modelling of a core melt in the lower plenum
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2015 (English)In: International Conference on Nuclear Engineering, Proceedings, ICONE, JSME , 2015Conference paper, Published paper (Refereed)
Abstract [en]

MELCOR contains a coupling interface based on the MPI-Standard, which enables the communication to other codes such as RELAP5 or GASFLOW. However a detailed knowledge of this coupling interface in MELCOR is necessary to use this possibility. Therefore, at the KIT the software tool DINAMO (Direct Interface for Adding Models) has been developed. This program contains the coupling routines as well as an interface to communicate with other programs. Using DINAMO it is also possible to utilize new or enhanced models for phenomena, which occur during a severe accident in a nuclear power plant, in MELCOR without modification of the MELCOR source code. In the present work MELCOR calculations of experiments in the LIVE-Facility are presented. The LIVE-Facility is used to simulate the behavior of a melt in the lower plenum of a reactor pressure vessel (RPV). For these calculations we coupled MELCOR via DINAMO with the Phase-Change Effective Convectivity Model (PECM), which has been developed at the KTH in Stockholm. Using the PECM it is possible to improve the prediction of a core melt in the lower plenum of a RPV in case of a core melt accident.

Place, publisher, year, edition, pages
JSME, 2015
Keywords
Coupled codes, LIVE, Lower head, MELCOR, PECM
National Category
Engineering and Technology
Identifiers
urn:nbn:se:kth:diva-184194 (URN)2-s2.0-84959049239 (Scopus ID)
Conference
23rd International Conference on Nuclear Engineering: Nuclear Power - Reliable Global Energy, ICONE 2015, 17 May 2015 through 21 May 2015
Note

QC  20160330

Available from: 2016-03-30 Created: 2016-03-30 Last updated: 2016-03-30Bibliographically approved
Organisations
Identifiers
ORCID iD: ORCID iD iconorcid.org/0000-0003-3132-7252

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