High-temperature Radiation Embrittlement Research Articles

Fracture mechanisms and properties for irradiated austenitic chromium-nickel steel over elevated temperature range and formulation of intergranular fracture criterion

The fracture mechanisms of austenitic 18Cr-9Ni steel have been investigated over the temperature range from 200° to 600 °C using the following states: (1) after neutron irradiation at temperature of 400 °C up to damage dose of 30 dpa; (2) after neutron irradiation and subsequent aging at 550 °C for 3000 h. The fracture properties and mechanisms have been determined under uniaxial tension of smooth round and notched round bars, i.e. for various stress triaxialities. Sharp decrease in the fracture strain and transition to intergranular fracture have been revealed for smooth round bars over high temperature range that indicates high temperature radiation embrittlement. Over the same temperature range the fracture strain is larger for notched round bars than for smooth round bars, moreover a portion of intergranular fracture is also larger. This finding is rather abnormal as the fracture strain usually decreases when stress triaxiality and intergranular fracture portion increase. On the basis of the obtained experimental results, for the first time, the fracture stress-controlled criterion and fracture model have been developed over temperature range of high temperature radiation embrittlement. The proposed model allows one to explain larger value of the fracture strain for notched round bars as compared with smooth round bars and also to predict fracture toughness over range of high temperature radiation embrittlement. The formulated criterion and model have been verified by comparison of the calculated and experimental values of fracture toughness for 18Cr-9Ni steel irradiated up to damage dose of (24–30) dpa. Experimental values of fracture toughness have been obtained from compact tension CT-0.5 specimens tested at 200 °C and 600 °C.

Fracture properties and mechanisms for irradiated austenitic steels over high temperature range and formulation of fracture criterion. Part 2. Fracture criterion and model

On the basis of experimental data considered in the first part of the paper the fracture criterion and fracture model are represented for high temperature radiation embrittlement of irradiated austenitic steels. The proposed model allows one to explain larger value of the fracture strain for notched cylindrical specimens as compared with smooth specimens as it has been found in the first part of the paper and also to predict fracture toughness over range of high temperature radiation embrittlement. The formulated criterion and model are verified by comparison of the calculated and experimental values of fracture toughness for 18Cr-9Ni steel irradiated up to damage dose of (24–30) dpa. Experimental values of fracture toughness are obtained from compact tension CT-0.5 specimens tested at 200°С and 600°С. The fracture toughness data for irradiated steel are represented also over temperature range from 20°С and 600°С.

Fracture properties and mechanisms for irradiated austenitic steels over high temperature range and formulation of fracture criterion. Part 1. Experimental results

The experimental study results of the fracture properties and mechanisms are represented over temperature range from 200 up to 600°С for austenitic 304 steel (18Cr–9Ni steel) in the following conditions: (1) after neutron irradiation at temperature of 400°С up to damage dose of 30 dpa; (2) after neutron irradiation and subsequent aging at 550°С for 3000 hours. The fracture properties and mechanisms are determined under uniaxial tension of standard smooth cylindrical specimens and notched cylindrical specimens, i.e. for various stress triaxialities. Sharp decrease in the fracture strain and transition to intergranular fracture are revealed over high temperature range forstandard smooth cylindrical specimens. Over the same temperature range the fracture strain is larger for notched cylindrical specimens than for smooth cylindrical specimens, moreover a portion of intergranular fracture is also larger. This finding is rather abnormal as the fracture strain usually decreases when stress triaxiality and intergranular fracture portion increase. The obtained results are used for formulation of fracture criterion and elaboration of fracture model for high temperature radiation embrittlement that is considered in the second part of the paper.

The effect of neutron irradiation in a helium atmosphere on the mechanical properties of 18Cr10NiTi austenitic steel

The microstructure and mechanical properties of 18Cr10NiTi austenitic steel were studied after high-temperature neutron irradiation of samples in a helium atmosphere. It was revealed during irradiation, helium penetrates into steel samples forming bubbles distributed on dislocations and along grain boundaries. Character of the temperature dependence of the mechanical characteristics of irradiated steel indicates a high-temperature radiation embrittlement of steel. The character of the mechanical properties temperature dependence change of irradiated steel points to its high-temperature radiation embrittlement.

High-Temperature Radiation Embrittlement of Neutron-Irradiated Austenitic Stainless Steels 08Kh18N10T, Ei-847, EP-172, and ChS-68

The results of research on the influence of the chemical composition, cold-deformation, and irradiation temperature on the short-time mechanical properties of the austenitic stainless steels 08Kh18N10T, EI-847, EP-172, and ChS-68 after neutron irradiation in the BN-350 fast reactor to damaging dose 57–59 dpa in the range 400–490°C are reported. The investigation showed that high-temperature radiation embrittlement is most pronounced in 08Kh18N10T followed by EI-847 steel. The steels EP-172 and ChS-68 are less susceptible to high-temperature radiation embrittlement. After irradiation, the ultimate strength of the studied steels in the austenitized and cold-deformed state is almost identical. A difference of the plasticity of irradiated samples is observed only for ChS-68, when in the cold-deformed state the plasticity after testing at temperature above 600°C is higher than in the austenitized state. An increase of the irradiation temperature noticeably impacted only the high-temperature mechanical properties of the steels 08Kh18N10T and EI-847. The result of raising the irradiation temperature was that the onset of high-temperature radiation embrittlement shifted toward lower temperatures.

Study of the Long-Term Strength of Neutron-Irradiated Austenitic and Ferrite-Martensite Steel

The results of a study of the long-term strength of austenitic and ferrite-martensite steel after irradiation in the BR-10, BOR-60, BN-350, and BN-600 reactors are presented. It is shown that irradiation of austenitic-class steel (EI-847, 0Kh18N10T, and ChS-68) greatly reduces the time to failure and the plasticity. This behavior of austenitic steel is associated with the phenomenon of high-temperature radiation embrittlement. Neutron irradiation does not degrade the long-term mechanical characteristics of ferrite-martensite steel, which remain either at the level of the unirradiated material (EP-823, 05Kh12N2M steels) or surpass it (EP-450 steel).

Effect of boundary conditions on the kinetics of helium release from structural materials

Gaseous products of nuclear reactions (specifically, helium) play a significant part in altering the material properties upon irradiation. It is known that atoms of inert gases promote the generation and growth of pores in irradiated materials and affect phenomena such as swelling, high-temperature irradiation embrittlement, etc. Therefore, a study of the behavior of helium (its production, accumulation, retention, and release) within structural materials is fairly topical. In order to validate the methods of express imitation of accumulation and retention of helium within structural materials under reactor irradiation, we perform a comparative analysis of the spectra of the rate of gas release from samples of austenitic steel 0Kh16N15M3B that were saturated with helium in different ways, i.e., through irradiation in a cyclotron, a magnetic massseparation setup, the IRT-2000 reactor, the BOR-60 reactor, and using the so-called tritium trick technique. The effect of the presence of dislocations and grain boundaries on the release of helium from materials is evaluated. The results of the research conducted show that the kinetics of helium release from samples saturated with helium through the bombardment with alpha particles of different energies, which ensures the simultaneous introduction of helium and radiation defects (in wide ranges of helium concentration and radiation damage) into the material lattice, is similar to the kinetics of helium release from samples irradiated in reactors.

Thermal annealing as a method to predict results of high temperature irradiation embrittlement

In order to assess the validity of post-irradiation annealing as a method to predict results of high temperature irradiation a new analysis of experimental data has been performed revealing the combined influence of annealing temperature and impurities content on residual embrittlement after annealing. For 2CrMoV (WWER-440 reactor pressure vessel) steel with low contents of copper and phosphorus, the comparison of two embrittlement dependencies has been done: on irradiation temperature and post-irradiation annealing temperature. It is demonstrated that data for both the transition temperature shift after irradiation, ΔTk, and the residual transition temperature shift after post-irradiation annealing, ΔTres, fall within the same scatter band. A similarly close correlation is observed by comparison of yield strength increases after irradiation and after post-irradiation annealing.

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

A comparison between pearlitic 2CrMoV steel (WWER-440) and 9% Cr based ferritic–martensitic steels (EUROFER 97 and LA12TaLC) is presented as regards irradiation induced ductile–brittle transition temperature shifts. For neutron doses of 1.5–2 dpa and irradiation temperatures around 300 °C the transition temperature shifts for WWER-440 steel and EUROFER 97 welds are comparable. In the temperature range 350–500 °C the radiation embrittlement levels of both steels are low. Moreover, post-irradiation annealing is proposed as a promising method to predict results of high temperature irradiation embrittlement from existing lower temperature irradiation embrittlement data.

Mechanical properties of high-nickel alloys EP-753 and РЕ-16 after neutron irradiation to 54 dpa at 400–650 °С

Short-term mechanical properties and void swelling were investigated for high-nickel alloys РЕ-16 and three compositional variants of Russian alloy EP-753 and in various starting conditions after side-by-side irradiation in the BN-350 fast reactor at 400, 500, 600 and 650 °С to 54 dpa. For both alloys irradiation resulted in significant hardening and ductility reduction dependent on their chemical composition and initial heat treatment. At test temperatures equal to the irradiation values both alloys exhibited a high level of strength and satisfactory ductility. In the test temperature range of 550–650 °С the phenomenon of high-temperature irradiation embrittlement was observed.

Utilization of electron accelerators for simulation and studies of radiation effects on mechanical properties of fusion reactor materials

Based on fundamental research in the field of radiation damage of materials, a method is developed to simulate and investigate principal mechanical properties of materials with the help of high-energy electron and gamma-ray beams. This method allows a simultaneous simulation of radiation damage arising during use of materials in nuclear fission and fusion reactors with different helium impregnation-to-damage generation rate ratios. The method is used to simulate and study radiation hardening, low-temperature and high-temperature radiation embrittlement and radiation creep phenomena. A correlation is established between the reactor and simulation experiments. This allows one to forecast the changes of mechanical properties of materials under operating conditions.

Grain boundary cavities and cracks during high temperature irradiation embrittlement

This paper discusses a possible role of grain boundary cavities in the high temperature irradiation embrittlement (HTIE) of reactor structural materials. It is demonstrated that grain boundary voids alone can hardly provide the dominance of intergranular fracture mode during HTIE. An alternative mechanism based on the growth of helium accumulating cracks is discussed and the resulting predictions for material failure parameters are presented.

The contribution of grain boundary cavities and cracks to the high temperature irradiation embrittlement of structural materials

This paper deals with the effect of an infinite row of equal elliptic cavities (which include voids or cracks on a grain boundary as limiting cases) on the stress distribution near them. It is shown that the contemporary widespread theoretical models overestimate the role of grain boundary cavities (helium bubbles) in the high temperature irradiation embrittlement (HTIE). It is also shown that the most effective interaction is realized for the row of cracks and the collective effects in the ensemble of cracks located in some plane can lead, in contrast to the ensemble of cavities, to the new qualitative effects (for example to the loss of stability of growing cracks due to the inner loading only). The growth kinetics of grain boundary cracks due to gaseous impurity accumulation in them is investigated in more detail.

Effect of helium pores on the mechanical properties, the structure, the high-temperature recovery processes, and the fracture of an austenitic high-nickel alloy

High-temperature annealing of the irradiated specimens of austenitic chromium--nickel steels clearly demonstrates the characteristic features of the phenomenon of high-temperature radiation embrittlement. In spite of the fact that the low-temperature properties are restored after annealing, the high-temperature ductility of the irradiated specimens remains at an extremely low level. Structural studies show [ 1-3] that after annealing at 1000-1100*C, the strengthening structural defects (dislocation loops, vacancy-pores, finely dispersed phases etc.) are virtually absent in the irradiated specimens. This fact indicates that the effect of p0stannealing embrittlement is mainly related to helium formed during neutron irradiation as a result of the (n, a)-reactions. Even the early studies on the phenomenon of high-temperature radiation embrittlement (HTRE) convincingly showed that the high-temperature ductility of the specimens of the 20 ,--} Cr/20Ni [1,2] and 316 [3] steels subjected to neutron i r radia t ionup to a fluence ranging from 2 • 109.s m-9. up to 1 • 109.5 m-9. is not restored even after annealing for a period of 60 days [3]. Furthermore, only the low-temperature ductility is restored after annealing the specimens of the 304 and 316 steels that were irradiated in an EVR-'II reactor [4] up to a neutron fluence of 4 • 109.6 m-9.. When testing at 600 and 700"C, the ductility of the irradiated and annealed specimens was found to be 30% of the ductility of the unexposed specimens. The specific role of helium in the embrittlement of the irradiated specimens was demonstrated by the studies of Kramer [5] and Shino [6]; their studies showed that the ductility of the 304 and 316 steels containing 6 • 10-4% 40 • 10-4% helium (introduced by irradiating in a cyclotron) is not restored even after annealing for a period of 3000 h. Based on the calculations and the experimental studies carried out during seventies for determining the helium content in neutron-irradiated metals [7-9], it is possible to interpret the data concerning embrittlement [1-3] with reference to the concentration of helium accumulated in the material (without relating to the neutron fluence). Intensive studies have been carr ied out in the Soviet Union [10-12] on the role of helium in HTRE and the mechanism of helium-induced embrittlement. In contrast to the data of mechanical testing, the information concerning the structure of the irradiated and annealed specimens is quite limited and not systematic. It is known that after annealing (Tan n _> 650"C), irradiated specimens show helium bubbles; their size increases with increasing annealing temperature and their density increases with increasing helium concentration. However, the effect of helium-induced porosity on the mechanical properties of the annealed specimens has not been evaluated quantitatively in the experimental studies carried out up to now. In this context, the absence of the studies on the postannealing embrittlement of the specimens containing different concentrations of helium forms the main problem. In view of this, we studied the effect of the helium bubbles on the mechanical properties and the structure of the 03Kh20N45M4B alloy.

Grain-boundary deformation in nickel irradiated by α-particles

The results of a study of grain-boundary deformation in Ni doped with 10−6 at% He tested in the temperature range 20 to 1000 °C are given. Compared to non-irradiated metals more intense deformation is observed in the samples irradiated by α-particles. The data show the predominant role of grain-boundary deformation in the high-temperature radiation embrittlement process. [Russian Text Ignored].

Investigation of the state of helium implanted into molybdenum lattice by alpha-particle bombardment

Abstract Knowledge of the state and behaviour of helium in reactor components is needed in order to understand such phenomena as void swelling and high temperature radiation embrittlement. The X-ray diffraction method of lattice parameter measurement can provide useful information about the state of helium in radiation damaged lattices of metals if the effects of irradiation induced lattice defects—vacancies and interstitials as well as their agglomerates—on lattice parameter changes are taken into account. On the basis of comparison of TEM results on damage microstructures in molybdenum uniformly doped with helium by alpha-particle bombardment and those of neutron irradiated, the effect of helium state on molybdenum lattice parameter change has been extracted. It has been shown that at relatively low concentrations helium forms complexes with vacancies, HemVn, where m < n. As the concentration increases, this correlation between vacancy and helium atoms numbers changes to m>n, leading to considerable expansion of the lattice.

Relationship between the high-temperature radiation embrittlement of austenitic stainless steels and nickel alloys and focusing processes

Relationship between the high-temperature radiation embrittlement of austenitic stainless steels and nickel alloys and focusing processes

A contribution to the interpretation of high-temperature irradiation embrittlement

At temperatures above about 500 °C neutron irradiation produces high-temperature embrittlement of austenitic steels and nickel alloys due to the forming of helium in the lattice by (n, α)-transmutation processes. The author shows how this irradiation embrittlement can be interpreted with the aid of the model of the gas-filled wedge-type crack.

POSSIBLE FAST-REACTOR CANNING MATERIAL STRENGTHENED AND STABILIZED BY DISPERSION

The high-temperature irradiation embrittlement of steels and nickel alloys is probably due to helium precipitation in grain boundaries. It would be desirable to retain the helium as fine bubbles evenly distributed within the grains. A fine dispersion in the metal matrix may act as nucleation site for helium bubbles and it is proposed to produce by powder-metallurgy techniques afine dispersion of a ceramic oxide in a steel matrix. Ferritic chromium steel was chosen because it is known to be less susceptible to embrittlement than austenitic alloys and in this case the oxide dispersion must also improve the high-temperature mechanical properties.Methods of preparation are discussed and results of preliminary tensile tests performed on iron containing alumina, magnesia, and titania dispersions indicate the feasibility of the proposed solution.