Creep-fatigue evaluation of normalized and tempered modified 9Cr---1Mo

Creep-fatigue is a fatal failure mode of the high temperature structural materials of liquid metal fast breeder reactors (LMFBRs). In this report, two important issues are discussed for creep-fatigue evaluation of normalized and tempered modified 9Cr---1Mo (modified 9Cr---1Mo(NT)) steel which is a promising structural material for the steam generator of large-scale LMFBRs in Japan. Several evaluation methods based on the ductility exhaustion concept are discussed for the prediction of tension strain hold creep-fatigue damage of this material. A time-fraction type of linear damage summation concept based on a new ductility exhaustion theory is proposed from the point of view of its appropriate conservatism for time extrapolation and its simplicity.Also, a life reduction mechanism of low cycle fatigue with strain hold at the compression side is discussed, based on the data observed by a scanning type electron microscope. Creep damage or the tension mean stress caused by compression strain hold hardly reduce the low cycle fatigue life of this material. A new concept based on the location of oxidation on the test specimen surface can explain the reduction in low cycle fatigue life of modified 9Cr-1Mo(NT) steel.     

Operating temperatures for an FBR

Structural mechanics aspects related to operating temperatures of a typical pool type 500 MWe fast breeder reactor are discussed. The critical high temperature components are analysed in detail based on elastic, inelastic and viscoplastic deformation theories, and life is predicted in accordance with the rules of design code RCC-MR 87. Analysis indicates that the control plug is the most critical component in the reactor which limits the reactor outlet temperature to 820 K with a temperature rise of 160 K across the core.     

Compatibility of FBR materials with sodium

In order to incorporate a procedure for the evaluation of the sodium environmental effects on core and structural materials into the elevated temperature structural design guide lines for fast breeder reactors, R&D on the sodium compatibility of the materials has been in progress in Japan Atomic Energy Agency. This paper reviews corrosion behavior in the sodium of conventional austenitic and ferritic steel. Simultaneously, the corrosion and mechanical properties of the materials for advanced FBRs, 12Cr steel and ODS steels are summarized, including the results of recent research.     

径向步进倒料行波堆的数值研究

行波堆是一种可实现自持增殖-燃耗的新概念快堆,它可直接使用天然铀、贫铀、钍等可转换核材料,实现非常高的燃料利用率。基于行波堆的原理,提出了具有现实应用价值的径向步进倒料行波堆的概念,并将其与典型钠冷快堆的设计相结合,采用数值方法对由外而内的径向步进行波堆二维渐近稳态特性进行了研究。计算结果表明:渐近keff随倒料循环周期近似抛物线分布,而渐近燃耗随倒料循环周期线性增长,满足临界条件的倒料循环周期中最大燃耗可达38%;堆芯功率峰随着倒料循环周期的增长,从燃料卸出区(堆芯中心)向燃料导入区(堆芯外围)移动,功率峰值逐渐降低,在高燃耗情况下,靠近堆芯中心的轴向功率分布呈M形。     

Concept and basic performance of an in-pile experimental reactor for fast breeder reactors using fast driver core

The possibility of an in-pile experimental reactor for fast breeder reactors using a fast driver core is investigated. The driver core is composed of a particle bed with diluted fuel. The results of various basic analyses show that this reactor could perform as follows: (1) power peaking at the outer boundary of test core does not take place for large test core; (2) the radial power distribution in test fuel pin is expected to be the same as a real reactor; (3) the experiments with short half width pulse is possible; (4) for the ordinary MOX core, enough heating-up is possible for core damage experiments; (5) the positive reactivity effects after power burst can be seen directly. These are difficult for conventional thermal in-pile experimental reactors in large power excursion experiments. They are very attractive advantages in the in-pile experiments for fast breeder reactors.     

Prospects for the Development of Fast Breeder Reactors

Possible ways to improve nuclear power systems with fast breeder reactors and conditions for ensuring that such systems are competitive are discussed. Certain questions concerning schematic and structural improvements are examined. The results of a comparative analysis of sodium- and lead-cooled breeder reactors are presented. It is pointed out that for sodium-cooled reactors the corresponding informtion is due to many years of experience in developing, investigating, and operating experimental, test, and commercial reactors. There is no experience in developing lead-cooled reactors. A comparative analysis does not confirm that there are any advantages with respect to technical or economic performance for lead-cooled breeder reactors.     

An evaluation method for elastic-plastic buckling of cylindrical shells under shear forces

Since vessels of fast breeder reactors are relatively thin-walled, the prevention of buckling against seismic loading is one of the key issues in their structural design. Buckling of cylindrical vessels under shear forces occurs with a shear and/or bending mode and in the elastic-plastic region. In this paper, we propose a buckling strength equation for cylindrical vessels under static shear loads, which is developed on the basis of theoretical considerations for plasticity and shape imperfections. The effects caused by nonlinear distribution of stresses along the circumference in the elastic-plastic region are also considered as correction factors. The equation is validated by the existing and the present buckling test data, as well as FEM analyses. The safety factors to be employed in the design are proposed by evaluating the reliability of the proposed equation in the light of available test data.     

Prediction method for thermal stratification in a reactor vessel

The prediction method for thermal stratification phenomena in a fast breeder reactor is described. The focus of attention is placed on the applicability of water test results to predict thermal stratification phenomena in a real plant. The basic feature of thermal stratification was examined in a cylindrical plenum, using water and sodium as test fluids. The similitude relationship between a small-scale test and a real plant is discussed in order to understand the experimental results. The scale-model experiments for LMFBRs (liquid metal-cooled fast breeder reactors) were also performed to see the effects of a reactor configuration and reactor-trip operation condition. Then the magnitudes of the temperature gradient and the ascending speed of stratified interface in the hot plenum of LMFBRs were predicted, based on the results of the water scale-model.     

Sodium effects on mechanical performance and consideration in high temperature structural design for advanced reactors

Sodium environmental effects are key limiting factors in the high temperature structural design of advanced sodium-cooled reactors. A guideline is needed to incorporate environmental effects in the ASME design rules to improve the performance reliability over long operating times. This paper summarizes the influence of sodium exposure on mechanical performance of selected austenitic stainless and ferritic/martensitic steels. Focus is on Type 316SS and mod.9Cr-1Mo. The sodium effects were evaluated by comparing the mechanical properties data in air and sodium. Carburization and decarburization were found to be the key factors that determine the tensile and creep properties of the steels. A beneficial effect of sodium exposure on fatigue life was observed under fully reversed cyclic loading in both austenitic stainless steels and ferritic/martensitic steels. However, when hold time was applied during cyclic loading, the fatigue life was significantly reduced. Based on the mechanical performance of the steels in sodium, consideration of sodium effects in high temperature structural design of advanced fast reactors is discussed.     

Study on environmental effect on fatigue and creep-fatigue strength of 316FR stainless steel in sodium at elevated temperature

Low-carbon 316 stainless steel with medium-nitrogen (316FR) is considered as the principal structural material for next generation fast breeder reactor (FBR) plants in Japan. The material strength standard and the creep-fatigue life evaluation method for 316FR have been developed. However, they are based on the results of material tests in air, while actual structural material will be used mainly in liquid sodium environment in the plants. In order to clarify the environmental effect, cyclic bending tests were carried out with and without hold time in sodium. Tested materials were 316FR and conventional 304 and 316 stainless steels. Weld metal of 316FR was also tested. As a result, it was found that fatigue and creep-fatigue lives of 316FR in sodium were larger than those in air and no explicit consideration of the environmental effect is necessary in design. It was also found that the life evaluation method based on the ductility exhaustion concept is applicable to creep-fatigue life assessment in sodium.     

Development of inelastic constitutive model for austenitic stainless steel for design use

To prevent creep-fatigue failure or excessive deformation in high-temperature components of fast reactor plants, accurate estimation of inelastic deformation is essential. In performing inelastic analysis, employment of constitutive models, which can precisely reproduce inelastic deformation of the material is of critical importance. The authors have been engaged in the development of inelastic constitutive model for the use in structural design assessment of liquid metal-cooled fast reactor plants. Various improvements were made on the nonlinear hardening model proposed by Ohno and Wang, placing an emphasis on capability to simulate inelastic deformation behavior of austenitic stainless steels, under regular or irregular cyclic loading possibly with temperature variation and hold time. It was demonstrated that the model can simulate the inelastic deformation behavior under various loading conditions with a sufficient accuracy.     

Effect of sodium environment on the low cycle fatigue properties of modified 9Cr-1Mo ferritic martensitic steel

Modified 9Cr-1Mo ferritic steel is the material of current interest for the steam generator components of liquid metal cooled fast breeder reactors (LMFBRs). The steam generator has been designed to operate for 30-40 years. It is important to accurately determine the life of the components in the actual environment in order to consider the extension of life beyond the design life. With this objective in view, a programme has been initiated at our laboratory to evaluate the effects of flowing sodium on the LCF behaviour of modified 9Cr-1Mo steel. LCF tests conducted in flowing sodium environment at 823 K and 873 K exhibited cyclic softening behaviour both in air and sodium environments. The fatigue lives are significantly improved in sodium environment when compared to the data obtained in air environment under identical testing conditions. The lack of oxidation in sodium environment is considered to be responsible for the delayed crack initiation and consequent increase in fatigue life. Comparison of experimental lifetimes with RCC-MR design code predictions indicated that the design curve based on air tests is too conservative.     

Low-cycle fatigue of welded joints: coupled initiation propagation model

This paper deals with the low-cycle fatigue (LCF) design of welded structures, the aim being the critical analysis of the rule used in the RCC-MR [Design and construction rules for mechanical components of FBR nuclear islands, AFCEN, 1993], for the design and construction of fast breeder reactors. The study takes into account the evolution of the material behavior laws and damage accumulation during the fatigue loading. The adopted model consists of analyzing separately the behavior and the damage evolutions. It allows us to determine the damage ratio corresponding to initiation and propagation of a significant crack in order to determine the life duration. This model suggests the existence of a threshold level of loading, above which micro-cracks initiate. The initiation fatigue life can then be neglected below the threshold level. This work shows also that the RCC-MR rules are valid below this threshold load level.     

Development of simplified evaluation method for creep-fatigue crack propagation

In the design assessment of fast reactor plant components, prevention of crack initiation from defect-free structures is a main concern. However, existence of initial defects such as weld defects cannot be entirely excluded and this potential cracks are to be evaluated to determine if initiated cracks do not lead to component failure instantly. Therefore, evaluation of structural integrity in the presence of crack-like defects is also important to complement the formal design assessment. The authors have been developing a guideline for assessing long-term structural integrity of fast reactor components using detailed inelastic analysis and nonlinear fracture mechanics. This guideline consists of two parts, evaluation of defect-free structures and flaw evaluation. In the latter, creep-fatigue is considered to be one of the most essential driving force for crack propagation at high operating temperature exceeding 500 °C. The uses of J-integral-type parameters (fatigue J-integral range and creep J-integral) are recommended to describe creep-fatigue crack propagation behavior in the guideline. This paper gives an outline of the simplified evaluation method for creep-fatigue crack propagation.     

Experimental Study of Nuclear Characteristics of Large Axially Heterogeneous Core Using Fast Critical Assembly

The physics characteristics of large axially heterogeneous liquid-metal fast breeder reactors (LMFBRs), particularly the parameters for use in design and safety assessment, were examined using the JAERI fast critical assembly facility, arranged in Assembly XH-1, a partial mock-up of axially heterogeneous LMFBR. The properties measured were (1) criticality, (2) reaction rates and reaction rate ratios, (3) material sample worths, (4) sodium-void worths and (5) B4C control rod worths.

The results were compared with those of prior experiments with assemblies representing conventional homogeneous core. Confirmation was obtained of the typical nuclear characteristics attributed to axially heterogeneous LMFBRs, including flattening of the axial distribution of power and of the differential worth of control rod, as also lower sodium void worth.

Theoretical analyses paralleling the experiments, using JENDL-2 cross section library and JAERI standard calculation code system for fast reactor neutronics, resulted in some discrepancies, particularly for the internal blanket, in respect of plutonium sample worth, fission rate and fission rate ratio.     

Thermal hydraulic investigations of primary coolant pipe rupture in an LMFBR

High quality for primary coolant pipes in fast reactors is ensured through utmost care taken in the design and manufacture. Demonstration of high structural reliability of them by extensive experimental and theoretical studies renders the double-ended guillotine rupture (DEGR) of a primary pipe a highly improbable event. However, as a defense in depth approach instantaneous DEGR of one of the pipes has been considered in design. Thermal hydraulic analyses of this event in a typical liquid metal cooled fast breeder have been carried out to study its consequences and to establish the availability of safety margins. Various uncertainties relevant to the event have been analysed to evaluate the sensitivity of each parameter. For this purpose, one-dimensional plant dynamics studies using thermal and hydraulic models of core subassemblies and primary sodium circuit have been performed. Validity of the assumptions made in the one-dimensional model like, uniform flow through all subassemblies in core under pipe ruptured condition and non possibility of sodium boiling by flashing have also been investigated through detailed three-dimensional and pressure transient studies. Analyses indicate the availability of good margins against the design safety limits in all the parametric cases analysed.     

The interaction of rate and history-dependent effects and its significance for slow cyclic inelastic analysis at elevated temperature

Life prediction for creep and low-cycle fatigue interaction and the analysis of ratcheting phenomena are of great importance in the design of future high-temperature nuclear reactors. These problems involve slow cyclic load application, inelastic deformation, and are inherently nonlinear. Fortunately, finite element computer programs with their time following and nonlinear capabilities are now available for the efficient solution of these complicated problems. The task is now to ensure that a proper representation of the material deformation behavior, the constitutive equation, is used in these computer programs. A systematic study of structural metal deformation behavior in slow cyclic laboratory tests such as cyclic creep, cyclic relaxation, low-cycle fatigue with and without hold-time showed that rate and history dependence interact. A few examples of such interactions are given in this paper. It is further shown that constitutive equations based on the additive use of elastic, plastic, and creep strains are not capable of reproducing these interaction phenomena on principal grounds. New inelastic constitutive equations for metals must be developed and some presently pursued approaches are discussed briefly.     

Development of elevated temperature structural design guide for Japanese demonstration breeder reactor

Programs to develop the “elevated temperature structural design guide for the demonstration fast breeder reactor” (DDS) in Japan have been conducted since 1987. The DDS is to be developed on the basis of the “elevated temperature structural design guide for class 1 components of prototype fast breeder reactors” (ETSDG), by considering structural and material features of the demonstration fast breeder reactor (DFBR) and incorporating results of the latest R&D. This paper describes the progress of the R& D concept of the DDS, and discusses some typical results of current studies on the DDS.     

水冷动力堆用锆合金的疲劳

锆合金是水冷动力堆核燃料元件的包壳材料和堆芯的其它结构材料，在反应堆运行时，堆功率的波动和水冷却介质的流动使燃料组件及其它构件发生循环变形，在极端情况下出现破损。本文概述了堆内锆合金包壳循环变形的特点，并综述了锆合金的循环变形行为，循环变形下的组织结构演化，疲劳裂纹的扩展以及影响疲劳寿命的因素，在此基础上，针对高性能燃料元件的发展趋势，指出了有待进一步研究解决的问题。