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.     

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.     

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.     

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.     

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.     

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。     

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。     

Corium crust strength measurements

Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO₂, ZrO₂, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.     

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%.     

小破口引发的严重事故工况及事故缓解的研究

利用MAAP4程序对方家山核电站进行建模,针对事故后果较为严重的小破口事件进行了计算分析,得到了假设事故下电厂系统的反应以及相应的严重事故现象.对事故中发生的DCH(安全壳直接加热)现象和安全壳失效以及裂变产物向环境的释放进行了分析.随后,本文根据相关的严重事故管理导则和该事故的特点,对缓解该事故的策略进行了研究和计算...     

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.     

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.     

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.     

Controlling and cooling core melts outside the pressure vessel

For future reactors, the control and cooling of ex-vessel corium melts is under consideration to increase the passive safety features even for very unlikely severe accidents. In this context, different research activities are studying ex-vessel corium behaviour and control, including the implementation of a core cooling device outside the reactor pressure vessel in order to prevent basement erosion and to maintain the integrity of the containment. This paper describes current research on key phenomena which must be understood and quantified to be finally controlled by the cooling device. These are the release of corium melt from the pressure vessel, the temporary retention of the melt in the reactor cavity until melt through of the gate, spreading of the melt on a large surface, and finally the cooling and solidification of the melt by direct water contact. The experiments use high temperature melts which are similar to corium melts. Where necessary, models are developed to transfer the results to reactor scale.     

Corium spreading and coolability: CSC Project

In the context of severe accidents, large R&D efforts throughout the world are currently directed towards ex-vessel corium behaviour. Among the mitigation means which can be envisaged, the European industries and utilities are considering the implementation of a core-catcher outside the reactor pressure vessel in order to prevent basemat erosion and to stabilize and control the corium within the containment. The CSC project focused on two key phenomena for external core-catcher efficiency, reliability and safety: spreading and coolability. An experimental programme, covering different scenarios and including both simulant and real materials provided a lot of results which now constitute a large database and which enabled the qualification of computer codes.     

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。     

A use of prototypic material for the investigation of severe accident progression

An experimental research platform using corium melts is established for the understanding of safety related important phenomena during a severe accident progression. The research platform includes TROI facility for corium water interaction experiments and VESTA facility for corium-structural material interaction experiments. A cold crucible technology is adapted and improved for a generation of 5–100 kg of corium melts at various compositions. TROI facility is used for experiments to investigate premixing and explosion behaviors during a fuel coolant interaction process. More than 70 experiments using corium at various compositions were performed to simulate steam explosion phenomena in a reactor situation. The results indicate that the conversion efficiency of steam explosion for corium is less than 1%. VESTA facility is used to investigate molten corium-structural material interaction phenomena. VESTA facility consists of two cold crucibles. One crucible is used for the melting of charged material and pouring of corium melt. The other crucible is used for the corium-structural material interaction while providing an induction heating to simulate the decay heat. The results of an experiment on the interaction between corium melt and a specimen made of Inconel performed in the VESTA facility is reported.     

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.     

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.