Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor

In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant.The MARA 10 experimental test simulates a HCDA in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge.This paper presents a numerical simulation of the test with the EUROPLEXUS code and an analysis of the computed results. In particular, the evolution of the fluid flows and the deformations of the internal and external structures are analysed in detail. Finally, the current computed results are compared with the experimental ones and with previous numerical results computed with the SIRIUS and CASTEM-PLEXUS codes.     

Comparison of various models for the simulation of a Core Disruptive Accident in the MARA 10 mock-up

During a Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor melts down partially and that the interaction between hot molten fuel and relatively cold liquid sodium creates a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant.The MARA 10 experimental test simulates a Core Disruptive Accident in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge.This paper presents some models available within the EUROPLEXUS code to simulate a Core Disruptive Accident and an analysis of the computed results. In particular, results are compared with experimental measurements and previous numerical simulations carried out with the codes SIRIUS and CASTEM-PLEXUS.     

Computation of a core disruptive accident in the MARS mock-up

A hypothetical core disruptive accident in a liquid metal fast breeder reactor (LMFBR) results from the interaction between molten fuel and liquid sodium, which creates a high-pressure bubble of gas in the core. The violent expansion of this bubble loads and deforms the vessel and the internal structures. The MARS experimental test simulates a HCDA in a small-scale mock-up containing all the significant internal components of a fast breeder reactor. The mock-up is filled with water, topped by an argon blanket, and the explosion is generated by an explosive charge.This paper presents a numerical simulation of the test with the EUROPLEXUS code. The top closure is represented by massive structures and the main internal structures are described by shells. The current numerical results are described and compared with the experimental ones, and previous computations with the CASTEM-PLEXUS code.     

Liquid entrainment by an expanding core disruptive accident bubble—a Kelvin/Helmholtz phenomenon

The final stage of a postulated energetic core disruptive accident (CDA) in a liquid metal fast breeder reactor is believed to involve the expansion of a high-pressure core-material bubble against the overlying pool of sodium. Some of the sodium will be entrained by the CDA bubble which may influence the mechanical energy available for damage to the reactor vessel. The following considerations of liquid surface instability indicate that the Kelvin–Helmholtz (K–H) mechanism is primarily responsible for liquid entrainment by the expanding CDA bubble. First, an instability analysis is presented which shows that the K–H mechanism is faster than the Taylor acceleration mechanism of entrainment at the high fluid velocities expected within the interior of the expanding CDA bubble. Secondly, a new model of liquid entrainment by the CDA bubble is introduced which is based on spherical-core-vortex motion and entrainment via the K–H instability along the bubble surface. The model is in agreement with new experimental results presented here on the reduction of nitrogen-gas-simulant CDA bubble work potential. Finally, a one-dimensional air-over-water parallel flow experiment was undertaken which demonstrates that the K–H instability results in sufficiently rapid and fine liquid atomization to account for observed CDA gas-bubble work reductions. An important byproduct of the theoretical and experimental work is that the liquid entrainment rate is well described by the Ricou–Spalding entrainment law.     

Research and development studies on plant and core seismic behaviour for a fast reactor

This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA in co-operation with ANSALDO and ISMES for the seismic verification of the Italian PEC fast reactor test facility. More precisely, the paper focuses on the wide-ranging research and development programme that has been performed (and recently completed) on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the feed-backs on the design, necessary to satisfy the seismic safety requirements, are recalled. The general validity of the analyses in the framework of the research and development activities for nuclear reactors is also pointed out.     

Simulation of central sloshing experiments with smoothed particle hydrodynamics (SPH) method

Liquid sloshing phenomena can be observed whenever a liquid in a container has an unrestrained surface and can be excited. A particular type of sloshing motion can occur during the core meltdown of a liquid metal cooled reactor (LMR) and can lead to a compaction of the fuel in the center of the core possibly resulting in energetic nuclear power excursions. This phenomenon was studied in series of “centralized sloshing” experiments with a central water column collapsing inside the surrounding cylindrical tank. These experiments provide data for a benchmark exercise for accident analysis codes. To simulate “centralized sloshing” phenomena, a numerical method should be capable to predict the motion of the free surface of a liquid, wave propagation and reflection from the walls. In this study, a meshless method based on smoothed particle hydrodynamics (SPH) for the simulation of a 3D free surface liquid motion has been developed. The proposed method is applied to the simulation of “centralized sloshing” experiments. Simulation results are compared with the experimental results as well as with results of computations performed with the 3D code SIMMER-IV which is an advanced reactor safety analysis code that implements the traditional mesh-based numerical method. In a series of numerical calculations it is shown that overall motion of the liquid is in a good agreement with experimental observations. Dependence on the initial and geometrical symmetry is studied and compared with experimental data.     

Comparison of ICECO code predictions with flexible vessel experiments

Some comparisons of ICECO code predictions with experimental data concerning transient fluid-structure interaction are given. The test results are taken from flexible vessel experiments conducted by Stanford Research Institute under the direction of Argonne National Laboratory. Two different experiments are considered: one with a rigid core barrel, and one with a flexible core barrel. Both experiments are performed in simple reactor vessels with a well-defined energy source and simple boundary conditions. Correlations of pressures and impulses are made at all available gauge stations. The permanent deformations of the core barrel and the cylindrical vessels are compared with ICECO predictions. The effects of core barrel flexibility on the wave propagation and vessel deformation are also investigated. The agreement between the analysis and experiments is found to be quite good.     

Experiments at the mixing test facility ROCOM for benchmarking of CFD codes

For the validation of computational fluid dynamics (CFD) codes, experimental data on fluid flow parameters with high resolution in time and space are needed.Rossendorf Coolant Mixing Model (ROCOM) is a test facility for the investigation of coolant mixing in the primary circuit of pressurized water reactors. This facility reproduces the primary circuit of a German KONVOI-type reactor. All important details of the reactor pressure vessel are modelled at a linear scale of 1:5. The facility is characterized by flexible possibilities of operation in a wide variety of flow regimes and boundary conditions. The flow path of the coolant from the cold legs through the downcomer until the inlet into the core is equipped with high-resolution detectors, in particular, wire mesh sensors in the downcomer of the vessel with a mesh of 64 × 32 measurement positions and in the core inlet plane with one measurement position for the entry into each fuel assembly, to enable high-level CFD code validation. Two different types of experiments at the ROCOM test facility have been proposed for this purpose. The first proposal concerns the transport of a slug of hot, under-borated condensate, which has formed in the cold leg after a small break LOCA, towards the reactor core under natural circulation. The propagation of the emergency core cooling water in the test facility under natural circulation or even stagnant flow conditions should be investigated in the second type of experiment. The measured data can contribute significantly to the validation of CFD codes for complex mixing processes with high relevance for nuclear safety.     

ACME试验台架典型小破口工况试验及数值分析

对我国首个大型非能动堆芯冷却系统整体试验台架（ACME）中的典型小破口事故进行了试验及数值分析。分析结果表明：在ACME上开展的典型小破口试验,其事故序列及试验现象符合预期;RELAP5数值分析的主要结果能较好地反映试验现象,与试验结果吻合良好;堆芯棒束区相间摩擦模型的选用对堆芯坍塌液位的计算有较大影响,在不同阶段选用不同的模型可使计算结果更好地与试验值相匹配。     

Boiling water reactor with innovative safety concept: The Generation III+ SWR-1000

AREVA NP has developed an innovative boiling water reactor (BWR) SWR-1000 in close cooperation with German nuclear utilities and with support from various European partners. This Generation III+ reactor design marks a new era in the successful tradition of BWR and, with a net electrical output of approximately 1250 MWe, is aimed at ensuring competitive power generating costs compared to gas and coal fired stations. It is particularly suitable for countries whose power networks cannot facilitate large power plants. At the same time, the SWR-1000 meets the highest safety standards, including control of core melt accidents. These objectives are met by supplementing active safety systems with passive safety equipment of various designs for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. The plant is also protected against airplane crash loads.The functional capabilities and capacities of all new systems and components were successfully tested under realistic and conservative boundary conditions in large-scale test facilities in Finland, Switzerland and Germany.In general, the SWR-1000 design is based on well-proven analytical codes and design tools validated for BWR applications through recalculation of relevant experiments and independent licensing activities performed by authorities or their experts. The overview of used analytical codes and design tools as well as performed experimental validation programs is presented.Effective implementation of passive safety systems is demonstrated through the numerical simulation of transients and loss of coolant accidents (LOCAs) as well as through analytical simulation of a severe accident associated with the core melt. In the LOCA simulation presented the existing active core flooding systems were not used for emergency control: only passive systems were relevant for the analyses. Despite this - no core heat-up occurred. In the case of reactor core melting numerically is demonstrated that the molten core debris would be retained inside the reactor vessel due to the effective passive external water cooling of the vessel, keeping it completely intact.A short construction period of just 48 months from first concrete to provisional take over, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burn-up all contribute towards meeting economic goals. Realistic average availability for a plant lifetime of 60 years and 12 months cycle is 94.5%. Systems and plant design were reviewed by expert groups of European utilities. With the SWR-1000, AREVA NP has developed a design concept for a BWR plant that is now ready for commercial deployment and which fully meets the most stringent international requirements in terms of nuclear safety and nuclear regulatory.     

Experimental investigation on bubble characteristics entrained by surface vortex

The cover gas entrainment at the free surface of sodium coolant becomes one of the significant issues according to the compact sizing of reactor vessel in the latest reactor design. In the present study, some basic experiments for the gas entrainment due to the surface vortex were performed in order to obtain the fundamental knowledge about the entrained bubble size. Distributions of entrained bubble diameters in several experimental conditions were obtained from bubble images using an image processing technique. Velocity fields around vortices and surface dimple shapes (gas cores) due to surface vortices were measured to grasp those influences on bubble shapes. The result showed that mean equivalent diameters of bubbles were varied from 1.3 to 2.1 mm in the range of present experimental conditions. The bubble sizes were influenced by the thickness of gas core.     

LOCA工况下环形燃料元件内包壳外压屈曲试验研究

环形燃料一种安全高效的新型核燃料。为对环形燃料元件冷却剂丧失事故（LOCA）下整体受压失效形式的问题进行研究,将环形电加热棒、模拟芯块和试验件组装成试验装置,在空气环境中,以环形电加热棒外加热的方式,对环形燃料元件内包壳进行了外压屈曲试验,并将试验屈曲压力与Bresse?Bryan公式计算结果和特征值屈曲数值模拟分析结果进行了对比分析。结果表明：Bresse?Bryan公式计算结果除以安全系数m=2?5得到的结果高于试验结果而不够保守,试验结果分布于特征值屈曲数值模拟分析结果的1/5?1/3之间。本文结果可为环形燃料元件安全评价及后续工程化提供基础数据。     

Numerical study of in-vessel retention under the gallium–water external reactor vessel cooling system using MARS-LMR

To confirm the feasibility of the gallium–water IVR-ERVCS (in-vessel retention-external reactor vessel cooling system), this paper focuses on the numerical simulation of severe accidents in APR 1400 using MARS-LMR (multidimensional analysis of reactor safety-liquid metal reactor). To analyze the gallium-cooled systems, the properties of liquid gallium were added to the MARS-LMR code used in our previous work. In this system, the generated decay heat is transferred to liquid gallium through the reactor pressure vessel and then removed from the water pool as a heat sink. The numerical analyses results show that the temperature range of the liquid gallium is much lower than its boiling point and confirm the natural convection. Sensitivity studies were also performed by changing several parameters such as the initial temperature of gallium and water pool inventory and their results indicated that the working time of the gallium–water IVR-ERVCS depends on the inventory of the water pool. Because liquid gallium in this system does not have a phase change, unlike water, the gallium–water IVR-ERVCS can provide stable and reliable cooling capability. To solve the limitation due to critical heat flux in IVR-ERVCS and to ensure the sufficient thermal margin, it is confirmed that the gallium–water IVR-ERVCS can be a successful severe accident mitigation strategy in nuclear power plants.     

Experimental and numerical methods related to safety factors in the PEC fast nuclear reactor

In this paper the author summarizes the activity of structural analysis related to the safety of the PEC fast nuclear reactor. There are two principal aspects of safety concerning problems of structures: the localized incident and the hypothetical core disruptive accident (HCDA).With regard to the first point, the phenomenon is dependent on the hydrodynnamic and structural behaviour of the fuel elements. With regard to the HCDA, it is necessary that the reactor vessel is able to absorb the explosive energy, whereas the plug must not sustain movements such as to alter the overall seal of the installation.Given the complexity of the phenomena, therefore, it was considered necessary first of all to carry out numerous experimental tests on both full-size and reduced scale models. The experimental tests on the individual hexcan, on the group of seven hexcans and on the vessel were carried out at the EURATOM centre of Ispra, in the context of a collaboration agreement between ENEA and EURATOM.Some of the results of these tests are presented in this paper, together with relevant comparisions with the numerical values.     

Fluid-structure interaction analysis of a hypothetical core disruptive accident in LMFBRs

To ensure safety, it is necessary to assess the integrity of a reactor vessel of liquid-metal fast breeder reactor (LMFBR) under HCDA. Several important problems for a fluid-structural interaction analysis of HCDA are discussed in the present paper. Various loading models of hypothetical core disruptive accident (HCDA) are compared and the polytropic processes of idea gas (PPIG) law is recommended. In order to define a limited total energy release, a “5% truncation criterion” is suggested. The relationship of initial pressure of gas bubble and the total energy release is given. To track the moving interfaces and to avoid the severe mesh distortion an arbitrary Lagrangrian–Eulerian (ALE) approach is adopted in the finite element modeling (FEM) analysis. Liquid separation and splash from a free surface are discussed. By using an elasticity solution under locally uniform pressure, two simplified analytical solutions for 3D and axi-symmetric case of the liquid impact pressure on roof slab are derived. An axi-symmetric finite elements code FRHCDA for fluid-structure interaction analysis of hypothetical core disruptive accident in LMFBR is developed. The CONT benchmark problem is calculated. The numerical results agree well with those from published papers.     

考虑流固耦合效应的快堆堆本体抗震试验模化方法

为准确模拟快堆堆本体中液体晃动对主容器的作用，本文建立了一种考虑流固耦合效应的快堆堆本体抗震试验模化方法，不仅保证加速度相似比严格为1，还保证了流体与结构的质量比与原堆的相同。依照上述试验模化方法，分别设计了与快堆原型比尺为1∶25（大）和1∶50（小）两个缩比试验模型。为验证上述理论方法的有效性，对这两个模型进行了地震动力学数值模拟，并比较了大模型和小模型的模拟结果。比较结果表明，大、小模型的地震动响应参数比值满足推导得到的理论准则，从而通过数值试验方法验证了上述模化方法的有效性。该模化方法可为快堆堆本体抗震试验提供理论依据。     

Integrated modular water reactor (IMR) design

Integrated modular water reactor (IMR) has been developed as one of the advanced small-scale light water reactors, with a thermal output of 1000 MW. The IMR adopts natural circulation and self-pressurization in the primary cooling system, and a reactor vessel built-in steam generators. The core design has been performed using the current light water reactor technology. Thermal-hydraulic sensitivity analyses have been done from the viewpoint of the departure from nucleate boiling (DNB) limitation. The IMR core, with 97 21×21-type fuel assemblies and natural circulation in the primary coolant system, shows a good nuclear and thermal-hydraulic behavior and good allowable margins for the DNB phenomenon. The reactivity change with burnup is about 1%Δk by using burnable absorbers, and only 12 rod-cluster-controls are used through the operating cycle. The 20 m-height reactor vessel encloses steam generators in vapor and liquid portions. Plant dynamic analyses have been also performed in order to evaluate the IMR behavior from the viewpoints of plant operation and control. This study shows that the IMR will operate with enough margins for the core safety and will be stably controlled for load demand changes expected during normal operations.     

AP1000 ADS-4夹带卸压试验模拟分析

第4级自动降压系统（ADS-4）是AP1000极为重要的非能动安全设施。ADS-4能在AP1000小破口失水事故中为反应堆系统提供可控卸压。然而,大量的冷却剂可通过卸压过程中ADS-4夹带和上腔室夹带被带到安全壳中,从而引发堆芯裸露和堆芯熔化事故。为研究小破口事故中的ADS-4夹带卸压和上腔室夹带过程,在以AP1000为原型、按直径/高度比1∶5.6设计建造的ADS-4喷放卸压试验回路（ADETEL）中,研究了不同初始压力、压力容器混合液位和加热功率下的夹带和卸压行为,以及反应堆内部构件的夹带沉积效应。试验数据表明,大量的水在短时间内迅速通过ADS-4支管被夹带出来。液体的夹带率和压力容器混合液位的降低速率随系统初始压力的增加而增大。值得注意的是,在本试验特定工况下,初始压力为0.5 MPa时出现堆芯裸露。堆内构件对夹带量和压力容器混合液位无显著影响。     

铠装铂电阻液位测量传感器研究

压力容器内的水位是反应堆运行中的重要参数．基于发热体在液相和汽相介质中放热系数的显著差异．本文提出了一种由铠装铂电阻组成的液位测量传感器,并给出了理论分析结果和0．1～3．0MPa压力范围内的试验结果。结果表明,该传感器原理正确．结构可行．性能可靠。