Preliminary results of the experimental study of PFCs exposure to ELMs-like transient loads followed by high heat flux thermal fatigue

This paper deals with the empirical evaluation of the plasma facing components (PFCs) lifetime under transient events, such as type I edge localized modes (ELMs), and high heat flux (HHF) thermal fatigue expected during ITER normal operations and slow transient events. The first results of experiments which are ongoing in the frame of an EU/RF collaboration are presented, in particular the results of the first campaign of exposure to ELMs-like load and HHF thermal fatigue performed on PFC monoblock mock ups. For carbon-fiber composite material the erosion was determined by the analysis of the PAN fibres. The erosion of tungsten is reported as melt layer movement and crack formation. As a preliminary result of the experiments performed one can say that the mock up power handling capability seems so far not to be affected by the ELMs and HHF cycling tests.     

Thermal fatigue damage in monofilament reinforced copper for heat sink applications in divertor elements

In fusion reactor systems extreme conditions require materials with high temperature and radiation resistance. The divertor component consists of a plasma facing W plate attached to a Cu heat sink to extract the heat from the nuclear reaction chamber coolant. The Coefficient of Thermal Expansion (CTE) mismatch between the W plate and the Cu heat sink causes interface delamination reducing the long term stability of the divertor.To avert this problem, composites are developed as interlayer materials with a high thermal conducting Cu matrix reinforced with up to 50 vol.% SiC or W monofilaments to increase the mechanical strength and to reduce the CTE mismatch. Thermal stresses are transferred from the macroscopic interface between the components into the bulk of the composite. Oscillating micro stresses may lead to fiber delamination and matrix damage during thermal cycling. Different matrix alloys, fiber materials and interface designs are investigated.In situ neutron diffraction performed during thermal cycling show the effect of bonding strength on the stress amplitudes expected under service conditions. The long term stability is tested by measurements after further ex situ cycling. Thermal fatigue damage and its propagation are visualized by in situ as well as ex situ high resolution synchrotron tomography. The combination of both methods helps to understand the strain induced damage mechanisms. Weak bonding leads to delamination of the fiber-matrix interfaces. Strong bonding causes severe matrix deformation and damage. Fiber cracks originating from sample production cause accumulating thermal fatigue damage during thermal cycling.     

Evaluation of novel Ti-doped 3D carbon-carbon composites under transient thermal loads

3D Ti-doped and undoped carbon-carbon composites (CFCs) were exposed to transient thermal loads to simulate plasma disruptions, in the electron beam test facility JUDITH at different power densities and multiple shots in order to study the evolution in the behavior of the material. The thermal shock response of the undoped and Ti-doped materials was compared in order to study the influence of titanium carbide as dopant. The erosion itself is driven during the first shots by macroscopic erosion (brittle destruction), which is a result of thermally induced stresses. With increasing number of shots, no more brittle destruction is observed and the main erosion mechanism is sublimation due to local overheating. This is also confirmed by the decrease of the erosion rate with increasing the number of shots. The pitch fibers are hardly affected by the applied heat loads and they show almost no erosion, especially in the Ti-doped composite.     

Thermo-physical properties and transient heat transfer of concrete at elevated temperatures

The objective of this study is to produce our own experimental data of physical properties of domestic concrete used in Korean NPPs, and to study on the thermal behavior of concrete exposed to high temperature conditions. The compressive strength and chemical composition of the concrete used in the Yonggwang NPP units 3 and 4 were analyzed. The chemical composition of Korean concrete is similar to that of US basaltic concrete. The thermal properties of the concrete, such as density, conductivity, diffusivity, and specific heat were also measured with a wide temperature range of 20–1100 °C. Most thermo-physical properties of concrete decrease with an increase in temperature except for the specific heat, and particularly the conductivity and the diffusivity are a 50% lower at 900 °C as compared with the values at room temperature. The specific heat increases until 500 °C, decreases from 700 to 900 °C, and then increases again when temperature is above 900 °C. In this work, we also have performed CORCON analysis and MCCI experiments to simulate a transient thermal behavior of concrete exposed to high temperature conditions. The measured maximum downward heat flux to the concrete specimen was estimated to be about 2.1 MW m⁻² and the maximum erosion rate of the concrete to be 175 cm h⁻¹ with maximum erosion depth of about 2 cm. In the CORCON analysis, it is found that the concrete compositions have an important effect upon concrete erosion.     

Intelligent system for transient data collection and fatigue monitoring of pressurised water reactors nuclear steam supply system

Electricité de France (EDF), the French national electricity company, is operating 54 standardised pressurised water reactors. This about 500 reactor-years experience in nuclear stations operation and maintenance area has allowed EDF to develop its own strategy for monitoring of age-related degradations of NPP systems and components relevant for plant safety and reliability. After more than fifteen years of experience in regulatory transient data collection and seven years of successful fatigue monitoring prototypes experimentation, EDF decided to design a new system called SYSFAC (acronym for SYstème de Surveillance en FAtigue de la Chaudière) devoted to transient logging and thermal fatigue monitoring of the reactor coolant pressure boundary. The system is fully automatic and directly connected to the on-site data acquisition network without any complementary instrumentation. A functional transient detection module and a mechanical transient detection module are in charge of the general transient data collection. A fatigue monitoring module is aimed towards a precise surveillance of five specific zones particularly sensible to thermal fatigue. After a first step of preliminary studies, the industrial phase of the SYSFAC project is currently going on, with hardware and software tests and implementation. The first SYSFAC system will be delivered to the pilot power plant by the beginning of 1996. The extension to all EDF’s nuclear 900 MW is planned after one more year of feedback experience.     

核二级波纹管截止阀在瞬态热冲击作用下的数值模拟

为研究瞬态承压热冲击对核二级波纹管截止阀的结构强度和疲劳寿命的影响,基于流固耦合及热边界条件相关理论,通过Fluent和ANSYS有限元软件对核二级截止阀阀体进行热流固耦合分析,研究阀体监测点在不同时间点下温度场、热应力和疲劳寿命的变化,以及热冲击作用时间对疲劳寿命的灵敏度的影响。结果表明:瞬态承压热冲击对阀体的温度场、结构强度、疲劳寿命和敏感度影响巨大,必须消除这种影响,以此来保证核二级波纹管截止阀的高安全性和高可靠性。      

Fatigue monitoring and life cycle analysis of a recuperative heat exchanger in the primary circuit

The example of a recuperative heat exchanger is used to demonstrate the temperature monitoring and the thermal fatigue assessment of a component in the primary circuit of a PWR. After describing the geometry and the operating conditions, the placement of the thermocouples at locations of interest is shown. The temperature recordings show that the existing operational instrumentation was very slow and the fluid transient were actually much faster than anticipated. With a computer program allowing the reduction from outside wall to fluid temperature, the thermal inertia of the old instrumentation could be quantified. Consequently, old recordings from slow instrumentation could still be used. Thermal finite element computations are presented that yield quite favourable correlations between computations and measurements.     

基于运行数据的船用堆波动管热分层现象研究

基于运行数据将船用堆波动管热分层划分为升功率、降功率、变工况、小喷淋流量4类典型瞬态,对4类典型瞬态分别进行无量纲里查德森数（Ri）分析、瞬态工况数值模拟计算,得到波动管在4类典型瞬态下水平管段的热分层区间长度、持续时间和最大温差。结果表明,升功率和降功率瞬态热分层仅单次贯穿波动管,升功率瞬态的接头部位循环的热波动以及小喷淋流量瞬态水平段的长区间、长时间、大温差的热分层现象和变工况导致的热应力波动可能影响到波动管的安全。本文提出的基于运行数据的波动管热分层现象研究方法为后续热应力和热疲劳分析奠定了基础,同时可以为其他容积设备热分层研究提供参考。      

Numerical Analysis of Fatigue Behavior of ITER-Like Monoblock Divertor Interlayer Under Coupled Heat Loads

High-confinement mode is a very prominent operation style for future fusion device due to its unique advantages. However, the conjuncted edge localized modes (ELMs) are very difficult to control so that divertor plates are very prone to suffer both stationary high heat flux (HHF) loads of long-pulse operating mode and transient shock loads of ELMs. Most previous researches focus on degradation of plasma facing material (PFM), however, as a layer joining PFM and cooling tube, the soft copper interlayer suffers concentrated thermal stress loads due to mismatched thermal expansion of PFM and cooling tube. Its thermal fatigue behavior under such coupled loads is also of great significance to structural safety of divertor component. With such a motivation, the reduction effects on fatigue life time of a typical interlayer of monoblock divertor under series of coupled HHF and ELMs shock loading conditions are investigated. It is found that: (1) The transient shock feature of ELMs loading is propagated into interlayer with less sharp pattern. The increase of damage induced by coupled ELMs loading is limited in single cycle, while the accumulated damage of multiple consecutive coupled loading cycles is increased nonlinearly. (2) Under the coupled HHF and ELMs loading, the fatigue life time of interlayer is generally decreasing. The magnitude of decrease is increasing nonlinearly with the magnitude of ELMs peak and averaged heat flux. (3) For three characteristic parameters of ELMs shock loading such as frequency, duration and peak heat flux, the peak heat flux and frequency are two parameters more sensitive to determine coupled reduction effects on fatigue lifetime of the interlayer, while for high frequency case, time averaged heat flux takes the lead.     

Development of thermal fatigue damage in 1CrMoV rotor steel

Crack initiation endurances have been determined for a 1CrMoV rotor steel in uniaxial service cycle thermo-mechanical fatigue (TMF) tests formulated to simulate a range of steam turbine start cycles with a maximum temperature of 565 °C. The experimental details for these TMF tests are described.Post test inspection has been employed to characterise the associated thermal fatigue damage mechanisms for the steel which are observed to be dependent on the magnitude of the thermal transient in the TMF cycle.The lowest resistance to thermal fatigue damage development occurs in these tests when the conditions determine that the rate of creep damage accumulation below the surface exceeds the rate of fatigue crack development at the surface.     

三维结构瞬态温度,位移和热应力计算的等效时间法

本文提出了一种计算二维结构瞬态温度，位移和热应力的近似方法－等效时间法。利用事先得出的热边界幅值的单位变化所引起的结构响应的有限元分析结构。通过简单的代数运算即可求出同样热边界模式下热边界模式下热边界幅值的任意变化过程在结构中产生的响应。这种近似方法有很高的计算效率。通过算例表明，与详细的有限元计算结果相比，在工程允许的范围内误差满足工程需要。等效时间法可肜于对计算效率要求较高的复杂结构的实时分析     

Particle emissions from graphite during disruption heat load

It is regarded that the life time of a graphite divertor wall is determined by erosion during the disruptions. In the disruption heat load experiments, the erosion of graphite due to particle emissions has been observed to be extremely serious. The particle emission process is considered based on localized thermal stress at the surface. There is a critical heat flux for the particle emission to take place. Erosion depth observed in the experiment is also discussed using the present simple model.     

Fluid-structure interaction analysis for pressurizer surge line subjected to thermal stratification

Serious mechanical damages such as cracks and plastic deformations due to excessive thermal stress caused by thermal stratification have been experienced in several nuclear power plants. In particular, the thermal stratification in the pressurizer surge line has been addressed as one of the significant safety and technical issues. In this study, a detailed unsteady computational fluid dynamics (CFD) analysis involving conjugate heat transfer analysis is performed to obtain the transient temperature distributions in the wall of the pressurizer surge line subjected to stratified internal flows either during out-surge or in-surge operation. The thermal loads from CFD calculations are transferred to the structural analysis code which is employed for the thermal stress analysis to investigate the response characteristics, and the fatigue analysis is ultimately performed. In addition, the thermal stress and fatigue analysis results obtained by applying the realistic temperature distributions from CFD calculations are compared with those by assuming the simplified temperature distributions to identify some requirements for a realistic and conservative thermal stress analysis from a safety point of view.     

Creep—fatigue test of a thick-walled vessel under thermal transient loadings

To improve the damage evaluation methods in the design code for Fast Breeder Reactors (FBRs), a series of creep—fatigue tests of structural models under thermal transient loadings are going on at Oarai Engineering Center of the Power Reactor and Nuclear Fuel Development Corporation (PNC). Test models are designed to incorporate representative structures of components and pipings used in FBRs and are subjected to severer cyclic thermal transients than those experienced in FBRs. The test is planned to be continued until failure occurs. This paper describes the creep—fatigue test results and their damage evaluation for the first test model.A 40 mm thick vessel model made of SUS304 austenitic stainless steel was subjected to cyclic thermal transients, in which sodium at 600°C and 250°C flowed repeatedly. The period of each transient was 2 h. Cracks were observed at seven test portions in the model after 1002 cycles of the thermal transients.Elastic and inelastic analyses were performed to evaluate creep—fatigue damage and crack propagation. The safety margins included in the creep—fatigue design methods based on elastic analysis as well as those based on inelastic analysis are discussed. Finally fracture mechanics analyses were performed to explain the observed crack growth.     

Fuel rod model based on Non-Fourier heat conduction equation

In this paper we explore the applicability of a fuel rod mathematical model based on Non-Fourier transient heat conduction as constitutive law for the Light Water Reactors transient analysis (LWRs). In the classical theory of diffusion, Fourier law of heat conduction is used to describe the relation between the heat flux vector and the temperature gradient assuming that the heat propagation speeds are infinite. The motivation for this research was to eliminate the paradox of an infinite thermal wave speed. The time-dependent heat sources were considered in the fuel rod heat transfer model. The close of the Main Steam Isolated Valves (MSIV) transient in a Boiling Water Reactor (BWR) was analyzed by different relaxation times. The results show that for long-times the heat fluxes on the clad surface under Non-Fourier approach can be important, while for short-times and from the engineering point of view the changes are very small. Some results from transient calculations are examined.     

Effects of operating experience on nuclear plant design

Operating experience is utilized in formulating engineering principles for the safe and effective design of nuclear plants. Persistently difficult design areas, such as (a) flow induced vibration, (b) thermal fatigue, (c) unpredictable transient response, and (d) coolant chemistry related corrosion are singled out for special consideration.     

TACR热传输系统热工水力瞬态分析

钍燃料的利用对于缓解核燃料资源短缺具有重要意义,坎杜型反应堆(Canadian Deuterium Uranium,CANDU)在堆芯布置、中子利用效率及先进燃料循环方面具有较高的灵活性,使得其在CANDU反应堆中引入钍燃料循环更具现实意义。CANDU型反应堆中钍基燃料应用关键基础技术研究是加拿大与我国正在开展的合作课题,其中开发自主的CANDU堆堆芯热工水力设计和安全分析程序是钍基燃料应用必不可少的设计工作之一。本文针对CANDU型反应堆热传输系统结构特点,采用FORTRAN程序设计语言开发了适用于CANDU型反应堆热传输系统的热工水力瞬态分析程序CANTHAC(CANDU Thermal-Hydraulic Analysis Code)。利用CANTHAC对钍基先进CANDU堆(Thorium-based Advanced CANDU Reactor,TACR)进行了瞬态分析,计算工况包括满功率稳态、无保护蒸汽发生器(Steam Generator,SG)二次侧给水温度降低事故及完全失流事故。其中,满功率稳态计算结果与清华大学设计的钍基先进CANDU堆TACR设计值吻合较好,相对误差不超过2%,在可接受范围内;无保护SG二次侧给水温度降低事故及完全失流事故在计算条件下所得的燃料温度及系统压力等关键热工水力参数均在安全限值内,满足安全准则要求。程序为模块化编程,便于移植和改进,具有一定的通用性,为进一步研究工作奠定了基础。     

Behavior of graphite under heat load and in contact with a hydrogen plasma

Graphite is extensively used in large tokamaks today. In these machines the material is exposed to vacuum, to intense heat loads, and to the edge plasma. The use of graphite in such machines, therefore, depends on the outgassing behavior, the heat shock resistance, and thermochemical properties in a hydrogen plasma. Investigations of these properties made at different laboratories are described here.Experiments conducted at Sandia National Laboratories (SNL), Livermore, and the Max-Planck-Institut für Plasmaphysik (IPP) in Garching showed that the outgassing behavior of fine-grain reactor-grade graphite and carbon fiber composites depends on the pretreatment (manufacturing and/or storage). However, after proper outgassing the samples tested behave similarly in the case of fine-grain graphite, but the outgassing remains high for the carbon fiber composites.Heat shock tests have been made with the Electron Beam Test System (EBTS) at SNL, Albuquerque. Directly cooled graphite samples (FE 159 graphite brazed onto Mo tubes) showed no failure at a heat load of 700 W/cm², 20 s; or 10 kW, 1 s. Thermal erosion due to sublimination and particle emission from the graphite surface was observed. This effect is related to the surface temperature and becomes significant at temperatures above 2500°K. Fourteen different types of graphite were tested; the main differences among these samples were the different surface temperatures obtained under the same heating conditions. Cracking due to heat shocks was observed in some of the samples, but none of the carbon fiber composites failed.Thermochemical properties have been tested in the PISCES plasma generator at UCLA for ion energies of around 100 eV. The formation of C-H compounds was observed spectroscopically at sample temperatures of around 600°C. However, this chemical reaction did not lead to erosion as observed in beam experiments but to a drastic change of the surface structure due to redeposition. Carbon-hydrogen lines were still observed at sample temperatures of around 100°C. Under these conditions the erosion yield is high and in agreement with those measured in beam experiments.     

A computational fatigue analysis of cyclic thermal shock in notched specimens

Thermal shock induced fatigue plays a role in the assessment of the lifetime of different components in the primary cooling circuit of a nuclear plant. In spite of the implementation of substantial and costly safety factors, a few, unexpected cases of fatigue failure have occurred. Here we report on a laboratory experiment which mimics the thermal loading observed in such components. A finite element thermal stress analysis using a calibrated, elasto-plastic, combined kinematic-isotropic cyclic hardening material model is presented. The distribution of transient stresses and strains in the specimens subjected to cyclic thermal shock, are used to predict the number of cycles to crack initiation with a fatigue curve that has been calibrated experimentally with data from equivalent specimens under pure mechanical fatigue. Our results indicate that cyclic thermal shock induced ratcheting occurs locally near the tip of the notch in the specimens, and the potential impact on the number of cycles to crack initiation is explored.     

Testing of high heat flux components manufactured by ENEA for ITER divertor

ENEA is involved in the International Thermonuclear Experimental Reactor (ITER) R&D activities and in particular in the manufacturing of high heat flux plasma-facing components, such as the divertor targets. During the last years ENEA has manufactured actively cooled mock-ups by using different technologies, namely brazing, diffusion bonding and HIPping. A new manufacturing process that combines two main techniques PBC (Pre-Brazed Casting) and the HRP (Hot Radial Pressing) has been set up and widely tested.A full monoblock medium scale vertical target, having a straight CFC armoured part and a curved W armoured part, was manufactured using this process.The ultrasonic method was used for the non-destructive examinations performed during the manufacturing of the component, from the monoblock preparation up to the final mock-up assembling. The component was also examined by thermography on SATIR facility (CEA, France), afterwards it was thermal fatigue tested at FE200 (200 kW electron beam facility, CEA/AREVA France).The successful results of the thermal fatigue testing performed according the ITER requirements (10 MW/m², 3000 cycles of 10 s on both CFC and W part, then 20/15 MW/m², 2000 cycles of 10 s on CFC/W part, respectively) have confirmed that the developed process can be considerate a candidate for the manufacturing of monoblock divertor components. Furthermore, a 35-MW/m² Critical Heat Flux was measured at relevant thermal–hydraulics conditions at the end of the testing campaign.This paper reports the manufacturing route, the thermal fatigue testing results, the pre and post non-destructive examination and the destructive examination performed on the ITER vertical target medium scale mock-up.These activities were performed in the frame of EFDA contracts (04-1218 with CEA, 93-851 JN with AREVA and 03-1054 with ENEA).