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[en] The treatment of zirconium oxidation kinetics in severe accident (SA) codes has been the subject of many discussions and controversies in recent years. The main problem was the existence of several correlations which could lead to large differences in the calculated results. It appeared clearly that there was a need to converge towards a common understanding of the physical processes that must be modeled (oxygen diffusion, blanketing effect, etc.) and an agreed database among code developers and users. It would help reducing an important source of uncertainties in SA calculations. The kinetic correlation database, obtained as a result of examination of complementary experimental data in Parts I and II, is applied here to analyze a few high-temperature separate-effects tests and bundle experiments where Zry oxidation reaction played a dominant role. The ICARE/CATHARE computer code developed by IRSN is used to check the validity of the high-temperature correlations derived in Parts I and II. The physical modeling provided by the code includes detailed account of specific features of chemical interactions between fuel rod cladding and steam. In particular, high reaction rates at T > 2000 K are moderated by two effects, examined in Part I: steam blanketing during thin oxide layers growth and transition to oxidation of α-Zr(O) phase after total consumption of primary β-Zr in cladding metallic part. When applied to separate-effects tests, the evaluated parabolic correlations have shown their applicability to different types of temperature transients taking into account Zry oxidation specifics in rod geometry. The bundle integral experiments QUENCH-06 and PHEBUS B9+ did not lead to extremely large temperature excursions. Calculated temperatures, hydrogen production and oxide thickness, as well as parameters of melt relocation were found to agree well with experimentally measured values. As a result of this study, we believe that the new best-fitted correlations, obtained in agreement with available experimental data, can be used in further studies and can improve predictive power of the codes. The continuation of the current work will be the application of ICARE/CATHARE code with best-fitted Zry oxidation correlations to NPP accident scenarios
[en] Highlights: • Five reflooding tests have been carried out with an experimental bed, 500 mm in height and 500 mm in diameter, made of 4 mm stainless steel balls. • For the first time, such a large bed was heated practically homogenously. • The quench front velocity was determined according to thermocouple measurements inside the bed. • An analytical model, assuming a quasi-steady progression of the quench front, allows to predict the conversion ratio in most cases. • It appears that the efficiency of cooling can be increased only up to a certain limit when increasing the inlet water flow rate. - Abstract: During a severe accident in a nuclear power plant, the degradation of fuel rods and melting of materials lead to the accumulation of core materials, which are commonly, called “debris beds”. To stop core degradation and avoid the reactor vessel rupture, the main accident management procedure consists in injecting water. In the case of debris bed, the reflooding models used for Loss of Coolant Accident are not applicable. The IRSN has launched an experimental program on debris bed reflooding to develop new models and to validate severe accident codes. The PEARL facility has been designed to perform, for the first time, the reflooding of large scale debris bed (Ø540 mm, h = 500 mm and 500 kg of steel debris) in a pressurized containment. The bed is heated by means of an induction system. A specific instrumentation has been developed to measure the debris bed temperature, pressure drop inside the bed and the steam flow rate during the reflooding. In this paper, the results of the first integral reflooding tests performed in the PEARL facility at atmospheric pressure up to 700 °C are presented. Focus is made on the quench front propagation and on the steam flow rate during reflooding. The effect of water injection flow rate, debris initial temperature and residual power are also discussed. Finally, an analytical model providing the steam flow rate and the quench front velocity is proposed to interpret these results.
[en] The paper is devoted to the presentation of validation cases carried out for the models describing the corium behaviour in the “lower plenum” of the reactor vessel implemented in the V2.0 version of the ASTEC integral code, jointly developed by IRSN (France) and GRS (Germany). In the ASTEC architecture, these models are grouped within the single ICARE module and they are all activated in typical accident scenarios. Therefore, it is important to check the validity of each individual model, as long as experiments are available for which a single physical process is involved. The results of ASTEC applications against the following experiments are presented: FARO (corium jet fragmentation), LIVE (heat transfer between a molten pool and the vessel), MASCA (separation and stratification of corium non miscible phases) and OLHF (mechanical failure of the vessel). Compared to the previous ASTEC V1.3 version, the validation matrix is extended. This work allows determining recommended values for some model parameters (e.g. debris particle size in the fragmentation model and criterion for debris bed liquefaction). Almost all the processes governing the corium behaviour, its thermal interaction with the vessel wall and the vessel failure are modelled in ASTEC and these models have been assessed individually with satisfactory results. The main uncertainties appear to be related to the calculation of transient evolutions
[en] During the degradation of a nuclear core in a severe accident scenario, the high temperatures reached lead to the melting of materials. The formation of liquid mixtures at various elevations is followed by the flow of molten materials through the core. Liquid mixture may flow under several configurations: axial relocation along the rods, horizontal motion over a plane surface such as the core support plate or a blockage of material, 2D relocation through a debris bed, etc.. The two-dimensional relocation of molten material through a porous debris bed, implemented for the simulation of late degradation phases, has opened a new way to the elaboration of the relocation model for the flow of liquid mixture along the rods. It is based on a volume averaging method, where wall friction and capillary effects are taken into account by introducing effective coefficients to characterize the solid matrix (rods, grids, debris, etc.). A local description of the liquid flow is necessary to derive the effective coefficients. Heat transfers are modelled in a similar way. The derivation of the conservation equations for the liquid mixture falling flow (momentum) in two directions (axial and radial-horizontal) and for the heat exchanges (energy) are the main points of this new model for simulating melt progression. In this presentation, the full model for the relocation and solidification of liquid materials through a rod bundle or a debris bed is described. It is implemented in the ICARE/CATHARE code, developed by IPSN in Cadarache. The main improvements and advantages of the new model are: A single formulation for liquid mixture relocation, in 2D, either through a rod bundle or a porous debris bed, Extensions to complex structures (grids, by-pass, etc..), The modeling of relocation of a liquid mixture over plane surfaces. (author)
[en] In 2009, EDF started its project to extend the operating life of its Gen II PWRs beyond 40 years. It implied : •a specific program for ageing management, • a safety reassessment in light of the requirements applicable to new reactors (EPR) and the state of the art of nuclear technologies → Prevention of basemat melt-through in case of a severe accident was one issue considered in that framework. Then, post-Fukushima actions were launched in France and the importance of that issue was confirmed.
[en] In case of the loss of coolant accident (LOCA) the prime accident management procedure consists in injecting water in the reactor core, by means of various safety injection devices. Nevertheless, the success of a core reflood is not guaranteed because of possible negative effects. The success of a core reflood depends mainly on the core state (geometry and temperature), the primary pressure and the injection rate. The description of the reflooding phase, for any core configuration, is one of the objectives of ICARE/CATHARE V2 code. A new reflooding model has been implemented in the ICARE-CATHARE V2 code, improving significantly the tracking of quench front and the accuracy of heat transfer calculation in the transition boiling zone. This new model has been qualified on QUENCH 11 test showing that the model is also valid for very high temperature of claddings and when rods are slightly damaged. The new model changes the thermalhydraulic behaviour and results in a better prediction of hydrogen production. It confirms that the understanding of thermal-hydraulic phenomena is a key point in the safety management, as important as the kinetic of oxidation. At high temperature, kinetic rate is so fast that the reaction is limited by steam availability. The impact of the injection on the hydrogen production has been studied. The conclusion is that injecting water during a severe accident is not an obvious procedure. If the flow rate is too low, the hydrogen production may increase significantly
[en] The Institut de Protection et de Surete Nucleaire is carrying out a severe-accident program involving both experiments (PHEBUS FP, in particular) and computer code development (ICARE2 and ICARE/CATHARE as far as the primary circuit phenomenology is concerned). The goal is to gain knowledge on core degradation and fission product behavior for safety analysis. This process will build a validated primary circuit code to predict the consequences of a severe accident in a pressurized water reactor (PWR) up to a possible lower head failure (ICARE/CATHARE), thus providing the boundary conditions for ex-vessel corium behavior, in particular for spreading in the new-generation reactor core catcher
[en] The accurate description of the geometry of a severely damaged core and of debris beds expected to form during quenching as many physical models strongly depend on geometrical parameters. This relies on a good description of degradation phenomena and of debris bed formation. This paper, based on a bibliographic survey, is focussed on the geometrical characteristics of debris beds expected to form during reflooding. The consequences of a severe accident on the geometry of the reactor core will be examined in paragraph 2. The observations show that the top of the damage regions consist of debris particles of a wide range of sizes. To clarify this particle size distribution, data on fuel fragmentation under operating and LOCA conditions have been gathered and will be presented in paragraph 3. Then, in paragraph 4, we have looked for cladding failure criteria so as to determine whether the rapid cooling may lead to the collapse of the embrittled oxidized claddings. Using these data has finally enabled us to determine in paragraph 5 the extent and the composition of the debris bed formed during the reflooding of a French 900 MW PWR during a six inches LOCA
[en] To demonstrate the ability of ICARE/CATHARE V2 to simulate a severe accident in a PWR and notably to describe multi-dimensional effects occurring in the core during the LOCA and degradation phases, a severe accident in a simplified reactor has been studied and TMI-2 transient calculation has been initiated. (authors)
[en] Among the main uncertainties which prevent making reliable estimates of safety margins of in-vessel melt retention, the effects of material interactions is one of the most important. The corium which would initially relocate down to the lower plenum consists mainly of UO2 and Zr which is partially oxidized. But the presence of steel structures in the lower head and the melting of upper structures by radiative heat transfer from the molten pool implies that a significant amount of molten steel would be also present. Liquid steel and liquid UO2 are immiscible. The objective of this paper is to propose a model that is able to describe the transient mass transfers between layers due to the changes of density of metal and oxide phases. The novelty of the model is to take into account local conditions at the interface and not just a global equilibrium. In a first part, a simple thermochemical model is presented and the consequences on the mass transfers are explained. In a second part, a model for mass transfers between the three possible layers (heavy metal layer, oxide pool layer, and steel layer) is proposed. In the last part, typical evolutions of the thickness of the three layers are shown. Discussions about the respective effects of material properties and scenario parameters lead to identify some key-parameters which remain to be evaluated for an accurate description of the inversion process