Neutron irradiation-induced creep of helium pressurized 304L stainless steel capsules

Irradiation-induced creep and swelling have been measured on 1.5 m long pressurized capsules of solution annealed type 304L stainless steel at 385 °C to neutron doses of 45 dpa. The core-midplane results (fixed position) which have a constant average neutron energy and dose rate but varying time are compared to data taken along the length of the capsule which have constant time but varying average neutron energy and dose rates. Additionally, the effect of stress on swelling, the stress dependency of in-reactor creep and the correlation of irradiation-induced creep and swelling are analyzed utilizing the data generated in this experiment. The results of these analyses are then used as a basis for appraising current theories on irradiation creep.     

Stress state dependence of in-reactor creep and swelling. Part 2: Experimental results

Irradiation creep constitutive equations, which were developed in Part I, are used here to analyze in-reactor creep and swelling data obtained ca. 1977-1979 as part of the US breeder reactor program. The equations were developed according to the principles of incremental continuum plasticity for the purpose of analyzing data obtained from a novel irradiation experiment that was conducted, in part, using Type 304 stainless steel that had been previously irradiated to significant levels of void swelling. Analyses of these data support an earlier observation that all stress states, whether tensile, compressive, shear or mixed, can affect both void swelling and interactions between irradiation creep and swelling. The data were obtained using a set of five unique multiaxial creep-test specimens that were designed and used for the first time in this study. The data analyses demonstrate that the constitutive equations derived in Part I provide an excellent phenomenological representation of the interactive creep and swelling phenomena. These equations provide nuclear power reactor designers and analysts with a first-of-its-kind structural analysis tool for evaluating irradiation damage-dependent distortion of complex structural components having gradients in neutron damage rate, temperature and stress state.     

Simulation of irradiation hardening of Zircaloy within plate-type dispersion nuclear fuel elements

Within plate-type dispersion nuclear fuel elements, the metal matrix and cladding attacked continuously by fast neutrons undergo irradiation hardening, which might have remarkable effects upon the mechanical behaviors within fuel elements. In this paper, with the irradiation hardening effect of metal materials mainly considered together with irradiation growth effect of the cladding, the three-dimensional large-deformation constitutive relations for the metal matrix and cladding are developed. The method of virtual temperature increase in the previous studies is further developed to model the irradiation swelling of fuel particles; the method of anisotropic thermal expansion is introduced to model irradiation growth of the cladding; and a method of multi-step-temperature loading is proposed to simulate the coupling features of irradiation-induced swelling of the fuel particles together with irradiation growth of the cladding. Above all, based on the developed relationship between irradiation growth at certain burnup and the loaded virtual temperatures, with considering that certain burnup corresponds to certain fast neutron fluence, the time-dependent constitutive relation due to irradiation hardening effect is replaced by the virtual-temperature-dependent one which is introduced into the commercial software to simulate the irradiation hardening effects of the matrix and cladding. Numerical simulations of the irradiation-induced mechanical behaviors are implemented with the finite element method in consideration of the micro-structure of the fuel meat. The obtained results indicate that when the irradiation hardening effects are introduced into the constitutive relations of the metal matrix and cladding: (1) higher maximum Mises stresses for certain burnup at the matrix exist with the equivalent plastic strains remaining almost the same at lower burnups; (2) the maximum Mises stresses for certain burnup at the cladding are enhanced while the maximum equivalent plastic strains are reduced; and (3) the maximum first principal stresses for certain burnup at the matrix or the cladding are lower than the ones without the hardening effect, and the differences are found to increase with burnup; and the variation rules of the interfacial stresses are similar.     

UMo/Al弥散燃料板内裂变气体肿胀的静压效应研究

将核燃料的裂变气体肿胀与静水压力计算相耦合,并考虑重要的辐照蠕变,编制了定义其复杂力学本构关系的子程序。将定义各部分材料热-力学本构关系的用户子程序引入ABAQUS软件,获得了燃料板细观尺度下辐照-热-力耦合行为的计算模拟方法,并计算分析了核燃料裂变气体肿胀的静压效应。与不考虑裂变气体肿胀静压相关性的计算结果对比发现,在裂变气体肿胀计算中引入静压的影响,将使得核燃料颗粒内的辐照肿胀应变显著减小,引起板内最高温度降低,并减弱燃料颗粒和基体间的力学相互作用,减小燃料颗粒内的等效蠕变应变,致使基体内最大Mises应力和第一主应力减小。     

Simulation of irradiation-induced creep in nickel

An experimental technique has been developed to simulate neutron irradiation-induced creep by charged particle bombardment. The experimental apparatus permits on-line computer monitoring of experimental parameters while temperature, stress, and flux are maintained at the desired levels. A typical result obtained with a 0.38 mm (0.015 in.) thick, high-purity nickel specimen bombarded with 22 MeV deuterons at 224°C (435°F) and at a stress of 345 MPa (50.12 ksi) is presented. The result demonstrates that charged particle irradiation can successfully be used to simulate irradiation-induced creep reproducibly in materials whose thickness is typical of nuclear fuel cladding.     

On the analogy between thermally and irradiation induced creep

Employing an analogy between thermally induced and irradiation induced creep, physical arguments are used first to deduce a one-dimensional constitutive relation for metals under stress in a high temperature and high neutron flux field. This constitutive relation contains modified superposition integrals in which the temperature and flux dependence of the material parameters is included via the use of two reduced time scales; linear elastic, thermal expansion and swelling terms are also included. A systematic development based on thermodynamics, with the stress, temperature increment and defect density increment as independent variables in the Gibbs free energy, is then employed to obtain general three-dimensional memory integrals for strain; the entropy and coupled energy equation are also obtained. Modified superposition integrals similar to those previously obtained by physical argument are then obtained by substituting special functions into the results of the thermodynamic analysis, and the special case of an isotropic stress power law is examined in detail.     

Void swelling and irradiation creep relationships

Recent analytical and theoretical work on swelling enhanced irradiation creep and stress effects on swelling is reviewed. A proposed explanation for swelling enhanced irradiation creep involves consideration of the role of vacancy loops. Theoretical work leads to the development of a new relationship for swelling enhanced creep which predicts larger irradiation creep rates at high levels of swelling (>5%) than the original formulation. Consideration is given to an additional effect of stress on swelling which involves a stress effect on the incubation dose. A constitutive equation is presented to describe this phenomenon. Design related illustrations are presented for these high fluence irradiation induced phenomena.     

Effects of hold time and neutron irradiation on the low-cycle fatigue behaviour of type 316-CL and their consideration in a damage model

The creep fatigue behaviour of AISI type 316 L(N) plate material has been investigated in the temperature range of 450–750 °C by performing axial strain controlled tests with GRIM specimens. The creep and creep fatigue behaviour of austenitic stainless steel material is known to be prone to neutron irradiation-induced embrittlement. Therefore, the irradiation behaviour was studied by performing irradiation experiments in the High Flux Reactor (HFR) of Petten at 550 °C. A newly developed damage model for time-dependent damage was applied to describe the failure behaviour of AISI 316 L(N) in the cyclic tests performed.     

Irradiation creep and interrelation with swelling in austenitic stainless steels

The available experimental data on irradiation-induced creep in austenitic stainless steels are summarized and the existing theories reviewed. Attention is paid to the influence of material composition and pretreatments on irradiation creep. In particular the stress, flux, fluence and temperature dependencies are reported and possible correlations of irradiation creep with the microstructural evolution, the swelling behaviour and the precipitation kinetics of the materials are outlined. The consequences of stress effects connected with swelling for the irradiation-creep behaviour, especially the stress-dependence, are discussed.     

Effect of self-ion injection in simulation studies of void swelling

The rate theory of void swelling is generalized to incorporate the effect of excess interstitial production resulting from self-ion injection in simulation studies. The swelling rate is shown to be reduced at all temperatures, the fractional reduction being largest in the region where intrinsic recombination of irradiation-induced interstitials and vacancies dominates, provided there is no dislocation recovery at high doses. When such recovery processes are operative, however, the effect of self-ion injection is most important near the peak swelling temperature and can lead to a saturation in swelling with increasing dose which, in the model considered, would be absent under neutron irradiation. For this reason it is emphasized that the results of simulation experiments, without complementary microstructural data, should be viewed with caution as far as their quantitative relation to neutron damage is concerned.     

The relationship between carbide precipitation and the in-reactor deformation of type 316 stainless steel

The precipitation of carbides from type 316 stainless steel is associated with enhanced irradiation-induced creep and swelling. Carbide precipitation occurs in irradiated type 316 materials at lower temperatures than in their unirradiated counterparts. Once the precipitate is formed, there is little tendency for it to dissolve back into the matrix upon further irradiation; thus, an irreversible change in the material's response to neutron irradiation occurs. The degree of precipitation and its effects on in-reactor strain have been investigated, and the results were used to explain double peaks in diameter profiles of irradiated fuel elements.     

Development of Thermal/Irradiation Stress Analytical Code “VIENUS” for HTTR Graphite Block

A new thermal/irradiation stress analysis code “VIENUS” has been developed for the graphite block in the High-Temperature Engineering Test Reactor (HTTR). The VIENUS is a two- dimensional finite element visco-elastic analysis code to take account of graphite behavior under irradiation in detail. In the analysis, the effects of both fast neutron fluence and temperature on material properties are considered.

The code has been evaluated by the irradiation test results of the Peach Bottom fuel elements to confirm the thermal/irradiation stresses in the graphite block. It is clarified that the calculated results are able to estimate a tendency of the test results, and that both the irradiation- induced creep and dimensional change are the most important parameters in the thermal/irradiation stress analysis. From the present study, it is suggested that the VIENUS code is a useful tool to evaluate the thermal/irradiation stresses in the HTTR graphite blocks.     

In-reactor stress relaxation of selected metals and alloys at low temperatures

Sttess relaxation of bent beam specimens under fast neutron irradiation at 340 and 570 K has been studied for a range of materials, as follows: several stainless steels, a maraged steel, AISI4140, Ni, Inconel X-750, Ti, Zircaloy-2, Zr-2.5% Nb and Zr₃ Al. All specimens were in the annealed or solution-treated condition. Where comparisons were possible, the creep coefficients derived from the stress relaxation tests were found to be consistent with other studies of irradiation-induced creep. The steels showed the lowest rates of stress relaxation; the largest rates were observed with Zr-Nb, Ti and Ni. For most materials, the creep coefficient at 340 K was equal to or greater than that at 570 K. Such weak temperature dependence is not easily reconciled with existing models of irradiation creep based on dislocation climb, such as SIPA or climb-induced glide. Rate theory calculations indicate that because the vacancy mobility becomes very low at the lower temperature, recombination should dominate point defect annealing, resulting in a very low creep rate compared to that at the higher temperature. It is shown that the weak temperature dependence observed experimentally cannot be accounted for by the inclusion of more mobile divacancies in the calculation.     

Irradiation creep under 60 MEV alpha irradiation

Accelerator-produced charged-particle beams have advantages over neutron irradiation for studying radiation effects in materials, the primary advantage being the ability to control precisely the experimental conditions and improve the accuracy in measuring effects of the irradiation. An apparatus has recently been built at ORNL to exploit this advantage in studying irradiation creep. These experiments employ a beam of 60 MeV alpha particles from the Oak Ridge Isochronous Cyclotron (ORIC). The experimental approach and capabilities of the apparatus are described. The damage cross section, including events associated with inelastic scattering and nuclear reactions, is estimated. The amount of helium that is introduced during the experiments through inelastic processes and through backscattering is reported. Based on the damage rate, the damage processes and the helium-to-dpa ratio, the degree to which fast reactor and fusion reactor conditions may be simulated is discussed. Recent experimental results on the irradiation creep of type 316 stainless steel are presented, and are compared to light ion results obtained elsewhere. These results include the stress and temperature dependence of the formation rate under irradiation. The results are discussed in relation to various irradiation creep mechanisms and to damage microstructure as it evolves during these experiments.     

An upper-bound evaluation of interactive creep and growth in textured Zircaloy

Uniform strain-rate, upper-bounding constitutive modeling techniques are applied to the study of irradiation creep in textured hexagonal metals with a specific application to Zircaloy. A crystallite orientation distribution function is used to describe the distribution of grains in a polycrystal for use in strain and stress averaging. Stress dependent creep compliances are constructed for a system exhibiting slip on a discrete set of crystallographic slip systems. Superposition of creep and stress independent sources of internal strain, such as those produced by the irradiation growth mechanism, is studied. A model for irradiation growth is proposed which accounts for grain shape and size effects as well as non-uniform dislocation densities arising from variable fabrication history. The internal strains are shown to interact with those dependent on applied stress in a manner which alters the constitutive behavior.     

The temperature dependence of irradiation creep

The stress induced absorption mechanism (SIPA) of irradiation creep will be discussed by developing a new application of rate theory which emphasizes particularly the significance of the vacancy loops formed from the displacement cascades during fast neutron irradiation. A simple analytic result for the expected creep strain rate and various related numerical results will be discussed. The relation between irradiation creep and void swelling will be emphasized, particularly in relation to the significance of the vacancy emission processes from the vacancy loops.     

U-10Mo/Zr单片式燃料元件堆内稳态热-力学性能研究

本文建立了U-10Mo/Zr单片式燃料元件的辐照性能模型以及热-力学本构关系,采用有限元方法进行非均匀辐照场中燃料元件稳态热-力学性能的数值模拟,获得并分析了U-10Mo/Zr单片式燃料元件温度、形变和应力的分布特点及变化规律。研究结果表明,燃料芯体厚度增量在芯体和包壳结合面附近达到最大,主要受到燃料辐照蠕变的影响;在较低燃耗条件下,燃料芯体高温辐照肿胀模拟结果与低温辐照肿胀试验结果相当;燃料芯体边角区域和包壳端面外侧区域存在应力集中。      

Model of the Formation and Growth of a Gas-Filled Cavity in Boron Carbide under Irradiation

A model of the formation and growth of gas-filled pores in the cores of absorbing elements under neutron irradiation is described. The model is based on a definite similarity between the mechanisms of radiation damage in boron carbide and oxide fuel. For boron carbide, the key mechanism is radiation-stimulated coalescence of helium bubbles, which gives rise to the much larger swelling of boron carbide as compared with nuclear fuel. Regimes of the coalescence of helium bubbles as a function of the irradiation intensity, neutron spectrum, and radial position are investigated. It is shown that the model describes qualitatively well the basic development of gas-filled porosity under irradiation in fast and thermal reactors.