In the event of a severe core meltdown accident in a pressurised water reactor (PWR), core material can relocate into the lower head of the vessel resulting in significant thermal and pressure loads being imposed on the vessel. In the event of reactor pressure vessel (RPV) failure there is the possibility of core material being released towards the containment.On the basis of the loading conditions and the temperature distribution, the determination of the mode, timing, and size of lower head failure is of prime importance in the assessment of core melt accidents. This is because they define the initial conditions for ex-vessel events such as core/basemat interactions, fuel/coolant interactions, and direct containment heating. When lower head failure occurs (i) the understanding of the mechanism of lower head creep deformation; (ii) breach stability and its kinetic of propagation leading to the failure; (iii) and developing predictive modelling capabilities to better assess the consequences of ex-vessel processes, are of equal importance.The objective of this paper is to present an original characterization programme of vessel steel tearing properties by carrying out high temperature tearing tests on Compact Tension (CT) specimens.The influence of metallurgical composition on the kinetics of tearing is investigated as previous work on different RPV steels has shown a possible loss of ductility at high temperatures depending on the initial chemical composition of the vessel material. Small changes in the composition can lead to different types of rupture behaviour at high temperatures.The experimental programme has been conducted on various French RPV 16MND5 steels for temperatures ranging from 900 degrees C to 1100 degrees C. Comparisons between the tests performed on these various 16MND5 steels show that this approach is appropriate to characterize the difference in ductility observed at high temperatures.The aim of this experimental study is also to contribute to the definition of a tearing criterion by identifying, on the basis of CT results, the related material parameters at temperatures representative of the real severe accident conditions.This experimental campaign has been carried out in partnership with IRSN in the framework of a research programme whose purpose is to complete the mechanical properties database of 16MND5 steel and to model tearing failure in French RPV lower head vessels under severe conditions (Koundy et al., 2008). (C) 2011 Elsevier B.V. All rights reserved.

Within the SARNET network of excellence in the 6th Framework Programme of the European Commission, the severe accident integral code ASTEC, jointly developed by IRSN (France) and GRS (Germany), has been validated against international experiments to evaluate the suitability and capability of new or improved models implemented in successive code versions up to V1.3rev2, delivered in December 2007. This paper focuses on the code applications concerning circuit thermal-hydraulics and core degradation to integral and separate-effect experiments: for the CESAR thermal-hydraulic module, BETHSY 9.1 b, PACTEL ISP 33 and T2.1, PMK2-SBLOCA, LOFT-LP-FP-2; for the DIVA core degradation module, CORA-13 and -W2, QUENCH-11 and -13, LOFT-LP-FP-2, Phebus FPT-4, FARO L14 and L28, LIVE-L1, OLHF-1, FOREVER EC2. Besides, the TMI-2 accident has been analyzed using the CESAR and DIVA modules in a coupling mode.The emphasis was put on the following new or improved models: i.e. in CESAR, reflooding of an intact core, condensation in the pressurizer, sub-critical break flow correlation, and new pressurizer spray model: in DIVA, corium behaviour in the lower head and lower head mechanical failure.For thermal-hydraulics in the circuits, good results have been obtained with ASTEC on the three integral experiments that cover various thermal-hydraulic flow regimes: LOFT-LP-FP-2 in Western PWR geometry and the two PACTEL experiments in VVER-440 geometry. These good results have been confirmed by the validation done on several BETHSY integral tests.For core degradation, the ASTEC results are good for early-phase models of core heat-up, oxidation and hydrogen production (before any quenching phase) on different CORA and QUENCH experiments and on LOFT-LP-FP-2. For the in-vessel late-phase, the results can be considered as good regarding debris bed melting (Phebus FPT-4), corium fragmentation at slump into vessel lower plenum (FARO), molten pool behaviour in lower plenum (LIVE-L1), and vessel lower head mechanics (OLHF-1 and FOREVER EC2). Furthermore, the first two phases of the TMI-2 accident before core reflooding are very well calculated by ASTEC. The main remaining modelling weaknesses concern the reflooding of a degraded core and the corresponding hydrogen production. The implementation of detailed magma 2D relocation models in the new series of ASTEC V2 versions (the first one being released mid-2009) will allow a more realistic simulation of late phase phenomena up to the failure of the lower head. (C) 2009 Elsevier Ltd. All rights reserved.

In PWR severe accident scenarios, involving a relocation of corium (core melt) into the lower head, the possible failure mode of the reactor pressure vessel (RPV), the failure time, the failure location and the final size of the breach are regarded as key elements, since they play an important part in the ex-vessel phase of the accident.Both the LHF and OLHF experiments as well as the FOREVER experiments revealed that initiation of the failure is typically local. For the case of a uniform temperature distribution in the lower head, crack initiation occurs in the thinnest region and for the case of a non-uniform temperature distribution, it initiates at the highest temperature region. These experimental results can be modelled numerically (but more accurately with 3D finite element codes). The failure time predictions obtained using numerical modelling agree reasonably well with the experimental values.However, the final size of the failure is still an open issue. Analyses of both the LHF and OLHF experimental data (as well as of that from the FOREVER experiments) do not enable an assessment of the final size of the breach (in relation with the testing conditions and results).Indeed. the size of breach depends on the mode of crack propagation which is directly related to the metallurgical characteristics of the RPV steel. Small changes in the initial chemical composition of the vessel material can lead to different types of rupture behaviour at high temperatures. Different rupture behaviours were observed in the LHF and OLHF experiments using the SA533BI steel. Similar observations were previously noticed during a CEA material characterization programme on the 16MND5 steel. To determine crack propagation and final failure size, 3D modelling would thus be needed with an adequate failure criterion taking into account the variability in behaviour of the RPV material at high temperatures.This paper presents an outline of the methodology being used in a current research programme of IRSN, in partnership with CEA and INSA Lyon. The aim is to model crack opening and crack propagation in French RPV lower head vessels under severe accidents conditions. This programme was initiated in 2003 and is made up of five main sections, namely an inventory of the different French PWR lower head materials, metallurgical investigations to better understand the cause of mechanical behaviour variability that is observed and related to material microstructure, Compact Tension (CT) testing of specimens to characterize the tear resistance of the material, validation of the modelling using experiments on tube specimens and the development of a new failure criterion for the 3D finite element models. (c) 2008 Elsevier B.V. All rights reserved.

In the study of severe pressurized water reactor accidents, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are usually investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exists. This may lead to an out of vessel steam explosion or to direct heating of the containment; both which have the potential to lead to early containment failure.Within the framework of the OECD Lower Head Failure (OLHF) programme, a simplified model based on the theory of shells of revolution under symmetrical loading was developed by IRSN. After successfully interpreting some other representative experiments on lower head failures, the model was recently integrated into the European integral severe accident computer ASTEC code. The model was also used to obtain the thermo-mechanical behaviour of a 900-MWe pressurized water reactor lower head, subjected to transient heat fluxes under severe accident conditions.The main objective of this paper is to present: (I) the full mathematical formulations used in the development of the model, including their matrices and integrals defined by analytical expressions; (2) the two creep laws implemented, one for the American steel SA533B1 and one for the French steel 16MND5; and (3) the various numerical interpretations of experiments using the simplified model. This paper can be considered as a theoretical manual to aid users of the simplified model during modelling of lower head failures under severe accident conditions. One of the applications presented in this paper concerns the determination of a diagram representing the vessel time to failure as a function of the pressure level and the heat flux intensity. This information has been used by IRSN in probabilistic safety assessment and severe accident management analyses. (c) 2008 Elsevier B.V. All rights reserved.

Within SARNET, the corium topic covers all the behaviors of corium from early phase of core degradation to in or ex-vessel corium recovery with the exception of corium interaction with water, direct containment heating and fission product release. The corium topic regroups in three work packages the critical mass of competence required to improve significantly the corium behavior knowledge. The spirit of the SARNET networking is to share the knowledge, the facilities and the simulation tools for severe accidents, so to reach a better efficiency and to rationalize the R&D effort at European level. Extensive benchmarking has been launched in most of the areas of research. These benchmarks were mainly dedicated to the recalculation of experiments, while, in the next periods, a larger focus will be given to integral experiments or reactor applications. Eventually, all the knowledge will be accumulated in the ASTEC severe accident simulation code through physical model improvements and extension of validation database. This paper summarizes the progress that has been achieved in the frame of the networking activities. A special focus is placed on the melt pool and debris coolability and corium-concrete interaction, in which, the effects due to multidimensional geometries and heterogeneities has been shown, during SARNET, to play a crucial role and for which further research is still needed.

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 on-going 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. (c) 2008 Elsevier B.V. All rights reserved.

In PWR severe accident scenarios, involving a relocation of corium (core melt) into the lower head, the possible failure mode of the reactor pressure vessel (RPV), the failure time, the failure location and the final size of the breach are regarded as key elements, since they play an important part in the ex-vessel phase of the accident. In the framework of the LHF and OLHF programmes, the failure time and failure location predictions were obtained using numerical modelling and agreed reasonably well with the experimental values. However the final size of the failure is still an open issue. Analyses of both the LHF and OLHF experimental data (as well as of that from the Swedish FOREVER experiments) do not enable an assessment of the final size of the breach (in relation with the testing conditions and results). Indeed, the size of breach depends on the mode of crack propagation which is directly related to the variability in behaviour of the RPV material at high temperatures. To determine crack propagation and failure final size, 3D modelling would thus be needed with an adequate failure criterion which takes into account this material behaviour variability. This paper presents an outline of the methodology being used in a current research programme of IRSN, in partnership with CEA and INSA LYON. The aim is to model crack opening and crack propagation in French RPV lower head vessels under severe accidents conditions and to develop a new failure criterion for the 3D finite element models.

The USNRC/SNL OLHF program was carried out within the framework of an OECD project. This program consisted of four one-fifth scale experiments of a reactor pressure vessel (RPV) lower head failure (LHF) under well controlled internal pressure and large throughwall temperature differentials; the objectives were to characterize the mode, timing and size of a possible PWR lower head failure in the event of a core meltdown accident. These experiments should also lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all ex-vessel events. A large quantity of escaping corium may lead to direct heating of the containment or ex-vessel steam explosion. These are important issues due to their potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, a 2D semi-analytical model has been developed and used to simulate these experiments. The aim of this effort is to develop a simplified but well predicting code that can be then implemented in European integral severe accident computer codes (ASTEC, ICARE/CATHARE). This paper presents the detailed mathematical formulation of this simplified method which is used to interpret the experimental results. The axi-symmetric shell theory under internal pressure proposed by Timoshenko has been utilised. The solution to the equilibrium equations is presented, with particular attention to the Rabotnov analytical formula. The radius and the polar angle of the deformed structure have been written as analytical expressions in order to take the large displacements and large strains into account using our mathematical formulation. The Norton type creep law and the Kachanov damage law have been used. Several failure criteria were used in the calculations and their effect on the numerical results is discussed. This 2D semi-analytical model gives very satisfactory results when compared, with the experimental and numerical results that were presented recently in the Benchmark calculations based on the first test of the OLHF program. The performance of this model is also illustrated by its capacity to accurately simulate the deformation of the lower head, including the variation of wall thickness.

In order to characterize the timing, mode and size of a possible lower head failure (LHF) of the reactor pressure vessel (RPV) in the event of a core meltdown accident, several large-scale LHF experiments were performed under the USNRC/SNL LHF program. The experiments examined lower head failure at high pressures (10MPa in most cases) and with small throughwall temperature differentials. Another USNRC/SNL LHF program, called the OLHF program, has been undertaken in the framework of an OECD project. This was an extension of the first program and dealt with low and moderate pressures (2–5MPa) but with large through wall temperature differentials. These experiments should lead to a better understanding of the mechanical behavior of the reactor vessel lower head, which is of importance both in severe accident assessment and the definition of accident mitigation strategies. A well-characterized failure of the lower head is of prime importance for the evaluation of the quantity of core material that can escape into the containment, since this defines the initial conditions for all external-vessel events. The large quantity of escaping corium may lead to direct heating of the containment. This is an important severe accident issue because of its potential to cause early containment failure. The experiments also provide data for model development and validation. For our part, as one of the program partners, numerical modeling was performed to simulate these experiments. This paper presents a detailed description of three of our numerical models used for the simulation. The first model is a simplified semi-analytical approach based on the theory of a spherical shell subjected to internal pressure. The two other methods deal with 2D finite element (2D-FE) modeling: one combines the Norton–Bailey creep law with a damage model proposed by Lemaitre–Chaboche while the other uses only a creep failure criterion but takes into account thermo-metallurgical phase transformations. One of the objectives of the simulation is the development of a simplified and valid analytical code that can be implemented in integral severe accident computer codes. The numerical results are consistent with the experimental measurements. The effect on the numerical results of the multiphase transformation of the shell material and of the two failure criteria used, one involving necking (Considère's criterion) and the other involving creep damage (Lemaitre–Chaboche) is discussed.

In the study of severe accidents of nuclear pressurized water reactors, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exist. This may lead to direct heating of the containment or outer vessel steam explosion. These issues are important due to their early containment failure potential. Since the TMI-2 accident, many theoretical and experimental investigations, relating to lower head mechanical behaviour under severe thermo-mechanical loading in the event of a core meltdown accident have been performed. IRSN participated actively in the one-fifth scale USNRC/SNL LHF and OECD LHF (OLHF) programs. Within the framework of these programs, two simplified models were developed by IRSN: the first is a simplified 1D approach based on the theory of pressurized spherical shells and the second is a simplified 2D model based on the theory of shells of revolution under symmetric loading. The mathematical formulation of both models and the creep constitutive equations used are presented in detail in this paper. The corresponding models were used to interpret some of the OLHF program experiments and the calculation results were quite consistent with the experimental data. The two simplified models have been used to simulate the thermo-mechanical behaviour of a 900MWe pressurized water reactor lower head under severe accident conditions leading to failure. The average transient heat flux produced by the corium relocated at the bottom of the lower head has been determined using the IRSN HARAR code. Two different methods, both taking into account the ablation of the internal surface, are used to determine the temperature profiles across the lower head wall and their effect on the time to failure is discussed. Using these simplified models, a graph representing the time to failure as a function of the pressure level and the heat flux intensity has been determined; such information will be used in our probabilistic safety assessment and severe accident management analyses. Another motivation for the development of simplified models in IRSN is to obtain a simplified but well-predicting code that can then be integrated into integral severe accident computer codes.

The subject of this paper is to compare the results of the different calculations performed by the benchmark participants in the framework of the OECD Lower Head Failure (OLHF) program. The benchmark consists in the finite element (FE) calculation or in analytical calculations of the mechanical behavior of the OLHF-1 experiment. Seven participants from six countries and seven companies or organizations (AVN, VTT, GRS, UJV, SNL, IPSN and CEA) have performed the benchmark.The OLHF experiment program extends the NRC-sponsored SNL LHF program (NUREG/CR-5582) completed in 1998: these experiments where intended to simulate the thermal/mechanical loads to a 1/4.85-scale model of a reactor pressure vessel. The pressure vessel material (SA533B 1 steel) used in these experiments is prototypic of reactor PWR vessel material and has been well characterized by material property testing as part of this program. The OLHF tests advance the results of the previous testing program by examining the effects of large temperature differences across the vessel wall. Large temperature differences in excess of 150-400 K are more prototypic of accident conditions.Most of the participants performed 2-D axisymmetric analyses and doesn't study the crack opening. The global mechanical behaviour of OLHF-1 experiment is well represented but prediction of the maximum vertical displacement is not in good agreement with the experimental value. Failure time and location are in quite good agreement with experimental results but large discrepancies are observed on the mode of failure: creep or plasticity. To improve predictions, more investigation and work is needed on the choice of the failure criteria and failure mode. (C) 2003 Elsevier Science B.V. All rights reserved.

Powder metallurgy techniques are often used in the fabrication of finished and semi-finished articles. At the CEA, a cold followed by a hot isostatic pressing method is used to produce solid forms from various types of powder. Occasionally, during the hot isostatic pressing part of the process, buckling of the titanium envelope near to the sealed end leads to fracture of the can and incomplete powder consolidation. The aim of this paper is to investigate by numerical finite element simulation the fracture process. A 2-D shell element model with Fourier series taking into account plastic deformation of the can material has been considered and used to determine the buckling critical pressure and the corresponding buckling mode. The simulation has been used to eliminate failure of the can by modifying the temperature-pressure schedule or by changing the can design. The calculations show that reducing the sharp angles of the initial titanium can near the rupture area can resolve the buckling problem; however, this solution is not totally satisfactory due to the development of a zone of constriction (breaks and irregularities) in the compacted powder just behind the modified can wall. This geometrical defect leads to difficulties in machining the final product. A better solution to the problem is to increase the initial can temperature prior to application of the pressure. This leads to a can of enhanced ductility with a better ability to deform. This latter solution, which can be employed with or without can modifications, eliminates both the can’s buckling and the zone of constriction. These numerical results have been validated by recent tests performed in our laboratories.

The local approach criterion of fracture mechanics, initially developed by Beremin for brittle cleavage fracture, is applied here to A508 class 3 low-alloy ferritic steel. This criterion, based on the maximum principal stress and Weibull statistics, has previously been verified in the case of uniaxial tests. In this study, it is extended to multiaxial loading tests, that can lead to more significant levels of plastic strain, and thus permit a study of the effect of plastic strain on cleavage fracture. Uniaxial tests on axisymmetric notched tensile bars (AE2-6) were used to determine Beremin’s model parameters m and σu. The cleavage fracture behavior, described by these parameters, was then verified by multiaxial tension-torsion tests carried out on thin tubular specimens. Numerical simulations of the tension-torsion tests, by the finite element method, were also performed, taking into account the nonlinear geometrical effects and the specimen plastic buckling. The buckling critical loads were calculated and used to ascertain whether fracture was associated with the instability phenomenon. Beremin’s model is shown to correctly describe experimental data which are not affected by buckling.

The high-temperature design codes are presently considering the use of stress reduction factors for designing welded structures submitted to creep. These reduction factors are derived from creep tests which are generally made on small specimens and are not necessarily representative of large-size geometries. These codes are very likely overconservative, consequently uneconomical and need to be improved; an investigation to assess and quantify the supposed size effect is required. This paper presents an experimental and numerical study on creep behavior at 600°C of full-size welded joints taking into account real full-thickness of weldings. The material investigated is the austenitic stainless steel 316L(N) with manual metal arc welds using the 19 Cr 12 Ni 2 Mo electrode grade. The creep laws used in calculations are those obtained from tests using small specimens, but some coefficients of their theoretical formulation have been modified to obtain a better coherence with fullsize specimen data. Between small and large full-size specimens, experimental results show no significant difference in time to rupture, and the same location of fracture, at the center of the weldment, is observed. Finite element simulations performed for full-size welded joints provide rupture times that are consistent with measured values. The calculated percentage of the damaged volume in the weld metal as a function of load levels and of creep-time duration is studied; it shows that the creep-rupture times for high stress loading are determined with higher accuracy than for low stress loading.

In order to validate the long-term behavior of fast breeder reactor steam generator tubes, a creep test monitoring program has been undertaken. A series of over one hundred tests was performed on small specimen tubes which were cut from the as-received tube material of Incoloy 800, similar to those used in generator applications. In contrast to uniaxial conditions in which constant load conditions usually apply, multiaxial experiments were performed assuming constant stress conditions. Thin walled tube approximation was used to estimate the stresses across the tube wall thickness required for given internal tube pressures. This paper deals with a refined numerical modeling of stresses and strains in the pressurized test specimens, taking into account: internal pressure changes, nonlinear geometrical effects, and creep deformation during all the test procedure cycles. Comparison of the numerical results with the experimental measurements shows a good agreement which validates the multiaxial generalization of the RCC-MR strain creep laws which were used in the calculations. A multiaxial creep rupture criterion is discussed.

In order to get a better understanding of the role of various parameters on the creep collapse phenomenon of long circular tubes, an experimental and numerical program has been undertaken. It mainly aims at studying the influence of the behavior of the material composing the circular tubes and their geometrical characteristics on the flattening time. This paper presents the results of numerical simulations performed with a specific computer program using an analytical approach. They have been compared with the data obtained from creep tests performed in various conditions on preovalized tubes as well as with the solutions obtained from the finite element method. It is shown that a good agreement is achieved between experimental results and calculation results.

Summary Within SARNET, the CORIUM topic covers all the behaviour of corium from earl y phase of core degradation to in or ex-vessel corium recovery with the exception of corium interaction with water, direct containment heating and fission produ ct release. The CORIUM topic regroups in three work packages the critical mass of competence to improve significantly the corium behaviour knowledge. The spirit of the SARNET networking is to share the knowledge, the facilities and the simulation tools for severe accidents, so to reach a better ef ficiency and to rationalise the R&D effort at European level. Extensive benchmarking has been launched in most of the areas of research. These benchmarks were mainly dedicated to the recalculation of experiments, while, in the next periods, a larger focus will be given to integral e xperiments or reactor applications. Eventually, all the knowledge will be accumulated in the ASTEC severe accident simulation code through physical model improvements and extension of validation database. This paper summarizes the progress that has been ac hieved in the frame of the networking activities and underlines also the orientation of t he CORIUM work packages for the next period. A special focus is placed on the melt pool and debris coolability and corium-concrete interaction, in which, the effects due to multidimensional geometri es and heterogeneities has been shown, during SARNET, to play a crucial role and further r esearch is still needed.

Summary Within the SARNET project, the severe accident code ASTEC is under validation against experiments and analytical results from other integral and/or mechanistic codes. Many applications concerning circuit thermal-hydraulics and core degradation have been performed on various integral and separate-effect experiments for the CESAR thermal-hydraulic module (experiments: BETHSY 9.1 b, PACTEL ISP 33 and T2.1, PMK2, LOFT LP-FP-2) and the DIVA core degradation module (experiments: CORA-13 and -W2, QUENCH-11 and -13, LOFT LP-FP-2, Phébus FPT-4, FARO L14 and L28, LIVE-L1, OLHF-1, FOREVER EC2) for validation purposes. Besides, the TMI-2 accident has been analyzed using the CESAR and DIVA modules for validation. The suitability and capability of new or improved models implemented in successive code versions up to ASTEC V1.3R2 delivered in December 2007 have been evaluated. The emphasis of the new or improved models in CESAR concern reflooding of an intact core, condensation in the pressurizer, sub-critical break flow correlation, and a new pressurizer spray model. Improvements of DIVA mainly concern the models of the corium behavior in the lower head and of lower head mechanical failure. The paper will present and discuss the progress in ASTEC validation with respect to circuit thermal-hydraulics and core degradation, mainly focussed on the use of the version ASTEC V1.3R2. Good results have been obtained with CESAR on the integral LOFT LP-FP- 2 test and on the two PACTEL experiments which cover various thermal-hydraulic flow regimes. These good results have been confirmed by the intensive validation work done on BETHSY integral tests. The results are good for early-phase DIVA models of core heatup, oxidation and hydrogen production (before any quenching phase) on different CORA, QUENCH and LOFT LP-FP-2 experiments. For the late-phase DIVA models, the results can be considered as good regarding debris bed melting (Phébus FPT4), corium fragmentation at slump in vessel lower plenum (FARO), molten pool behaviour in lower plenum (LIVE-L1), and vessel lower head mechanics (OLHF-1 and FOREVER EC2). Furthermore, the first two phases of the TMI-2 accident before core reflood are very well calculated by ASTEC. Major remaining weaknesses are found in the late phase degradation and the reflooding of a degraded core. Implementation of improved debris bed and magma models is in progress for the ASTEC V2 version, which will allow a more realistic simulation of late phase phenomena up to the failure of the lower head.