Experimental simulation of corium dispersion phenomena in direct containment heating

In a direct containment heating (DCH) accident scenario, the degree of corium dispersion is one of the most significant factors responsible for the reactor containment heating and pressurization. To study the mechanisms of the corium dispersion phenomenon, a DCH separate effect test facility of 1:10 linear scale for Zion PWR geometry is constructed. Experiments are carried out with air-water and air-woods metal simulating steam and molten core materials. The physical process of corium dispersion is studied in detail through various instruments, as well as with flow visualization at several locations. The accident transient begins with the liquid jet discharge at the bottom of the reactor pressure vessel. Once the jet impinges on the cavity bottom floor, it immediately spreads out and moves rapidly to the cavity exit as a film flow. Part of the discharged liquid flows out of the cavity before gas blowdown, and the rest is subjected to the entrainment process due to the high speed gas stream. The liquid film and droplet flows from the reactor cavity will then experience subcompartment trapping and re-entrainment. Consequently, the dispersed liquid droplets that follow the gas stream are transported into the containment atmosphere, resulting in containment heating and pressurization in the prototypic condition. Comprehensive measurements are obtained in this study, including the liquid jet velocity, liquid film thickness and velocity transients in the test cavity, gas velocity and velocity profile in the cavity, droplet size distribution and entrainment rate, and the fraction of dispersed liquid in the containment building. These data are of great importance for better understanding of the corium dispersion mechanisms.     

Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TM1–2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating.     

全厂断电时安全壳直接加热的概率研究

安全壳直接加热(DCH)是导致安全壳早期失效的潜在因素,本文应用基于风险导向的事故分析方法(ROAAM),对堆芯碎片中UO_2的质量和Zr的氧化份额的概率密度分布抽样,对安全壳直接加热模型TCE(Two-cell Equilibrium)编程,将抽样结果带人TCE模型中计算,得到安全壳压力峰值的累积概率分布和安全壳失效概率,研究压水堆全厂断电始发事故下轴封破口面积不同的情况对下封头失效后安全壳压力峰值的影响。其中TCE模型的输入数据由严重事故分析程序计算给出。     

High pressure corium melt quenching tests in FARO

First-of-a-kind experimental data on the quenching of large masses of corium melt of realistic composition when poured into pressurised water at reactor scale depths are presented and discussed. The tests involved 18 and 44 kg of a molten mixture 80 w% UO₂-20 w% ZrO₂, which were delivered by gravity through a nozzle of diameter 0.1 m to 1 m depth nearly saturated water at 5.0 MPa. The objective was to gain early information on the melt/water quench process previous to tests that will involve larger masses of melt (1.50 kg of mixtures UO₂---ZrO₂---Zr). Particularly, pressures and temperatures were measured both in the gas phase and in the water. The results show that significant quenching occurred during the melt fall stage with 30% to 42% of the melt energy transferred to the water. About two-thirds of the melt broke up into particles of mean size of the order of 4.0 mm. The remaining one-third collected still molten in the debris catcher but did not produce any damage to the bottom plate. The maximum downward heat flux was 0.8 MW m². The maximum vessel overpressurisation, i.e. 1.8 MPa, was recorded with 44 kg of melt poured into 255 kg of water and a gas phase volume of 0.875 m³. No steam explosions occurred.     

Closure of the direct containment heating issue for Zion

This paper, which was originally published in more detail (M.M. Pilch, M.D. Allen, D.L. Knudsen, D.W. Stamps and E.L. Tadios, Rep. NUREG/CR-6075, Supplement 1, 1994b (Sandia National Laboratories, Albuquerque, NM)), provides closure of the direct containment heating (DCH) issue for the Zion plant. It incorporates the comments and suggestions of the peer reviewers of NUREG/CR-6075 (M.M. Pilch, H. Yan, and T.G. Theofanous, Rep. NUREG/CR-6075, SAND93-1535, 1994a (Sandia National Laboratories, Albuquerque, NM)) and specifically includes assessments of four new splinter scenarios defined in working group meetings and modeling enhancements recommended by the working groups. In the four new scenarios, consistency of the initial conditions has been implemented by using insights from systems-level codes. was used to analyze three short-term station blackout cases with different leak rates. In all three cases, the hot leg or surge line failed well before the lower head and thus the primary system depressurized to a point where DCH was no longer considered a threat. However, these calculations were continued to lower head failure in order to gain insights that were useful in establishing the initial and boundary conditions. The most useful insights are that the reactor coolant system pressure is low at vessel breach, metallic blockages in the core region do not melt and relocate into the lower plenum, and melting of upper plenum steel is correlated with hot leg failure. The output was used as input to to assess the containment conditions at vessel breach. The containment-side conditions predicted by are similar to those originally specified in NUREG/CR-6075.The methodology originally developed in NUREG/CR-6075 (M.M. Pilch, H. Yan, and T.G. Theofanous, Rep. NUREG/CR-6075, SAND93-1535, 1994a (Sandia National Laboratories, Albuquerque, NM)) was used to analyze the new splinter scenarios. Some modeling enhancements in response to working group discussions were implemented for these analyses. The entrainment of hydrogen pre-existing in the atmosphere into a burning jet was examined more carefully. In addition, the impact of DCH-induced deflagrations on DCH loads was quantified. A new computational tool—the two-cell equilibrium—Latin hypercube sampling (TCE-LHS) code—was developed for this effort to perform Monte Carlo sampling of the scenario distributions. The TCE-LHS code was benchmarked against the original Scenario I calculations in NUREG/CR-6075 performed using the code, which is based on the method of discrete probability distributions. The results were in excellent agreement.The analyses of the new scenarios showed no intersection of the load distributions and the containment fragility curves, and thus the containment failure probability was negligible for each scenario. These supplemental analyses complete closure of the DCH issue for Zion.     

A two-cell equilibrium model for predicting direct containment heating

This paper discusses two adiabatic equilibrium models. Assessment and validation of the separate effects (kinetic) models and the parameters (i.e. particle size) that control them are not required. The first, a single-cell equilibrium model, places a true upper bound on direct containment heating (DCH) loads. This upper bound, when compared with the entire DCH database, often far exceeds experiment observations by a margin too large to be useful in reactor analyses. The single-cell model is used as a conceptual seed for a two-cell model. A two-cell equilibrium (TCE) model is developed that captures the dominant mitigating features of containment compartmentalization and the noncoherence of the entrainment and blowdown processes. The existing DCH database has been used to extensively validate the TCE model. DCH loads are shown to be insensitive to physical scale and details of the subcompartment geometry. A simple model is developed to predict the coherence of debris dispersal and reactor coolant system blowdown. The coherence ratio is independent of physical scale and only weakly dependent on cavity design.     

Comparison of and TCE calculations for direct containment heating of Surry

This paper presents the results of several CONTAIN code calculations used to model direct containment heating (DCH) loads for the Surry plant. The results of these calculations are compared with the results obtained using the two-cell equilibrium (TCE) model for the same set of initial and boundary conditions. This comparison is important because both models have been favorably validated against the available DCH database, yet there are potentially important modeling differences. The comparisons are to quantitatively assess the impact of these differences. A major conclusion of this study is that, for the accident conditions studied and for a broad range of sensitivity cases, the peak pressures predicted by both TCE and are well below the failure pressure for the Surry containment.     

Investigations into the physical phenomena and mechanisms that effect direct containment heating loads

Integral direct containment heating (DCH) experiment results are presented. The results are analyzed and discussed for the insights they have given into understanding the important physical phenomena and mechanisms that effect DCH loads to the containment. Particular attention is paid to (1) debris dispersal from the cavity and containment structure trapping, (2) hydrogen production and combustion, (3) the importance of difference in corium simulants used in integral DCH experiments and (4) corium debris quenching by flooded cavities. It is found that much has been learned about DCH phenomena that can be used for modeling and assessing potential containment loads.     

Analysis of aerosol deposition in PHEBUS containment during FPT-1 experiment

The international PHEBUS-FP programme was initiated in 1988 and performed by the French Institute de Protection et de Sûreté Nucléaire (IPSN) to investigate the key phenomena of severe water reactor accident. The main objective of the programme is to study the release, transport and retention of fission products in an in-pile facility under conditions representative of a severe accident in a light water reactor. The Lithuanian Energy Institute has joined the programme in 2005 and most of the efforts were directed to investigation of containment phenomena.This paper presents overview of the analyses performed to investigate aerosol transport and deposition phenomena in PHEBUS containment during FPT-1 test. A lumped parameter code COCOSYS was used for the analysis. Parametric analyses were performed to investigate the influence of aerosol density, solubility and diffusive boundary layer thickness on the deposition rate and deposition distribution of particles in PHEBUS containment. The performed analysis showed only minor influence of the selected parameters on the results of FPT-1. Also it was observed that the diffusive boundary layer thickness should not be defined too small.     

Modeling pressure suppression pool hydrodynamics in the ABWR containment

This paper focuses on the assessment of pressure suppression pool hydrodynamics in the advanced boiling water reactor (ABWR) containment under design-basis, loss-of-coolant accident (LOCA) conditions. The paper presents a mechanistic model for predicting various suppression pool hydrodynamics parameters. A phenomena identification and ranking table (PIRT) applicable to the ABWR containment pool hydrodynamics analysis is used as a basis for the development of the model. The highly ranked phenomena are represented by analytic equations or empirical correlations. The best estimate and several sensitivity calculations are performed for the ABWR containment using this model. Results of the sensitivity calculations are also presented that demonstrate the influence of key model parameters and assumptions on the pool hydrodynamics parameters. A comparison of model predictions to the results of the licensing analyses shows reasonable agreement. Comparison of the results of the proposed model to experimental data shows that the model predicted top vent clearance time, the pool swell height, and the bubble breakthrough elevation are within 10% of the data. The predicted pool surface velocity and the liquid slug thickness are within 30% of the measurements, which is considered adequate given the large uncertainties in the experimental measurements.     

Direct containment heating integral effects tests in geometries of European nuclear power plants

The DISCO test facility at Forschungszentrum Karlsruhe (FZK) has been used to perform experiments to investigate direct containment heating (DCH) effects during a severe accident in European nuclear power plants, comprising the EPR, the French 1300 MWe plant P’4, the VVER-1000 and the German Konvoi plant. A high-temperature iron–alumina melt is ejected by steam into scaled models of the respective reactor cavities and the containment vessel. Both heat transfer from dispersed melt and combustion of hydrogen lead to containment pressurization. The main experimental findings are presented and critical parameters are identified.The consequences of DCH are limited in reactors with no direct pathway between the cavity and the containment dome (closed pit). The situation is more severe for reactors which do have a direct pathway between the cavity and the containment (open pit). The experiments showed that substantial fractions of corium may be dispersed into the containment in such cases, if the pressure in the reactor coolant system is elevated at the time of RPV failure. Primary system pressures of 1 or 2 MPa are sufficient to lead to full scale DCH effects. Combustion of the hydrogen produced by oxidation as well as the hydrogen initially present appears to be the crucial phenomenon for containment pressurization.     

An approach to numerical simulation and analysis of molten corium coolability in a boiling water reactor lower head

This paper discusses an approach for application of the computational fluid dynamics (CFD) method to support development and validation of computationally effective methods for safety analysis, on the example of molten corium coolability in a BWR lower head. The approach consists of five steps designed to ensure physical soundness of the effective method simulation results: (i) analysis and decomposition of a severe accident problem into a set of separate-effect phenomena, (ii) validation of the CFD models on relevant separate-effect experiments for the reactor prototypical ranges of governing parameters, (iii) development of effective models and closures on the base of physical insights gained from relevant experiments and CFD simulations, (iv) using data from the integral experiments and CFD simulations performed under reactor prototypic conditions for validation of the effective model with quantification of uncertainty in the prediction results and (v) application of the computationally effective model to simulate and analyze the severe accident transient under consideration, including sensitivity and uncertainty analysis. Implementation of the approach is illustrated on a so-called effective convectivity model for simulation of turbulent natural convection heat transfer and phase changes in a decay-heated corium pool. It is shown that detailed information obtained from the CFD simulations are instrumental to ensure the effective models capture safety-significant local phenomena, e.g. the enhanced downward heat flux in the vicinity of a cooled control rod guide tube.     

LOCA steam condensation loads in BWR Mark II pressure suppression containment system

Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates ( 30 kg/m²·s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.     

The LOCA air-injection loads in BWR Mark II pressure suppression containment systems

Large-scale blowdown tests were conducted to investigate the thermal-hydrodynamic response of a boiling-water reactor (BWR) Mark II pressure suppression containment system to a postulated loss-of-coolant accident. This paper presents the test results on the early blowdown transients, where air in the drywell is injected into the pressure suppression pool and induces various hydrodynamic loads onto the containment pressure boundary and internal structures. The test data are compared to predictions by analytical models used for the licensing evaluation of the hydrodynamic loads to assess these models.     

Simulation of containment thermal-hydraulics in the Marviken Blowdown 16 experiment with ASTEC and CONTAIN codes

The boiling heavy water reactor Blowdown 16 experiment, which was performed in the Marviken experimental facility, was simulated with the ASTEC and CONTAIN codes. The main purpose of the work was the assessment of the codes for simulating thermal-hydraulic phenomena in a BWR containment at accident conditions. Simulated pressures, atmosphere temperatures and wetwell pool masses are compared to experimental measurements. The results show that both codes satisfactorily reproduced the overall containment thermal-hydraulic behaviour. The simulations also allow a more detailed understanding of the governing mechanisms during the performed experiment.     

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods.An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m³ and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1:64 compared to the 1200 MW PWR plant Biblis A.Up to now the test facility has been used for four trial runs and nine PWR LOCA experiments with single- and double-ended pipe ruptures of 100 mm dia. in a steam generator compartment and in the nozzle compartment. The initial conditions of the pressurized water in the coolant circuit before rupture were 140 bar and 290°C. About 0.1 sec after the rupture the flow rate at the site of rupture reaches its maximum of about 400 kg/sec (single-ended rupture) and 800 kg/sec (double-ended rupture). From the compartment where the rupture takes place a water-steam-air mixture streams through openings into the other compartments of the containment. Differential pressures between the compartments were measured with maximums of up to a few bar 0.15–0.5 sec after rupture, depending on the positions of rooms and transducers.Approximately 30–40 sec after rupture the blowdown has finished and the pressure in the containment has reached about 4–5 bar. The maximum pressure in a model containment is lower and the decrease of the pressure by condensation is faster than in a full-scale containment, due to the greater ratio of inner surface area to volume of a model containment. During blowdown the temperature of the containment atmosphere rises to about 150°C. Several minutes later the temperature of the concrete walls has increased non-uniformly causing considerable stress in the walls. Approximately 30 min after rupture measurements on the outside of the outer containment wall show a temperature-caused strain of about 30–60% of the maximum pressure-caused strain. A comparison between experiments and calculations shows discrepancies indicating the need for further development of calculational methods.     

Numerical simulation of the combined effects of plasma heating and neutron heating loads on the ITER first wall

To understand the combined effect of plasma heating and neutron heating loadings, the distributions of temperature, stress, and strain in different two-dimensional first wall panel models under normal ITER operation condition were simulated using finite element method. The maximum temperature occurs at the Be armor, and reaches 461 °C. High thermal stresses (in the range of 80-200 MPa) are found at the interface between the Be armor and the CuCrZr layer. The maximum thermal stress reaches 324 MPa in the SS316L cooling tube (20 mm diameter), exceeding its yield strength and resulting in a maximum strain of about 1.7% at the tube inner surface. These simulation results are useful for the design and operation of ITER.     

Experimental simulation study on start-up of the 5 MW nuclear heating reactor

An experimental simulation study on the start-up of a low temperature, natural circulation nuclear heating reactor (5 MW developed by the Institute of Nuclear Energy of Tsinghua University, Beijing) is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instability, namely geysering, flashing instability and low steam quality density wave instability on the start-up are described. The mechanism of flashing instability, which has never been well studied in this field, is especially interpreted. Based on the study of these instabilities, it is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: increasing of initial pressure by means of a noncondensable gas (N₂), start-up of the reactor at this pressurized condition (single-phase regime operation), and transition to a lower pressure, boiling operation. Three transition methods are discussed. As a result of these studies, the method of transition with low heat flux and low inlet subcooling is proposed. A stable start-up process of the 5MW reactor is achieved by careful selection of the thermohydraulic parameters.     

燃料棒内氦气压力无损检测的数值模拟

用数值模拟的方法对一种用温度测量进行燃料棒内氦气压力无损检测的方法进行了模拟计算,该方法的基本原理是氦气在不同压力下具有不同的传热特性.采用二维差分模型,编制了用于计算燃料棒内瞬态二维温度分布的程序RODTRAN,计算模拟具有不同氦气压力下元件棒在一端固定热源温度加热后所形成的温度分布随时间的响应特性.通过用RODTRAN程序计算各种不同压力情况下的燃料棒动态传热特性,发现利用x=14.5 cm处的包壳表面升温速率可以推算燃料棒内的氦气压力,氦气压力的测量精度可小于5%,也就是说可以区分1.9 MPa和2.0 MPa的压力差别,此时的最大温差可达0.5℃,同时也发现压紧弹簧段的温度响应比铀芯块段要快.所得到的结论,可为氦气压力无损检测装置的设计提供很好的技术支持.     

Chord length sampling correction analysis for dispersion fuel in Monte Carlo simulation

Dispersion fuels,knowned for their excellent safety performance,are widely used in advanced reactors,such as high-temperature gas-cooled reactors.Compared with deterministic methods,the Monte Carlo method has more advantages in the geometric modeling of stochastic media.The explicit modeling method has high computational accuracy and high computational cost.The chord length sampling(CLS)method can improve computational efficiency by sampling the chord length during neutron transport using the matrix chord length's probability density function.This study shows that the excluded-volume effect in realistic stochastic media can introduce certain deviations into the CLS.A chord length correc-tion approach is proposed to obtain the chord length correction factor by developing the Particle code based on equivalent transmission probability.Through numerical analysis against reference solutions from explicit modeling in the RMC code,it was demonstrated that CLS with the proposed correction method provides good accuracy for addressing the excluded-volume effect in realistic infinite stochastic media.