MELCOR/MACCS simulation of the TMI-2 severe accident and initial recovery phases, off-site fission product release and consequences

MELCOR has become the preferred code of the Swiss nuclear industry and of PSI for severe accident analysis, on account of its integrated systems-level approach and validation against experiments and more detailed codes, while MACCS is commonly used by safety authorities for independent assessment of off-site consequences, in particular health effects. The present work arises out of a programme to assess MELCOR independently using empirical data consistent with the recommendations of the OECD/CSNI validation matrix for core degradation codes. The MELCOR 1.8.5RD calculations are based on a model for phases 1 and 2 provided by the code developers but with a simplified thermal hydraulic noding in certain regions and the inclusion of a simple representation of the fission product release and transport pathways. The model has also been extended to simulate phases 3, 4, and the continuing initial period of core recovery and stabilisation. These calculations are a first attempt to demonstrate a MELCOR–MACCS capability to simulate the whole plant accident sequence beyond phase 4, including the containment response and off-site consequences arising from fission product release from the containment. Emphasis is placed on the overall accident evolution and whole plant response, rather than the detailed behaviour. Results are compared with observed and deduced data for the major accident signatures and rough estimates for exposure based on off-site monitoring. The results provide a good basis for the NPP analysis foreseen.     

Fluid dynamic analysis of a catalytic recombiner to remove hydrogen

Catalytic reacting surfaces in recombiners are a reliable way to remove hydrogen as well as other burnable gases like CO in a passive way from the containment atmosphere of a nuclear power plant (NPP) during an accident. Industrial mature designs are ready to be installed in large dry containments to act as a mitigation measure preferably in the case of severe accidents. Experiments have been carried out to study manifold aspects of recombiners like the efficiency of hydrogen removal, start-up conditions, poisoning, oxygen starvation, steam and water impact and others. Mostly the global behaviour of a given device in a larger environment has been investigated in order to demonstrate the effectiveness and to facilitate the derivation of simplified models for long-term severe accident analyses. There are a number of reasons to look inside a recombiner to understand the interaction of chemistry and flow. This can help in understanding the dependencies of non-measurable variables (e.g. reaction rate), of local surface temperatures and more. It also offers possibilities to increase the chemical efficiency by optimising the geometry properly. Computational fluid dynamics (CFD) codes are available to be used as development tools to include the specifics of catalytic surface reactions. The present paper describes the use of the code system CFX (CFX 4.1 Flow Solver User Guide. 1995, Computational Fluid Dynamics Services, AEA Technology plc, Oxfordshire, UK) for creating a recombiner model. Finally its comparison with existing test data is discussed.     

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.     

Integral large scale experiments on hydrogen combustion for severe accident code validation-HYCOM

A joint research project was carried out in the EU Fifth Framework Programme, concerning hydrogen risk in a nuclear power plant. The goals were: Firstly, to create a new data base of results on hydrogen combustion experiments in the slow to turbulent combustion regimes. Secondly, to validate the partners CFD and lumped parameter codes on the experimental data, and to evaluate suitable parameter sets for application calculations. Thirdly, to conduct a benchmark exercise by applying the codes to the full scale analysis of a postulated hydrogen combustion scenario in a light water reactor containment after a core melt accident. The paper describes the work programme of the project and the partners activities. Significant progress has been made in the experimental area, where test series in medium and large scale facilities have been carried out with the focus on specific effects of scale, multi-compartent geometry, heat losses and venting. The data were used for the validation of the partners CFD and lumped parameter codes, which included blind predictive calculations and pre- and post-test intercomparison exercises. Finally, a benchmark exercise was conducted by applying the codes to the full scale analysis of a hydrogen combustion scenario. The comparison and assessment of the results of the validation phase and of the challenging containment calculation exercise allows a deep insight in the quality, capabilities and limits of the CFD and the lumped parameter tools which are currently in use at various research laboratories.     

使用FCI-CMFD软件分析岭澳二期核电厂直接安全壳加热事故

采用分析熔融物与冷却剂反应(FCI)的三维多相流数值计算软件MC3D,建立岭澳二期核电厂模型,对严重事故下可能发生的直接安全壳加热(DCH)现象进行了模拟和分析。为准确预测事故现象,本文结合全厂断电事故后期参数与岭澳二期核电厂核岛几何模型,模拟事故过程。计算得出了事故下安全壳内气体温度场、熔滴体积份额场、速度场及压力随时间的变化。结果表明：直接安全壳加热事故会在短时间内引起安全壳内压力和局部温度的迅速上升。     

先进压水堆核电厂氢气控制策略分析研究

新建核电厂的设计必须做到“实际消除”早期与大量放射性释放的可能性,氢气燃爆导致的安全壳失效是必须要“实际消除”的严重事故工况之一。因此对各种消氢措施的特点进行分析研究,建立联合消氢策略评价方法,可为先进压水堆核电厂氢气控制策略选择设计评价提供支持手段。根据严重事故管理中对氢气控制策略的考虑,研究安全壳内局部位置的可燃性是相关设计评价的关键问题。根据可燃性准则、火焰加速准则、燃爆转变准则,本文使用三维CFD程序对典型严重事故工况下安全壳蒸汽发生器隔间内的可燃性及氢气风险进行模拟分析。研究结果表明,虽然喷放源项中有大量水蒸气,蒸汽发生器隔间中仍有较大区域处于可燃限值以内,合理布置的点火器能在设计中点燃并消除氢气。本研究建立的分析方法能用于对核电厂氢气控制策略选择设计的评价。     

Validation of the CFX4 CFD code for containment thermal-hydraulics

In order to determine the risk associated with the presence of hydrogen in a nuclear power plant containment during a hypothetical severe accident, thermal hydraulic codes are used. Amongst other codes, NRG uses the commercial computational fluid dynamics code CFX4 for this purpose. Models to describe condensation have been implemented by user coding. This paper describes these models. In addition, an overview is given of validation activities with the CFX4 model. Experimental results from the following sources have been used: the Kuhn condensation model; the PHEBUS test facility; the PANDA test facility; and the TOSQAN, MISTRA, and THAI test facilities within the OECD International Standard Problem 47. The CFX4 model predictions are fairly good. Deviations originate primarily from the applied wall treatment. Several recommendations for further development are therefore proposed.     

Comparison of hydrogen generation for TVSM and TVSA fuel assemblies for water water energy reactor (VVER)-1000

This paper presents the results received during investigation of hydrogen generation for both types fuel assemblies—the old modernistic type of fuel assemblies (TVSM) and recently installed new one alternative type of fuel assemblies (TVSA) in case of severe accident. There are some differences between both types FAs. They have different geometry as well as different burnable poisons. To investigate behavior of new fuel assemblies during the severe conditions it have been performed comparison of fuel behavior of old type TVSM fuel assembly to new one TVSA.To perform this investigation it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) VVER 1000. The model was developed by Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS) for investigation of severe accident scenarios and Probabilistic Safety Analyses (PSA) level 2. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides release outside fuel cladding.It was accepted criteria for vessel integrity about hydrogen concentration to be 8%. This criterion was based on the decision of RSK (Germany commission for reactor safety).Generally based on the received results it was made conclusion that using both types of fuel assemblies it was not disturbance safety conditions of NPP.     

CFD simulations of hydrogen combustion in a simplified EPR containment with CFX and REACFLOW

The prediction of over-pressures and temperatures that are generated by hydrogen explosions in case of a severe nuclear accident is a crucial stage of the safety analysis of the containment. The investigation presented in this paper is a continuation of the numerical studies of validation and benchmarking that were carried out in the European co-sponsored project HYCOM. In the present work, numerical simulations of hydrogen deflagrations within a simplified, real-scale European Pressure Reactor (EPR) containment have been performed with two CFD codes, CFX4 and REACFLOW. The analysis has been focused not only on overpressure peaks and pressure oscillations, but also on pressure differences between the two sides of the same wall of internal compartments. Different geometrical configurations have been considered in term of presence of vents between internal compartments and in term of vents number, size and position. Single and multiple ignition points have also been taken into account. The paper describes the main results of the investigation and it is a demonstration of how CFD modelling can provide significant indications for real-scale safety applications within the limits of uncertainty of the accident scenarios.     

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.     

核电站安全壳内氢气扩散和燃烧的分析程序GASFLOW及其应用

介绍了由美国洛斯阿拉莫斯实验室(LANL)和德国卡尔斯鲁厄研究中心(FzK)共同开发的三维计算流体力学程序GASFLOW的基本数学物理模型和数值计算方法。该程序主要用于分析核电站严重事故下安全壳内氢气、水蒸气扩散分布和燃烧。列举了该程序在德国Konvio型压水堆氢气安全分析中的应用。     

百万千瓦级压水堆严重事故下局部隔间氢气风险分析

核安全法规要求控制严重事故下核电厂安全壳内的氢气浓度。除安全壳整体外,局部隔间的氢气浓度同样是关注的重点。本文采用一体化严重事故分析程序对百万千瓦级压水堆核电厂安全壳局部隔间进行建模,分析了不同事故下的氢气风险。结果表明,严重事故下部分隔间短时间内可能存在燃烧风险。本文对降低燃烧风险的方法进行分析计算和筛选,得出的结论可以为安全壳隔间的设计优化提供参考依据。     

International standard problem on containment thermal–hydraulics ISP47: Step 1—Results from the MISTRA exercise

The understanding of hydrogen distribution during severe accidents in a nuclear reactor containment is still an open issue. Several containment thermal–hydraulics international standard problems (ISP) have been conducted to address this topic. However, the predictions made by the available lumped parameter or CFD computer codes were generally not satisfactory. Therefore, a new exercise was launched in 1999 using new state-of-the-art experimental facilities TOSQAN, MISTRA and ThAI that included sophisticated 3D instrumentation and well-controlled boundary conditions. Predictive capabilities of important and still uncertain phenomena such as wall condensation, natural circulation and gas stratification are assessed. In addition, comparison between lumped parameter (LP) and CFD codes and assessment of the capability of CFD codes to deal with scaling effects are performed. This article reports on the part of the exercise which concerns the MISTRA facility including experimental results and blind benchmark exercises.     

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。     

Assessment of passive autocatalytic recombiners' capabilities in case of Zr oxidation during station blackout scenario for TVSA and TVSM fuel assemblies of WWER 1000

This paper presents the investigation of Passive Autocatalytic Recombiners' (PARs) capabilities for hydrogen recombination in case of Station blackout scenario. The assessment was performed for both types of WWER fuel assemblies – the old Modernistic type of fuel assemblies (TVSM) and recently installed new Alternative type of fuel assemblies (TVSA) in Kozloduy NPP. The main difference between both types of fuel assemblies is the different geometries, masses of internals materials as well as different burnable poisons. To investigate the PARs' capabilities it has been performed comparison of fuel behaviour of both types of fuel assemblies.To perform this analysis it has been used MELCOR “input model” for Kozloduy Nuclear Power Plant (KNPP) WWER-1000. The model was developed at the Institute for Nuclear Research and Nuclear Energy (INRNE) for investigation of severe accident scenarios. The model provides a significant analytical capability for the Bulgarian technical specialists, working in the field of the NPP safety, for analysis of core and containment damaged states and the estimation of radionuclides' release outside fuel cladding.To assess the PARs' capabilities it was used the acceptance criterion for containment integrity to be 8% hydrogen concentration. This criterion was based on the decision of RSK (Germany commission for reactor safety).Generally, based on the performed analyses it was made a conclusion that using both types of fuel assemblies it was not disturbance of PARs' capabilities and safety criterion of NPP.     

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。     

不同产氢速率对安全壳内氢气分布影响的数值模拟

利用计算流体力学软件（CFX）,初步研究了严重事故下氢气在安伞壳空间内的流动特性,分析了不同产氢速率对安全壳内氢气分布的影响。结果表明：各种氢气释放速率情况下,氢气分布的基本趋势一致;不同的产氢速率对氢气分布的影响主要体现在氢气运动到安伞壳穹顶时所形成的涡旋小同,氢气释放速率低的序列,氢气容易滞留在穹顶,然后向下慢慢扩散,分布较均匀;氢气释放速率高的序列的氢气运动方向性强,容易向下空间运动,分布的区域集中些,分层现象明显。     

Experimental simulation of the Paks-2 cleaning tank incident through separate effect and integral tests

Following the ex-core cleaning of VVER type fuel assemblies, a severe fuel damage took place at the Paks NPP. Assemblies with high decay heat remained without cooling for 7 h. As a common result of high temperature oxidation and final water quenching the assemblies were broken and the fuel rods fragmented.In order to improve the understanding of the phenomena that took place during the Paks-2 incident, several series of separate effect and integral tests have been carried out. The separate effect tests investigated the phenomena related to the upward shift of fuel assemblies, to the role of the spacer grid and the spring in the loss of cladding integrity, and to the importance of high hydrogen content in the vessel's atmosphere on the embrittlement of zirconium. Integral tests were carried out in the CODEX facility to simulate the whole scenario of the incident using electrically heated fuel rods.     

Simulation of hydrogen distribution in an Indian Nuclear Reactor Containment

The management of hydrogen in a Nuclear Reactor Containment after LOCA (Loss Of Coolant Accident) is of practical importance to preserve the structural integrity of the containment. This paper presents the results of systematic work carried out using the commercial Computational Fluid Dynamics (CFD) software FLUENT to assess the concentration distribution of hydrogen in a typical Indian Nuclear Reactor Containment. In order to obtain an accurate estimate of hydrogen concentration distribution, a suitable model for turbulence closure is required to be selected. Using guidelines from the previous studies reported in the literature and a comparative simulation study using simple benchmark problems, the most suitable turbulence model for hydrogen mixing prediction was identified. Subsequently, unstructured meshes were generated to represent the containment of a typical Indian Nuclear Reactor. Analyses were carried out to quantify the hydrogen distribution for three cases. These were (1) Uniform injection of hydrogen for a given period of time at room temperature, (2) Time varying injection as has been computed from an accident analysis code, (3) Time varying injection (as used in case (2)) at a high temperature. A parametric exercise was also carried out in case (1) where the effect of various inlet orientations and locations on hydrogen distribution was studied. The results indicate that the process of hydrogen dispersal is buoyancy dominated. Further for typical injection rates encountered following LOCA, the dispersal is quite poor and most hydrogen is confined to the fuelling machine vault.     

Analysis of hydrogen flame acceleration in APR1400 containment by coupling hydrogen distribution and combustion analysis codes

This study was conducted as part of the construction of an integrated system to mechanistically evaluate flame acceleration characteristics in a containment of a nuclear power plant during a severe accident. In the integrated analysis system, multi-dimensional hydrogen distribution and combustion analysis codes are used to consider three-dimensional effects of the hydrogen behaviors. GASFLOW is used for the analysis of a hydrogen distribution in the containment. For the analysis of a hydrogen combustion in the containment, an open-source CFD (computational fluid dynamics) code OpenFOAM is chosen. Data of the hydrogen and steam distributions obtained from a GASFLOW analysis are transferred to the OpenFOAM combustion solver by a conversion and interpolation process between the solvers. The combustion solver imports the transferred data and initializes the containment atmosphere as an initial condition of a hydrogen combustion analysis. The turbulent combustion model used in this study was validated by evaluating the F22 test of the FLAME experiment. The coupled analysis method was applied for the analysis of a hydrogen combustion during a station blackout accident in an APR1400. In addition, the characteristics of the flame acceleration depending on a hydrogen release location are comparatively evaluated.