Institut De Radioprotection Et De Surete Nucleaire Research Articles

Validation of the extended COCOSYS cable fire model based on a generic FLASH-CAT approach

Fire simulations and analytical validation approaches have gained more and more relevance, in particular with respect to fire safety assessment of operating nuclear power plants (NPPs). For providing an added value to the regulatory authorities and the technical safety experts authorized by them suitability and applicability of the analytical tools applied such as fire simulation codes need be known and demonstrated.The COCOSYS (Containment Code System) code developed by Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) gGmbH contains i.e. models for simulating cable as well as liquid fuel fires. In NPPs, cables represent a non-negligible fire load as initial fire source or contributor to fire spreading.To overcome some generic deficiencies of the existing so-called simple cable fire model in COCOSYS the model has been improved based on a more generic FLASH-CAT approach. The extended model uses a time dependent mass loss rate (MLR) density profile, can consider composite materials to regard the increase of the heat of combustion during the fire development and is able to consider under-ventilated conditions.The model was validated on several experiments performed in open atmosphere and confined environment by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and conducted within the PRISME 2 and PRISME 3 Projects, launched by the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA). One main objective is to provide one single set of input parameters for each cable type, which can be used under different ventilation conditions. The paper presents capabilities and remaining limitations of the new model.

Recently revisited TH issues and associated R&D A French perspective

This paper presents an overview of some recent evolutions in the field of thermal-hydraulics (TH) within nuclear safety studies in France, therefore restricted to Pressurized Water Reactors (PWR) design, from the perspective of the French Technical Support Organization IRSN (Institut de Radioprotection et de Sûreté Nucléaire). These evolutions can concern either the hypothetical accident scenario (recently raised questions, modified rulemaking, new methodology) or the development of recent sources of knowledge (experimental programs, advanced numerical simulation tools).

Thermal scattering law for ice based on neutron time-of-flight experiments carried out at the SEQUOIA spectrometer at the Oak Ridge National Laboratory

Precise estimation of neutron thermalization in moderators relies on high-fidelity thermal scattering cross-section data governed by the thermal scattering law (TSL). The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) has been working on the development of improved TSL for light water ice. Many polymorphic phases of light water ice exist depending on the thermodynamic conditions. The most common type of ice at standard pressure and temperature below water freezing point (273.15 K) is ice-Ih. It is essential to have high-resolution experimental double differential data for developing and/or validating TSL for moderator materials.Existing experimental double-differential scattering data for ice-Ih are extremely sparse and of limited quality. New high-quality double-differential measurements for ice-Ih over multiple temperatures and incident neutron energies would directly support the validation and improvement of ice-Ih TSL models for criticality safety applications. Series of time-of-flight (TOF) inelastic neutron scattering experiments on ice-Ih at temperatures starting at 271 K and down to 6 K, have been carried out at the SEQUOIA spectrometer at the Spallation Neutron Source (SNS) at the Oak Ridge National Laboratory (ORNL), United States. The experiments have been performed for incident neutron energies, Ei= 11, 55, 160, 250, and 600 meV, to explore different excitation energies in the vibrational phonon spectrum. This paper presents the thermodynamic conditions and the details of the TOF measurements on ice-Ih and the derived phonon spectrum from the experimentally measured double differential data. A study of the variation of the phonon spectrum of ice-Ih as a function of temperature is highlighted, and a preliminary TSL evaluation is developed based on the experimental phonon spectrum.

A 3D model to solve U-tube steam generator secondary side thermal-hydraulics with coupled primary-to-secondary side heat transfer

The U-tubes of steam generators (SG) are concerned by vibration issues. The vibration behavior is driven by the local secondary flow (velocity, void fraction, two-phase pattern). Fluid-elastic instability (FEI) can be considered as the most severe vibration phenomenon, since it features high vibration amplitudes and tube-to-tube contacts, which can rapidly cause tube rupture. In order to predict if the fluid forces exerted along the tube bundle can trigger a FEI, numerical tools able to calculate the secondary side of a SG are needed. In the frame of its mission of assessing methods adopted by utilities for demonstrating the safe operation of nuclear facilities, the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) has developed such a numerical model. The model is based on a 3D homogeneous drift-flux formulation of the SG steam-water mixture, together with a porous media approach to account for the tube bundle. The model is implemented in the ANSYS Fluent CFD code. The SG primary side is solved using a dedicated meshing. The secondary side and the primary side meshes communicate through heat transfer, which is fundamental to obtain consistent numerical results. The details of the numerical approach are described in the first part of this paper. Then, numerical versus experimental results comparisons in terms of heat transfer are provided for a SG test facility: numerical and experimental results are in good agreement. The model was then applied to a real SG: numerical results are in good agreement with nominal operating parameters.

CIEMAT WBC capabilities for responding in case of nuclear and radiological emergencies

Nuclear and radiological accidents can result in internal exposures of workers (including first responders) and general population. For this reason it is necessary precise and rapid methods are developed in order to ensure capabilities for monitoring and dose assessments of internally contaminated persons. The Whole Body Counter (WBC) of the Center for Energy, Environmental and Technological Research (CIEMAT) in Madrid, Spain, has implemented techniques to determine radioiodine in the thyroid and gamma emitters throughout the body of various age groups of children and adults. In addition, the WBC laboratory has developed a new in vivo calibration and measurement method with germanium detectors in case of contamination of gamma emitting radionuclides in wounds at different depths.Two detection systems have been calibrated accordingly: a Fastscan counter, consisting of two NaI(Tl) detectors, and a low-energy germanium (LEGe) detection system (four detectors), which is highly efficient at low gamma energies.A family of Lucite thyroid neck phantoms for children and adults was designed and manufactured at CIEMAT for thyroid calibration purposes. A set of whole-body phantoms of the BOMAB type was fabricated with polyethylene, and filled with distilled water in an acid medium and with a known gamma emitter mixture for in vivo monitoring calibration. In addition simulators of contaminated wounds with circular pieces of Ethyl Vinyl Acetate (EVA) were provided by Institut de Radioprotection et de Sûreté Nucléaire (IRSN) from France.Efficiency curves depending on energy are obtained in the different counting geometries, which facilitate calculating the activities of the gamma emitting radionuclides retained in the organism when monitoring real contaminated persons. In case of intake of radionuclides retained in the thyroid or in the whole body, a curve is obtained for each age specific phantom. Moreover in case of contamination in wounds a set of efficiency curves was developed depending on the contamination depth.The detection limits (DL) of the in vivo measurement techniques for radionuclides have been calculated from the measurements of blank phantoms or persons according to ISO 28218 standard obtaining results for each method and age. Using the limits of detection for the estimation of the minimum committed effective dose, the values obtained were lower than 1 mSv.y−1, which is below the limit for the public recommended by ICRP.

Reactivity Initiated Accident transient testing on irradiated fuel rods in PWR conditions: The CABRI International Program

The CABRI International Program (CIP) tests irradiated UO2 or MOX fuels submitted to Reactivity Initiated Accidents (RIA) representative power pulses in prototypical PWR thermal-hydraulic conditions. CIP is managed by The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) within a OECD/NEA framework. Experiments are conducted in the CABRI reactor operated by CEA.For CIP, CABRI benefits from a new pressurized water loop. An important refurbishment program enhanced the facility safety and upgraded the experimental equipment such as the non-destructive examination bench IRIS and the Hodoscope on-line fuel motion monitoring system. Specific test devices with appropriate innovative instrumentation follow the test rod behavior during the transient.The successful first test in the pressurized water loop demonstrated the CABRI capability to perform fully instrumented RIA tests in PWR conditions and to provide highly valuable results for RIA phenomena modelling (boiling crisis, post-failure events), for code validation, and for assessing PWR safety criteria.

Unresolved resonance range processing and probability tables generation in the GAIA2 system

The objective of this paper is to present the work carried out at the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to process nuclear data in the unresolved resonance range (URR). Recently, a great deal of effort has been devoted at IRSN to develop an independent nuclear data processing system, GAIA2. First, a nuclear data storing object, independent from the ENDF-6 format, has been implemented in order to transmit information between the components of a module-based scheme. Then, the generation of probability tables in the URR has been added as an independent module named TOP (as Table Of Probability), and tested on a selected set of benchmarks. The methods used and the results are discussed, and some limitations in the manner to construct the tables are pointed out.

Analysis of the time-of-flight neutron scattering cross-section data for light water measured at the SEQUOIA spectrometer, Spallation Neutron Source (SNS)

Thermal neutron scattering cross-section data for light water available in the major nuclear data libraries observes significant differences especially at reactor operating temperatures. During the past few years there has been a renewed interest in reviewing the existing thermal scattering models and generating more accurate and reliable thermal scattering cross sections using existing experimental data and in some cases based on Molecular Dynamics (MD) simulations. There is a need for performing new time-of-flight experiments at high temperatures and pressures, to have a better understanding of the physics involved in the scattering process that could help improve the existing TSL data.Lack of experimental thermal scattering data for light water at high temperatures led to a new measurement campaign within the INSIDER project at the Institut de radioprotection et de sûreté nucléaire (IRSN). Double differential scattering cross section for light water have been measured at the SEQUOIA spectrometer based at the Spallation Neutron Source (SNS), Oak Ridge National Laboratory, United States. Several measurements have been carried out at different temperatures and pressures corresponding to liquid light water. Measurements at five different incident neutron energiesEi(8, 60, 160, 280 and 800 meV) have been carried out to help exploring different regions of the frequency spectrum. This paper presents the analysis of the dynamic structure factor and the derived frequency spectrum of light water. The analysis of the experimental data would provide one with better confidence, the behavior of thermal scattering cross sections for light water at high temperatures, knowledge of which is very important for the design of novel reactors as well as existing pressurized water reactors.

Assessment of Radio-Induced Damage in Endothelial Cells Irradiated with 40 kVp, 220 kVp, and 4 MV X-rays by Means of Micro and Nanodosimetric Calculations

The objective of this work was to study the differences in terms of early biological effects that might exist between different X-rays energies by using a mechanistic approach. To this end, radiobiological experiments exposing cell monolayers to three X-ray energies were performed in order to assess the yields of early DNA damage, in particular of double-strand breaks (DSBs). The simulation of these irradiations was set in order to understand the differences in the obtained experimental results. Hence, simulated results in terms of microdosimetric spectra and early DSB induction were analyzed and compared to the experimental data. Human umbilical vein endothelial cells (HUVECs) were irradiated with 40, 220 kVp, and 4 MV X-rays. The Geant4 Monte Carlo simulation toolkit and its extension Geant4-DNA were used for the simulations. Microdosimetric calculations aiming to determine possible differences in the variability of the energy absorbed by the irradiated cell population for those photon spectra were performed on 10,000 endothelial cell nuclei representing a cell monolayer. Nanodosimetric simulations were also carried out using a computation chain that allowed the simulation of physical, physico-chemical, and chemical stages on a single realistic endothelial cell nucleus model including both heterochromatin and euchromatin. DNA damage was scored in terms of yields of prompt DSBs per Gray (Gy) and per giga (109) base pair (Gbp) and DSB complexity was derived in order to be compared to experimental data expressed as numbers of histone variant H2AX (γ-H2AX) foci per cell. The calculated microdosimetric spread in the irradiated cell population was similar when comparing between 40 and 220 kVp X-rays and higher when comparing with 4 MV X-rays. Simulated yields of induced DSB/Gy/Gbp were found to be equivalent to those for 40 and 220 kVp but larger than those for 4 MV, resulting in a relative biological effectiveness (RBE) of 1.3. Additionally, DSB complexity was similar between the considered photon spectra. Simulated results were in good agreement with experimental data obtained by IRSN (Institut de radioprotection et de sûreté nucléaire) radiobiologists. Despite differences in photon energy, few differences were observed when comparing between 40 and 220 kVp X-rays in microdosimetric and nanodosimetric calculations. Nevertheless, variations were observed when comparing between 40/220 kVp and 4 MV X-rays. Thanks to the simulation results, these variations were able to be explained by the differences in the production of secondary electrons with energies below 10 keV.

Atmospheric modeling and source reconstruction of radioactive ruthenium from an undeclared major release in 2017

In October 2017 unusual 106Ru detections across most of Europe prompted the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) to analyze the event in order to locate the origin and identify the magnitude of the release. This paper presents the inverse modeling techniques used during the event to achieve this goal. The method is based on a variational approach and consists of using air concentration measurements with the ldX long-range dispersion model included in the IRSN's C3X operational platform. The method made it possible to quickly identify the southern Urals as the most likely geographical origin of the release. Despite uncertainties regarding the starting date of the release, calculations show that it potentially began on 23 September, while most of the release was emitted on 26 September. Among the nuclear plants identified in the southern Urals, the Mayak complex is that from which the dispersion of the 106Ru plume is most consistent with observations. The reconstructed 106Ru source term from Mayak is ∼250 TBq. In total, it was found that for 72% of the measurements simulated and observed air concentration agreed within a factor of 5. In addition, the simulated deposition of 106Ru agrees with the observed deposition. Outside the southern Urals, the simulations indicate that areas with highest deposition values are located in southern Scandinavia and southeastern Bulgaria and are explained by rainfall events occurring while the plume was passing over.

Nuclear data and applications at the nuclear safety and radioprotection institute: Analysis, evaluation and application

The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) is the French public service expert in nuclear and radiation risks with activities that cover all the related scientific and technical issues. IRSN provides expertise in support of radiation protection and nuclear safety risks. A great deal of research work carried out at IRSN concerns the improvement of computational tools and data used on a daily basis by nuclear safety assessment experts. Particularly in this paper the approach used to improve data quality and its uncertainties is presented. Computation tools and nuclear data evaluation methodology are discussed.

Experimental and numerical study of water spray system for a fire event in a confined and mechanically ventilated compartment

SummaryThe present work addresses the application of a water spray system in case of a fire event in large‐scale experiments for nuclear safety issues. It focuses on the interaction between a water spray system and a stratified smoke layer due to a pool fire in a mechanically ventilated enclosure. This study is supported by a set of four large‐scale tests and one numerical simulation with a 3D CFD software, named CALIF3S/ISIS, and developed by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN). The modelling used in this paper is based on an Eulerian‐Lagrangian approach. The fire tests are performed in a 165 −m3 mechanically ventilated single room. The fire is a lubricant oil pool fire of about 400 kW. The ventilation flow rate is 2550 m3.h−1 and corresponds to a renewal rate of 15.5 h−1. The spray nozzles are deluge and sprinkler type. The test parameters are the water flow rate, the time of activation, and the duration of activation. Based on the large‐scale experiments and the numerical simulation, four typical physical mechanisms have been enlightened. The first one corresponds to the cooling of the gas phase that is the straightforward consequence of the heat transfer exchange between the water droplets and the surrounding gas. The second effect is the process of gas mixing and homogenization induced by the water spraying system. The gas concentrations (O2, CO2) in the upper and lower parts of the room tend to the same level. The third effect is the significant increase of the fire heat release rate (HRR), up to 25 %, when the water spray is activated. Then, the last noteworthy effect is the occurrence of gas pressure peaks when the water spray is activated or shut off, consequence of the sudden change of the gas temperature. The processes of gas cooling and fire HRR increase are showed to be the main causes of these variations of gas pressure.

The Ruthenium-106 plume over Europe in 2017: a source-receptor model to estimate the source region

Between the end of September and the beginning of October 2017, low but detectable air concentrations of Ruthenium-106 (R106u) were measured throughout the European atmosphere. To date, the source has not been clearly identified. This study aims at constraining the location of the most likely emission source, by combining HYSPLIT back trajectories and a multi-site source-receptor model. The model follows a concentration weighted trajectory (CWT) approach, based on a R106u measurement dataset gathered by the International Atomic Energy Agency (IAEA) shortly after the event and shared by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN). The back trajectories are computed using Global Data Assimilation System (GDAS) and ERA-Interim meteorological data, with horizontal resolution from 1.5° to 0.5°. The effect of the starting altitude was also investigated. The source-receptor model results show that the most likely emission source region extends from Ukraine to the West to the Russian Volga and Ural regions to the East, in accordance with two previous studies. The analyses also show that the identified source region is relatively insensitive to the type of meteorological data, their horizontal resolution and values of starting altitude. The model precision is found to increase with increasing number of receptors. The method is fully open-source, requires low computational resources and is therefore suited to estimate the source origin of radioactive pollutants on a real time basis, providing an additional tool to more sophisticated atmospheric models.

Experimental and numerical analysis of fire scenarios involving two mechanically ventilated compartments connected together with a horizontal vent

SummaryThis work deals with an experimental and numerical investigation of a fire scenario involving two rooms mechanically ventilated and connected together with a horizontal vent. The objective is to improve the understanding of the physical phenomena and to assess the capability of computational fluid dynamics (CFD) numerical simulations to predict flow field for such a fire scenario. The study is based on a set of large‐scale fire experiments performed in the framework of the OECD PRISME‐2 project in the DIVA multi‐room facility of the Institut de Radioprotection et de Sûreté Nucléaire (IRSN) and of numerical simulations performed with the ISIS CFD code. The fire scenario consists of two rooms, one above the other, mechanically ventilated and connected to each other with a horizontal vent of 1 m2. The fire is a heptane pool fire located in the lower room. The analysis focuses on the coupling between the burning rate, the flow at the vent, and the configuration of the mechanical ventilation. Several regimes of combustion are encountered from well‐ventilated steady fire to under‐ventilated unsteady and oscillatory fire. The results show that the burning rate is controlled by both the mechanical ventilation and the downward flow from the vent. The numerical simulations highlight the specific pattern of the oxygen concentration field induced by the downward flow at the vent.

Extending the reactivity initiated accident (RIA) fuel performance code SCANAIR for boiling water reactor (BWR) applications

In this paper, capabilities of the SCANAIR transient fuel performance code are evaluated and extended for boiling water reactor (BWR) fuel low temperature cladding failure predictions and high temperature thermal hydraulics modelling. The SCANAIR code, developed by the Institut de Radioprotection et de Sûreté Nucléaire (IRSN), is designed for modelling the behaviour of a single fuel rod during reactivity initiated accident (RIA) in a pressurized water reactor (PWR). In a previous study (Arffman et al., 2012), new BWR cladding material property correlations were developed and implemented into SCANAIR. Here, SCANAIR’s ability to predict BWR cladding failures due to pellet-cladding-mechanical interaction (PCMI) is evaluated by modelling the NSRR FK test series. SCANAIR is found to give correct predictions with reasonably good accuracy when applied to a larger dataset of several tests. As the standard thermal hydraulics model in SCANAIR is one-dimensional and able to model single phase coolant only, the simulation of a BWR RIA, the control rod drop accident, is not possible when the bulk boiling regime is reached. In this study, the code’s application field is broadened to consider bulk boiling in BWRs. In the chosen approach, SCANAIR is coupled with an external thermal hydraulics code. For that, VTT’s in-house general thermal hydraulics code GENFLO has been used. The first demonstration simulations show promising results.

Use of integral experiments for the assessment of a new235U IRSN-CEA evaluation

The Working Party on International Nuclear Data Evaluation Co-operation (WPEC) subgroup 29 (SG 29) was established to investigate an issue with the 235 U capture cross-section in the energy range from 0.1 to 2.25 keV, due to a possible overestimation of 10% or more. To improve the 235 U capture crosssection, a new 235 U evaluation has been proposed by the Institut de Radioprotection et de Surete Nucleaire (IRSN) and the CEA, mainly based on new time-of-flight 235 U capture cross-section measurements and recent fission cross-section measurements performed at the n_TOF facility from CERN. IRSN and CEA Cadarache were in charge of the thermal to 2.25 keV energy range, whereas the CEA DIF was responsible of the high energy region. Integral experiments showing a strong 235 U sensitivity are used to assess the new evaluation, using Monte-Carlo methods. The k eff calculations were performed with the 5.D.1 beta version of the MORET 5 code, using the JEFF-3.2 library and the new 235 U evaluation, as well as the JEFF-3.3T1 library in which the new 235 U has been included. The benchmark selection allowed highlighting a significant improvement on k eff due to the new 235 U evaluation. The results of this data testing are presented here.

Comparative modelling approaches of hydro-mechanical processes in sealing experiments at the Tournemire URL

In this paper, a comparative modelling exercise from the DECOVALEX-2015 project is presented. The exercise is based on in situ experiments, performed at the Tournemire Underground Research Laboratory (URL), run by the IRSN (Institut de Radioprotection et de Surete Nucleaire), in France. These experiments aim at identifying conditions (e.g. technical specifications, design, construction, and defects) that will affect the long-term performance of swelling clay-based sealing systems, which is of key importance for the safety of underground nuclear waste disposal facilities. A number of materials are being considered as seals; the current work focusses on a 70/30 MX80 bentonite–sand mixture initially compacted at a dry density of 1.94 Mg/m3. The performance of the sealing plug involves at least three different important components, which are the hydro-mechanical behaviour of the bentonite–sand core, the overall permeability of the surrounding argillite, and the influence of the technological gap between the core and the argillite. Two particular tests have been selected for a comparative modelling exercise: the WT-1 test, which was designed to study the rock mass permeability, and the PT-A1 test, which aimed at quantifying the evolution of the hydro-mechanical field within the bentonite–sand core. A number of independent teams have worked towards modelling these experiments, using different codes and input parameters calibrated on additional small-scale laboratory experiments. Their results are compared and discussed.

The VATO project: An original methodology to study the transfer of tritium as HT and HTO in grassland ecosystem

Tritium (3H) is mainly released into the environment by nuclear power plants, military nuclear facilities and nuclear reprocessing plants. The construction of new nuclear facilities in the world as well as the evolution of nuclear fuel management might lead to an increase of 3H discharges from the nuclear industry. The VATO project was set up by IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and EDF (Electricité de France) to reduce the uncertainties in the knowledge about transfers of 3H from an atmospheric source (currently releasing HT and HTO) to a grassland ecosystem. A fully instrumented technical platform with specifically designed materials was set up downwind of the AREVA NC La Hague reprocessing plant (Northwest of the France). This study, started in 2013, was conducted in four main steps to provide an hourly data set of 3H concentrations in the environment, adequate to develop and/or validate transfer models. It consisted first in characterizing the physico-chemical forms of 3H present in the air around the plant. Then, 3H transfer kinetics to grass were quantified regarding contributions from various compartments of the environment. For this purpose, an original experimental procedure was provided to take account for biases due to rehydration of freeze-dried samples for the determination of OBT activity concentrations in biological samples. In a third step, the 3H concentrations measured in the air and in rainwater were reconstructed at hourly intervals. Finally, a data processing technique was used to determine the biological half-lives of OBT in grass.

A Generalized turbulent dispersion model for bubbly flow numerical simulation in NEPTUNE_CFD

The NEPTUNE_CFD code, based upon an Eulerian multi-fluid model, is developed within the framework of the NEPTUNE project, financially supported by EDF (Electricité de France), CEA (Commissariat à l’Energie Atomique et aux Energies Alternatives), IRSN (Institut de Radioprotection et de Sûreté Nucléaire) and AREVA-NP. NEPTUNE_CFD is mainly focused on Nuclear Safety applications involving two-phase water-steam flows, like two-phase Pressurized Shock (PTS) and Departure from Nucleate Boiling (DNB).Many of these applications involve bubbly flows, particularly, for application to flows in PWR fuel assemblies, including studies related to DNB.Considering a very usual model for interfacial forces acting on bubbles, including drag, virtual mass and lift forces, the turbulent dispersion force is often added to moderate the lift effect in orthogonal directions to the main flow and get the right dispersion shape.This paper presents a formal derivation of this force, considering on the one hand, the fluctuating part of drag and virtual mass, and on the other hand, Turbulent Pressure derivation obtained by comparison between Lagrangian and Eulerian description of bubbles motion. An extension of the Tchen’s theory is used to express the turbulent kinetic energy of bubbles and the two-fluid turbulent covariance tensor in terms of liquid turbulent velocities and time scale. The model obtained by this way, called Generalized Turbulent Dispersion Model (GTD), does not require any user parameter.The model is validated against Liu & Bankoff air-water experiment, Arizona State University (ASU) experiment, DEBORA experiment and Texas A&M University (TAMU) boiling flow experiments.

Past and Future Research at IRSN on Corium Progression and Related Mitigation Strategies in a Severe Accident

Reactor core degradation and in-vessel and ex-vessel corium behavior have been major research topics for the last three decades to which Institut de Radioprotection et de Sûreté Nucléaire (IRSN) strongly contributed by the coordination of or the contribution to large research programs and through the development and validation of the severe accident (SA) ASTEC code. In recent years, the balance of research efforts has trended toward analyses of pros and cons and assessments of mitigation measures. The outcomes of risk significance analysis [including fuel-coolant interaction (FCI), hydrogen combustion, and molten core–concrete interaction (MCCI) risks] performed in France and corium behavior research are described. The focus these days is on (1) in-vessel melt retention (IVMR) strategies for future reactor concepts and the need to establish the reliability of such strategies when implemented in existing reactors and (2) in-containment corium cooling for existing reactors.This paper summarizes the main achievements and remaining issues related to understanding and modeling of (1) reflooding of a degraded core where, despite substantial knowledge gained through research programs, additional efforts are required to establish the efficiency of such a measure and the associated risks for largely degraded cores; (2) corium behavior in the reactor pressure vessel (RPV) lower head where, despite the Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) MASCA program results, efforts remain necessary to predict RPV thermal loadings resulting from corium layer evolution and RPV resilience with and without IVMR measures (internal and/or external cooling); (3) FCI for which, despite the OECD/NEA SERENA program results, the knowledge is not sufficient to assess with confidence the induced risk of containment failure; and (4) MCCI, where the knowledge on corium cooling in the containment by top and/or bottom water flooding is insufficient to formulate conclusions regarding the efficiency of such measures. Of particular interest for top flooding are the water ingress and corium eruption processes. Specifically for top flooding, respective impacts of water ingress and corium eruption processes remain to be quantified in reactor conditions.In support of these activities, substantial efforts are also being conducted at IRSN to constantly improve and validate nuclear material property databases that are key tools for corium behavior analysis.This paper describes ongoing and future research programs performed at IRSN or internationally with IRSN coordination or participation to tackle the remaining issues and summarizes expected progress in modeling for SA codes, in risk analysis and in SA management.