Quench front progression in a superheated porous medium: experimental analysis and model development

In case of severe accident in a nuclear reactor, the fuel rods may be highly damaged and oxidized and finally collapse to form a debris bed. Removal of decay heat from a debris bed is a challenging issue because of the difficulty for water to flow inside. Currently, IRSN has started experimental program PEARL with two experimental facilities PRELUDE and PEARL, to investigate the reflood process at high temperature, for various particle sizes. On the basis of PRELUDE experimental results, the thermal hydraulic features of the quench front have been analysed and the intensity of heat transfers was estimated. From a selection of experimental results, a reflooding model was improved and validated. The model is implemented in the code ICARE-CATHARE developed by IRSN which is used for severe accident reactor analysis

Quench front progression in a superheated porous medium: experimental analysis and model development

In case of accident at a nuclear power plant, water sources may not be available for a long period of time and the core heats up due to the residual power. Any attempt to inject water during core degradation can lead to quenching and further fragmentation of core material. The fragmentation of fuel rods and melting of reactor core materials may result in the formation of a "debris bed". The typical particle size in a debris bed might reach few millimeters (characteristic length-scale: 1-5 mm). The two-phase flow model for reflood of the degraded core is briefly introduced in this paper. It is implemented into the ICARE-CATHARE code, developed by IRSN (Institut de radioprotection et de sûreté nucléaire), to study severe accident scenarios in pressurized water reactors. Currently, the French IRSN sets up two experimental facilities to study debris bed reflooding, PEARL and PRELUDE, and validate safety models. The PRELUDE program studies the complex two phase flow (water/steam), in a porous medium (diameter 180 mm, height 200 mm), initially heated to a high temperature (400 °C or 700 °C). On the basis of the experimental results, thermal hydraulic features at the quench front have been analyzed. The two-phase flow model shows a good agreement with PRELUDE experimental results

Final Interpretation Report of the PHEBUS test FPT0: Bundle Aspects

In this paper, the actual status of understanding of the dominant bundle degradation processes is presented. Here, mainly the results reported in the last years in the Bundle Interpretation Circles organised by JRC/IE and IRSN (Institut de Radioprotection et de Surete Nucleaire, Cadarache) are summarised. For the extensive and detailed computational analyses the commonly used severe accident codes such as ICARE, MELCOR, SCDAP/RELAP and ATHLET-CD are used. For the analysis of fission product release from the FPT0 bundle, specific codes such as SVECHA and XMPR were used as well.JRC.F.4-Nuclear design safet

Long-Wave Infrared Digital Holography

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VIKTORIA experiments on sump filtration during a Loss Of Coolant Accident (DBA)

International audienceDuring a Loss Of Coolant Accident, in French PWR's, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the Refueling Water Storage Tank (RWST), water is taken from the sump in the lower part of the reactor building. A filtering system is implemented to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the ECCS and CSS systems. IRSN has launched an experimental R&D project investigating the clogging of the sump filter by integral tests performed in the VIKTORIA loop. The main objectives are to investigate the head loss of the filter (physical and chemical clogging) for prototypic upstream debris source term in relevant thermal hydraulic conditions (water temperature and flow velocity on the filter surface) in compliance with the temperature profile and the chemistry of the water in the sump during a LOCA transient. For that, the VIKTORIA loop was equipped successively with two types of 2 m 2 filters use in French 900MWe NPP's. The tests highlight the settling of the largest particles (concrete, painted chips) and part of the fibers; the transport of debris (roughly 55 to 65 % of the injected debris source term) leads to the physical clogging of the filter. The debris carried to the filter generate at 80°C (with chemistry) a very quick increase of the pressure drop across the filter (≈ 7 kPa) that could be due to rapid chemical effects further to fibers corrosion. At the end of the 80°C plateau, a pressure drop increase was observed due to the temperature decrease in agreement with the water viscosity evolution. The two types of filters (rectangular cartridges or planar types) behave very differently with rather low head losses for the second type; ≈ 1,5 kPa. Nevertheless, downstream debris source term appears to be more significant with the planar filter (compared to the filter with cartridges) during the first hour of injection before the fibrous debris bed formation. We have also observed more stabilized "cake", during the (7 days) medium term evolution compared to that for rectangular cartridges filter due to motion of debris inside cartridges

COAL experiments investigating the reflooding of a 7 X 7 rod bundle during a Loss Of coolant Accident: Effect of a partially blocked area with ballooned rods

International audienceDuring a loss of the coolant of the primary circuit (LOCA) in a pressurized water reactor, the drying of the fuel assemblies may lead to the fuel temperature increase and deformation of the fuel rod claddings. In addition to the restriction of the flow area, the fragmented irradiated fuel relocation within the ballooned zone leads to an increase of the local residual power. The COAL experiments will focus on the coolability issue of a partially deformed fuel assembly during water injection with the safety systems using a 7x7 bundle of electrically heated rods. These experiments are part of the PERFROI project launched by IRSN with the support of the french "Agence Nationale pour la Recherche" (ANR), EDF and the US-NRC. The effect of the flow blockage {intact geometry up to long ballooning (100 to 300 mm) with different blockage ratios (80 to 90%)} will be evaluated for various power, mass flow rates and different pressures representative to Large (LBLOCA at 0.3 MPa) and Medium break size (MBLOCA : 0.5 to 3 MPa) configuration. Relocation of fragmented fuel in the balloons are taken into account by a local increase of the power by a factor of 1.5. This paper presents the thermal hydraulics parameters and the main results of the experiments performed in a facility of the STERN Laboratories (Canada). The presence of the balloons increases significantly the Peak Cladding Temperature. These results will be used to improve and validate the heat exchange models of thermal hydraulics codes dealing with the complex reflooding processes in such configuration

Coal experiments investigating the reflooding of a 7 x 7 rods bundle during a loss of coolant accident thermalhydraulics results

International audienceDuring an accident causing loss of the coolant of the primary circuit (LOCA) in a pressurized water reactor, partial or complete drying of the fuel assemblies may lead to the fuel temperature increase and significant deformation of the fuel rod claddings. In addition to the restriction of the flow area, the fragmented irradiated fuel relocation within the ballooned zone leads to an increase of the local residual power. The COAL experiments are part of the PERFROI project launched by "Institut de Radioprotection et de Sureté Nucléaire" with the support of the French National Agency of Research (ANR), EDF and the US-NRC. In this framework, IRSN has designed and developed those specific experiments focusing on the coolability issue of a partially deformed fuel assembly during the cooling phase by water injection with the safety systems using a 7x7 bundle of electrically heated rods. The effect of the flow blockage (intact geometry up to long ballooning with different blockage ratios 80 to 90%) will be evaluated for various mass flow rates and different pressures representative to Large (LBLOCA) and Medium break size (MBLOCA) configuration. This paper presents the thermal hydraulics parameters and the preliminary results of the experiments

VIKTORIA experiments investigating the filtering system in the sump of a pwr after a Loss Of Coolant Accident : Part II Downstream effects

International audienceDuring a Loss Of Coolant Accident, water is injected by the Emergency Core Cooling System (ECCS) to remove decay heat from the core and by the Containment Spray System (CSS) to maintain containment integrity which are needed to ensure the long term management of the accident. After the drainage of the RWST (Refueling Water Storage Tank), water is taken from the containment sump in the lower part of the nuclear reactor building. A filtering system is implemented at the bottom of the containment to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint particles, and to minimize the amount of debris entering in the ECCS and CSS systems. Nevertheless, a part of the debris may pass through the strainer and be transported to the core, which might affect the fuel assembly coolability (downstream effects). The aim of the tests performed in the VIKTORIA facility is to investigate the downstream effects on a fuel assembly mock-up including a representative lower support plate, a bottom nozzle with its anti-debris grid, one spacer grid and one mixing grid. The analyses of the VIKTORIA experiments, presented in this paper, give very useful results regarding the impact of the "Downstream Debris Source Term", the thermal hydraulic conditions (flow rate, temperature), the chemical products in the water and the type of fibers composing the debris source term, on the debris collected on the grids and on the resulting head losses

VIKTORIA EXPERIMENTS INVESTIGATING THE FILTERING SYSTEM IN THE SUMP OF A PWR AFTER A LOSS OF COOLANT ACCIDENT : Part I Physical /chemical effects on strainer head loss evolution

International audienceDuring a Loss Of Coolant Accident, water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the RWST (Refueling Water Storage Tank), water is taken from the containment sump in the lower part of the nuclear reactor building. A filtering system is implemented at the bottom of the containment to collect debris produced by the pipe break as well as other latent materials, such as fiberglass, paint and concrete particles, and to minimize the amount of debris entering in the ECCS and CSS systems. Consequently, one of the major issues to be assessed is the plugging of the filtering system due to physical and chemical conditions which can lead to an inadequate net positive suction head (NPSH) margin for the ECCS and CSS pumps and can affect the mechanical integrity of the strainers. The "Institut de Radioprotection et de Sûreté Nucléaire" has launched an experimental R&D project investigating the possible plugging of the sump filter by integral tests performed in the VIKTORIA loop. The analyses of the experiments give very useful results regarding sedimentation, transport of debris and physical plugging, as well as the impact of chemical effects on strainer head loss evolution

Operation Extension of 900MWe NPPs: French TSO Main Conclusions regarding Long Term Sump Performance after a Loss of Coolant Accident

International audienceDuring a Loss Of Coolant Accident (LOCA), water is injected by the Emergency Core Cooling System (ECCS) to ensure the long-term core coolability and by the Containment Spray System (CSS) to remove residual heat and to maintain containment integrity. After the drainage of the RWST (Refueling Water Storage Tank), water is taken from the containment sump in the lower part of the nuclear reactor building. A filtering system is implemented at the bottom of the containment to collect debris produced by the pipe break and the degradation of the ambient conditions inside the reactor building, and to minimize the amount of debris entering in the ECCS and CSS systems. Consequently, one of the major issues is the clogging of the filtering system due to physical and chemical conditions which can lead to an inadequate net positive suction head (NPSH) margin for the ECCS and CSS pumps and can affect the mechanical integrity of the strainers. Furthermore, despite the filtering system, a part of the debris bypassing the strainers is transported to the core. That might alter the fuel assembly coolability (downstream effects), which constitutes the second major issue. Since the Baresbäck incident in Sweden in 1992, which raised questions concerning the risk of strainers clogging at international level, sump performance has become a major concern in France, where this issue is common to all NPPs. In the framework of the operation extension of the French 900MWe NPPs beyond 40 operating years, EDF presented its safety demonstration based on studies and new experimental program on strainer clogging and fuel assembly coolability. IRSN conducted the evaluation of the EDF safety case and performed simultaneously its own tests on the VIKTORIA loop in Slovakia. IRSN provided its technical position in March 2020. The paper will present the French background and French Technical Safety Organization (IRSN) conclusions on sumps performance issue for the 4th periodic safety review of 900MWe NPPs dated the end of March, 2020