Accidents In Light Water Reactors Research Articles

BISON validation of FeCrAl cladding mechanical failure during simulated reactivity-initiated accident conditions

A reactivity-initiated accident (RIA) is a postulated design basis accident in light water reactors (LWRs) in which a rapid reactivity insertion induces a fission rate increase and a fuel pellet temperature rise. During RIA, the fuel pellet thermally expands and may cause pellet-cladding mechanical interaction (PCMI). Separate effects PCMI tests were performed on C26M FeCrAl cladding tube samples, introducing biaxial stress via a well understood modified burst test (MBT) system. A high-speed camera in the MBT system captured the projections, covering a 360° view of the cladding deformation and enabling a digital image correlation (DIC) method to quantify the surface strains with high fidelity. Representative hot zero-power RIA mechanical loading conditions were applied to the sample, and the test duration ranged from 20 to 500 ms at an average temperature of 573 K. BISON finite element–based fuel performance code modeling was performed against the high-fidelity DIC data produced from the MBTs. Validation calculations were conducted with 2D models and systematically compared with test data of the burst time, burst pressure, burst hoop strains, and hoop strain rates. Based on the behaviors from the separate effects test, BISON calculations satisfactorily predicted the cladding deformation behaviors. Sensitivity analysis was conducted to identify highly influential mechanical properties responsible for the cladding failure behavior during the MBT experiments. The results highlight the significance of the cladding's mechanical strength in governing cladding strain, followed by the significance of the mechanical interactions between the pellet and the cladding.

Monte Carlo method with SNBCK nongray gas model for thermal radiation in containment flows

During a postulated severe accident in light water reactors, containment flows are characterized by relatively low velocities and modest temperature differences after depressurization of the reactor cooling system. Owing to the low gas temperature level and small temperature differences, gas radiation heat transfer was neglected in CFD simulations of hydrogen transport and mixing in containment flows. During the natural mixing and dilution processes, thermal radiation can affect the fluid temperatures and thereby influence the buoyancy forces in the presence of large amounts of steam. In order to quantify the effect of thermal radiation, an efficient Monte Carlo method with statistical narrow band correlated-k (SNBCK) nongray gas model has been implemented to solve the radiative transfer equation within the framework of containmentFOAM. The tailored Monte Carlo solver is validated against the OECD/NEA SETH-2 PANDA ST1 experiments, and it is proven that gas radiation has a significant effect on the gas temperature field and the helium-steam mixing process. The results clearly indicate that gas radiation cannot be neglected in containment flows even at modest temperature differences (ΔT<50K).

Numerical Study on the Effect of Jet Cross Section Shape on the Corium Jet Breakup Behavior With Lattice Boltzmann Method

In a core meltdown accident in light water reactors, molten corium may drop into the lower plenum of the pressure vessel and interact with water, which is called fuel–coolant interaction (FCI). The behavior of the corium jet breakup in water during FCIs is important for the in-vessel retention strategy and has been extensively studied. While in previous studies, the jet cross-section shapes are naturally assumed to be circular, which is actually not always the case, in this study, the breakup processes of the corium jets with four different elliptical cross-section shapes and three different penetration velocities are simulated with color-gradient lattice Boltzmann method. The effect of the cross-section shape on the hydrodynamic breakup behavior of the corium jet is analyzed in detail. It is found that the effect of the cross-section shape on the jet penetration depth is very limited. With the increase in the aspect ratio under the same penetration velocity, the jet breakup length decreases gradually. In general, the dimensionless corium surface area increases with the increase in the aspect ratio for the jets under the same penetration velocity.

Application of the machine learning technique for the development of a condensation heat transfer model for a passive containment cooling system

A condensation heat transfer model is essential to accurately predict the performance of the passive containment cooling system (PCCS) during an accident in an advanced light water reactor. However, most of existing models tend to predict condensation heat transfer very well for a specific range of thermal-hydraulic conditions. In this study, a new correlation for condensation heat transfer coefficient (HTC) is presented using machine learning technique. To secure sufficient training data, a large number of pseudo data were produced by using ten existing condensation models. Then, a neural network model was developed, consisting of a fully connected layer and a convolutional neural network (CNN) algorithm, DenseNet. Based on the hold-out cross-validation, the neural network was trained and validated against the pseudo data. Thereafter, it was evaluated using the experimental data, which were not used for training. The machine learning model predicted better results than the existing models. It was also confirmed through a parametric study that the machine learning model presents continuous and physical HTCs for various thermal-hydraulic conditions. By reflecting the effects of individual variables obtained from the parametric analysis, a new correlation was proposed. It yielded better results for almost all experimental conditions than the ten existing models.

COMSOL Simulation for Design of Induction Heating System in VULCAN Facility

The experimental facility VULCAN was setup to study the fuel-coolant interaction (FCI) phenomena in a postulated severe accident of light water reactors. The heating system is important for the facility to prepare molten material in a crucible. This article is concerned with the design of the heating system, which applies electromagnetic induction heating method. The COMSOL code was employed to simulate the induction heating characteristics of a graphite crucible under different current and frequency of the work coil (inductor). Given a frequency, the relationship between the crucible’s average temperature and the inductor’s current is obtained, which is instrumental to select the power supply of the induction heating system. Meanwhile, the skin effect of induction heating is analyzed to guide the choice of frequency and inductor of the heating system. According to the simulation results, the induction heating system of frequency 47 kHz is suitable for the experiment, with a good agreement in temperature between the measured and the predicted.

Corium Experimental Thermodynamics: A Review and Some Perspectives

More than 30 years ago a specialist meeting was held at Joint Research Center Ispra (Italy) from 15 to 17 January 1990 to review the current understanding of chemistry during severe accidents in light water reactors (LWR). Let us consider that, at the end of the 1980s, thermodynamics introduced in the severe accident codes was really poor. Only some equilibrium constants for a few simple reactions between stoichiometric compounds were used as well as some simple correlations giving estimates of solidus and liquidus temperatures. In the same time, the CALPHAD method was developed and was full of promise to approximate the thermodynamic properties of a complex thermochemical system by the way of a critical assessment of experimental data, a definition of a simple physical model and an optimisation procedure to define the values of the model parameters. It was evident that a nuclear thermodynamic database had to be developed with that new technique to obtain quite rapidly prominent progress in the knowledge of thermochemistry in the severe accident research area. Discussions focused on the important chemical phenomena that could occur across the wide range of conditions of a damaged nuclear plant. The most pressing need for improved chemical models is identified with condensed phase mixtures to model the corium progression. This paper reviews more than 30 years of experimental data production in the field of corium thermodynamics. This work has been conducted through multiple international programs (EURATOM, ISTC, OECD) as well as through more specific studies conducted at the national scale. This research has been capitalised in specific databases such as NUCLEA and TAF-ID, databases developed at IRSN and at CEA, respectively, and are now used in degradation models of the severe accident simulation codes. This research is presented in this paper. In the conclusion, we outline the research perspectives that need to be considered in order to address today’s and tomorrow’s issues.

Uniform corrosion of FeCrAl cladding tubing for accident tolerant fuels in light water reactors

Iron-Chromium-Aluminum (FeCrAl) alloys are candidate materials for light water reactor accident tolerant fuel (ATF) rod cladding because of the unparalleled resistance of these alloys to attack by superheated steam in the case of a loss of coolant accident. Since FeCrAl alloys were never used before in a reactor environment, it was important to characterize their corrosion resistance under simulated reactor normal operation conditions. Tests were conducted for APMT and C26M alloys in three ~300°C water systems (two containing hydrogen and one containing oxygen). Results show that after 12 months of immersion the FeCrAl alloys lose a small amount of mass. The resistance to corrosion is mostly given by the development on the surface of a protective oxide mostly rich in Cr and Al. The mass loss was lower in APMT than in C26M because of the higher Cr content in the former. After 12 months residence in the ~300°C water, the FeCrAl specimens maintained their unequaled resistance to attack by 1200°C steam.

Influence of coupled factors on premixing and fragmentation of mild thermal interaction

After core melt jets drop into the water pool, they break up into several parts and premix with water, and subsequently fine-fragmentation may take place, leading to steam explosions. Besides, mild thermal interaction can also occur in fuel–coolant interactions (FCIs) during core melt accidents in light water reactors. However, this phenomenon is not fully understood owing to the various coupled mechanisms involved. Therefore, we conduct an experimental study to understand the influence of the underlying factors and thermal interaction mechanisms on the premixing and fragmentation processes of FCIs within a confined space. Based on the morphology method, the coupled effects of melt mass, water level, and melt superheating temperature are investigated, quantified in terms of jet breakup length, solidification time, and fragments size distribution. Based on the corresponding pressure and temperature histories, two types of influencing factors are studied in detail: the effect of melt properties coupled with the water level; and the effect of melt superheating temperature coupled with the water level. Finally, a new bubble dynamic phenomenon called the premix expansion is featured based on Rayleigh’s equation. These experimental data on mild thermal interaction are expected to be of significance in the development of models and validation of codes.

2D MPS method analysis of ECOKATS-V1 spreading with crust fracture model

Understanding the ex-vessel spreading behavior of molten core materials (corium) on the reactor containment floor is of great importance for designing, managing, and evaluating severe accidents of light water reactors. Among some simulated ex-vessel corium spreading experiments, temporal termination and restart of the flow (stop-and-go) were highlighted in the ECOKATS-V1 experiment, in which the simulant oxide melt (consisted of 41% of Al2O3, 24% of SiO2, 19% of CaO, and 16% of FeO) was released into a 1-D channel, and the spreading behavior seemed to have temporarily stopped and restarted following crust formation and breach at the spreading leading edge. The preceding study developed crust formation and breach models for the Lagrangian-based Moving Particle Semi-Implicit (MPS) method to analyze such spreading. However, the applicability of the models was limited due to numerical instability. In this study, the timestep control of the MPS method is revised by newly introducing particle number density control, which can efficiently reduce timestep to prevent instability when the crust is formed or breached in low viscous flow. The new method is successfully applied to ECOKATS-V1 spreading with different crust formations and breach conditions. Namely, with different Solidification Viscosity Threshold (SVT) and the Crust Fracture Stress Threshold (CFST) of the melt. The former determines the threshold viscosity above which the fluid is regarded as solid (crust). The latter determines the mechanical strength of the crust at the leading edge of the spreading front. There remain issues for future studies to further develop the method for more quantitative discussions.

Recent Advances in Protective Coatings for Accident Tolerant Zr-Based Fuel Claddings

Zirconium-based alloys have served the nuclear industry for several decades due to their acceptable properties for nuclear cores of light water reactors (LWRs). However, severe accidents in LWRs have directed research and development of accident tolerant fuel (ATF) concepts that aim to improve nuclear fuel safety during normal operation, operational transients and possible accident scenarios. This review introduces the latest results in the development of protective coatings for ATF claddings based on Zr alloys, involving their behavior under normal and accident conditions in LWRs. Great attention has been paid to the protection and oxidation mechanisms of coated claddings, as well as to the mutual interdiffusion between coatings and zirconium alloys. An overview of recent developments in barrier coatings is introduced, and possible barrier layers and structure designs for suppressing mutual diffusion are proposed.

Numerical simulation on jet breakup in the fuel-coolant interaction using smoothed particle hydrodynamics

In a severe accident of light water reactor (LWR), molten core material (corium) can be released into the wet cavity, and a fuel-coolant interaction (FCI) can occur. The molten jet with high speed is broken and fragmented into small debris, which may cause a steam explosion or a molten core concrete interaction (MCCI). Since the premixing stage where the jet breakup occurs has a large impact on the severe accident progression, the understanding and evaluation of the jet breakup phenomenon are highly important. Therefore, in this study, the jet breakup simulations were performed using the Smoothed Particle Hydrodynamics (SPH) method which is a particle-based Lagrangian numerical method. For the multi-fluid system, the normalized density approach and improved surface tension model (CSF) were applied to the in-house SPH code (single GPU-based SOPHIA code) to improve the calculation accuracy at the interface of fluids. The jet breakup simulations were conducted in two cases: (1) jet breakup without structures, and (2) jet breakup with structures (control rod guide tubes). The penetration depth of the jet and jet breakup length were compared with those of the reference experiments, and these SPH simulation results are qualitatively and quantitatively consistent with the experiments.

Tolerant fuel for VVER reactors

The main factor of destruction of fuel rods in accidents with loss of coolant is associated with the vapor-zirconium reaction occurring between the fuel rod shell and the coolant (water). Improving the reliability of fuel cells can be obtained by modifying or replacing the fuel shell, materials that do not interact with the coolant during normal operation and in emergencies. The loss of coolant accident is a design-basis accident in light water reactors. The postulated accident requires an analysis of the double guillotine break in a main primary coolant pipe, which allows the coolant to freely discharge out of the primary system into the containment of the building. In the present work, the heat stored in the active zone of the reactor, the heat capacity of the fuel and the shell temperatures are determined for ATF materials U-10Mo and U3Si2. The calculated values are compared with the usually used fuel UO2. The results showed beneficial effect of using ATF materials. The above quantities are decreased dramatically for ATF, promising to replace UO2 with ATF fuels. This reduces the probabilities of accidents and oxidation in the nuclear reactors. The calculations are done for nuclear power plant, type VVER-1200.

Modeling of the corium crust of a stratified corium pool during severe accidents in light water reactors

During Severe Accidents (SAs) in Light Water Reactors (LWRs), the core heat up due to the lack of fuel cooling in the Reactor Pressure Vessel (RPV) may lead to the formation of a so called corium pool (molten core oxidic and metallic liquid materials) relocating in the lower head of the RPV. Within the framework of SA, one strategy to minimize risks of containment failure is the In Vessel Retention (IVR) strategy. To demonstrate the feasibility of such a strategy and to ensure the vessel integrity during the accident, one of the key points is to correctly evaluate the thermal load applied on the RPV wall by the lower head corium pool. Among all the complex coupled transient phenomena taking place in the RPV, the possible solidification of the corium pool, forming the so-called corium crust, has been shown by experimental and numerical results to be a key phenomenon for the IVR strategy. However, in most SA codes, the corium crust is only taken into account as a stationary model playing the role of a thermal resistance between the corium pool and the RPV wall. In this paper, a transient corium crust model is presented which explicitly takes into account the mass, temperature, composition and thermodynamic properties of the crust and pool material. Numerical results, in the context of the IVR strategy, highlight the impact of the transient corium crust modeling in comparison to a stationary modeling. The proposed model opens new perspective of modeling such as the dissolution of the crust and/or its thermo-mechanical behavior.

Mechanical failure of fresh nuclear grade iron–chromium–aluminum (FeCrAl) cladding under simulated hot zero power reactivity initiated accident conditions

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens’ outer surfaces. Experiments were conducted at approximately 275 °C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8–3.4%, corresponding to quasi-static energy depositions of approximately 110–260 calories per gram UO2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

Evaluations of the performance of multi-metallic layered composite cladding for the light water reactor accident tolerant fuel

Multi-metallic layered composite (MMLC) fuel cladding is designed to survive a longer period of time during accident scenarios for light water reactor (LWR) fuel. One proposed MMLC cladding concept joining Zircaloy to steel via interaction barrier layers will increase corrosion resistance, but result in a neutron absorption penalty. This cladding could also result in negative impacts due to mismatches between the mechanical behavior of its layers. In this study, the mechanical performance of the MMLC has been evaluated in two ways: small scale thermo-mechanical simulations and full-length fuel rod simulations. Both were carried out using the BISON fuel performance code. The small scale simulations predicted a reduction in residual stress and elastic deformation compared with standard Zircaloy cladding. The fuel rod simulations predicted decreased creep strain, larger pellet-clad gaps, and possible plastic deformation under long time use when compared to Zircaloy and stainless-steel cladding alone. The performance of various gap widths and layer thicknesses were compared to assist in the design of MMLC cladding; we recommend an MMLC with a Zircaloy/steel layer thickness ratio of at least 1.75 and an initial gap width of 40 μm.

A modified theta projection model for creep behavior of RPV steel 16MND5

During a hypothetical severe accident of light water reactors, the reactor pressure vessel (RPV) could fail due to its creep under the influence of high-temperature corium. Hence, modelling of creep behavior of the RPV is paramount to reactor safety analysis since it predicts the transition point of accident progression from in-vessel to ex-vessel phase. In the present study we proposed a new creep model for the classical French RPV steel 16MND5, which is adapted from the “theta-projection model” and contains all three stages of a creep process. Creep curves are expressed as a function of time with five model parameters θi(i=1−4 and m). A model parameter dataset was constructed by fitting experimental creep curves into this function. To correlate the creep curves for different temperatures and stress loads, we directly interpolate the model’s parameters θi(i=1−4 and m) from this dataset, in contrast to the conventional “theta-projection model” which employs an extra single correlation for each θi(i=1−4 and m), to better accommodate all experimental curves over the wide ranges of temperature and stress loads. We also put a constraint on the trend of the creep strain that it would monotonically increase with temperature and stress load. A good agreement was achieved between each experimental creep curve and corresponding model’s prediction. The widely used time-hardening and strain-hardening models were performing reasonably well in the new method.

Comparative evaluation of response reliability during accidents in light water reactors and fast reactors

Comparative evaluation of response reliability during accidents in light water reactors and fast reactors

Mechanism of loop seal clearance occurring at U-shaped pipe with air-water two-phase flow

Main purposes of the LOCAL (LOop seal Clearance separate effect test of ATLAS) facility are to investigate the effect of cross-over leg (COL) diameter at the loop seal clearance (LSC), to evaluate the two-phase flow regime through high speed visualization, and to suggest the mechanistic models for the LSC prediction at Small Break-Loss of Coolant Accident (SB-LOCA) in Light Water Reactors (LWRs). In this paper, we introduce a key mechanism of the onset of the LSC occurring at U-shaped COL with air-water two-phase flow condition. The onset of LSC is coupled with the flow regime transition at vertical upward and horizontal section of COL. Slug flow with single Taylor bubble results in decrease of pressure difference at vertical upward section and causes rapid flow regime transition at horizontal section from the stratified to the slug. The critical velocity of gas leading to flow regime transition at horizontal section can explain the onset of LSC and high speed visualization can provide the clear understanding of LSC phenomena at the local point of view.

Analysis of natural circulation behaviors and flow instabilities of passive containment cooling system design for advanced PWR using MARS-KS code

Several design basis accidents such as loss of coolant accident (LOCA) and main steam line break (MSLB) of pressurized light water reactor could threaten the integrity of containment building by increasing containment temperature and pressure. Although containment spray systems (CSS) have been installed for the depressurization of containment against the accidents, iPower nuclear power plant requires additional passive safety system design to provide for malfunction of CSS. Therefore, passive containment cooling system (PCCS), which consists of passive containment cooling tank (PCCT), heat exchanger modules, and connection pipelines, is under development. The natural circulation behavior of the PCCS is an important hydraulic phenomenon for the guarantee of decay heat removal capacity, required to satisfy the safety acceptance criteria. Especially, flow instabilities, which could threaten the structural integrity and enlarge the performance uncertainty of the system, have been reported for multi-channel heat exchangers operating with natural circulation like present PCCS design of iPower reactor. In this study, the natural circulation behaviors of iPower PCCS according to various operating conditions and design parameters (elevation of returning pipeline) were analyzed by MARS-KS code to observe the flow instability phenomenon inside the system. Two types of flow instabilities, flashing-induced instability and density wave oscillation were observed in the PCCS with different heights of returning pipeline. Through additional analysis on flow stability map of PCCS design, effects of mitigation strategies (system pressure and flow resistance) on natural circulation behavior of the system were quantified. The analysis results provide physical insight on possible flow instability and its mitigation strategies, that would be crucial information on designing natural circulation-driven multi-channel heat exchanging system consisting of more than two heat exchanger modules.

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

Oxidation kinetics and microstructural evolution of cold sprayed Cr coatings on Zircaloy-4 at 1130–1310 °C in flowing steam at atmospheric pressure have been studied. The study is aimed at understanding of the response of Cr coated Zr-alloy under a steam environment and high temperatures pertinent to design basis accidents (DBAs) and beyond design basis accidents (BDBAs) in light water reactors (LWRs). Surface morphology, microstructure, and phases of post-oxidation test samples were characterized using Scanning Electron Microscopy (SEM), x-ray diffraction (XRD), and Scanning Transmission Electron Microscopy (STEM). Growth kinetics of the Cr-oxide scale and interdiffusion layers between the Cr coating and the Zr-alloy substrate were quantified from cross-sectional SEM images. Cross-sectional analysis showed that the Cr coatings offered a 50 times reduction in oxidation rate over bare Zircaloy-4 in a 1310 °C steam environment. Oxidation kinetics at 1130 °C followed parabolic law (i.e., n ∼ 0.5 in power law kinetics) but at 1230 °C and 1310 °C the value of n was suppressed to below 0.5 possibly due to the volatilization of Cr species at the two highest temperatures. Interdiffusion at coating/substrate interface resulted in formation of a brittle Cr2Zr or Zr(Fe, Cr)2 intermetallic compound layer on the order of micrometer in thickness and scattered Cr-rich precipitates were observed well below the interface within the Zr-alloy substrate after cooling. These experimental results could provide data to LWR system simulation codes for better estimation of coping time in the event of accidents.