Seismic isolation of pool-type tanks for the storage of nuclear spent fuel assemblies

Seismic isolation of pool-type nuclear spent fuel storage tanks requires careful investigation of dynamic behavior of the fluid–structure–isolator interaction system to satisfy the requirements of safety functions and the prevention of nuclear criticality. This paper presents the investigation, results and discussions on the seismic design considerations of isolated pool-type tanks for the storage of nuclear spent fuel assemblies. A three-dimensional boundary element-finite element method is presented for the analysis of the fluid–structure–isolator systems in time domain. Scaled model tests were performed to verify the numerical method and to study the dynamic behavior of isolated pool-type storage tanks. Important factors affecting the dynamic behavior of tanks with a fixed base are further investigated as is the case for isolated tanks using base isolators with different mechanical properties. The base isolators are the high damping rubber-bearing type and are modeled using a bilinear analysis model. Based on the numerical analysis and experimental results, some conclusions and discussions on the design considerations for isolated storage tanks are presented. In general, it is shown that careful selection of mechanical properties of the isolators with a certain lower limit on the effective frequency can guarantee the reduction of the dynamic responses of the storage tanks and the enhancement of the stability of stored spent fuel assemblies against earthquake excitations.     

Conceptual evaluation of fluid–structure interaction effects coupled to a seismic event in an innovative liquid metal nuclear reactor

The seismic response analysis of such liquid storage systems, especially liquid metal reactors, as for example the eXperimental Accelerator Driven System (XADS), was examined taking into account mainly the coupling effects of the fluid–structure interaction and their influence on its relevant internal systems and components.Therefore this paper deals with the structural analyses of the seismically induced hydrodynamic responses, in the event of a safe shutdown earthquake (SSE), and the free oscillation (known as sloshing waves) of a metal liquid coolant as well as the dynamic buckling effects on involved structures.To the mentioned purpose the interaction and coupling effects among the main reactor vessel structures and the primary coolant response were investigated by means of a numerical evaluation (with a qualified finite element code) because of the lack of analytical linear theories that in any case are not adequate to describe all the complex phenomena related to the seismic loading.For the numerical modelling procedure, 3D finite element models were set up to analyse the propagation of seismic waves as well as its derived structural effects, such as the fluid steep waves motion, the local buckling bulges, etc., taking into account the geometrical and material nonlinearities of the RPV and the considered simplified internals.The obtained numerical results in terms of stress intensity and of the capability of the structures to resist relevant seismic loads are, thus, presented and discussed. Moreover the performed analyses allowed to highlight the structures mostly affected by the assumed loading conditions in order to achieve data useful for an upgrading of the design geometry, if any, for the considered reactor.     

Dynamic analysis of a nuclear reactor with fluid–structure interaction: Part I: Seismic loading, fluid added mass and added stiffness effects

The present paper is related to the dynamic (seismic) analysis of a naval propulsion ground prototype (land-based) nuclear reactor with fluid–structure interaction modelling. Many numerical methods have been proposed over the past years to take fluid–structure phenomenon into account in various engineering domains, among which nuclear engineering in seismic analysis. The purpose of the present paper is to make a comparative study of these methods on an industrial case, namely the pressure vessel and internals of a nuclear reactor. A simplified model of the pressure vessel and the internal structure is presented; fluid–structure interaction is characterised by added mass, added stiffness and coupling effects. The basic principles of the mathematical techniques for fluid–structure modelling and dynamic methods used in the analysis are first presented and then applied to compute the eigenmodes and the dynamic response of the fluid–structure coupled system with various numerical procedures (quasi-static, spectral and temporal approaches). Numerical results are presented and discussed; fluid–structure interaction effects are highlighted. As a main conclusion, added mass effects are proved to have a significant influence on the dynamic response of the nuclear reactor.     

采用组合质量-弹簧模型的快堆卧式贮钠罐晃动分析

在核电工程中广泛使用各类形式的储液容器,储液容器的抗震分析必须考虑液体晃动的影响。针对矩形储液容器,不同于传统的单向质量-弹簧模型将液体晃动对容器侧壁、底部的作用都等效为对侧壁的作用,本文提出一种组合质量-弹簧模型及计算公式,模拟了液体晃动分别对容器的侧壁、底部的作用。组合质量-弹簧模型在三维有限元模型上的加载位置更加合理,容器底部的应力结果更加真实。利用组合质量-弹簧模型对中国实验快堆的卧式贮钠罐进行横向晃动有限元计算,算例表明了计算结果的可靠性。组合质量-弹簧模型为储液容器的有限元抗震分析提供了一种有效的方法。     

An advanced seismic analysis of an NPP powerful turbogenerator on an isolation pedestal

This paper presents a detailed seismic analysis of a powerful high-speed Russian turbine within a Nuclear Power Plant. Although dozens of these turbines have worked reliably since the 1970s worldwide, until the last decade, only simplified structural analyses were available due to the turbines’ complicated overall structure and internal design. The current analysis considers the detailed geometry of the turbine itself and the vibration and seismic isolation system within the turbine's pedestal and the full range of operational, accident and seismic loads like high pressure, outside loads induced by pipelines and so on.To solve the problem of the turbine seismic qualification, the following steps have been taken. The first step was to create detailed finite element models of the turbine's high and low pressure parts and rotor system with bearings. Using such models, corresponding simplified models were developed to be included into the coupled model of the system: “Building–Vibroisolation Pedestal–Turbine” (BVT). The second step was the analysis of that coupled system. Soil–structure interaction was considered using actual soil conditions. Three components of time history acceleration were used to define seismic excitation. As the result of BVT system analysis, a full picture of time history displacements and loads was determined. At the same time, a problem of rotor gaps was solved. In the final step, determined loads were applied to the detailed models of the turbine's parts for seismic qualification of the whole structure.     

Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the “force equivalent area” (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques.     

Improved method to demonstrate the structural integrity of high density fuel storage racks

Reracking of existing fuel pools to the maximum extent is desirable from an economical point of view. This goal can be achieved by minimizing the gaps between the spent fuel storage racks. Since the rack design is aimed at enabling consolidated fuel rod storage, additional requirements arise with respect to the design and the structural analysis. The loads resulting from seismic events are decisive for the structural analysis and require a specially detailed and in-depth analysis for high seismic loads. The verification of structural integrity and functionality is performed in two phases. In the first phase the motional behavior of single racks, rows of racks and, where required, of all racks in the pool is simulated by excitation with displacement time histories under consideration of the fluid–structure interaction (FSI). The displacements from these simulations are evaluated, while the loads are utilized as input data for the structural analysis of the racks and the pool floor. The structural analyses for the racks comprise substantially stress analyses for base material and welds as well as stability analyses for the support channels and the rack outside walls. The analyses are performed in accordance with the specified codes and standards.     

Dynamic analysis of a nuclear reactor with fluid–structure interaction: Part II: Shock loading, influence of fluid compressibility

The present paper is related to the dynamic (shock) analysis of a naval propulsion (on-board) reactor with fluid–structure interaction modelling. In a previous study, low frequency analysis has been performed; the present study deals with high frequency analysis, i.e. taking into account compressibility effects in the fluid medium. Elasto-acoustic coupling effects are studied and described in the industrial case. The coupled problem is formulated using the so-called (u, p, φ) formulation which yields symmetric matrices. A modal analysis is first performed on the fluid problem alone, then for the coupled fluid–structure problem in the following cases: (i) with incompressible fluid; (ii) with compressible fluid at standard pressure and temperature conditions; (iii) with compressible fluid at the operating pressure and temperature conditions. Elasto-coupling effects are then highlighted, in particular through the calculation of an elastic energy ratio. As a general conclusion, compressibility effects are proved significant in the dynamic response of the reactor in the high frequency range.     

Seismic behaviour and design of steel coupling beams in a hybrid coupled shear wall systems

The hybrid coupled shear wall system is more efficient and economical than individual structural walls because the steel coupling beams connected shear walls significantly increase the strength, stiffness and energy dissipation capacity of the system. In this study, experimental studies on the steel coupling beam were carried out. The main test variables were the ratios of the coupling beam strength to the connection strength. In addition, the seismic design methods are presented for steel coupling beam–wall connection and shear critical and flexure critical steel coupling beams in hybrid coupled shear wall system consisting of steel coupling beams and reinforced concrete shear walls. Finally, this paper provides background for design guidelines in hybrid coupled shear walls that include steel coupling beam–wall connections and steel coupling beams.     

Coupling effect between liquid sloshing and flexible fluid-filled systems

Current practice in seismic design of flexible liquid-filled systems is reviewed. A coupled fluid-structure finite element method which considers the sloshing effect is developed for the seismic analysis of liquid-filled systems of various geometries with and without internal components. An analysis of the dynamic interaction between the structural vibration and liquid sloshing is also presented. Both rigid and flexible fluid-tank systems of different configurations are considered. Results demonstrate that tank flexibility can affect the amplitude of the free surface wave and hence the sloshing pressure and structural response. This result is consistent with the perturbation analysis. The dynamic interaction depends on (1) the ratio of natural frequency between fluid sloshing and the fluid-tank system and (2) the ratio of the effective areas of the fluid-structure interface and free surface of the fluid. Hence it is expected that in analyzing tanks with flexible internal components, this coupling effect can be more pronounced.     

Soil-structure interaction of nuclear reactor structures considering through-soil coupling between adjacent structures

The theoretical problem concerning the influence of through-soil coupling between adjacent structures on the seismic loading of nuclear reactors has been investigated by considering a soil-structure interaction model in which several three-dimensional flexible structures are bonded to an elastic half-space. These structures, which are allowed to be either similar or dissimilar, are modeled as conventional discrete systems mounted on separate base slabs of close proximity. For the purpose of this study, it is assumed that the stiffness of any structure such as piping connecting the adjacent buildings is negligible.For purposes of comparison, the seismic responses of structural masses are determined both with and without the influence of nearby structures. Both transient and steady-state results are presented and discussed for some typical simplified two- and three-structure complexes. Emphasis is placed on the effects of through-soil coupling on the dynamic response of the system rather than actual magnitudes of response which have previously been treated for plants erected on a single base slab. The significant findings are that nuclear power plants can be designed to achieve a reduction in seismic loads due to interaction with neighboring structures. Conversely, improper plant design and layout may result in mutual reinforcement of resonances with increased loads.     

A modified equilibrium void distribution model applicable to subchannel-scale vapor–liquid cross flow model for conventional square and tight lattice BWR fuel bundles

As a part of the advanced subchannel code development project sponsored by Ministry of Economy, Trade and Industry, Japan, this paper describes improvement of the equilibrium void distribution model that is a main part of the vapor–liquid cross flow model.The three-component cross flow (TCCF) model is defined as the present framework that separates contributions of diversion, turbulent mixing and void drift. The Lahey's void settling model is introduced to express the latter two components. Based on the high-resolution air–water database and other published data of steam-water tests, general trends of vapor–liquid cross flow processes are examined. It can be assumed that subchannel void distributions are dominated by the three major effects, i.e. the fluid dynamic effect, the geometrical effect and the narrow gap effect.The equilibrium void distribution model is modified to include the above-mentioned three effects. Three characteristic parameters are assigned for each of the three effects and they are identified experimentally as functions of the void fraction. Multi-dimensional lattice geometries are incorporated based on the two-dimensional flow network model. The network equation is constructed by mapping the equilibrium void balance problem into the force-deflection problem. The resultant models are verified based on equilibrium void distribution data obtained by Sadatomi and Kawahara.     

Dynamic buckling experiments of fluid-structure-coupled co-axial thin cylinder

The purpose of this paper is to clarify dynamic buckling behaviours such as buckling mode and buckling pressure for thin cylindrical shells immersed in fluid subjected to seismic excitations. For this purpose, dynamic buckling experiments of thin cylindrical shells placed inside a rigid liquid container are carried out using a shaking table. These shells and the container are intended to represent thermal baffles and a main vessel of a fast breeder reactor, respectively. The fluid pressure caused by horizontal excitation induces buckling deformation which involves flower-shaped deformation, which is a type of external pressure buckling. The buckling pressure is measured with various types of the test cylinders under seismic excitations and this pressure is confirmed to agree with static buckling pressure predicted by static buckling analysis. It is also found that sub-harmonic vibration occurs under a certain sinusoidal excitation inducing a sudden increase in response displacement at a lower pressure level than the buckling pressure under seismic excitations. Based on these experiments, it is pointed out that, in seismic design, to prevent the buckling of thermal baffles, static buckling analyses can be used as long as sub-harmonic vibration does not occur.     

Calculation of equivalent static loads and its application

All the forces in the real world act dynamically on structures. Since dynamic loads are extremely difficult to handle in analysis and design, static loads are usually utilized with dynamic factors. Generally, the dynamic factors are determined from design codes or experience. Therefore, static loads may not give accurate solutions in analysis and design and structural engineers often come up with unreliable solutions. Two different methods are proposed for the transformation of dynamic loads into equivalent static loads (ESLs). One is an analytical method for exact ESLs and the other is an approximation method. The exact ESLs are calculated to generate identical response fields such as displacement and stress with those from dynamic loads at a certain time. Some approximation methods are proposed in engineering applications, which generate similar response fields from dynamic loads. They are divided into the displacement-based approach and the stress-based approach. The process is derived and evaluated mathematically. Standard examples are selected and solved by the proposed method and error analyses are conducted. Applications of the method to structural optimization are discussed.     

Construction materials for reactor buildings for decommissioning and recycling

A conceptual fluid–steel structure was studied to investigate the seismic characteristics of its use in the reactor building of nuclear power plants. The results of the earthquake response analysis of the conceptual fluid–steel structure showed that the structure had the same seismic safety ability as conventional reactor buildings. Applying the fluid–steel structure to a rector building, results in the following advantages: more elastic and light weight building materials, reducing the decommissioning wastes; the ability to recycle the structure materials because the fluid in the steel structure can be discharged the steel can be reused easily; the fluid in the steel structure has the possibility of reducing the seismic response of the structure by the sloshing damper effect. Further study is encouraged by this results.     

Design and corrosion study of a closed spallation target module of an accelerator-driven system (ADS)

At the Forschungszentrum Karlsruhe (FZK) the characteristics of an accelerator-driven subcritical reactor system (ADS) are critically evaluated, mainly with respect to the potential of transmutation of minor actinides and long-lived fission products, to the feasibility and to safety aspects. The work is concentrating on system design, neutronics, thermalhydraulics, safety, materials and corrosion. This article describes the FZK approach to design a closed 4 MW(th) spallation target module with a solid beam window and eutectic lead–bismuth (Pb–Bi) as spallation material and cooling fluid, which is going to be implemented in the FZK three-beam concept of an ADS. This multi-beam concept shows significant improvements towards single-beam concepts from the literature with respect to power distribution in the subcritical blanket and thermal loads of heat removal from the beam window and the spallation region. For some selected martensitic and austenitic steels, corrosion tests in static lead are performed to examine their suitability as structural or window materials. Alloying aluminum into the surface layer by high-power electron beam treatment, corrosion can be reduced to nearly zero. One prerequisite to minimize corrosion is a proper oxygen control system (OCS) via the gas-phase to set the oxygen concentration in the liquid Pb–Bi. The dynamic behaviour of this oxygen control system is described. Finally, the KArlsruhe Lead LAboratory (KALLA) is introduced, the objectives of which are technological, thermal-hydraulic and corrosion investigations into the beam window, the spallation target module and the primary system of an ADS.     

高温气冷堆蓄电池组地震易损性研究

为验证核电厂发生地震外部事件时的电力安全,需要对蓄电池组进行抗震鉴定试验。本文以高温气冷堆(HTR)核电厂安全级蓄电池组为研究对象、以安全级蓄电池组抗震鉴定试验数据和工程经验为基础,通过识别、量化蓄电池组的地震易损性变量,并应用基于试验的易损性分析法推导出地震易损性曲线和高置信度低失效概率(HCLPF)抗震能力。研究结果表明,安全级蓄电池组的抗震能力远高于核电厂设计基准地震动需求。     

CAST3M/ARCTURUS: A coupled heat transfer CFD code for thermal–hydraulic analyzes of gas cooled reactors

The safety of gas cooled reactors (High Temperature Reactors (HTR), Very High Temperature Reactors (VHTR) or Gas Cooled Fast Reactors (GFR)) must be ensured by systems (active or passive) which maintain loads on component (fuel) and structures (vessel, containment) within acceptable limits under accidental conditions. To achieve this objective, thermal–hydraulics computer codes are necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Some key safety questions are related to the evaluation of decay heat removal and containment pressure and thermal loads. This requires accurate simulations of conduction, convection, thermal radiation transfers and energy storage. Coupling with neutronics is also an important modeling aspect for the determination of representative parameters such as neutronics coefficient (Doppler coefficient, Moderator coeffcient, …), critical position of control rods, reactivity insertion aspects, …. For GFR, the high power density of the core and its necessary reduced dimension cannot rely only on passive systems for decay heat removal. Therefore, forced convection using active safety systems (gas blowers, heat exchangers, …) are highly recommended. Nevertheless, in case of station black-out, the safety demonstration of the concept should be guaranteed by natural circulation heat removal. This could be performed by keeping a relatively high back-up pressure for pure helium convection and also by heavy gas injection. So, it is also necessary to model mixing of different gases, the on-set of natural convection and the pressure and thermal loads onto the proximate or guard containment. In this paper, we report on the developments of the CAST3M/ARCTURUS thermal–hydraulics (Lumped Parameter and CFD) code developed at CEA, including its coupling to the neutronics code CRONOS2 and the system code CATHARE. Elementary validation cases are detailed, as well as application of the code to benchmark problems such as the HTR-10 thermal–hydraulic exercise. Examples of containment thermal–hydraulics calculations for fast reactor design (GFR) are also detailed.     

乏燃料贮存格架流固耦合参数的数值模拟

乏燃料贮存格架是储存乏燃料组件的重要设备。在地震载荷下,其响应是非线性的,可能产生滑移、颠覆等。发生地震时,存在于格架间隙内的流体耗散了结构的能量,保证了格架的完整性。本文使用3/10缩比模型,利用CFD软件Fluent进行了乏燃料贮存格架2D瞬态分析。计算过程中利用动网格方法模拟格架强迫振动,并进行了参数不确定性分析。利用CFD瞬态流体力分别获得了双Ⅱ区、双Ⅰ区格架附加质量矩阵。利用同轴圆柱体附加质量的计算解与解析解进行对比验证,证明了本文计算方法的准确性。本文计算所得的附加质量矩阵可为乏燃料贮存格架结构动态软件提供流固耦合参数。     

快堆堆芯抗震实验研究与有限元分析进展

堆芯的安全评价是快中子增殖反应堆抗震设计的一个重要问题。发生地震时,应该确保堆芯组件的结构完整性和核电厂能按要求紧急停堆。数百根堆芯组件之间存在着间隙,组件与堆芯支承处也存在间隙,整个堆芯被液钠包围,堆芯的抗震计算比较困难。本文重点介绍近年来法国、日本、意大利以及中国等国家针对快堆做过的一系列实验和理论研究进展情况。