Core Meltdown Accident Research Articles

Experimental simulation of downward molten material relocation by jet ablation of structures and fuel coolant interaction in a fast reactor

In the remote chance of a core melt-down accident in a fast reactor, the molten corium from the fuel subassembly can come out in the form of a liquid jet due to the prevailing transient pressure. This high temperature jet is capable of causing local breaches on horizontal solid structures such as the grid plate with which they come into contact inside the main vessel. After penetration through the plates, the molten jet undergoes fragmentation in the bulk coolant available below the grid plate, in the lower plenum and settles on the core catcher. Experimental investigation of molten material relocation in the downward direction is carried out with a wood's metal jet impinging on a pair of solid wood's metal plates. The wood's metal alloy is heated electrically to about 300 °C inside the melt chamber and released through a nozzle to impinge on the dry plates mounted on rigid supports and allowed to pierce through the plates. The jet then gets quenched in bulk water contained in the experimental vessel. The trajectory of vertical jet impinging on the set of two parallel horizontal plates is captured using a high speed camera. Thermocouples are placed on the top and bottom surface of the plates to assess the temperature evolution. The breach area formed on the plates by the hot molten jet is assessed from the experiments. The initial pressure of the liquid jet is varied from 0 bar to 2 bar gauge pressure. The mass of the debris collected on the breached wood's metal target plates and the stainless steel collector plate gives evidence that the entire molten mass of jet does not reach the bottom collector plate but gets distributed on all the plates. The particle size distribution is obtained for the fragmented wood's metal debris settled on the collector plate. It is observed that the mass median diameter of the fragmented particles and porosity of the debris bed decreases as the jet release pressure is increased. An empirical correlation between the particle size represented by the mass median diameter and the initial pressure of the jet is obtained.

Safest Roadmap for Corium Experimental Research in Europe

Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) severe accident (SA) with reactor core melting and formation of hazardous material system known as corium. The main objective of the project is to establish coordinated activities, enabling the development of a common vision and severe accident research roadmaps for the next years, and of the management structure to achieve these goals. In this frame, a European roadmap on severe accident experimental research has been developed to define research challenges to contribute to further reinforcement of Gen II and III NPP safety. The roadmap takes into account different SA phenomena and issues identified and prioritized in the analyses of severe accidents at commercial NPPs and in the results of the recent European stress tests carried out after the Fukushima accident. Nineteen relevant issues related to reactor core meltdown accidents have been selected during these efforts. These issues have been compared to a survey of the European SA research experimental facilities and corium analysis laboratories. Finally, the coherence between European infrastructures and R&D needs has been assessed and a table linking issues and infrastructures has been derived. The comparison shows certain important lacks in SA research infrastructures in Europe, especially in the domains of core late reflooding impact on source term, reactor pressure vessel failure and molten core release modes, spent fuel pool (SFP) accidents, as well as the need for a large-scale experimental facility operating with up to 500 kg of chemically prototypic corium melt.

Numerical simulation of Post Accident Heat Removal in a typical pool-type SFR under severe core relocation scenario

Under a hypothetical core meltdown and relocation scenario in Sodium cooled Fast Reactor (SFR), there may be a chance for reactor main vessel failure if the core debris are allowed to settle at the bottom of this vessel. To avoid this, an in-vessel core catcher assembly is installed in pool type SFRs just above the bottom wall of the main vessel. The function of this assembly is to collect, retain and passively cool the core debris by assisting natural convection decay heat removal from these debris. In this work, the passive decay heat removal from the core debris settled on single-tray-type design of core catcher has been numerically simulated under different amounts of core relocation using a validated 2-D axisymmetric CFD model. The maximum temperature experienced by the debris bed, core catcher plate and the welded joint between the core support structure and the main vessel (i.e., triple point) are evaluated in order to estimate the amount of core relocation that can be safely handled by the primary containment system under Post Accident Heat Removal (PAHR) scenario. The numerical study revealed that the single-tray-type design of core catcher can potentially assist the passive cooling even under the settling of debris material with heat generation equivalent to 60% of the full core on the core catcher. In addition to this, the effect of the size of opening created on grid plate during core relocation on Post Accident Heat Removal (PAHR) is studied for different amounts of core relocation. From this numerical study, the thermal capability of single-tray type in-vessel core catcher of SFR is assessed. The predictions are useful in determining the safe thermal margins available under core relocation scenarios and extending the functionality of single-tray-type core catcher for handling Whole Core Meltdown Accident (WCA) scenario in future SFRs.

Analysis of composition and temperature of debris bed based on settling characteristics of relocated core debris in pool type SFR

The hypothetical core meltdown accident in sodium cooled fast reactors (SFR) is a severe accident situation during which the energetic dispersion of core materials and their relocation may pose a threat to the integrity of reactor containment. During core meltdown accident, part of the molten core materials relocated downwards would be fragmented in the lower plenum sodium pool and settle over the in-vessel core catcher plate and forms a particulate debris bed. The recriticality potential and coolability of the debris bed are strongly influenced by the composition of fragmented particles of different core materials and their size distribution. In this paper, simplified models have been developed to evaluate the incipient conditions i.e., composition profile and initial temperature of the particulate debris bed formed over an in-vessel core catcher plate following the core meltdown accident in pool type SFR. The particle mixture settling process subsequent to the fragmentation of molten core materials in sodium pool has been analyzed to evaluate the composition profile of the debris bed. A 1-D model for particle motion through the stagnant fluid under the influence of gravity has been considered to evaluate the settling characteristics of core materials debris (mainly UO2 and stainless steel) in liquid sodium for a wide range of possible sizes. By extending this model, the differential settling behaviour of mixture of UO2 and stainless steel particles in liquid sodium pool has been investigated. The composition profile of the debris bed along its thickness has been evaluated for the material relocation scenario considered with various mass ratios of stainless steel and UO2 mixture using the published experimental results based particle size distributions. Further, a 1-D transient conduction model is implemented to estimate the temperature of debris particles when they settle on the core catcher plate. Simulations revealed that the particles will be quenched to near the temperature of the liquid sodium in lower pool while reaching the core catcher. The results are useful in assessing the coolability characteristics of debris bed formed on the core catcher and its recriticality potential in the analysis of the Post Accident Heat Removal.

Numerical investigation on the dissolution kinetics of ZrO2 by molten zircaloy using MPS method

During the postulated core meltdown accident of the nuclear reactor which employs the dispersion plate-type fuel, the UO2 microspheres would be dissolved by the molten zirconium, and the dissolution process is accelerated by the mass diffusion phenomenon. In this study, the dissolution model, based on the Fick’s second law and the dissolution criteria, was developed and incorporated into the Moving Particle Semi-implicit (MPS) method. To validate the dissolution model, the one-dimensional mass diffusion problem was simulated by the improved MPS code. The study of numerical convergence was also performed based on the sensitivity analysis of the particle size. The distribution of the oxygen predicted by the improved MPS agreed well with the analytical solution for the particle size of 0.5mm. Then the improved MPS was adopted to analyze the dissolution kinetics of the three-dimensional ZrO2-Zr system in this study. The oxygen concentration and the interface displacement of the ZrO2-Zr system were considered, and the simulation results indicate that the behavior of ZrO2 dissolution caused by molten zircaloy follows the parabolic law. The improved MPS method is proved to be capable of simulating the dissolution process and the induced phase change.

Physical properties of core-concrete systems: Al2O3-ZrO2 molten materials measured by aerodynamic levitation

During a molten core–concrete interaction, molten oxides consisting of molten core materials (UO2 and ZrO2) and concrete (Al2O3, SiO2, CaO) are formed. Reliable data on the physical properties of the molten oxides will allow us to accurately predict the progression of a nuclear reactor core meltdown accident. In this study, the viscosities and densities of molten (ZrO2)x(Al2O3)1-x (x = 0.356 and 0.172) were measured using an aerodynamic levitation technique. The densities of two small samples were estimated from their masses and their volumes (calculated from recorded images of the molten samples). The droplets were forced to oscillate using speakers, and their viscosities were evaluated from the damping behaviors of their oscillations. The results showed that the viscosity of molten (ZrO2)x(Al2O3)1-x compared to that of pure molten Al2O3 is 25% lower for x = 0.172, while it is unexpectedly 20% higher for x = 0.356.

설계기준초과지진 하의 원전 배관 구조건전성 평가를 위한 변형률 기반 방법

Following the 2011 Fukushima Nuclear Power Plant accident, the IAEA has issued a revised version of the Nuclear Safety Standard for beyond design basis earthquake to consider the core meltdown accident. In Korea, relevant laws and regulations were also revised to consider beyond design basis earthquake to nuclear components. In this paper, CAV, an seismic damage factor that determines the restart of nuclear power plant after operating breakdown earthquake, is proposed for extension to the beyond design basis earthquake. For pipings not satisfying the beyond design basis earthquake condition, several evaluation methods are suggested, such as strain-based evaluation methods, simple nonlinear analysis method and cumulative damage evaluation method.

Physical properties of molten core materials: Zr-Ni and Zr-Cr alloys measured by electrostatic levitation

It is important to understand the behaviors of molten core materials to investigate the progression of a core meltdown accident. In the early stages of bundle degradation, low-melting-temperature liquid phases are expected to form via the eutectic reaction between Zircaloy and stainless steel. The main component of Zircaloy is Zr and those of stainless steel are Fe, Ni, and Cr. Our group has previously reported physical property data such as viscosity, density, and surface tension for Zr-Fe liquid alloys using an electrostatic levitation technique. In this study, we report the viscosity, density, and surface tension of Zr-Ni and Zr-Cr liquid alloys (Zr1-xNix (x = 0.12 and 0.24) and Zr0.77Cr0.23) using the electrostatic levitation technique.

Simulation of the in-vessel retention device heat-removal capability of AP-1000 during a core meltdown accident

Westinghouse Electric Company’s design for the Advanced Pressurized Water Reactor (AP-1000) incorporates a special in-vessel retention (IVR) device to enhance the capability of removing heat through the outer wall of the reactor vessel during a core meltdown. External Reactor Vessel Cooling (ERVC) is established by flooding the reactor cavity during a severe accident. The amount of heat removed is limited by the critical heat flux (CHF) at the outer surface of the vessel wall. In this paper, the CHF correlations developed specifically for this purpose are reviewed. The RELAP5-3D thermal-hydraulic code is used to simulate the natural convection flow within the water channel of the IVR device. The results of RELAP5-3D simulations are substituted into ULPU, SULTAN, SBLB, and KAIST critical heat flux correlations to determine the safety margin of the AP-1000 IVR design. Among these four correlations, the CHFs predicted by the ULPU and SBLB correlations are insensitive to the thermal-hydraulic conditions of the coolant channel. The CHF predicted by SULTAN is the lowest among the four correlations. The results demonstrate that the predicted CHF increases with the angle. Nevertheless, the ratio of CHF to heat flux is lowest around the upper quarter of the vessel. The heat load to the lower head is based on two bounding melt configurations. Configuration I involves a stratified light metallic layer on top of a molten ceramic pool, and Configuration II represents the condition that an additional heavy metal layer forms below the ceramic pool. The impact of corium pool configuration on the safety performance of IVR is very significant. Under Configuration II, as predicted by the SULTAN and KAIST correlations, the performance of the current AP-1000 IVR design is marginal. The results of the RELAP5-3D IVR simulation show that the total mass flow rate in the coolant channel increases sharply as the widths of the annular channels vary from 0.0762 to 0.762m. The mass-flow rate decreases gradually when the gap width increases beyond 0.762m. The KAIST CHF correlation predicts that CHF is maximal when the gap width is around 0.4572m and decreases gradually when the gap width increases.

A Fukushima-Like Disaster in New York City and Possible Issues Associated with the Extreme Event Management System

People living in communities surrounding the Indian Point Nuclear Generating Plant (IPNGP) inevitably face potential risks associated with a nuclear plant issue, which could become a potential extreme event. The study searches for answers to the three research questions: (1) What is the population makeup of those residing nearby the IPNGP (2) Which racial and ethnic groups are more vulnerable to the potential risks associated with the IPNGP (3) Who would be exposed to the radioactive plume if the IPNGP experienced a core-meltdown accident? The study utilizes Radiological Assessment System for Consequence Analysis (RASCAL) computer software code to project the radioactive plume dispersion when the IPNGP encounters a core-meltdown accident. Findings show that there are between 1.39 million and 19.85 million people that could possibly be exposed to the radioactive plume under four different typifying weather conditions in a calendar year. The study suggests this potential risk of an extreme event should be taken into consideration for “flexible adaptation approach” for building a resilient and sustainable New York City, rather than the one where a serious disaster could impact thousands of lives, create millions of dollars in damage, and create another scenario where emergency response is tested and fails to meet the requirements of appropriate sources to respond and do their part accordingly and appropriately to protect and save the lives of residents.

Analysis of containment venting strategy under severe accident conditions

Under severe accidents without containment heat removal, the containment integrity can be challenged due to over-pressurization by the steam and non-condensable gas generation. Containment filtered venting has been considered as an effective measure to avoid or delay the containment failure by over-pressurization. In order to minimize the environmental effects and maximize the effectiveness of filtered venting, it is critical to initiate venting in timely manners with sufficient discharge flow rate. It is also important to optimize the vent line size to prevent additional risk of leakage and to minimize the space and the cost requirement. In this study, the various venting strategies are investigated with respect to possible release flow characteristics. First of all, the accident scenarios representing the core meltdown accident with containment pressurization are selected by reviewing relevant literature. With those scenarios, thermal hydraulic behaviors in the containment and the discharge flow depending on the different venting strategies (i.e. vent line size and vent initiation pressure) are analyzed. MAAP5 model for the OPR1000 Korean nuclear power plant has been used for simulation.

Thermophysical properties of molten core materials: Zr–Fe alloys measured by electrostatic levitation

ABSTRACTIn order to investigate the progression of a core meltdown accident, it is necessary to understand the behavior of molten core materials. Zr–Fe alloys are one of the low-melting-temperature liquid phases that are thought to form in the early stages of bundle degradation. The objective of this study is to measure the thermophysical properties of Zr–Fe liquid alloys. Alloy samples with a composition of Zr1−xFex (x = 0.12, 0.24, and 0.50) were synthesized by arc melting, and their density, viscosity, and surface tension were measured using an electrostatic levitation technique. The results indicate that the density of Zr–Fe liquid alloys can be estimated by a linear combination of the measured or extrapolated densities of pure Zr and Fe. The viscosities of the Zr–Fe liquid alloys can be roughly estimated by extrapolating those of Zr to lower temperatures, although this method tends to underestimate the viscosity of alloys, especially for eutectic compositions. The values of the Zr–Fe liquid alloys’ surface tensions are close to those of pure Zr.

Robotic suction nozzle design optimization for severe accident portable air-cleaning system

A new portable filtered air-suction system (PoFASS) is proposed in this paper to prevent the release of radioactive fission products into the atmosphere from severe core meltdown accidents in nuclear power plants (NPPs), which has become very important after the Fukushima accident. The PoFASS consists of a robot arm with a vent line, a robot body equipped with a radioactive filter, and a robot suction nozzle that is the shape of a hand with fingers. By controlling the angle of the robot fingers on the robot suction nozzle, it can change its own shape to completely adapt to cover the damaged sections found in NPPs, and consequently, radioactive release to the atmosphere is effectively prevented. Surveying all the possible shapes of damaged sections in non-containment structures, six distinctive cases were identified. The minimum number of robot fingers that can cover all six cases of damaged sections was mathematically determined to be four, but in the design of the suction nozzle, six robot fingers were included to take into consideration the extension limit of the extensible cover of the suction nozzle as well as the area to be covered at the damaged sections. A complete mathematical model was developed to validate the covering capacity of the designed suction nozzle for all six cases. The minimum number of robot arm rods to access to any damaged section in NPPs was also determined to be three; hence, the PoFASS can access the hatch door of a containment buildings or the damaged sections of containment buildings, turbine buildings, and auxiliary buildings in NPPs. Experimental verification for the proposed PoFASS was successfully done with a prototype suction nozzle for the six cases of damaged sections.

P26] Surface tension calculation of Si based and Zr based alloys in the Si-Zr system

The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2–PuO2–ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2–PuO2–ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

High temperature investigation of the solid/liquid transition in the PuO2–UO2–ZrO2 system

The solid/liquid transitions in the quaternary U-Pu-Zr-O system are of great interest for the analysis of core meltdown accidents in Pressurised Water Reactors (PWR) fuelled with uranium-dioxide and MOX. During a severe accident the Zr-based cladding can become completely oxidised due to the interaction with the oxide fuel and the water coolant. In this framework, the present analysis is focused on the pseudo-ternary system UO2–PuO2–ZrO2. The melting/solidification behaviour of five pseudo-ternary and one pseudo-binary ((PuO2)0.50(ZrO2)0.50) compositions have been investigated experimentally by a laser heating method under pre-set atmospheres. The effects of an oxidising or reducing atmosphere on the observed melting/freezing temperatures, as well as the amount of UO2 in the sample, have been clearly identified for the different compositions. The oxygen-to-metal ratio is a key parameter affecting the melting/freezing temperature because of incongruent vaporisation effects. In parallel, a detailed thermodynamic model for the UO2–PuO2–ZrO2 system has been developed using the CALPHAD method, and thermodynamic calculations have been performed to interpret the present laser heating results, as well as the high temperature behaviour of the cubic (Pu,U,Zr)O2±x-c mixed oxide phase. A good agreement was obtained between the calculated and experimental data points. This work enables an improved understanding of the major factors relevant to severe accident in nuclear reactors.

Investigation of fragmentation phenomena and debris bed formation during core meltdown accident in SFR using simulated experiments

The event of a severe core melt down accident, resulting in the relocation of the active core is analyzed as a part of the nuclear reactor safety research in order to ensure safe removal of decay heat. Molten Fuel Coolant Interaction (MFCI) and debris bed configuration on the core catcher plate assumes importance in assessing the post accident heat removal capability. The key factors affecting the coolability of the debris bed are the bed porosity, morphology of the fragmented particles, degree of spreading/heaping of the debris on the core catcher and the fraction of lump formed. A well defined debris bed is helpful in fixing a prototypical source term for the PAHR studies. Towards this, a series of experiments on fragmentation kinetics and subsequent debris bed formation is conducted with molten Wood's metal (an alloy of Bi 50%, Pb 25%, Sn 12.5% and Cd 12.5% with melting point of 346K) in water simulant system. The experiments are carried out using 2kg, 5kg and 20kg melt inventories. The particle size distribution obtained for the fragmented debris is fit using an Upper Limit Log Normal (ULLN) distribution. The dependence of particle size distribution on initial melt temperature and interaction height is quantified by correlating them to the key parameters i.e. shape factor and location factor in the ULLN expression. Morphology of the debris particles is investigated to understand the fragmentation mechanisms involved. Three major mechanisms of fragmentation are identified namely melt entrainment mechanism, boundary layer stripping and hydrodynamic breakup due to capillary forces. Finally, an approach to quantify the stratification and spreading behaviour of debris on the collector tray is presented by analysing the particle size and mass percentage in different radial zones of the collector tray.

Numerical simulation of passive heat removal under severe core meltdown scenario in a sodium cooled fast reactor

A sequence of highly unlikely events leading to significant meltdown of the Sodium cooled Fast Reactor (SFR) core can cause the failure of reactor vessel if the molten fuel debris settles at the bottom of the reactor main vessel. To prevent this, pool type SFRs are usually provided with an in-vessel core catcher above the bottom wall of the main vessel. The core catcher should collect, retain and passively cool these debris by facilitating decay heat removal by natural convection. In the present work, the heat removal capability of the existing single tray core catcher design has been evaluated numerically by analyzing the transient development of natural convection loops inside SFR pool. A 1-D heat diffusion model and a simplified 2-D axi-symmetric CFD model are developed for the same. Maximum temperature of the core catcher plate evaluated for different core meltdown scenarios using these models showed that there is much higher heat removal potential for single tray in-vessel SFR core catcher compared to the design basis case of melting of 7 subassemblies under total instantaneous blockage of a subassembly. The study also revealed that the side pool of cold sodium plays a significant role in decay heat removal. The maximum debris bed temperature attained during the initial hours of PAHR does not depend much on when the Decay Heat Exchanger (DHX) gets operational, and it substantiates the inherent safety of the system. The present study paves the way for better understanding of the thermal hydraulics of Post Accident Heat Removal (PAHR) towards extending the functionality of core catcher for handling Whole Core Meltdown Accident (WCA) scenarios in future SFRs.

Ablation characteristics of special concrete due to an impinging zirconium-dioxide melt jet

Jet impingement experiments were performed to investigate the ablation characteristics of special concrete, which has been developed as one of the candidate protecting materials for the EU-APR1400 ex-vessel core catcher. In order to simulate the jet impingement phenomenon owing to the reactor vessel failure during a severe core meltdown accident, the experimental facility was established and the experimental conditions were determined based on parametric studies. The special concrete specimen was manufactured in accordance with the standard procedures, and its microstructures and physicochemical properties were analyzed to verify the requirements for the qualification. An induction melting technique in a cold crucible was employed to generate the zirconium-dioxide melt as a simulant of the corium melt. The special concrete was ablated uniformly over the impact area by jet impingement, and the average ablation depth was measured to be 4.33mm. The average ablation rate in depth was evaluated as 1.59mm/s using the temperature measurements of the specimen. As compared with the predictions by the models based on the convective and radiative heat transfer analysis, both the measured ablation rate and depth were estimated appropriately within the bounds of their limits. However, the convective heat transfer model turned out to predict the ablation characteristics of the special concrete more reasonably during the jet impingement even though some water content within the special concrete could lead to a sudden generation of the steam layer through which the material ablation is attenuated substantially by the radiative heat transfer.

An estimate of the order of magnitude of the explosion during a core meltdown-compaction accident for heavy liquid metal fast reactors: A disquieting result updating the Bethe–Tait model

Criticality and recriticality considerations in heavy liquid metal fast reactors (HLMFRs) after a hypothetical core meltdown accident are discussed. Although many aspects of system behaviour in such scenarios can be deduced directly from the classical theory of sodium-cooled fast reactors (SFRs), certain ideas that have been accepted as true for SFRs cannot be extrapolated to HLMFRs without sufficiently careful thought. In this paper, we are concerned, as in SFRs, with fuel compaction, but with one important difference: there would be no boiling of the surrounding heavy liquid metal pool. Utilizing a Bethe–Tait model, it is shown that, due to the power flattening effect of the heavy liquid metal, explosive excursions at least an order of magnitude higher than for SFRs in similar situations are conceivable.

SARNET hydrogen deflagration benchmarks: Main outcomes and conclusions

In case of a core melt-down accident in a light water nuclear reactor, hydrogen is produced during reactor core degradation and released into the reactor building. This subsequently creates a combustion hazard. A local ignition of the combustible mixture may generate standing flames or initially slow propagating flames. Depending on geometry, mixture composition and turbulence level, the flame can accelerate or be quenched after a certain distance. The loads generated by the combustion process (increase of the containment atmosphere pressure and temperature) may threaten the integrity of the containment building and of internal walls and equipment. Turbulent deflagration flames may generate high pressure pulses, temperature peaks, shock waves and large pressure gradients which could severely damage specific containment components, internal walls and/or safety equipment. The evaluation of such loads requires validated codes which can be used with a high level of confidence.Currently, turbulence and steam effect on flame acceleration, flame deceleration and flame quenching mechanisms are not well reproduced by combustion models usually implemented in safety tools and further model enhancement and validation are still needed. For this purpose, two hydrogen deflagration benchmark exercises have been organised in the framework of the SARNET network. The first benchmark was focused on turbulence effect on flame propagation. For this purpose, three tests performed in the ENACCEF facility were considered. They concern vertical flame propagation in an initially homogenous mixture with 13vol.% hydrogen content and different geometrical configurations. Three blockage ratios of 0, 0.33 and 0.6 were considered to generate different levels of turbulence. The second benchmark objective was the investigation of the diluting effect on flame propagation. Thus, three tests performed in the ENACCEF facility using the same blockage ratio of 0.63 and three different initial gas compositions (with 10, 20 and 30vol.% diluents) have been considered. Since ENACCEF runs at ambient temperature, a surrogate to steam was used consisting of a mixture of 0.6He+0.4CO2 on molar basis.This paper aims to present the benchmarks conclusions regarding the ability of LP and CFD combustion models to predict the effect of turbulence and diluent on flame propagation.