Aerosol retention in the flooded steam generator bundle during SGTR

A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out.To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated.Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with high gas flow rates carrying aerosol particles with them. However, compared to particle retention in the water close to the tube break, the effect of droplet entrainment on particle transport was small.     

Experimental study on the blowdown load during the steam generator feedwater line break accident in the evolutionary pressurized water reactor

Integral effect tests using the ATLAS facility were performed to obtain the thermal-hydraulic parameters such as dynamic and static pressures, local temperatures, and flow rates during a feedwater line break of a steam generator. The break of a feedwater line was simulated using a double rupture disc assembly in order to satisfy the requirements for the break opening time of around a few milliseconds. In the present study, estimated break opening time was less than 1.5 ms and broken areas were 48.1% and 93.4% of the feedwater line, respectively. The maximum dynamic pressures of about 1.57 bar were obtained inside of feedwater box that was closest to the break location of the feedwater line. After the break of the feedwater line, propagation of the pressure wave along the distance from the break location inside the steam generator was clearly and pertinently observed in all the tests. From a structural integrity point of view, however, the risk induced by this maximum dynamic load could be treated to be insignificant.     

Simulation of a gas jet entering the secondary side of a steam generator during a SGTR sequence: Validation of a FLUENT 6.2 Model

This paper summarizes the major insights gained as a result of gas jets entering a tube bundle from either a guillotine or a fish-mouth breach of a steam generator tube. This scenario is highly relevant in nuclear safety since it determines the potential retention of radioactive particles during risk-dominant sequences, the so-called Steam Generator Tube Rupture (SGTR) sequences. The scenario has been modeled with the FLUENT 6.2 code and its predictions have been proven to be grid independent and consistent with the experimental data available. The topology of the jets and the influence of the inlet mass flow rate (from 75 to 250 kg/h) have been studied in terms of velocity profiles.The results show that the breach shape heavily determines the jet topology. Both jets initially describe a quasi-parabolic trajectory, which is affected by the presence of the tubes. A guillotine breach generates a jet with azimuthal symmetry, which vanishes for the fish-mouth breach configuration. In this case, jet expands azimuthally in a pseudo-triangular way with a small angle. This fact diminishes the momentum loss across the bundle, so that for the same inlet mass flow rate the fish-mouth jet penetration is higher than the guillotine one. The normalized maximum radial and axial velocities of the jet from the guillotine breach are found to be self-similar with respect to inlet mass flow rate along the tube row position and axial distance to the breach, respectively. However, in absolute terms higher penetrations are found at higher mass flow rates.     

CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam

Steam generator tube rupture(SGTR)accident is an important scenario needed to be considered in the safety analysis of lead-based fast reactors.When the steam generator tube breaks close to the main pump,water vapor will enter the reactor core,resulting in a two-phase flow of heavy liquid metal and water vapor in fuel assemblies.The thermal-hydraulic prob-lems caused by the SGTR accident may seriously threaten reactor core's safety performance.In this paper,the open-source CFD calculation software OpenFOAM was used to encapsulate the improved Euler method into the self-developed solver LBEsteamEulerFoam.By changing different heating boundary conditions and inlet coolant types,the two-phase flow in the fuel assembly with different inlet gas content was simulated under various accident conditions.The calculation results show that the water vapor may accumulate in edge and corner channels.With the increase in inlet water vapor content,outlet coolant velocity increases gradually.When the inlet water vapor content is more than 15％,the outlet coolant temperature rises sharply with strong temperature fluctuation.When the inlet water vapor content is in the range of 5-20％,the upper part of the fuel assembly will gradually accumulate to form large bubbles.Compared with the VOF method,Euler method has higher computational efficiency.However,Euler method may cause an underestimation of the void fraction,so it still needs to be calibrated with future experimental data of the two-phase flow in fuel assembly.     

Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.     

Evaluation of a steam generator tube rupture accident in an accelerator driven system with lead cooling

A postulated steam generator tube rupture (SGTR) accident in a lead cooled accelerator driven transmuter (ADT) is investigated. The design of the ADT without intermediate loops bears the risk of water/steam blasting into the primary coolant. As a consequence a nuclear power excursion could be triggered by steam ingress into the ADT core which has a significant positive void worth. A thermal coolant–coolant interaction (CCI) might initiate a local core voiding too and additionally could lead to sloshing of the lead pool with mechanical impact of the heavy liquid on structures. The steam formation will also lead to a pressurization of the cover gas. The problems related to an SGTR are identified and investigated with the SIMMER-III accident code.     

Fission-product chemistry and aerosol behaviour in the primary circuit of a pressurized water reactor under severe accident conditions

This review considers fission-product chemistry and aerosol behaviour in the primary circuit of a pressurized water reactor (PWR) during severe accidents. Three key accident sequences (V, TMLB' and S₂D) are considered, and their principal thermal-hydraulic and physical characteristics affecting chemistry behaviour are identified. The inventories, chemical forms and timing of fission products released from the fuel are summarized together with the major sources of structural materials and their release characteristics. The chemistry of each main fission-product species within the primary circuit is reviewed from available experimental and thermodynamic data and/or theoretical predictions. Modelling studies of primary circuit fission-product behaviour are reviewed briefly and the principal requirements for further study assessed with respect to experimental and modelling programmes currently in progress.     

On the nature of aerosols produced during a severe accident of a water-cooled nuclear reactor

Particle behaviour depends strongly on classic characteristics, e.g., size, and less macroscopic ones involving structure and composition these being especially important in situations of strong differential forces on a particle, i.e., surface impact or intensely-shearing flows. The former situation may lead to particle deposition or break-up and re-entrainment (with potential accident-management implications). This paper reviews information on aerosols from prototypical experiments identifying common features and typical variations. It emerges that a particle comprising one-third metal, one-third metal oxide and one-third a mixture of fission-product species would not be out of place in any potential reactor-accident sequence. Particle shapes appear relatively compact without branching chain-like structures. On size and structure, aerosols in the upstream part of the primary circuit would comprise a near-lognormal population with AMMD no more than 2 μm and geometric standard deviation around 2, particles comprising agglomerates of highly-coordinated clusters as small as 0.1 μm. In the containment, aerosols can typically be represented by primary-circuit particles and their agglomerates though particular circumstances (core–concrete interaction, hot-leg accident sequence) can alter this simple picture.     

Three-dimensional behaviors of the hydrogen and steam in the APR1400 containment during a hypothetical loss of feed water accident

During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by an active reaction of the fuel-cladding and the steam in the reactor pressure vessel and released with the steam into the containment. In order to mitigate hydrogen hazards which could possibly occur in the NPP containment, a hydrogen mitigation system (HMS) is usually adopted. The design of the next generation NPP (APR1400) developed in Korea specifies that 26 passive autocatalytic recombiners and 10 igniters should be installed in the containment for a hydrogen mitigation. In this study, an analysis of the hydrogen and steam behavior during a total loss of feed water (LOFW) accident in the APR1400 containment has been conducted by using the computational fluid dynamics (CFD) code GASFLOW. During the accident, a huge amount of hot water, steam, and hydrogen is released into the in-containment refueling water storage tank (IRWST). The current design of the APR1400 includes flap-type openings at the IRWST vents which operate depending on the pressure difference between the inside and outside of the IRWST. It was found from this study that the flaps strongly affect the flow structure of the steam and hydrogen in the containment. The possibilities of a flame acceleration and a transition from deflagration to detonation (DDT) were evaluated by using the Sigma–Lambda criteria. Numerical results indicate that the DDT possibility was heavily reduced in the IRWST compartment by the effects of the flaps during the LOFW accident.     

Non-uniform flow distribution in the steam generator U-tubes of a pressurized water reactor plant during single- and two-phase natural circulations

A flow model is formulated to analyze the single- and two-phase flow characteristics in the primary side of a vertical, inverted U-tube steam generator. Using the flow model, a pressure drop–mass flow rate curve is derived. It is shown that the curve has a negative slope and, thus, flow excursion instability can occur under certain low-flow conditions. The stability criterion and its physical interpretation are given. Nonuniform flow behaviors in multiple U-tubes of different tube lengths are also discussed.     

Predicting the accumulated number of plugged tubes in a steam generator using statistical methodologies

A steam generator (SG) plays a significant role not only with respect to the primary-to-secondary heat transfer but also as a fission product barrier to prevent the release of radionuclides. Tube plugging is an efficient way to avoid releasing radionuclides when SG tubes are severely degraded. However, this remedial action may cause the decrease of SG heat transfer capability, especially in transient or accident conditions. It is therefore crucial for the plant staff to understand the trend of plugged tubes for the SG operation and maintenance. Statistical methodologies are proposed in this paper to predict this trend. The accumulated numbers of SG plugged tubes versus the operation time are predicted using the Weibull and log–normal distributions, which correspond well with the plant measured data from a selected pressurized water reactor (PWR). With the help of these predictions, the accumulated number of SG plugged tubes can be reasonably extrapolated to the 40-year operation lifetime (or even longer than 40 years) of a PWR. This information can assist the plant policymakers to determine whether or when a SG must be replaced.     

Distribution of liquid sodium in the inlet plenum of steam generator in a Fast Breeder Reactor

Experimental and Computational Fluid Dynamics (CFD) investigations have been carried out on a 1/5th scale model of the inlet plenum of steam generator (SG) used in the Fast Breeder Reactor (FBR) technology. The distribution of liquid sodium in the inlet plenum of the steam generator strongly affects the thermal as well as mechanical performance of the steam generator. In the present work, flow distribution in a scaled down model has been investigated. Various strategies adopted for obtaining uniform flow distribution have been evaluated. Experiments have been conducted to measure the axial and radial velocity distributions using Ultrasonic Velocity Profiler (UVP) under a variety of geometries. Computational Fluid Dynamics (CFD) studies have been carried out for various geometries. On the basis of these experiments and CFD simulations, various flow distribution devices have been compared.     

A fuzzy decision tree method for fault classification in the steam generator of a pressurized water reactor

This paper extends a method previously introduced by the authors for building a transparent fault classification algorithm by combining the fuzzy clustering, fuzzy logic and decision trees techniques. The baseline method transforms an opaque, fuzzy clustering-based classification model into a fuzzy logic inference model based on linguistic rules which can be represented by a decision tree formalism. The classification model thereby obtained is transparent in that it allows direct interpretation and inspection of the model. An extension in the procedure for the development of the fuzzy logic inference model is introduced to allow the treatment of more complicated cases, e.g. splitted and overlapping clusters. The corresponding computational tool developed relies on a number of parameters which can be tuned by the user to optimally compromise the level of transparency of the classification process and its efficiency. A numerical application is presented with regards to the fault classification in the Steam Generator of a Pressurized Water Reactor.     

Evaluation of the effect of perforations in the collector vessel of a steam generator on the critical stress intensity coefficient

An approach to calculating the effective critical stress intensity coefficient for mainline cracks in the perforation zone of the first-loop collector of a VVER steam generator is examined. The energy criterion is chosen as the condition for crack stability: the decrease of the elastic deformation energy which is due to the opening of a growing crack should not exceed the change of the surface energy of the crack in the process. The difference between the critical stress intensity coefficient for a mainline crack in the continuous and perforated regions is analyzed. Relations making it possible to lower the critical stress intensity coefficient as a function of the parameters of the perforation (spacing, diameter of openings) as well as the dimensions and orientation of the mainline crack in the perforation zone are presented.     

Two-phase natural circulation flow of air and water in a reactor cavity model under an external vessel cooling during a severe accident

As a part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of the Advanced Power Reactor (APR) 1400, a Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale (HERMES-HALF) experiment has been performed by using the non-heating method of an air injection. This large-scale experiment uses a half-height and half-sector model of the APR1400. This experiment has been analyzed to verify and evaluate the experimental results by using the RELAP5/MOD3 computer code. The RELAP5/MOD3 results have shown that the water circulation mass flow rate is very similar to the experimental results of the HERMES-HALF, in general. Increases in the water inlet area and the water level in the reactor cavity lead to an increase in the water circulation mass flow rate. The effects of an air injection mass flow rate and the water outlet area on the water circulation mass flow rate are dependent on the water inlet area size. As the water outlet moves to a lower position, the water circulation mass flow rate increases slowly.     

Effect of turbulent natural convection on sodium pool combustion in the steam generator building of a fast breeder reactor

A computational model is proposed to simulate sodium pool combustion considering the effect of turbulent natural convection in a vented enclosure of the steam generator building (SGB) of a fast breeder reactor. The model is validated by comparing the simulated results with the experimental results available in literature for sodium pool combustion in a CSTF vessel. After validation, the effects of vents and the location of the pool on the burning rate of sodium and the associated heat transfer to the walls are studied in an enclosure comparable in size to one floor of the steam generator building. In the presence of ventilation, the burning rate of sodium increases, but the total heat transferred to the walls of the enclosure is reduced. It is also found that the burning rate of sodium pool and the heat transfer to the walls of the enclosures vary significantly with the location of sodium pool.