H. Willschütz - Academia.edu (original) (raw)

Papers by H. Willschütz

10th International Conference on Nuclear Engineering, Volume 1, 2002

ABSTRACT To get an improved understanding and knowledge of the melt pool convection and the vesse... more ABSTRACT To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments are currently underway at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behaviour of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. Due to the multi axial creep deformation of the vessel with a non-uniform temperature field these experiments are on the one hand an excellent source of data to validate numerical creep models which are developed on the basis of uniaxial creep tests. On the other hand the results of pre-test calculations can be used to optimize the experimental procedure and by supporting decision making during the experiment. For that, a Finite Element model is developed based on a multi-purpose code. After post-test calculations for the FOREVER-C2 experiment, pre-test calculations for the forthcoming experiments are performed. Additionally metallographic post test investigations of the experiments are conducted to improve the numerical damage model and to adjust the correlation between the metallographic observations and the calculated damage. Taking into account both - experimental and numerical results - gives a good opportunity to improve the simulation and understanding of real accident scenarios. After analysing the calculations, it seems to be advantageous to introduce a vessel support which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it is possible to install an absolutely passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down. (authors)

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreak-tor (LWR) ist es n... more Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreak-tor (LWR) ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Es wurden bereits eine Reihe von Experimenten durchge-führt, welche Erkenntnisse hierüber liefern sollen. Vom Institut für Sicherheitsforschung des FZR wurde ein Finite-Elemente-Modell er-stellt, das sowohl die Temperaturfeldberechnung für die Wand als auch die elasto-plastische Mechanik der Behälterwand beschreibt. Dabei wurde ein fortgeschrittenes Modell für das Kriechen und für die Materialschädigung entwickelt und an Hand von experimentellen Daten validiert. Die thermischen und mechanischen Berechnungen sind rekursiv und sequentiell gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Das Modell wurde für Voraus- und Nachrechnungen der FOREVER-Exper...

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reac... more Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluate...

Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible f... more Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Numerous experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work /REM 1993/, /THF 1997/, /CHU 1999/. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Fi-nite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation a...

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendi... more Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgen...

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendi... more Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Es wurden bereits eine Reihe von Experimenten durchgeführt, welche Erkenntnisse hierüber liefern sollen. Begleitend wurden in Einzelversuchen Materialeigenschaften ermittelt, sowie theoretische und numerische Arbeiten durchgeführt. Für die Simulation von Experimenten zum Versagen der Bodenkalotte, wie OLHF oder FOREVER, ist es notwendig, Kriechen und Plastizität zu berücksichtigen. Gleichzeitig müssen geeignete Modelle das Temperaturfeld in der Behälterwand für die mechanischen Rechnungen bereitstellen. Vom Institut für Sicherheitsforschung des FZR wird ein Finite-Elemente-Modell erstellt, das sowohl die Temperaturfeldberechnung für die Wand als auch die elasto-plastische Mechanik der Behälterwand modelliert. Die bisher durchgeführten Ar...

To get an improved understanding and knowledge of the processes and phenomena during the late pha... more To get an improved understanding and knowledge of the processes and phenomena during the late phase of a core melt down accident the FOREVER-experiments (Failure of Reactor Vessel Retention) are currently underway. These experiments are simulating the lower head of a reactor pressure vessel under the load of a melt pool with internal heat sources. The geometrical scale of the experiments is 1:10 compared to a common Light Water Reactor. During the first series of experiments the Creep behaviour of the vessel is investigated. Due to the multi-axial creep deformation of the three-dimensional vessel with a non-uniform temperature field these experiments are on the one hand an excellent possibility to validate numerical creep models which are developed on the basis of uniaxial creep tests. On the other hand the results of pre-test calculations can be used for an optimized experimental procedure. Therefore a Finite Element model is developed on the basis of the multi-purpose commercial c...

For calculations of Lower Head Failure experiments like FOREVER it is necessary to model creep an... more For calculations of Lower Head Failure experiments like FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a nu- merical creep data base

Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formati... more Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of Reactor Vessel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences...

Nuclear Engineering and Design, 2004

ABSTRACT Scaled coupled melt pool convection and vessel creep failure experiments are being perfo... more ABSTRACT Scaled coupled melt pool convection and vessel creep failure experiments are being performed in the FOREVER program at the Royal Institute of Technology, Stockholm. These experiments are simulating the lower head of a pressurized reactor vessel under the thermal load of a melt pool with internal heat sources and a specified internal pressure. Due to the multi-axial creep deformation of the three-dimensional vessel with a prototypic non-uniform temperature field these experiments offer an excellent opportunity to validate numerical creep models. A Finite Element Model is developed and using the Computational Fluid Dynamic module, the melt pool convection is simulated and the temperature field within the vessel wall is evaluated. The transient structural mechanical calculations are then performed applying a new creep modeling procedure. Additionally, the material damage is evaluated considering the creep deformation as well as the prompt plastic deformation.After post-test calculations for the FOREVER-C2 experiment, pre-test calculations for the forthcoming experiments are performed. Taking into account both—experimental and numerical results—gives a good opportunity to improve the simulation and understanding of real accident scenarios. After analyzing the results of the calculations, it seems to be advantageous to provide a vessel support, which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it may be advantageous to install a passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.

Nuclear Engineering and Design, 2001

... during the late phase of the experiment when many thermocouples (TC), at high temperature loc... more ... during the late phase of the experiment when many thermocouples (TC), at high temperature locations, failed. ... Much greater creep strains were obtained during 3 h of operation at temperature and pressure ... The analysis presented in this work is based on the latter experiment. ...

Nuclear Engineering and Design, 2005

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations ... more The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and

Nuclear Engineering and Design, 2008

A good understanding of the mechanical behaviour of the reactor pressure vessel (RPV) lower head ... more A good understanding of the mechanical behaviour of the reactor pressure vessel (RPV) lower head is necessary both for severe accident assessment and for the definition of appropriate accident mitigation strategies. Indeed, a well-characterized failure of the lower head leads to a better evaluation of the quantity and kinetics with which core material can escape into the containment. These are the initial conditions for several ex-vessel events such as direct heating of the containment or molten core-concrete interaction. In this context, the objectives of the joint ongoing work of the WP10-2 group of SARNET are: (1) improvement of predictability of the time, mode and location of RPV failure; (2) development of adequate models with the ultimate aim of being included into integral codes; (3) interpretation/analysis of experiments with models/codes combined with sensitivity studies; and (4) better understanding of the breach opening process in order to better characterize the corium release into the containment. Different approaches are considered: a simplified but well predicting model recently implemented in the severe accident Astec and Icare-Cathare codes, and viscoplasticity models implemented in the Cast3m, Ansys and Code Aster finite element codes. Several failure criteria are considered: stress criterion, strain criterion and damage evaluation (coupled way or post-evaluation). In this paper, the OLHF-1 experiment has been used to assess the models, to perform sensitivity studies and to evaluate failure criteria that could be applied in the case of reactors. All the partners performed 2D axisymmetric analyses, allowing the evaluation of time, mode and location of vessel failure. Nevertheless, CEA conducted further 3D calculations in order to study crack propagation and the corresponding results will be presented separately at the end of the paper. The numerical formulation of the different models used is given and a comparison of experimental and numerical results is presented. The paper also shows the progress made with the objective of defining failure criteria that can be used for reactor vessel applications.

Annals of Nuclear Energy, 2003

Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible... more Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be considered for a determination of the loadings on the containment. For pre-and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed in which the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties is performed. For the consideration of the tertiary creep stage and for the evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in three levels: starting with the simulation of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called ''tube-failure-experiments'' are modeled: the RUPTHER-14 and the ''MPA-Meppen''-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German

Annals of Nuclear Energy, 2008

This paper presents methods to compute J-integral values for cracks in two-and three-dimensional ... more This paper presents methods to compute J-integral values for cracks in two-and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.

Annals of Nuclear Energy, 2006

Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible fai... more Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible failure mode of the reactor pressure vessel (RPV) and its failure time have to be investigated for a determination of the load conditions for subsequent containment analyses. Worldwide several experiments have been performed in this field accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model (FEM) has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre-and post-test calculations for the FOREVER test series representing the lower head RPV of a pressurised water reactor (PWR) in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. In this paper the differences between the results of a simple coupled and a recursive coupled FE-simulation are highlighted. Due to the thermal expansion at the beginning and the accumulating creep strain later on the shape of the melt pool and of the vessel wall are changing. Despite of the fact that these relative small geometrical changes take place relatively slowly over time, the effect on the temperature field is rather significant concerning the mechanical material behaviour and the resulting failure time. Assuming the same loading conditions, the change in the predicted failure time between the simple and the recursive coupled model is in the order of magnitude of the total failure time of the simple model. The comparison with results from the FOREVER-experiments shows that the recursive coupled model is closer to reality than the one-way coupled model.

Jahrestagung Kerntechnik, 2002

Scaled coupled melt pool convection and vessel creep failure experiments are being performed in t... more Scaled coupled melt pool convection and vessel creep failure experiments are being performed in the FOREVER program at the Royal Institute of Technology, Stockholm. These experiments are simulating the lower head of a pressurized reactor vessel under the thermal load of a ...

qucosa.de

Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failu... more Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed ...

10th International Conference on Nuclear Engineering, Volume 1, 2002

ABSTRACT To get an improved understanding and knowledge of the melt pool convection and the vesse... more ABSTRACT To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments are currently underway at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behaviour of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences in a depressurization scenario. Due to the multi axial creep deformation of the vessel with a non-uniform temperature field these experiments are on the one hand an excellent source of data to validate numerical creep models which are developed on the basis of uniaxial creep tests. On the other hand the results of pre-test calculations can be used to optimize the experimental procedure and by supporting decision making during the experiment. For that, a Finite Element model is developed based on a multi-purpose code. After post-test calculations for the FOREVER-C2 experiment, pre-test calculations for the forthcoming experiments are performed. Additionally metallographic post test investigations of the experiments are conducted to improve the numerical damage model and to adjust the correlation between the metallographic observations and the calculated damage. Taking into account both - experimental and numerical results - gives a good opportunity to improve the simulation and understanding of real accident scenarios. After analysing the calculations, it seems to be advantageous to introduce a vessel support which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it is possible to install an absolutely passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down. (authors)

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreak-tor (LWR) ist es n... more Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreak-tor (LWR) ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Es wurden bereits eine Reihe von Experimenten durchge-führt, welche Erkenntnisse hierüber liefern sollen. Vom Institut für Sicherheitsforschung des FZR wurde ein Finite-Elemente-Modell er-stellt, das sowohl die Temperaturfeldberechnung für die Wand als auch die elasto-plastische Mechanik der Behälterwand beschreibt. Dabei wurde ein fortgeschrittenes Modell für das Kriechen und für die Materialschädigung entwickelt und an Hand von experimentellen Daten validiert. Die thermischen und mechanischen Berechnungen sind rekursiv und sequentiell gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Das Modell wurde für Voraus- und Nachrechnungen der FOREVER-Exper...

Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reac... more Considering the late in-vessel phase of an unlikely core melt down scenario in a light water reactor (LWR) with the formation of a corium pool in the lower head of the re-actor pressure vessel (RPV) the possible failure modes of the RPV and the time to failure have to be investigated to assess the possible loadings on the containment. In this work, an integral model was developed to describe the processes in the lower plenum of the RPV. Two principal model domains have to be distinguished: The temperature field within the melt and RPV is calculated with a thermodynamic model, while a mechanical model is used for the structural analysis of the vessel wall. In the introducing chapters a description is given of the considered accident scenario and the relevant analytical, experimental, and numerical investigations are discussed which were performed worldwide during the last three decades. Following, the occur-ring physical phenomena are analysed and the scaling differences are evaluate...

Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible f... more Considering the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Numerous experiments have been performed accompanied with material properties evaluation, theoretical, and numerical work /REM 1993/, /THF 1997/, /CHU 1999/. For pre- and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Fi-nite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed where the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation a...

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendi... more Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Vom Institut für Sicherheitsforschung des FZD wurden Finite-Elemente-Modelle erstellt, die sowohl die Temperaturfeldberechnung für die Wand als auch die elastoplastische Mechanik der Behälterwand beschreibt. Die thermischen und mechanischen Berechnungen sind gekoppelt. Das Modell ist in der Lage, Versagenszeit und Versagensposition eines Behälters mit beheiztem Schmelzepool zu berechnen. Es existieren Modelle für die Druckwasserreaktortypen KONVOI und WWER-1000. Es wurden prototypische Szenarien mit und ohne externe Flutung des RDB untersucht, wobei die homogen und die segregierte Schmelzepoolkonfiguration betrachtet wurden. Zusätzlich wurde eine bruchmechanische Bewertung des Thermoschocks, der durch die externe Flutung entsteht, vorgen...

Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendi... more Bezüglich eines hypothetischen Kernschmelzeszenarios in einem Leichtwasserreaktor ist es notwendig, mögliche Versagensformen des Reaktordruckbehälters sowie Versagenszeiträume zu untersuchen, um die Belastung für das Containment bestimmen zu können. Es wurden bereits eine Reihe von Experimenten durchgeführt, welche Erkenntnisse hierüber liefern sollen. Begleitend wurden in Einzelversuchen Materialeigenschaften ermittelt, sowie theoretische und numerische Arbeiten durchgeführt. Für die Simulation von Experimenten zum Versagen der Bodenkalotte, wie OLHF oder FOREVER, ist es notwendig, Kriechen und Plastizität zu berücksichtigen. Gleichzeitig müssen geeignete Modelle das Temperaturfeld in der Behälterwand für die mechanischen Rechnungen bereitstellen. Vom Institut für Sicherheitsforschung des FZR wird ein Finite-Elemente-Modell erstellt, das sowohl die Temperaturfeldberechnung für die Wand als auch die elasto-plastische Mechanik der Behälterwand modelliert. Die bisher durchgeführten Ar...

To get an improved understanding and knowledge of the processes and phenomena during the late pha... more To get an improved understanding and knowledge of the processes and phenomena during the late phase of a core melt down accident the FOREVER-experiments (Failure of Reactor Vessel Retention) are currently underway. These experiments are simulating the lower head of a reactor pressure vessel under the load of a melt pool with internal heat sources. The geometrical scale of the experiments is 1:10 compared to a common Light Water Reactor. During the first series of experiments the Creep behaviour of the vessel is investigated. Due to the multi-axial creep deformation of the three-dimensional vessel with a non-uniform temperature field these experiments are on the one hand an excellent possibility to validate numerical creep models which are developed on the basis of uniaxial creep tests. On the other hand the results of pre-test calculations can be used for an optimized experimental procedure. Therefore a Finite Element model is developed on the basis of the multi-purpose commercial c...

For calculations of Lower Head Failure experiments like FOREVER it is necessary to model creep an... more For calculations of Lower Head Failure experiments like FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a nu- merical creep data base

Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formati... more Assuming the hypothetical scenario of a severe accident with subsequent core meltdown and formation of a melt pool in the reactor pressure vessel (RPV) lower plenum of a Light Water Reactor (LWR) leads to the question about the behavior of the RPV. One accident management strategy could be to stabilize the in-vessel debris configuration in the RPV as one major barrier against uncontrolled release of heat and radio nuclides. To get an improved understanding and knowledge of the melt pool convection and the vessel creep and possible failure processes and modes occurring during the late phase of a core melt down accident the FOREVER-experiments (Failure Of Reactor Vessel Retention) have been performed at the Division of Nuclear Power Safety of the Royal Institute of Technology Stockholm. These experiments are simulating the behavior of the lower head of the RPV under the thermal loads of a convecting melt pool with decay heating, and under the pressure loads that the vessel experiences...

Nuclear Engineering and Design, 2004

ABSTRACT Scaled coupled melt pool convection and vessel creep failure experiments are being perfo... more ABSTRACT Scaled coupled melt pool convection and vessel creep failure experiments are being performed in the FOREVER program at the Royal Institute of Technology, Stockholm. These experiments are simulating the lower head of a pressurized reactor vessel under the thermal load of a melt pool with internal heat sources and a specified internal pressure. Due to the multi-axial creep deformation of the three-dimensional vessel with a prototypic non-uniform temperature field these experiments offer an excellent opportunity to validate numerical creep models. A Finite Element Model is developed and using the Computational Fluid Dynamic module, the melt pool convection is simulated and the temperature field within the vessel wall is evaluated. The transient structural mechanical calculations are then performed applying a new creep modeling procedure. Additionally, the material damage is evaluated considering the creep deformation as well as the prompt plastic deformation.After post-test calculations for the FOREVER-C2 experiment, pre-test calculations for the forthcoming experiments are performed. Taking into account both—experimental and numerical results—gives a good opportunity to improve the simulation and understanding of real accident scenarios. After analyzing the results of the calculations, it seems to be advantageous to provide a vessel support, which can unburden the vessel from a part of the mechanical load and, therefore, avoid the vessel failure or at least prolong the time to failure. This can be a possible accident mitigation strategy. Additionally, it may be advantageous to install a passive automatic control device to initiate the flooding of the reactor pit to ensure external vessel cooling in the event of a core melt down.

Nuclear Engineering and Design, 2001

... during the late phase of the experiment when many thermocouples (TC), at high temperature loc... more ... during the late phase of the experiment when many thermocouples (TC), at high temperature locations, failed. ... Much greater creep strains were obtained during 3 h of operation at temperature and pressure ... The analysis presented in this work is based on the latter experiment. ...

Nuclear Engineering and Design, 2005

The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations ... more The assessment of reactor vessel integrity (ARVI) project involved a total of nine organizations from Europe and USA. The work consisted of experiments and analysis development. The modeling activities in the area of structural analyses were focused on the support of EC-FOREVER experiments as well as on the exploitation of the data obtained from those experiments for modeling of creep deformation and the validation of the industry structural codes. Work was also performed for extension of melt natural convection analyses to consideration of stratification, and mixing (in the CFD codes). Other modeling activities were for (1) gap cooling CHF and (2) developing simple models for system code. Finally, the methodology and data was applied for the design of IVMR severe accident management scheme for VVER-440/213 plants. The work was broken up into five packages. They were divided into tasks, which were performed by different partners. The major experimental project continued was EC-FOREVER in which data was obtained on in-vessel melt pool coolability. In previous EC-FOREVER experiments data was obtained on melt pool natural convection and lower head creep failure and rupture. Those results obtained were related to the following issues: (1) multiaxial creep laws for different vessel steels, (2) effects of penetrations, and (3) mode and location of lower head failure. The two EC-FOREVER tests reported here are related to (a) the effectiveness of gap cooling and (b) water ingression for in vessel melt coolability. Two other experimental projects were also conducted. One was the COPO experiments, which was concerned with the effects of stratification and

Nuclear Engineering and Design, 2008

A good understanding of the mechanical behaviour of the reactor pressure vessel (RPV) lower head ... more A good understanding of the mechanical behaviour of the reactor pressure vessel (RPV) lower head is necessary both for severe accident assessment and for the definition of appropriate accident mitigation strategies. Indeed, a well-characterized failure of the lower head leads to a better evaluation of the quantity and kinetics with which core material can escape into the containment. These are the initial conditions for several ex-vessel events such as direct heating of the containment or molten core-concrete interaction. In this context, the objectives of the joint ongoing work of the WP10-2 group of SARNET are: (1) improvement of predictability of the time, mode and location of RPV failure; (2) development of adequate models with the ultimate aim of being included into integral codes; (3) interpretation/analysis of experiments with models/codes combined with sensitivity studies; and (4) better understanding of the breach opening process in order to better characterize the corium release into the containment. Different approaches are considered: a simplified but well predicting model recently implemented in the severe accident Astec and Icare-Cathare codes, and viscoplasticity models implemented in the Cast3m, Ansys and Code Aster finite element codes. Several failure criteria are considered: stress criterion, strain criterion and damage evaluation (coupled way or post-evaluation). In this paper, the OLHF-1 experiment has been used to assess the models, to perform sensitivity studies and to evaluate failure criteria that could be applied in the case of reactors. All the partners performed 2D axisymmetric analyses, allowing the evaluation of time, mode and location of vessel failure. Nevertheless, CEA conducted further 3D calculations in order to study crack propagation and the corresponding results will be presented separately at the end of the paper. The numerical formulation of the different models used is given and a comparison of experimental and numerical results is presented. The paper also shows the progress made with the objective of defining failure criteria that can be used for reactor vessel applications.

Annals of Nuclear Energy, 2003

Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible... more Investigating the hypothetical core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be considered for a determination of the loadings on the containment. For pre-and post-test calculations of Lower Head Failure experiments like OLHF or FOREVER it is necessary to model creep and plasticity processes. Therefore a Finite Element Model is developed at the FZR using a numerical approach which avoids the use of a single creep law employing constants derived from the data for a limited stress and temperature range. Instead of this a numerical creep data base (CDB) is developed in which the creep strain rate is evaluated in dependence on the current total strain, temperature and equivalent stress. A main task for this approach is the generation and validation of the CDB. Additionally the implementation of all relevant temperature dependent material properties is performed. For the consideration of the tertiary creep stage and for the evaluation of the failure times a damage model according to an approach of Lemaitre is applied. The validation of the numerical model is performed by the simulation of and comparison with experiments. This is done in three levels: starting with the simulation of single uniaxial creep tests, which is considered as a 1D-problem. In the next level so called ''tube-failure-experiments'' are modeled: the RUPTHER-14 and the ''MPA-Meppen''-experiment. These experiments are considered as 2D-problems. Finally the numerical model is applied to scaled 3D-experiments, where the lower head of a PWR is represented in its hemispherical shape, like in the FOREVER-experiments. An interesting question to be solved in this frame is the comparability of the French 16MND5 and the German

Annals of Nuclear Energy, 2008

This paper presents methods to compute J-integral values for cracks in two-and three-dimensional ... more This paper presents methods to compute J-integral values for cracks in two-and three-dimensional thermo-mechanical loaded structures using the finite element code ANSYS. The developed methods are used to evaluate the behavior of a crack on the outside of an emergency cooled reactor pressure vessel (RPV) during a severe core melt down accident. It will be shown, that water cooling of the outer surface of a RPV during a core melt down accident can prevent vessel failure due to creep and ductile rupture. Further on, we present J-integral values for an assumed crack at the outside of the lower plenum of the RPV, at its most stressed location for an emergency cooling (thermal shock) scenario.

Annals of Nuclear Energy, 2006

Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible fai... more Postulating an unlikely core melt down accident for a light water reactor (LWR), the possible failure mode of the reactor pressure vessel (RPV) and its failure time have to be investigated for a determination of the load conditions for subsequent containment analyses. Worldwide several experiments have been performed in this field accompanied with material properties evaluation, theoretical, and numerical work. At the Institute of Safety Research of the FZR a finite element model (FEM) has been developed simulating the thermal processes and the viscoplastic behaviour of the vessel wall. An advanced model for creep and material damage has been established and has been validated using experimental data. The thermal and the mechanical calculations are sequentially and recursively coupled. The model is capable of evaluating fracture time and fracture position of a vessel with an internally heated melt pool. The model was applied to pre-and post-test calculations for the FOREVER test series representing the lower head RPV of a pressurised water reactor (PWR) in the geometrical scale of 1:10. These experiments were performed at the Royal Institute of Technology in Stockholm. In this paper the differences between the results of a simple coupled and a recursive coupled FE-simulation are highlighted. Due to the thermal expansion at the beginning and the accumulating creep strain later on the shape of the melt pool and of the vessel wall are changing. Despite of the fact that these relative small geometrical changes take place relatively slowly over time, the effect on the temperature field is rather significant concerning the mechanical material behaviour and the resulting failure time. Assuming the same loading conditions, the change in the predicted failure time between the simple and the recursive coupled model is in the order of magnitude of the total failure time of the simple model. The comparison with results from the FOREVER-experiments shows that the recursive coupled model is closer to reality than the one-way coupled model.

Jahrestagung Kerntechnik, 2002

Scaled coupled melt pool convection and vessel creep failure experiments are being performed in t... more Scaled coupled melt pool convection and vessel creep failure experiments are being performed in the FOREVER program at the Royal Institute of Technology, Stockholm. These experiments are simulating the lower head of a pressurized reactor vessel under the thermal load of a ...

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Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failu... more Considering the unlikely core melt down scenario for a light water reactor (LWR) a possible failure mode of the reactor pressure vessel (RPV) and its failure time has to be investigated for a determination of the loadings on the containment. Worldwide several experiments have been performed ...