CONTAIN-LMR程序中池式钠火事故分析计算模型的验证

CONTAIN-LMR是针对以液态钠为冷却剂的反应堆而开发的安全壳事故一体化分析程序。我国目前的CONTAIN-LMR程序版本为2000年左右从法国引进,还未进行过面向工程设计的系统性地程序开发和验证。本文主要针对CONTAIN-LMR程序中模拟池式钠火事故的分析模型进行详细分析,并采用国际上的池式钠火实验进行验证,实验验证结果表明CONTAIN-LMR程序可以较准确地模拟池式钠火事故造成的钠工艺间内的温度、压力升高及放射性钠气溶胶行为。本文的研究结果初步表明CONTAIN-LMR程序可用于钠冷快堆的钠火事故分析。     

非能动安全壳冷却系统空气流道自然循环的比例分析与设计

在考虑建设试验台架经济性的前提下,缩小比例的单项和整体效应试验台架对研究和开发大型先进压水堆核电站及其分析验证程序都具有重要意义。非能动安全壳冷却系统（PCS）壳外空气流道内的自然循环在安全壳非能动冷却性能中发挥着重要的作用。本文从自然循环的数学模型出发,推导出了单项和整体效应试验台架的比例设计方法。在给定壳内热流密度的条件下,通过PCCSAP-3D程序对CAP1400非能动安全壳的2/5比例单项效应试验理想比例台架（ISF）进行模拟。结果表明,本比例分析与设计方法以及在降低高度台架上模拟自然循环是可行的。     

The TONUS CFD code for hydrogen risk analysis: Physical models, numerical schemes and validation matrix

The French Atomic Energy Commission (CEA) and the Institute for Radiological Protection and Nuclear Safety (IRSN) are developing a hydrogen risk analysis code, called the TONUS code, which incorporates both lumped-parameter (LP) and computational fluid dynamics (CFD) formulations. In this paper we present the governing equations, numerical strategy and schemes used for the CFD approach.Several benchmark exercises based on experimental results obtained on large-scale facilities, such as MISTRA, TOSQAN and RUT, are presented. They have been used as verification and assessment procedures for modelling and numerical approaches of the code. Specific emphasis is given to the sensitivity analysis of the computed results with respect to numerical and physical parameters. The powerful Design-Of-Experiments technique for sensitivity analysis is successfully applied to the ISP47 MISTRA test case.The TONUS CFD code is presently used to support the hydrogen risk assessment for the European Pressurized Reactor (EPR) plant and to investigate the impact of the two-room concept on hydrogen distribution in the EPR containment.     

Modeling of atmosphere stratification in containments of nuclear power plants using lumped-parameter code

One of the current high priority safety issues of nuclear power plants is atmosphere stratification in the containments. It requires extensive experimental and numerical investigations. Numerical investigation of stratification using lumped-parameter codes is complicated due to inherent limitations of the codes. These limitations have to be taken into account in developing nodalisation of the containments.The paper presents M5 experiment simulations using lumped parameter code COCOSYS. Experiment M5 was performed in the MISTRA test facility in the frames of FP-6 project SARNET (Severe Accident Research Network). Obtained results are presented in the context of evaluation of the lumped-parameter code COCOSYS capabilities to model the formation of stratified atmosphere and to give recommendations for the development of nodalisations in similar cases using any lumped-parameter code. The paper describes the MISTRA test facility and the performed experiment M5, the gas flow model of the COCOSYS code, the developed nodalisations, and the obtained results. The performed analysis showed that detailed vertical nodalisation and special treatment of injected gas flow is required for simulation of gas mixing phenomena.     

Scaling of passive condenser system for separate effect testing

The passive containment cooling system (PCCS) of the simplified boiling water reactor (SBWR) is a passive condenser system designed to remove energy from the containment for long term cooling period after a postulated reactor accident. Depending on pressure condition and noncondensable (NC) gas fraction in drywell (DW) and suppression pool (SP), three different modes are possible in the PCCS operation namely the forced flow, cyclic venting and complete condensation modes. The prototype SBWR has total of six condenser units with each unit consisting of hundreds of condenser tubes. Simulation of such prototype system is very expensive and complex. Hence a scaling analysis is used in designing an experimental model for the prototype PCCS condenser system. The motive for scaling is to achieve a homologous relationship between an experiment and the prototype which it represents. A scaling method for separate effect test facility is first presented. The design of the scaled test facility for PCCS condenser is then given. Data from the test facility are presented and scaling approach to relate the scaled test facility data to prototype is discussed.     

Study on local hydrogen behaviors in a subcompartment of the NPP containment

Hydrogen control in the case of severe accidents has been required by nuclear regulations to ensure the integrity of containment after TMI accidents. Up to now, many experiments have been conducted to estimate the distribution of hydrogen during accidents in nuclear power plants. In this article, we proposed a computer code named HYCA3D developed to calculate the local hydrogen distribution with three-dimensional time-dependent governing equations, which can simulate the transport of multiple species. Also, local hydrogen behavior has been experimentally investigated in a cylindrical multi-subcompartment mixing chamber, measuring the local concentration in various conditions. Hydrogen is simulated by helium in the experiments. The proposed code was verified with these experimental results, followed by pre-tests with EPRI/HEDL standard problems. The calculation results show good agreement with the experimental data.     

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.     

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.     

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).     

Development of an Integrated Computer Code for Simulating Processes Occurring in the Containment Shell of Nuclear Power Plants with VVéR Reactors

The results of a competition for the development and verification of a more accurate domestic integrated computer code for simulating processes occurring in the containment shell of a nuclear power plant with a VVÉR reactor are examined. The basic requirements for the competing codes, the results of the calculations performed on the basis of the standard safety problem, a comparative analysis of the codes presented, and the steps in the development and verification of the KUPOL-M code are presented.     

The use of the COMPACT code for calculation of long term environmental effects as a result of high temperature discharges into confined areas

COMPACT is a computer code used to provide long term thermo-hydraulic safety analysis both for buildings associated with light water nuclear reactors and other similar structures. This paper discusses the main features of the code and the validation programme, which has been put in place to establish its ability to predict accurately the environmental conditions within the containment of a reactor. In particular the experience gained in modelling the hydrogen tests performed in the HDR reactor building is described.     

Simulation of containment cooling with outside spray after a core meltdown

In the most severe hypothetical loss of coolant accident, the reactor core melts and falls into the containment sump, there evaporating much of the sump water and raising the pressure within the containment building. One possible method to remove the decay heat is to cool the steel containment shell with an outside spray system. To perform the structural analysis needed to confirm the integrity of the containment, the thermal profile in the containment wall must first be found. The purpose of this work is to develop a computer code to calculate this transient profile. Other aspects such as hydrogen build-up are not considered in this code.The method uses relationships for the natural convection-partial condensation phenomena occurring at the containment internal surfaces, iteratively coupled to a one-dimensional heat balance at the wall to solve for the wall temperature as a function of angular position. A differential calculation as a function of time treats the thermodynamic changes within the containment as quasi-steady state. The result is a quick-running code capable of analyzing the temperature transient for several hours following the LOCA with a few minutes of computing time.     

GOTHIC局部氢气风险三维分析研究

目前的氢气风险分析中,主要采用一体化严重事故分析程序进行分析计算。日本福岛事故后,对氢气风险分析提出了更高的要求。为了实现对集总参数程序的有益补充,本文开展了GOTHIC程序氢气风险三维分析的研究。利用GOTHIC建立了局部氢气风险三维分析模型,在模型验证的基础之上,对典型严重事故序列下的氢气风险进行三维分析研究。研究表明:安全壳上部空间气流混合较好,氢气分层并不是非常明显;对于核电厂压力容器直接注射(DVI)管道破口所在的非能动堆芯冷却系统隔间B(PXS-B),由于破口以下部分区域被水淹没,破口以上区域的氢气浓度较高,但氢气风险较小。     

GOTHIC局部氢气风险三维分析研究

Integrated severe accident code is used to analyze the hydrogen risk in current safety assessment. After Fukushima accident, higher requirements are placed on hydrogen risk analysis. In order to supplement the lumped parameter analysis, three dimension hydrogen risk analysis method using GOTHIC is studied. Local three dimension hydrogen risk model is constructed by GOTHIC. Based on model validation, typical severe accident cases are chosen to analyze the hydrogen distribution. The results show that, hydrogen and other gas are mixed well in the upper compartment of the containment, and hydrogen stratification phenomenon is not obvious. For DVI rupture accident in PXS-B, the lower area of the break is flooded, and the hydrogen concentration for the upper area of the break is large, however, the hydrogen risk is little.     

局部空间内氢气爆炸程序开发及应用

针对反应堆安全壳或厂房局部空间内氢气爆炸过程,利用Fortran 90语言开发了氢气爆炸数值分析程序。采用单步反应模拟氢气与空气的化学反应,采用5阶精度的WENO求解对流项,时间步进采用3阶精度的龙格-库塔方法,对局部二维空间内氢气/空气/水蒸气预混气的爆炸过程进行了数值模拟。采用开发的程序计算了两种典型的激波管问题以验证程序的准确性,并用该程序分析了带隔间的沸水反应堆厂房局部空间内的氢气爆炸过程。计算结果表明：爆炸过程中最大的压力峰值来源于冲击波与反射波之间的碰撞,最大的冲击波压力和温度高达7.5 MPa和3 245 K。由此可得,安全壳内的氢气爆炸过程可能会威胁到安全壳的完整性,导致放射性物质释放。     

Experimental modelling of hypothetical core disruptive accidents

A common feature to reactor containment programmes is the use of detailed models to furnish data for design and safety assessment purposes. Despite the great strides which have been made in computational methods it is expected that the experimental approach will have a continuing role. It is therefore still pertinent to review the basis of such experiments, to see how they could be improved, and to see how well model experiments describe other processes occurring during an hypothetical core disruptive accident (HCDA).Numerous papers have described experiments on detailed models of a fast reactor scheme, and in all these, the sodium coolant of the reactor is replaced by water in the model for obvious practical reasons, but the scaling consequences of this change seem to have been given little attention. Therefore the object of this paper is to review the fundamentals of the scaling process, and then to discuss in more detail the effects of changing the working fluid in HCDA experiments.It is shown that the usual practice of using a geometrically scaled model, water as the working fluid, and a charge of the same characteristics as expected in the reactor excursion results in an inexact simulation, requiring somewhat uncertain corrections before the data can be used for the reactor case. An alternative possibility which is discussed in this paper would be to model the compressible characteristics of the sodium and the results could then be applied directly to the reactor scale using well defined scaling factors. This proposal, however, does require detailed changes to the experimental model and to the charge, but neither of these is expected to give undue difficulty.Modelling of an HCDA normally refers to modelling of the compressible fluid/structure interaction but in recent years interest has grown in other processes, such as heat and mass transfer. By looking at the appropriate dimensionless numbers in the model and reactor, the possibilities of using scale experiments to investigate certain features can be gauged. It is concluded that with experiments using water as the working fluid many processes associated with heat and mass transfer will not be modelled correctly and therefore special experiments have to be devised. For the same reason, caution should be used in extrapolating to the reactor heat and mass transfer data from experiments designed to reproduce structure deformation and loading.Although the modelling of compressible fluid/structure interactions is without doubt the main interest at the present time, other processes can be modelled without difficulty. In the example given, it is shown that buoyancy effects can be modelled provided an incompressible fluid simulation is sufficient. This simulation requires a low pressure charge such as might be provided by the evaporation of FREON released from a frangible container.     

GASFLOW程序液膜模型开发及初步验证

本文基于三维CFD安全壳程序GASFLOW开发了热构件壁面上的液膜覆盖与蒸发模型。通过选定AP1000大破口事故序列,采用耦合了液膜模型的GASFLOW程序分析了AP1000核电厂安全壳内温度压力响应及其非能动安全壳冷却系统(PCS)的性能,并与相同事故序列下WGOTHIC、MELCOR、CONTAIN等程序的计算结果进行比较。结果表明,耦合了液膜模型的GASFLOW程序可用于分析PCS的热工水力行为,其基本功能满足计算需要。     

Simulation of hydrogen mitigation in catalytic recombiner. Part-II: Formulation of a CFD model

This paper aims at formulation of a model compatible with CFD code to simulate hydrogen distribution and mitigation using a Passive Catalytic Recombiner in the Nuclear power plant containments. The catalytic recombiner is much smaller in size compared to the containment compartments. In order to fully resolve the recombination processes during the containment simulations, it requires the geometric details of the recombiner to be modelled and a very fine mesh size inside the recombiner channels. This component when integrated with containment mixing calculations would result in a large number of mesh elements which may take large computational times to solve the problem. This paper describes a method to resolve this simulation difficulty. In this exercise, the catalytic recombiner alone was first modelled in detail using the best suited option to describe the reaction rate ( [Prabhudharwadkar et al., 2005] and [Prabhudharwadkar et al., 2011]). A detailed parametric study was conducted, from which correlations for the heat of reaction (hence the rate of reaction) and the heat transfer coefficient were obtained. These correlations were then used to model the recombiner channels as single computational cells providing necessary volumetric sources/sinks to the energy and species transport equations. This avoids full resolution of these channels, thereby allowing larger mesh size in the recombiners. The above mentioned method was successfully validated using both steady state and transient test problems and the results indicate very satisfactory modelling of the component.     

钠池内混合对流的研究

利用二维模拟方法对快中子反应堆钠池内混合对流现象进行了系统研究。提出了柱坐标系广义环向过流线度概念，以解决池内几何引起的三维效应问题。提出了时间子步法用于紊流方程的求解，辅助速度场边界条件处理的独特见解，并介绍如何处理流场中的质量源项，使得流函数方法同样可以用于存在质量源项的流场计算。作为校正例子，选用方形空腔水模拟实验方案，应用理论分析成果对空腔内混合对流现象进行计算，分析空腔内分层流形成的规律，并对一个典型快堆热池内混合对流现象进行了校正计算。     

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.