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.     

High pressure simulation experiment on corium dispersion in direct containment heating

Purdue 1/10 scale direct containment heating separate effects experiments under a reactor vessel pressure up to 14.2 MPa are presented. With the test facility scaled to the Zion PWR geometry, these tests are mainly focused on the corium dispersion phenomenon in order to obtain a better understanding of the dominant driving mechanisms. Water and woods metal have been used separately to simulate the core melt, the reactor vessel being pressurized with nitrogen gas analogous to the steam in the prototypic case. The entire test transient lasted for a few seconds, and the liquid dispersion in the test cavity occurred within only 0.5 s. To synchronize the data acquisition and blowdown transient, the test initiation was triggered by breaking two rupture discs in the liquid/gas delivery system. Parameters characterizing the liquid transport were obtained via various instruments. Important information about the mean size and size distribution of the dispersed droplets in the test cavity, the liquid film flow transient, the subcompartment trapping, and the liquid carry-over to the containment has been obtained. These results, along with data from a previous low pressure (1.4 MPa) experiment carried out at Purdue University, form a solid database for further theoretical analysis.     

Hydrogen management and overpressure protection of the containment for future boiling water reactors

New design and evaluation method for hydrogen management of containment atmosphere have been developed for application in the future boiling water reactor (BWR). These are intended as a part of consideration of severe accidents in the course of design so as to assure a high level of confidence that a large release of radioactivity to the environment that may result in unacceptable social consequences can reasonably be avoided. Emphasis on hydrogen management and protection against overpressure failure is based on the insights from probabilistic safety assessments (PSAs) that late phase overpressure (and associated leakage) and molten corium concrete reaction (MCCI) need attention to ensure that containment remains intact, in case energetic challenges to the containment such as DCH (direct containment heating) or FCI (fuel coolant interactions) are practically eliminated by design or resolved from risk standpoint of view. The authors studied the use of palladium-coated tantalum for hydrogen removal from containment atmosphere in order to avoid pressurization of the containment with small free volume by non-condensable gas and steam. Its effectiveness for ABWR (advanced boiling water reactor) containment was evaluated using laboratory test data. Although further experimental studies are necessary to confirm its effectiveness in real accident conditions, the design is a promising option and one that could be backfitted upon necessity to existing plants for which pressure retaining capability cannot be altered. Also new evaluation method for flammability control under severe accident conditions was developed. This method employes a realistic assessment of the amount of oxygen and hydrogen gases generated by radiolytic decomposition of water under severe accident conditions and their subsequent transport from water to containment atmosphere.     

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.     

Effect analysis of the intentional depressurization on fission product behavior during TMLB' severe accident

It has been found that the pressure in the reactor coolant system (RCS) remains high in some severe accident sequences at the time of reactor vessel failure, with the risk of causing direct containment heating (DCH).Intentional depressurization is an effective accident management strategy to prevent DCH or to mitigate its consequences. Fission product behavior is affected by intentional depressurization, especially for inert gas and volatile fission product. Because the pressurizer power-operated relief valves (PORVs) are latched open, fission product will transport into the containment directly. This may cause larger radiological consequences in containment before reactor vessel failure. Four cases are selected, including the TMLB' base case and the opening one, two and three pressurizer PORVs. The results show that inert gas transports into containment more quickly when opening one and two PORVs,but more slowly when opening three PORVs; more volatile fission product deposit in containment and less in reactor coolant system (RCS) for intentional depressurization cases. When opening one PORV, the phenomenon of revaporization is strong in the RCS.     

Direct containment heating experiments for Vandellos and ASCo nuclear power plants

The ongoing IPE studies for the Vandellos and ASCo nuclear power plants require evaluation of accident phenomena that have been perceived to potentially challenge containment integrity including direct containment heating (DCH). Analyses and scaled experiments performed to date indicated that the lower containment structures play a substantial role in mitigating the extent of DCH given a high pressure melt ejection. Since the geometry is judged to be of major importance, linearly scaled experiments were conceived and conducted to evaluate the role of such structures in the Vandellos and ASCo specific configurations. The Vandellos test configuration with an initally dry cavity and significant exhaust area for the instrument tunnel resulted in the dispersal of a majority of the debris from the instrument tunnel into the lower compartment. The test of the ASCo configuration with an initially wet reactor cavity and limited exhaust area from the instrument tunnel exhibited the retention of the majority of the debris within the instrument tunnel and reactor cavity. The observed pressure responses in these scaled experiments for the seal table room, lower containment vessel, and upper containment vessel were all less than the containment design basis pressure. These test results contribute to the existing technical basis for concluding that direct containment heating would not represent a challenge to the integrity of these containments.     

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.     

An analysis of containment responses during a station blackout accident

An analysis of the responses of the containment during a station blackout accident is performed for the APR1400 nuclear power plant using MELCOR 2.1. The analysis results show that the containment failure occurs at about 84.14 h. Prior to the failure of the reactor vessel, the containment pressure increases slowly. Then, a rapid increase of the containment pressure occurs when a large amount of hot molten corium is discharged from the reactor pressure vessel to the cavity. The molten corium concrete interaction (MCCI) is arrested when water is flooded over a molten corium in the cavity. The boiling of water in the cavity causes a fast increase in the containment pressure. During the early phase of the accident, a large amount of steam is condensed inside the containment due to the presence of the heat structures. This results in a mitigation of a containment pressure increase. During the late phase, the containment pressure increases gradually due to the addition of steam and gases from an MCCI and water evaporation. It was found that two-thirds of the total mass of steam and gases in the containment is from an MCCI and one-third of the mass is from water evaporation.     

In-vessel retention of corium at the Loviisa plant

In-vessel retention of corium has been approved to be part of the severe accident management strategy for IVO's Loviisa plant. The approach selected takes advantage of the unique features of the plant such as a low power density, a reactor pressure vessel (RPV) without penetrations at the bottom, and ice-condenser containment which ensures a flooded cavity in all risk significant sequences. The thermal analyses, which are supported by an experimental program, demonstrate that, in Loviisa, the molten corium on the lower head of the RPV is externally coolable with wide margins. This paper summarizes the approach, the thermal analyses and the plant modifications being implemented.     

Ex-vessel fuel-coolant interaction experiment in the DISCO facility in the LACOMECO project

In the frame of the LACOMECO (large scale experiments on core degradation, melt retention and containment behavior) project of the 7th European Framework Program, a test in the DISCO (dispersion of corium) facility was performed in order to analyze the phenomena which occur during an ex-vessel fuel–coolant interaction (FCI). The test is focused on the premixing phase of the FCI with no trigger used for explosion phase. The objectives of the test were to provide data concerning the dispersion of water and melt out of the pit, characterization of the debris and pressurization of the reactor compartments for scenarios, where the melt is ejected from the reactor pressure vessel (RPV) under pressure. The experiment was performed for a reactor pit geometry close to a French 900 MWe reactor configuration at a scale of 1:10. The corium melt was simulated by a melt of iron–alumina with a temperature of 2400 K. A containment pressure increase of 0.04 MPa was measured, the total pressure reached about 0.24 MPa. No spontaneous steam explosion was observed. About 16% of the initial melt (11.62 kg) remained in the RPV vessel, 60% remained in the cavity mainly as a compact crust. The fraction of the melt transported out of the pit was about 24%.     

An evaluation of a direct cooling method for the ex-vessel corium stabilization

An evaluation of the ex-vessel core catcher system of a sample advanced light water reactor was presented. The core catcher was designed to cool down the molten corium through a combined injection of water and gas from the bottom of the molten corium, which could be effective in the reduction of rapid steam generation. By using the MELCOR code, a scenario analysis was performed for a representative severe accident scenario of the ALWR, that is, the 6-in. large break loss of coolant accident without safe injection. The spreading characteristics of ejected corium at vessel breach were asymptotically evaluated on the core catcher horizontal surface. The composition of the molten corium, the decay power level, and the sacrificial concrete ablation depth with time were obtained by a sacrificial concrete ablation analysis. The corium cooling history in the core catcher during the coolant injection was evaluated to calculate the temporal steam generation rate by considering an energy conservation equation. These were used as the major inputs for the temporal calculations of containment pressure which was performed by using the GASFLOW code. Several cases with change of water and gas injection rates were calculated. It was confirmed that the bottom water/gas injection system was an effective corium cooling method in the ex-vessel core catcher to suppress the quick release of steam.     

Detailed Evaluation of RCS Boundary Rupture during High-Pressure Severe Accident Sequences

A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture.     

三代压水堆核电厂DCH参数敏感性研究

安全壳直接加热（DCH）是导致安全壳早期超压的主要贡献之一,严重威胁安全壳完整性,并可能造成放射性物质早期大量不可控释放。本文以我国某三代压水堆为研究对象,首先基于风险导向的事故分析方法（ROAAM）,利用双隔间平衡（TCE）模型编写程序计算典型事故工况下的DCH载荷;其次结合安全壳失效概率曲线得出DCH现象造成的安全壳失效概率;最后对计算程序中不易得到的参数或经验值等不确定性较大的参数进行敏感性分析,归纳敏感性分析结果,找出敏感参数的不确定因素。结果表明：熔融物质量、堆腔几何设计、安全壳布置设计会直接影响DCH后果。     

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。     

An experimental study on quenching of a radially stratified heated porous bed

The quenching characteristics of a volumetrically-heated particulate bed composed of radially stratified sand layers were investigated experimentally in the POMECO facility. The sand bed simulates the corium particulate debris bed which is formed when the molten corium released from the vessel fragments in water and deposits on the cavity floor during a postulated severe accident in a light water reactor (LWR). The electrically-heated bed was quenched by water from a water column established over top of it, and later also with water coming from its bottom, which was circulating from the water overlayer through downcomers. A series of experiments were conducted to reveal the effects of the size of downcomers, and their locations in the bed, on the quenching characteristics of the radially stratified debris beds. The downcomers were found to significantly increase the bed quenching rate. To simulate the non-condensable gases generated during the MCCI, air and argon were injected from the bottom of the bed at different flow rates. The effects of gas flow rate and its properties on the quenching behaviour were observed. The results indicate that the non-condensable gas flows reduce the quenching rate significantly. The gas properties also affect the quenching characteristics.     

Modelling of PWR lower head failure under severe accident loading using improved shells of revolution theory

In the study of severe pressurized water reactor accidents, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are usually investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exists. This may lead to an out of vessel steam explosion or to direct heating of the containment; both which have the potential to lead to early containment failure.Within the framework of the OECD Lower Head Failure (OLHF) programme, a simplified model based on the theory of shells of revolution under symmetrical loading was developed by IRSN. After successfully interpreting some other representative experiments on lower head failures, the model was recently integrated into the European integral severe accident computer ASTEC code. The model was also used to obtain the thermo-mechanical behaviour of a 900-MWe pressurized water reactor lower head, subjected to transient heat fluxes under severe accident conditions.The main objective of this paper is to present: (1) the full mathematical formulations used in the development of the model, including their matrices and integrals defined by analytical expressions; (2) the two creep laws implemented, one for the American steel SA533B1 and one for the French steel 16MND5; and (3) the various numerical interpretations of experiments using the simplified model. This paper can be considered as a theoretical manual to aid users of the simplified model during modelling of lower head failures under severe accident conditions. One of the applications presented in this paper concerns the determination of a diagram representing the vessel time to failure as a function of the pressure level and the heat flux intensity. This information has been used by IRSN in probabilistic safety assessment and severe accident management analyses.     

延展表面凝结液膜的动量积分方法

采用动量积分方法分析压水堆发生失水事故时在安全壳的内表面上的液膜凝结、再浸润和蒸发过程。由凝结液膜的质量和动量守恒方程导出了凝结液膜在延展表面的子午线方向平均速度的积分-微分方程。假设液膜以层流的方式流动，把导出的积分-微分方程变成容易进行数值积分的液膜速度的一阶常微分方程，由此求得液膜厚度分布。液膜能量守恒方程的解给出了安全壳内壁面的温度分布。     

非能动安全壳冷却系统膜状冷凝强化换热设计

非能动安全壳冷却系统(PCCS)能在反应堆发生事故时将安全壳内部的热量及时导出,避免安全壳因超温、超压而失效。为强化换热,本文设想在安全壳内部安装阻隔带和液滴收集装置,通过降低层流区液膜厚度、扰动不可凝气体隔离层并充分利用湍流的换热强化作用,降低总的换热热阻,提高换热效率。以AP1000为例,依托GDLM模型对改进前后安全壳的换热情况进行分析,结果表明,通过安装阻隔带和液滴收集装置,能降低安全壳壁面的液膜厚度,提高壁面热流量,从而实现强化换热。     

PCS表面液膜覆盖率对安全壳完整性影响分析

钢制安全壳是防止严重事故工况下放射性物质向环境释放的最后一道屏障,因此有必要研究分析事故条件下安全壳外液膜覆盖率对安全壳完整性影响,以得到安全壳在事故工况下的失效裕度。应用非能动安全壳分析程序,建立了大功率非能动反应堆非能动安全壳冷却系统(Passive Containment Cooling System,PCS)的热工水力模型,并以冷段双端剪切事故为基准研究对象,分别研究了水分配器单一故障和出水管堵管叠加水分配器故障两种事故工况。分析结果表明,两种事故工况在液膜覆盖率大于35%时,均不会出现短期安全壳超压超温失效;事故后24 h,液膜覆盖率低于45%时,安全壳出现长期冷却失效。此次研究得出结论:在流量大于61.76 m3·h-1、安全壳液膜覆盖率大于45%时,事故发生后24 h安全壳不会失效。