Experimental study on hydrogen behavior at a subcompartment in the containment building

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of the nuclear containment building after Three Miles Island (TMI) accidents. Up to now, many experiments have been conducted to estimate the distribution of hydrogen during accidents in nuclear power plants. In this study, local hydrogen behavior has been experimentally investigated in a cylindrical multi-subcompartment mixing chamber of the SNU (Seoul National University) hydrogen mixing facility, measuring the local concentration in various conditions and mixture injection locations. Hydrogen is simulated by helium in the experiments. Results showed remarkably different local behavior of helium in experiments of several conditions, and the local analysis for hydrogen concentration rather than the lumped compartment analysis, used widely in most plants, would be important to ensure the equipment survivability or to determine the positions of ignitors.     

Improvement of HYCA3D code and experimental verification in rectangular geometry

Concerns about the local hydrogen behavior in a nuclear power plant (NPP) containment during severe accidents have increased with the 10CFR50.34(f) regulation after TMI accident. Consequently, investigations on the local hydrogen behaviors under severe accident conditions were required. An analytical model named HYCA3D was developed at Seoul National University (SNU) to predict the thermodynamics and the three dimensional behavior of a hydrogen/steam mixture, within a subdivided containment volume following hydrogen generation during a severe accident in NPPs. In this study, the HYCA3D code was improved with a steam condensation and spray model, and verified with hydrogen mixing experiments executed in a SNU rectangular mixing facility. Helium was used to simulate hydrogen in both the calculations and the experiments. The calculation results show good agreement with the experimental data.     

严重事故下核电站安全壳内氢气分布及控制分析

使用安全壳分析程序CONTAIN计算分析了百万千瓦级压水堆核电站严重事故下安全壳内的氢气浓度分布.分别对一回路冷段大破口失水(LB-LOCA)叠加应急堆芯冷却系统(ECCS)失效(不包括非能动的安注箱)事故和全厂断电(SBO)叠加汽轮机驱动的应急给水泵失效事故两个严重事故序列进行了计算.计算结果表明,不同严重事故下,安全壳各隔间对氢气控制系统的要求不同.氢气控制系统的设计必须满足不同事故下的法规要求,提高电站的安全性.     

Assessment of the GOTHIC code for prediction of hydrogen flame propagation in small scale experiments

With the rising concerns regarding the time and space dependent hydrogen behavior in severe accidents, the calculation for local hydrogen combustion in compartment has been attempted using CFD codes like GOTHIC. In particular, the space resolved hydrogen combustion analysis is essential to address certain safety issues such as the safety components survivability, and to determine proper positions for hydrogen control devices as e.q. recombiners or igniters. In the GOTHIC 6.1b code, there are many advanced features associated with the hydrogen burn models to enhance its calculation capability.In this study, we performed premixed hydrogen/air combustion experiments with an upright, rectangular shaped, combustion chamber of dimensions 1 m × 0.024 m × 1 m. The GOTHIC 6.1b code was used to simulate the hydrogen/air combustion experiments, and its prediction capability was assessed by comparing the experimental with multidimensional calculational results. Especially, the prediction capability of the GOTHIC 6.1b code for local hydrogen flame propagation phenomena was examined. For some cases, comparisons are also presented for lumped modeling of hydrogen combustion. By evaluating the effect of parametric simulations, we present some instructions for local hydrogen combustion analysis using the GOTHIC 6.1b code. From the analyses results, it is concluded that the modeling parameter of GOTHIC 6.1b code should be modified when applying the mechanistic burn model for hydrogen propagation analysis in small geometry.     

国产先进压水堆严重事故下氢气行为及控制系统分析

严重事故下的氢气控制是核电厂安全需要考虑的重要问题之一。采用一体化严重事故分析程序对国产先进压水堆核电厂进行系统建模,选取大破口触发的严重事故序列,对严重事故工况下的氢气产生情况及氢气控制系统的性能进行分析评价。结果表明:大破口事故序列下氢气的产生主要有两个阶段,分别是早期锆包壳与水反应产生氢气及堆芯熔融物迁移至下腔室产生氢气,其中燃料包壳的氧化是产氢的主要阶段,氢气释放时间较早,氢气产生速率较大。氢气控制系统的设计能够有效缓解可能的氢气风险,满足相关法规标准的安全要求,确保安全壳的完整性。     

Hydrogen and steam distribution following a small-break LOCA in large dry containment

The hydrogen deflagration is one of the major risk contributors to threaten the integrity of the containment in a nuclear power plant, and hydrogen control in the case of severe accidents is required by nuclear regulations. Based on the large dry containment model developed with the integral severe-accident analysis tool, a small-break loss-of-coolant-accident （LOCA） without HPI, LPI, AFW and containment sprays, leading to the core degradation and large hydrogen generation, is calculated. Hydrogen and steam distribution in containment compartments is investigated. The analysis results show that significant hydrogen deflagration risk exits in the reactor coolant pump （RCP） compartment and the cavity during the early period, if no actions are taken to mitigate the effects of hydrogen accumulation.     

基于CFD的安全壳局部隔间非能动氢气复合器布置方案研究

非能动氢气复合器用于压水堆核电厂严重事故条件下安全壳内氢气的消除。通过计算流体力学(CFD)方法能够给出事故条件下非能动氢气复合器周围三维流场和温度场的分布。基于CFD程序根据非能动氢气复合器消氢公式,计算非能动氢气复合器进出口的气体流量和气体组分,并作为非能动氢气复合器的边界条件,开展三维空间内非能动氢气复合器消氢速率和氢气分布情况研究。结果表明：简化的非能动氢气复合器模拟方案能很好地模拟非能动氢气复合器样机的消氢效果;对安全壳内局部隔间开展非能动氢气复合器消氢效果研究发现,在相同环境条件下,非能动氢气复合器布置在较高位置与布置在较低位置相比,布置在较高位置时,非能动氢气复合器具有更高的消氢速率,隔间整体氢气浓度较低,但是非能动氢气复合器布置在较高位置时出现隔间底部局部氢气聚集的情况。     

Improvements in a CFD code for analysis of hydrogen behaviour within containments

The use of CFD codes for the analysis of the hydrogen behaviour within NPP containments during severe accidents has been increasing during last years. In this paper, the adaptation of a commercial multi-purpose code to this kind of problem is explained, i.e. by the implementation of models for several transport and physical phenomena like: steam condensation onto walls in presence of non-condensable gases, heat conduction, fog and rain formation, material properties and criteria for assessing the hydrogen combustion regime expected. The code has been validated against several experiments in order to verify its capacity to simulate the following phenomena: plumes, mixing, stratification and condensation. Moreover, two tests in an integral large enough experimental facility have been simulated, showing that the well-mixed and stratified conditions of the test were reproduced by the code. Finally, an example of a plant application demonstrates the ability of the code in this kind of problems.     

Multi-compartment hydrogen deflagration experiments and model development

Hydrogen deflagration in a multi-compartment containment geometry can produce drastic flame accelerations and local pressure build-ups much higher than expected from single-compartment experiments in the same concentration range. This is attributed to the so-called jet ignition effect occurring downstream of the vent openings in the partition walls inside the containment building. The various parameters affecting jet ignition have been investigated in a comprehensive experimental program. Modelling of the multi-compartment deflagration effects has been achieved with different degrees of perfection by lumped-parameter codes, by distributed-parameter code simulation, and by a combination of both.     

Experimental examinations of the 2D-thermal-hydraulic code FRECON for single phase natural convection flows in complex structures

An experimental program for the verification of the 2D-thermal-hydraulic code FRECON is being carried out. This code serves for the simulation of single phase natural convective flows in complex structures. As a special application it has already been used in the analyses of severe LWR accidents. For this, the code has been coupled with the KESS code system describing the core behaviour during such an accident. While strong improvements have been reached meanwhile concerning the FRECON code itself as well as its application to the reactor problem, experimental verification has not yet been performed. Such a verification is strongly required because of the basic assumption of a quasi-continuum approach for the solid structures in FRECON and the problem of describing local characteristic properties of the structures within this approach.In contrast to integral experiments aiming to simulate the real reactor situation as closely as possible, the present experiments aim to check specific structures and aspects in detail. Thus, in a first step, natural convective flows in a rod bundle with imposed solid temperature profiles are being analysed. In addition to temperature measurements in the solid as well as in the gas, the flow patterns are visualized by means of light-weight glass spheres as tracers, illuminated with a laser-light-sheet. After recording the tracers by a video system the analysis of the flow patterns is made either directly at the TV-monitor or by using the digital image processing technique PIDV (Particle Image Displacement Velocimetry). First comparisons between code calculations and experimental results support the code development in general, but need further refinement.     

Analysis of hydrogen risk mitigation with passive autocatalytic recombiner system in CPR1000 NPP during a hypothetical station blackout

Hydrogen safety has attracted extensive concern in severe accident analysis especially after the Fukushima accident. In this study, a similar station blackout as happened in Fukushima accident is simulated for CPR1000 nuclear power plant (NPP) model, with the computational fluid dynamic code GASFLOW. The hydrogen risk is analyzed with the assessment of efficiency of passive autocatalytic recombiner (PAR) system. The numerical results show that the CPR1000 containment may be damaged by global flame acceleration (FA) and local detonation caused by hydrogen combustion if no hydrogen mitigation system (HMS) is applied. A new condensation model is developed and validated in this study for the consideration of natural circulation flow pattern and presence of non-condensable gases. The new condensation model is more conservative in hydrogen risk evaluation than the current model in some compartments, giving earlier starting time of deflagration to detonation transition (DDT). The results also indicate that the PAR system installed in CPR1000 could prevent the occurrence of the FA and DDT. Therefore, HMS such as PAR system is suggested to be applied in NPPs to avoid the radioactive leak caused by containment failure.     

Analysis of steam and hydrogen distributions with PAR mitigation in NPP containments

The 3-D-field code, GASFLOW is a joint development of Forschungszentrum Karlsruhe and Los Alamos National Laboratory for the simulation of steam/hydrogen distribution and combustion in complex nuclear reactor containment geometries. GASFLOW gives a solution of the compressible 3-D Navier–Stokes equations and has been validated by analysing experiments that simulate the relevant aspects and integral sequences of such accidents. The 3-D GASFLOW simulations cover significant problem times and define a new state-of-the art in containment simulations that goes beyond the current simulation technique with lumped-parameter models. The newly released and validated version, GASFLOW 2.1 has been applied in mechanistic 3-D analyzes of steam/hydrogen distributions under severe accident conditions with mitigation involving a large number of catalytic recombiners at various locations in two types of PWR containments of German design. This contribution describes the developed 3-D containment models, the applied concept of recombiner positioning, and it discusses the calculated results in relation to the applied source term, which was the same in both containments. The investigated scenario was a hypothetical core melt accident beyond the design limit from a large-break loss of coolant accident (LOCA) at a low release location for steam and hydrogen from a rupture of the surge line to the pressurizer (surge-line LOCA). It covers the in-vessel phase only with 7000 s problem time. The contribution identifies the principal mechanisms that determine the hydrogen mixing in these two containments, and it shows generic differences to similar simulations performed with lumped-parameter codes that represent the containment by control volumes interconnected through 1-D flow paths. The analyzed mitigation concept with catalytic recombiners of the Siemens and NIS type is an effective measure to prevent the formation of burnable mixtures during the ongoing slow deinertization process after the hydrogen release and has recently been applied in backfitting the operational German Konvoi-type PWR plants with passive autocatalytic recombiners (PAR).     

严重事故下安全壳卸压箱隔间氢气浓度场模拟

根据MELCOR程序对全厂断电诱发的严重事故下安全壳内各隔间的氢气浓度分布的计算结果,参考美国联邦法规关于氢气控制和风险分析的标准,分析安全壳内氢气的燃烧风险。结果表明：安全壳内平均氢气浓度不会导致整体性氢气燃烧,但存在局部燃烧的风险。通过CFD程序对氢气浓度较高的卸压箱隔间进行氢气释放和空间气体流动过程的模拟,得到更细致的卸压箱隔间内氢气浓度场分布,给出氢气聚集区域的准确位置,为采取严重事故缓解措施,设计氢复合器布置方案提供了参考依据。     

Accident analysis in LMFBR fuel rods by the fenht code

This paper describes the fenht code capability related to the safety requirements in nuclear plants. The code solves the non-linear transient heat-transfer problem for the fuel element of a nuclear reactor, in order to simulate any accidental, operational and emergency power transient with arbitrary initial conditions. The temperature distribution in the fuel, gap and cladding is obtained by a finite-element technique based on minimizing the thermal potential with respect to the temperature vector at the nodes of the finite elements. The non-linear differential matricial equation is linearized by an iterative procedure and solved by the Crank—Nicholson method. Also the thermoelastic stresses in the cladding are valued, by the usual Hooke's law. The code has been applied to the analysis of two reference accidents (incidental power transients) occurring in a liquid-metal fast-breeder reactor (LMFBR); the results are reported and briefly discussed.     

Studies validating a sodium purification system for fast reactors

The sources of impurities entering the sodium in fast reactors were investigated. The analysis showed that oxygen and hydrogen can be removed from the sodium by using cold traps in all operating regimes of a nuclear power plant as well as hot traps. An operating regime preventing hydrogen accumulation in the first-loop cold trap is proposed for a system purifying the first two loops. A computer code for calculating the impurity mass transfer is perfected. Test calculations showed that the procedure developed and the code are both serviceable. The deviation of the computational results from the experimental data is about 30% on average. For a built-in purification system, it is essential to develop a cold trap with a large impurity capacity. It is shown on the basis of experiments that such cold traps can in principle be developed. Thermohydraulic and mass-transfer codes must be developed in order to realize this possibility.     

Synopsis and outcome of the QUENCH experimental program

The paper gives an overview of the main outcome of the QUENCH program launched in 1997 at the Karlsruhe Institute of Technology (KIT), formerly Karlsruhe Research Center (FZK). The research program comprises bundle experiments as well as complementary separate-effects tests. The focus of the experiments performed from 1997 to 2009 was on scenarios of severe accidents whereas that of the future test program will be on large-break loss-of-coolant accidents (LOCA) in the frame of design-basis accidents, and debris coolability, in the frame of severe accidents. The major objective of the program is to deliver experimental and analytical data to support the development and validation of quench and quench-related models as used in code systems that model severe accident progression in light water reactors.So far, 15 integral bundle QUENCH experiments with 21-31 electrically heated fuel rod simulators of 2.5 m length have been conducted. The following parameters and their influence on bundle degradation and reflood have been investigated: degree of pre-oxidation, temperature at initiation of reflood, flooding rate, influence of neutron absorber materials (B₄C, AgInCd), air ingress, and influence of the type of cladding alloy.In six tests, reflooding of the bundle led to a temporary temperature excursion driven by runaway oxidation of zirconium alloy components and resulting in release of a significant amount of hydrogen, typically two orders of magnitude greater than in those tests with “successful” quenching in which cool-down was rapidly achieved. Considerable formation, relocation, and oxidation of melt were observed in all tests with escalation. The temperature boundary between rapid cool-down and temperature escalation was typically in the range of 2100-2200 K in the “normal” quench tests, i.e. in tests without absorber and/or steam starvation. Tests with absorber and/or steam starvation were found to lead to temperature escalations at lower temperatures.All phenomena occurring in the bundle tests have been investigated additionally in parametric and more systematic separate-effects tests. Oxidation kinetics of various cladding alloys, including advanced ones, have been determined over a wide temperature range (873-1773 K) in different atmospheres (steam, oxygen, air, and their mixtures). Hydrogen absorption by different zirconium alloys was investigated in detail, recently also using neutron radiography as non-destructive method for determination of hydrogen distribution in claddings. Furthermore, degradation mechanisms of absorber rods including B₄C and AgInCd as well as the oxidation of the resulting low-temperature melts have been studied. Steam starvation was found to cause deterioration of the protective oxide scale by thinning and chemical reduction.The most recent topic of the QUENCH program has been investigation of the behavior of advanced cladding materials (ACM) in comparison with the classical Zircaloy-4. Although separate-effects tests have shown some differences in oxidation kinetics, the influence of the various cladding alloys on the integral bundle behavior during oxidation and reflooding was only limited.     

A study on the hydrogen recombination rates of catalytic recombiners and deliberate ignition

A study is being carried out by the Department of Nuclear and Mechanical Constructions at the University of Pisa on catalytic recombiners and on deliberately induced weak deflagration. These are the most practical methods for recombining hydrogen released into large nuclear containments during severe accidents. The recombination rates of different types of catalytic device were obtained from a thorough analysis of published experimental data. The main parameter that affects the effectiveness of these devices seems to be the molar density of the deficiency reactant rather than its volumetric concentration. The recombination rate of weak deflagrations in vented compartments has been assessed with experimental tests carried out in a small-scale glass vessel. Through a computerized system of analysis of video recordings of the deflagrations, the flame surface and the burnt gas volume were obtained as functions of time. These values of flame surface and burnt gas volume were used as inputs for a computer code to calculate the recombining rate, the burning velocity and the pressure transient in the experimental test. The code is being validated with a methodology principally based on a comparison of the measurements of pressure with the calculated values.     

Model to simulate high temperature oxidation kinetics of zircaloy-4: Parabolic and linear behaviour

A calculation code to simulate parabolic and linear behaviour of Zircaloy-4 oxidation between 600 and 862°C in water vapour was developed. This code consists of solving the diffusion equations by the finite-difference method. This method in its explicit version was employed previously, but this type of calculation becomes impracticable with present-day computers when attempts are made to simulate long-term experiments (24 h). This is why the implicit finite-difference method is proposed here: this method has the advantage of drastically reducing the calculation time. The code allowed us to calculate the relationship between the oxygen mass in theα-phase to the total oxygen mass, the oxide thickness and the diffusion profile of oxygen in the α-phase.The results obtained with the model are compared with experimental data existing in the literature for Zircaloy-4, although it could be applied to pure zirconium or other zirconium alloys if more experimental data were available. The singular behaviour of the diffusion profiles in the α-phase during linear kinetics is particularly analyzed.This work is part of a programme to study the oxide-metal interface movement during vapour oxidation of Zircaloy-4 subjected to temperature transients. Knowledge of this is of vital importance for evaluating mechanical properties of fuel claddings during possible loss of coolant accidents in nuclear power reactors.     

核热推进系统氢气物性及流动换热模型分析

为开展关于核热推进反应堆堆芯的稳态热工水力计算,基于现有针对压水堆的系统分析程序,添加了氢气的物性模型及流动换热和摩擦阻力关系式,并采用公开文献中的数据进行验证。结果表明采用上述模型计算得到的结果与参考值符合较好,二次开发的程序适用于氢气的流动换热计算。针对一种折流式核热推进反应堆堆芯,使用该系统程序建模并计算,得到了堆芯的流量、焓升等分布情况。研究结果表明,对于折流式核热推进反应堆,内外堆芯燃料元件之间的导热会增强堆芯释热不均,对堆芯的稳态热工水力特性有较大影响,堆芯物理方案的设计应结合热工水力方面的计算。本研究可为核热推进系统内氢气流动换热计算提供借鉴。     

Scaling and extrapolation of hydrogen distribution experiments

Important physical processes within a containment system, which govern the long-term behaviour under severe accident conditions have been analyzed with respect to the scaling of relevant test rigs. This analysis has been performed under contract with the Commission of the European Communities on the basis of the equations processed within a typical long term containment analysis code like the CONTAIN code.An improved set of conservation equations for the involved components (air, hydrogen, vapor etc.), has been subject to a detailed dimensional analysis. This resulted in a set of dimensionless parameter groups which determine the similarity requirements necessary to warrant simple extrapolation of the results to full size reactor containments.The discussion of the physical meaning of the parameter groups showed that steady-state and transient events could possibly require different similarity criteria. For the establishment of a long term natural circulation pattern the strong coupling with a preceding LOCA blowdown process has been demonstrated. Inevitable distortions of mixing effects for hydrogen may be expected for small scaled experiments, if it is not possible to perform experiments with a thoroughly scaled distribution of heat sources and heat sinks. The hydrogen source terms into the containment must be properly related to a volumetric scaling concept. The understanding of the transport process is strongly dependent on a detailed analysis of the Strouhal Number to avoid misinterpretation of the results. Only limited overall similarity seems achievable.The implications of the similarity requirements with the applied empirical correlations or constants in combination with a chosen nodalisation concept have been addressed. Code verification is based on comparison of selected measured with calculated parameters which leads to conclusions concerning an optimal choice of correlations and the proper nodalisation of the test rigs. Heat transfer correlations and local flow resistance determination are important empirical elements for a successful reanalysis of experiments. Facility dependent code verification can only be avoided, if the dimensional dependence of the empiricism in code application is assessed.