Estimation of manufacturing defects in the creep range

Various concepts are available for the estimation of crack-like manufacturing defects of components operated in the creep regime. All these methods raise questions as to their practicable application respectively transferability to the component. Discussing these problems resulted in the demand for the development of a “Two-Criteria-Diagram” which can represent practicable load conditions and make statements on the failure modus.     

The dependence of the high temperature mechanical properties of austenitic stainless steels on implanted helium

High temperature helium embrittlement effects on the creep properties of AISI 316 SS (solution annealed + aged) and DIN 1.4970 SS (solution annealed + cold worked + aged) have been investigated. The generation of helium due to (n. α) nuclear reactions in a fusion reactor environment has been simulated by homogeneous helium implantation at a cyclotron. The creep rupture tests with various applied tensile stresses have been carried out at 1023 K. (316 SS) and 1079 K (1.4970 SS). respectively, with four differently treatly sets of samples: (1) unimplanted controls; (2) after room temperature implantation of 100 appm He; (3) after implantation of 100 appm He at test temperature; (4) creep tested at high temperature during implantation (“in-beam”) with implantation rates of 10–100 appm He/b. In contrast to the ductile behaviour with transgranular failure of the unimplanted controls, all He-implanted samples showed brittle, intergranular early failure. The embrittlement effect was enhanced for the “in-beam” tested samples. The difference between the different treated sets of samples can be related to different bubble microstructures investigated by TEM. In addition, a comparison to reactor data for the DIN 1.4970 SS is presented.     

Fusion neutron test facility requirements for interactive effects in structural and high-heat-flux components

A relevant design data base is needed for structural components in near-term and commercial fusion devices. A high-flux, high-fluence fusion neutron test facility is required for testing the failure mechanisms and lifetime-limiting features for first wall, blanket, and high-heat-flux components. We describe here the key aspects of the fusion environment which influence the response of structural and high-heat-flux components. In addition to test capabilities for fundamental radiation-effects phenomena, e.g., swelling, creep, embrittlement, and hardening, it is shown that the facility must provide an adequate range of conditions for accelerated tests to study the limitations on component lifetime due to the interaction between such fundamental phenomena. In high-heat-flux components, testing of the failure mechanisms of duplex structures is shown to require maintenance of an appropriate temperature gradient in the 14-MeV neutron field. Thermal stresses are shown to result in component failure, particularly when the degradation in the thermal conductivity and mechanical properties by irradiation are considered. Several factors are discussed for assessment of the failure modes of the first wall and blanket structures. These are displacement-damage dose and dose rate, the amount of helium gas generated, the magnitude of irradiation and thermal creep, prototypical temperature and temperature-gradient distributions, module geometry, and external mechanical constraints.     

高温含缺陷结构与时间相关的失效评定图

针对高温含缺陷结构的稳态蠕变情况,在裂纹扩展控制参量JT积分的基础上建立了一个与时间相关的失效评定图,给出了与之相关的参量Kr、Lr、Lr^max,σ0.2^3以及等时应力应变曲线的定义和确定方法,详细讨论了材料的蠕变断裂韧性Kmat^c的试验测定方法、参数估算法以及图算法,并据此建立了2．25CrlMo钢—系列的高温失效评定曲线研究结果表明：当环境温度低于材料蠕变温度时．服役时间对2．25CrlMo钢的高温失效评定曲线影响不大,可以采用R6的通用失效评定曲线进行粗略评定;当环境温度高于材料蠕变温度时,服役时间越长、温度越高高温失效评定曲线与R6通用失效评定曲线相比变化越大．这时必须采用与时间相关的失效评定图。     

Thermo-mechanical analysis of supercritical pressure light water-cooled fast reactor fuel rod by FEMAXI-6 code

Thermo-mechanical behaviors of supercritical pressure light water cooled fast reactor (SWFR) fuel rod and cladding have been investigated by FEMAXI-6 (Ver.1) code with high enriched MOX fuel at elevated operating condition of high coolant system pressure (25 MPa) and high temperature (500 °C in core average outlet temperature). Fuel rod failure modes and associated fuel rod design criteria that is expected to be limiting in SWFR operating condition have been investigated in this fuel rod design study. Fuel centerline temperature is evaluated to be 1853 °C and fission gas release fraction is about 45% including helium production. Cumulative damage fraction is evaluated by linear life fraction rule with time-to-rupture correlation of advanced austenitic stainless steel. In a viewpoint of mechanical strength of fuel cladding against creep rupture and cladding collapse at high operation temperature, currently available stainless steels or being developed has a potential for application to SWFR. Admissible design range in terms of initial gas plenum pressure and its volume ratio are suggested for fuel rod design The stress ranges suggested by this study could be used as a preliminary target value of cladding material development for SWFR application.     

Potential failure mode and effects analysis for the ITER NB injector

The failure mode and effects analysis (FMEA) is a widely used analytical technique that helps in identifying and reducing the risks of failure in a system, component or process.The application of a systematic method like the FMEA was deemed necessary and adequate to support the design process of the ITER NBI (neutral beam injector). The approach adopted was to develop a FMEA at a general “system level”, focusing the study on the main functions of the system and ensuring that all the interfaces and interactions are covered among the various subsystems. The FMEA was extended to the whole NBI system taking into account the present design status. The FMEA procedure will be then applied to the detailed design phase at the component level, in particular to identify (or define) the ITER Class of Risk.Several important failure modes were evidenced, and estimates of subsystems and components reliability are now available. FMEA procedure resulted essential to identify and confirm the diagnostic systems required for protection and control, and the outcome of this analysis will represent the baseline document for the design of the NBI and NBTF integrated protection system.In the paper, rationale and background of the FMEA for ITER NBI are presented, methods employed are described and most interesting results are reported and discussed.     

Irradiation creep in non-fissile metals and alloys

The principal methods used in measuring irradiation creep in non-fissile metals and alloys are described and the limitations of the techniques emphasised. The theoretical models of irradiation creep are surveyed and the experimental data on thermal and fast reactor core component materials, such as zirconium alloys and austenitic steels, are reviewed. In particular, the effects of compositional and metallurgical variables and irradiation parameters (temperature, dose and dose rate) on the magnitudes of the irradiation creep are assessed. Finally, the additional theoretical studies required to further the understanding of the phenomenon and the experimental work necessary for optimising the design and operation of thermal and fast reactors are summarised.     

Theory of enhanced high temperature cyclic creep in polycrystalline material

A theory of high temperature cyclic stress creep rate enhancement is developed that is based on the mechanism of the athermal generation of self-interstitial atoms. It is first shown that interstitial atoms cannot be generated athermally at a fast enough rate to affect the creep rate under a constant stress. Under cyclic stressing it may be possible for athermally-generated interstitial atoms to produce an enhanced creep rate. The effect can only be important in the high temperature region from about one-third to one-half the melting temperature of the material and then only for fine-grained material that deforms in double or multiple slip. The predicted effect is larger the higher is the frequency of the cyclic stress. This creep enhancement effect should be taken into account in the design and material selection of the first-wall material of fusion reactors that operate in pulsed modes, such as reactors using lasers.     

高温翅片管的长时蠕变屈曲分析及评定方法研究

由于翅片管处于高温环境和外压载荷下,需要考虑其发生蠕变屈曲失效的风险。本文对翅片管在高温环境下的蠕变屈曲分析及评定方法进行了研究,提出了一种基于塑性本构和蠕变本构的有限元长时蠕变屈曲分析方法,并通过数值拟合,获得了高温屈曲的失效评定图以及失效评定公式,提出了一种方便应用于工程的快速评定方法。针对翅片管结构,将该方法的评定结果与规范中的屈曲分析评定结果进行对比,验证了该方法的可行性。同时研究了在有压力波动的情况下,结构的临界屈曲时间与载荷历程的关系,为复杂结构和复杂载荷工况的蠕变屈曲分析奠定了基础。     

Bottom nozzle failure mechanism of water reactor pressure vessel under severe accident conditions

Most of past studies devoted to the creep rupture of a nuclear reactor pressure vessel (RPV) lower head under severe accident conditions, have focused on global deformation and rupture modes. Limited efforts were made on local failure modes associated with penetration nozzles as a part of TMI-2 vessel investigation project (TMI-2 VIP) in 1990s. However, it was based on an excessively simplified shear deformation model. In the present study, the mode of nozzle failure has been investigated using data and nozzle materials from Sandia National Laboratory's lower head failure experiment (SNL-LHF). Crack-like separations were revealed at the nozzle weld metal to RPV interfaces indicating the importance of normal stress component rather than the shear stress in the creep rupture. Creep rupture tests were conducted for nozzle and weld metal materials, respectively, at various temperature and stress levels. Stress distribution in the nozzle region is calculated using elastic–viscoplastic finite element analysis (FEA) using the measured properties. Calculation results are compared with earlier results based on the pure shear model of TMI-2 VIP. It is concluded from both LHF-4 nozzle examination and FEA that normal stress at the nozzle/lower head interface is the dominant driving force for the local failure. From the FEA for the nozzle weld attached in RPV, it is shown that nozzle welds failure occur by displacement controlled fracture of nozzle hole not by load controlled fracture of internal pressure. Considering these characteristics of nozzle weld failure, new concept of nozzle failure time prediction is proposed.     

Analyses of a restrained growth graphite irradiation creep experiment

Graphite is used as a moderator, reflector and structural component in pebble bed and prism High Temperature Reactors (HTRs). It is fortunate to reactor designers that irradiated graphite shows remarkably high creep behaviour under the influence of fast neutron irradiation at temperatures far below those required for significant creep strains to be generated in unirradiated graphite. This creep behaviour is important in the design of nuclear graphite reactor cores because the self-induced shrinkage stresses generated in typical core components during irradiation can be relieved. However, there are no reliable data on high fluence irradiation creep and the mechanistic understanding of the irradiation creep is insufficiently developed to reliably extrapolate to the high fluences expected of graphite in future HTR designs. The understanding of irradiation creep is further complicated because it has been experimentally observed that irradiation creep strain in graphite modifies other properties in particular the coefficient of thermal expansion. In addition modified changes in Young's modulus in crept specimens have been reported and it has also been postulated that irradiation creep may also modify dimensional change. The assessment of irradiation creep in graphite components is based on empirical laws derived from data obtained from small samples irradiated in a materials test reactor. However, due to the complicated irradiation rigs required and the amount of dimensional and property measurements needed to be taken, constant stress irradiation creep experiments are difficult and very expensive to carry out successfully. However, restrained creep experiments are simple to implement, less expensive and can be easily included as part of other, more conventional irradiation graphite experimental programmes. However, in the past, the disadvantage of these experiments has been that the results have been difficult to interpret using the then available analytical methods. In this paper the restrained creep experiment is revisited and analysed numerically and the possible benefit of using a restrained creep experiment in future graphite irradiation experiments is investigated. It is shown that a numerical simulation of the restrained creep experiment behaviour would be an essential tool to ensure that the stress within the specimen remains within defined limits so that specimen failure can be avoided.     

Numerical simulation of creep damage for low alloy steel welded joint considering as-welding residual stress

A model to calculate the welding temperature and residual stress was built using finite element code ABAQUS, and a subroutine of creep damage was also developed. Based on the coupling of welding residual stress and creep damage, the welding residual stress and creep damage of a tube made of Cr5Mo steel were simulated. This method can obtain the distributions of complex residual stress, creep damage and stress relaxation, which provide a reference for discussing the effect of residual stress on creep damage. The results show that the welding residual stress is very large at initial stage, then it is relaxed in a short time at high temperature. The distribution of creep and damage is mainly decided by the as-welding residual stress. Welding residual stress has a great effect on the creep and damage, which provides a reference for the design and life prediction of high temperature component.     

Structural design of high temperature metallic components

In the high temperature engineering test reactor (HTTR), even at normal operation the service temperatures of class 1 metallic components reach temperatures above 900 °C when exposed to primary helium coolant of 950 °C. For these components, Hastelloy XR, which is the improved version of Hastelloy X, was developed and used for high temperature application.Some of the high temperature materials and their service temperatures, including Hastelloy XR, used for the class 1 and reactor internal metallic components of the HTTR are very well beyond the well-established Japanese elevated temperature structural design guideline. Moreover, at very high temperatures, where creep deformation is significant, the component design based on elastic analysis is impossible. Therefore, many research works on structural mechanics behavior were carried out to establish a high temperature structural design guideline and creep analysis methods. This paper reviews structural design of the high temperature components for the HTTR made of Hastelloy XR, 2 1/4Cr–1Mo steel, austenitic stainless steels SUS321TB and SUS316, and 1Cr–0.5Mo–V steel.     

Study on creep-fatigue failure prediction methods for type 304 stainless steel

Creep-fatigue failure is one of the principal failure modes to be avoided in elevated-temperature components of liquid metal fast breeder reactor (LMFBR) plants. To prevent this failure during the plant life with sufficient confidence, accurate and reliable methods should be employed for evaluating creep-fatigue endurance. A number of creep-fatigue tests have been conduced to establish a reliable creep-fatigue design methodology applicable to LMFBR plants in the last two decades but the conditions of these tests are generally far from those expected in actual plants. For the purpose of studying the characteristics of various creep-fatigue life prediction methods in conditions closer to actual plant conditions, the authors initiated creep and creep-fatigue tests for type 304 austenitic stainless steel with a special emphasis on tests with longer durations than past tests. Interim results are summarized in this paper. Two representative life prediction methods, linear damage fraction rule and ductility exhaustion method, were then applied to these test conditions. It was found that both methods can predict the failure lives with reasonable accuracy. Some comparisons were made regarding the characteristics of these two methods.     

Finite element analysis for creep failure of coolant pipe in light water reactor due to local heating under severe accident condition

During severe accident of a light water reactor (LWR), the piping of the reactor cooling system would be damaged when the piping is subjected to high internal pressure and very high temperature, resulted from high temperature gas generated in a reactor core and decay heat released from the deposit of fission products. It is considered that, under such a condition, short-term creep at high temperatures would cause the piping failure. For the evaluation of piping integrity under a severe accident, a method to predict such high temperature short-term creep deformation should be developed, using a creep constitutive equation considering tertiary creep. In this paper, the creep constitutive equation including tertiary creep was applied to nuclear-grade cold-drawn pipe of 316 stainless steel (SUS316), based on the isotropic damage mechanics proposed by Kachanov and Ravotnov. Tensile creep test data for the material of a SUS316 cold-drawn pipe were used to determine the coefficients of the creep constitutive equation. Using the constitutive equation taking account of creep damage, finite element analyses were performed for the local creep deformation of the coolant piping under two types of conditions; uniform temperature (isothermal condition) and temperature gradient of circumferential direction (non-isothermal condition). The analytical results show that the damage variable integrated into the creep constitutive equation can predict the pipe failure in the test performed by Japan Atomic Energy Research Institute, in which failure occurred from the outside of the pipe wall.     

Fundamental principles for a nuclear design and structural analysis code for HTR components operating at temperatures above 800°C

With reference to the special characteristics of an HTR plant for the supply of nuclear process heat, the investigation of the fundamental principles to form the basis for a high temperature nuclear structural design code has been described. As examples, preliminary design values are proposed for the creep rupture and fatigue-behaviour. The linear damage accumulation rule is for practical reasons proposed for the determination of service life, and the difficulties in using this rule are discussed. Finally, using the data obtained in structural analysis, the main areas of investigation which will lead to improvements in the utilization of the materials are discussed. Based on the current information, the working group “Design Code” believes that a service life of 70 000 h for the heat-exchanging components operating at above 800°C can be.     

基于耦合CFD-FEM方法的严重事故下RPV蠕变失效风险评估

核电站严重事故发生后,反应堆压力容器（RPV）的剩余固壁在高温差、内压、熔池重量等的作用下可能发生蠕变失效。本文以CPR1000 RPV为研究对象,基于FLUENT软件二次开发求解反应堆压力容器下封头烧蚀温度场,然后基于ANSYS Workbench开展耦合CFD-FEM力学分析,求解严重事故下RPV烧蚀温度场稳定后72 h内的等效应力、等效塑性应变和等效蠕变应变,并评估了RPV的蠕变失效风险。结果表明：当堆坑注水等措施投运后,RPV剩余固壁在72 h内不会发生蠕变失效和塑性变形失效,有效卸压可明显提升RPV结构完整性的安全裕度。     

Risk Assessment of Creep Failure for RPV under Severe Accident Condition Based on Coupled CFD-FEM Method

After a reactor core melt accident, creep failure may occur in the residual solid wall of the reactor pressure vessel (RPV) under the influence of high temperature difference, internal pressure and the weight of the molten pool. In this work, the CPR1000 RPV was used as a research object. The ablation temperature field of the lower head of RPV was solved through the secondary development of the FLUENT software. And then, a CFD-FEM coupling analysis was carried out based on ANSYS Workbench software. The equivalent stress, the equivalent plastic strain and the equivalent creep strain of the RPV within 72 h under severe accident after the wall ablation and temperature field distribution formed stably were calculated. The risk of creep failure of the RPV was evaluated. The results show that when the reactor pit water injection measure puts into operation, the residual solid wall of the RPV will not experience creep failure and plastic deformation failure within 72 h, and besides, the pressure relief can significantly increase the safety margin of the structural integrity of the RPV.     

The application of leak before break concept to W7-X target module

Fusion is the energy production technology, which could potentially solve problems with growing energy demand of population in the future. Wendelstein 7-X (W7-X) is an experimental stellarator of the helias type fusion reactor currently being built in Greifswald, Germany. This experimental stellarator is a complex structure, such as nuclear power plants and high level of safety requirements should be used for structural integrity analysis. It is thus not possible to obtain simple solutions for general cases, therefore sophisticated methods are necessary for the analysis. Inside the Plasma Vessel (PV) of W7-X there is a number of different components such as pipes, divertors, baffles and targets. A guillotine failure of one component is very dangerous for structural integrity of surrounding components located in PV. For this reason it is very important to evaluate possibility to apply “leak before break” (LBB) concept for W7-X. The LBB concept is widely used in the nuclear industry to describe the idea that in the piping carrying the coolant of a power reactor a leak will occur before a catastrophic break will occurred. LBB allows to conduct the structural design without considering the loads due to postulated line breaks.The LBB analysis was made for the case when plasma vessel is operating in “baking” mode. “Baking” is the mode, when the cooling system is working as a warming system and it heats the plasma vessel structures up to 160 °C in order to release the absorbed gases from the surfaces and to pump them out of the plasma vessel before plasma operation.The LBB analysis was performed for most loaded component of target module. According to the results of the analysis it is possible to conclude that target module 1H fulfils the LBB requirements.     

Investigation and analysis on ITER in-vessel coils’ raw-materials

The ITER in-vessel coils (IVCs) consist of 27 coils edge localized modes (ELM) and 2 coils vertical stabilization (VS) which are all mounted on the vacuum vessel wall behind the shield modules. The IVCs design and manufacturing work is being conducted in between Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and Princeton Plasma Physics Laboratory (PPPL). Because the position of ELM and VS coils is close and face to the plasma, the IVCs must undergo a severe environment, such as the high dose of radiation and high operation temperature, thus the conventional electrical insulation materials cannot be used. And the technology of “Stainless Steel Jacketed Mineral Insulated Conductor” (SSMIC) is deemed as the best choice to provide the necessary radiation resistance and compatibility strength in ITER's vacuum vessel. While mineral insulated conductor technology is not new, and is similar to the mineral insulated cable used in industrial. Some difficulties still need to be solved, such as searching for the proper raw-materials to make sure that the conductor have the properties of high current carrying capability, the necessary radiation resistance, the proper strength, at the same time, it must be come true in manufacture technology. This paper described the analysis of the materials for VS and ELM coil conductor.