The effect of a temperature change on the in-reactor creep of 20% cold worked AISI 316 stainless steel

Results of a temperature change experiment designed to investigate the effects of prior irradiation temperature on the irradiation creep of 20% cold worked AISI 316 stainless steel are reported. The data indicate that this material exhibits an in-reactor “temperature memory” effect. After a temperature change, the material creeps as if it were still subjected to its prior irradiation temperature. Specimens subjected to a decrease in irradiation temperature are found to swell more rapidly than materials subjected to a constant temperature irradiation.     

The effect of neutron and He-irradiation on the crack modes of type 316 stainless steel

The objective of this investigation is to determine the crack opening mode (Mode I and Mode II) during in situ HVEM tensile testing and how it is influenced by neutron and helium irradiation, and test temperature. Mode II crack opening was observed as grain boundary sliding initiated by a predominantly Mode I crack steeply intersecting the grain boundary. Mode II crack opening was absent in neutron- or helium-irradiated specimens tested between 400°C and room temperature, but could be restored by a post-irradiation anneal.     

Defect development in neutron irradiated type 316 stainless steel

The effects of fast neutron irradiation on the defect development in unstressed solution treated Type 316 stainless steel were investigated by transmission electron microscopy. The irradiation conditions investigated covered the fluence range from 0.75 to $\text{5.1 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E > 0.1}\text{MeV}$) and temperatures from 380 to 850°C. Empirical equations were developed relating the void volume, void number density, mean void size, Frank faulted loop diameter, Frank loop number density and dislocation density with the neutron fluence and irradiation temperature. Void nucleation changes from homogeneous at low irradiation temperature (? 400°C) to heterogeneous at higher temperatures in that voids are preferentially associated with irradiation induced rod shaped precipitates. The void number density decreases while the void diameter increases with irradiation temperature. The total faulted loop line length per unit volume and dislocation density increases with fluence and decreases with temperature. The Frank loop diameter increases and number density decreases with temperature. The range of temperature in which Frank faulted loop formation occurs decreases with neutron fluence.     

Effect of thermal aging on the room temperature tensile properties of AISI type 316LN stainless steel

Round tensile specimens of AISI type 316LN stainless steel, thermally aged at 1123 K for 0, 2, 10, 25, 100, 500 and 1000 h, were tested for tensile properties at room temperature at a strain rate of 7.7 × 10⁻³ s⁻¹. The changes in tensile properties were correlated to the transmission electron microscopic studies. The various stages of nitrogen repartitioning including Cr–N cluster formation, intragranular and subsequent cellular precipitation of Cr₂N were found to have a strong influence on the yield strength (YS) and ductility of the material. However, the changes in ultimate tensile strength (UTS) with aging were negligible. The results of electrochemical extraction of secondary phases clearly indicated a two-slope behavior. X-ray diffraction analysis of electrochemically extracted residue suggested that the initial smaller sloped line corresponded to the precipitation of the Cr₂N phase while the line with larger slope at longer aging time corresponded to the domination of chi phase precipitation.     

Failure of type 316 stainless steel cladding during thermal transients

To obtain a basic understanding of the deformation and failure behaviors of 20% cold-worked type 316 stainless steel cladding during reactor over-power and loss-of-coolant transients, the mechanical response of the cladding was investigated, using a true 2 to 1 stress-state transient tube-burst method at a heating rate of 5.6°C/s (10°F/s).The uniform diametral strain to failure of the tubular cladding specimens was uniquely defined, based on strain profiles of the failure specimens and high-speed motion pictures taken during the tests. The effects of initial hoop stress, specimen length and metallurgical condition on the transient tube-burst response were studied. The heat-to-heat variation of the test results were examined.The uniform diametral strains to failure for the three heats of cladding tested in the present study were found to be consistently higher than the strains reported by Hunter et al.     

Nickel-ion bombardment of annealed and cold-worked type 316 stainless steel

Bombardment with high doses of 5 MeV nickel ions has produced swellings as high as 90% and 60%, respectively, in annealed and 20% cold-rolled Type 316 steels. The steels contained 15 ppm of cyclotron-injected helium. Swellings were determined by both transmission electron microscopy and by a step-height method that measures the total swelling integrated along the ion path. The swelling in annealed Type 316 has a pronounced peak in the vicinity of 625°C, which is about 155°C higher than the peak swelling temperature in-reactor. The magnitudes of the swelling, void densities and void sizes produced in annealed Type 316 by nickel ions and in-reactor at the respective peak swelling temperatures are similar and it is concluded that the nickel ion bombardments provide an acceptable simulation of in-reactor behavior. Using the high dose ion results to guide extrapolation of presently available EBR-II data to higher fluences leads to the prediction that the swelling of annealed Type 316 steel at the peak swelling temperature will reach 40% at $\text{2 × 10}{}^{\mathrm{p23}}\text{n/cm}{}^2\text{(E > 0.1 MeV)}$ in EBR-II core, and 70% at $\text{3 × 10}{}^{23}\text{n/cm}{}^2$. These fluences in EBR-II correspond to 155 and 230 dpa respectively. Twenty percent reduction by cold-rolling reduces the ion produced swelling by 35% at 625°C and by 50% at 575°C.     

Dependence of irradiation creep on temperature and atom displacements in 20% cold worked type 316 stainless steel

Irradiation creep studies with pressurized tubes of 20% cold worked Type 316 stainless steel were conducted in the Second Experimental Breeder Reactor. These studies have shown that as atom displacements are extended above 5 dpa and temperatures are increased above 375°C, the irradiation induced creep rate increases with both increasing atom displacements and increasing temperature. The stress exponent for irradiation induced creep remained near unity. Irradiation induced effective creep strains up to 1.8% were observed without specimen failure.     

Sorption of gaseous tritium on the surface of type 316 stainless steel

The sorption of gaseous tritium on the type 316 stainless steel was studied. The stainless steel was first contacted with gaseous tritium, and then the remaining tritium was evacuated. During a gradual etching from the surface by an acid solution, the tritium was released as HTO with a fraction of HT. They were radioassayed separately. The HTO mostly originates from the tritium present on the outer-most surface and about 90% of it could be released easily into water. However, the rest is sorbed tightly and remained in the surface layer. A fraction of the sorbed-tritium will diffuse atomically through the surface layer into the bulk of stainless steel and is released as HT by etching. The activation energy of the diffusion was determined as 32.8 kJ/mol.     

Crack propagation characteristics in neutron irradiated type 316 stainless steel microspecimens

In order to better relate the macroscopic mechanical behavior of irradiated alloys to their associated microstructural condition, unirradiated and neutron irradiated microspecimens were tensile tested at 25–600°C in a quantitative load elongation stage while under continuous observation in a high voltage electron microscope (HVEM). The microtensile specimens, 40 μ m thick, of type 316 stainless steel were irradiated at ambient temperature to a fluence of 1 × 10²² n/m² with 14 MeV neutrons in the Lawrence Livermore Rotating Target Neutron Source II (RTNS) facility.Crack angles, directions and length plotted against total specimen elongation were used to describe the manner in which a crack progressed through each specimen. Rapid crack propagation is accompanied by rapidly changing crack angles and direction and conversely slow propagation corresponds to slowly changing variables. A graph of cumulative crack length plotted against total elongation exhibits a slope which increases as specimen ductility decreases. This graph reflects changes due to the effect of neutron irradiation.     

Behavior of annealed type 316 stainless steel under monotonic and cyclic biaxial loading at room temperature

This paper addresses the elastic-plastic behavior of type 316 stainless steel, one of the major structural alloys used in liquid-metal fast breeder reactor components. The study was part of a continuing program to develop a structural design technology applicable to advanced reactor systems. Here, the behavior of solution annealed material was examined through biaxial stress experiments conducted at room temperature under radial loadings (√3τ = σ) in tension-torsion stress space. The effects of both stress limited monotonic loading and strain limited cyclic loading were determined on the size, shape, and position of yield loci corresponding to a small offset strain (10 microstrain) definition of yield.In the present work, the aim was to determine the extent to which the constitutive laws previously recommended for type 304 stainless steel are applicable to type 316 stainless steel. It was concluded that for the conditions investigated, the inelastic behavior of the two materials are qualitatively similar. Specifically, the von Mises yield criterion provides a reasonable approximation of initial yield behavior and the subsequent hardening behavior, at least under small offset definitions of yield, is to the first order kinematic in nature.     

Radiation-induced stress relaxation in high temperature water of type 316L stainless steel evaluated by neutron diffraction

Weld beads on plate specimens made of type 316L stainless steel were neutron-irradiated up to about 2.5 × 10²⁵ n/m² (E > 1 MeV) at 561 K in the Japan Material Testing Reactor (JMTR). Residual stresses of the specimens were measured by the neutron diffraction method, and the radiation-induced stress relaxation was evaluated. The values of σ_x residual stress (transverse to the weld bead) and σ_y residual stress (longitudinal to the weld bead) decreased with increasing neutron dose. The tendency of the stress relaxation was almost the same as previously published data, which were obtained for type 304 stainless steel. From this result, it was considered that there was no steel type dependence on radiation-induced stress relaxation. The neutron irradiation dose dependence of the stress relaxation was examined using an equation derived from the irradiation creep equation. The coefficient of the stress relaxation equation was obtained, and the value was 1.4 (×10⁻⁶/MPa/dpa). This value was smaller than that of nickel alloy.     

The effect of pulsed irradiation on the swelling of 316 stainless steel in fusion reactors

A time-dependent rate theory formulation has been used to study the effects of pulsed irradiation on point defect and void behavior at elevated temperatures. It is found that point defects in pulsed tokamaks, θ-pinchs and inertial confinement fusion reactors (ICFR) display non-steady-state behavior. The pulsed nature of the irradiation has been shown to produce considerable deviations from steady-state void growth behavior at high temperatures (0.3 T_m to 0.5 T_m). In particular, the amount of swelling in the first-wall can be reduced for ICFR pulsing conditions and pulse widths ranging from a nanosecond to a microsecond. The amount of reduction increases with increased pellet yield at a fixed operating temperature, geometry and ICFR plant power output.     

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.     

Effect of surface oxide layer on deuterium permeation behaviors through a type 316 stainless steel

Effect of surface oxide layer on the hydrogen isotope permeation was studied. Iron oxide was uniformly formed in the oxide layer, although chromium was limited at the interface between the oxide layer and bulk SS-316. The permeation behavior of deuterium for oxidized SS-316 was compared with that for unoxidized SS-316 at temperature range of 333–673 K. The deuterium permeability for the oxidized SS-316 was reduced 1/10–1/20 times as high as that for unoxidized one. However, the activation energy of deuterium permeation as gas form for oxidized SS-316 was almost the same as that for unoxidized SS-316 and was 0.64 eV, which was almost consistent with the sum of activation energies for diffusion and solubility. This fact indicates that the deuterium permeation is diffusion limited. The permeability of deuterium as water form was almost constant even if heating temperature is high, showing that the deuterium was permeated through bulk SS-316 and react with oxygen at the oxide layer as water desorption, which is controlled by the permeation flux of deuterium and oxygen concentration on the surface of oxide layer in downstream side.     

Creep failure time of 20% cold-worked type 316 stainless steel

The Robinson failure criterion is examined for its accuracy to predict the creep failure time of 20% cold-worked type 316 stainless steel under uniaxial and multiaxial stresses. Observed changes of slope in the log-log plots of maximum tensile stress versus isothermal rupture life are neglected, and large errors in prediction of creep failure time are found. A failure criterion that best explains the experimental behavior of 20% cold-worked type 316 stainless steel in uniaxial and multiaxial creep conditions is developed by incorporating the effective stress responsible for crack initiation and the maximum tensile stress responsible for crack propagation into the creep failure time equation.     

Tensile properties of annealed type 316 stainless steel after EBR-II irradiation

The effects of fast reactor neutron irradiation on the tensile properties of annealed Type 316 stainless steel were determined over a wide range of irradiation temperatures and reactor fluences. These effects are described as a function fluence and temperature to $\text{7 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E}{}_{\mathrm{n}}\text{> 0.1}\text{MeV}$) at 430 to 820 °C. The usual flow stress increase and ductility decrease were observed with increasing neutron fluence. Strengthening decreases continuously from 480 °C to ≈ 700 °C with no hardening at or above 760 °C. Elongation values increase with temperature in the 430–540 °C range and generally decrease with temperature above 540 °C.