Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72–1.92 × 10²⁰n/cm² (E>1 MeV) and 2.03×10²¹n/cm² (E>1 MeV) at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 ×10²¹ n/cm² (E>1 MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

Irradiation induced stress-relaxation of previously irradiated 304 stainless steel in a fast flux environment

Results of a recent fast flux neutron irradiation experiment in EBR-II designed to determine the effects of high levels of prior irradiation (to 10²³ n/cm², $\text{E}$ > 0.1 MeV) on the irradiation creep of type 304 stainless steel at 800° F are reported. The primary conclusion drawn from the data is that the steady state creep coefficient increases by a factor of 8 as the specimen fluence increases from 0 to 10.0 × 10²² n/cm² ($\text{E}$ > 0.1 MeV). The irradiation creep coefficients are consistent with a linear variation in creep rate with swelling rates over the entire data range. The restrictions that the experimental results place on the choice of a theoretical model for irradiation creep are cited.     

The effect of temperature and neutron irradiation upon the fatigue-crack propagation behavior of ASTM A533-B steel

The fatique-crack propagation behaviour of A533-B steel was studied within the framework of linear-elastic fracture mechanics. Tests were conducted at 75° F (24° C) and 550° F (288°C) on unirradiated material, and on material irradiated at 550° F to $\text{2.3 – 2.8 × 10}{}^{19}\text{n/cm}{}^2$ and $\text{5.3 – 5.7 × 10}{}^{19}\text{n/cm}{}^2\text{(E > 1 MeV)}$. In general, at the cyclic frequency used (600 cpm), neither temperature nor neutron irradiation had a significant effect on the fatigue-crack propagation.     

Voids resulting from fast neutron irradiation of a stabilized stainless steel

A 16 Cr-13 Ni-niobium stabilized stainless steel (Type 1.4988) was irradiated as fuel pin cladding in the DFR reactor. The irradiation temperatures ranged from 300 to 650 °C. Neutron fluences from 3.3 to 4.0 n/cm²$\text{(E > 0.1 MeV)}$ were achieved. Swelling behavior was studied in this material by transmission electron microscopy, immersion density determinations and diametral change measurements.Void formation was observed in the temperature range 360–610 °C. Void concentration values at lower temperatures (< 480 °C) are comparable to those cited for the unstabilized stainless steels AISI 304 and 316, however, they decrease more strongly with increasing irradiation temperatures. The mean and maximum diameters show a pronounced maximum at about 530 °C. The decrease of the void diameters at higher temperatures can be explained by the nature of the cavities observed. Annealing experiments with specimens irradiated at 610 °C have revealed a bubble or a mixed void-bubble character for these cavities.Swelling in type 1.4988 was found to be lower than that reported previously for types 304 and 316 at comparable irradiation and material conditions. This is especially pronounced at higher temperature. Diameter changes of the pin at $\text{T >- 610 °C}$ cannot be explained by void-formation.     

Effect of fast-neutron irradiation on mechanical properties of stainless steels: AISI types 304, 316 and 347

The effect of low-temperature (< 100° C) fast-neutron irradiation on the room-temperature tensile and hardness properties of stainless steels, AISI Types 304, 316, and 347, was investigated up to a fluence of 1.43 × 10²⁰ n/cm² (E > 1 MeV). Several methods were used for analysis of results and the approach using the irradiation-induced increase in yield stress, Δσ = σ ? σ, where σ_i and σ are the yield stresses of irradiated and unirradiated specimens, respectively, proved to be the best for describing irradiation-hardening. Below saturation fluence, ≈ 4?5 X 10¹⁹n/cm² (E > 1 MeV), it was shown that $\text{Δσ ∝(øt)}\text{1}\text{2}$ in agreement with Seeger's model. Yield points were observed at a fluence of _{1.3 × 10¹⁹ n/cm²} (E > 1 MeV) and above. The results are discussed in relation to transmission electron microscopy results of irradiated materials. The relation between irradiation-induced changes in yield stress and Vickers hardness was described by ΔH = KΔσ, where K = 2.82 for AISI Type 304, and 3 for both AISI Type 316 and AISI Type 347.     

Effects of pre-irradiation treatments and minor composition modification on neutron irradiation embrittlement of 316 stainless steel

The effects of the combination of heavy cold work and low temperature aging (873–1073 K, 100 h), and minor composition modification on the irradiation embrittlement of 316 stainless steel were investigated. The samples were irradiated by JMTR at JAERI to the dose level of 2.5 × 10²⁴ n/m²$\text{(E > 1 MeV)}$ at 823 and 923 K and tensile tested between R.T. and 1023 K. The embrittlement was compared from the standpoint of ductility survival ratio. The lowering of carbon content caused severer high temperature helium embrittlement in the solution treated condition. The heavy cold work and low temperature aging treatments could not improve the high temperature embrittlement compared with the solution treated condition. Titanium addition was beneficial especially for the reduction of the irradiation temperature sensitivity to the high temperature ductility.     

In-reactor stress relaxation in bending of 20% cold-worked 316 stainless steel

A test is in progress to measure in-reactor stress relaxation of 20% cold-worked 316 stainless steel in bending through the sequential irradiation in the experimental reactor EBR-II of two materials capsules. This paper details the results from the first capsule, the second capsule has completed irradiation and is awaiting examination. Approximately 50% total relaxation was measured following irradiation to $\text{2 × 10}{}^{21}\text{n/cm}{}^2$, $\text{E > 0.1}\text{MeV}$ at 370°C. The extent of relaxation was independent of specimen orientation to rolling direction, independent of initial stress level and only weakly dependent on neutron dose at this fluence level.     

The effect of fast neutron irradiation upon the fatigue-crack propagation behavior of two austenitic stainless steels

The effect of fast neutron irradiation ($\text{454° < T}{}_{\mathrm{irr}}\text{< 477° C}$) to a fluence of $\text{9 × 10}{}^{21}\text{n/cm}{}^2$ ($\text{E > 0.1 MeV}$) on the fatigue-crack growth behavior was investigated for annealed Type 304 and 20% coldworked Type 316 stainless steels using linear-elastic fracture mechanics techniques. Irradiation to this fluence had little or no effect upon the crack growth behavior of annealed Type 304 at a test temperature of 427° C, nor upon the behavior of 20% cold-worked Type 316 at test temperatures of 427° C and 538° C. Irradiation to this fluence did tend to decrease crack growth rates slightly, relative to unirradiated material, in annealed Type 304 at a test temperature of 538° C.     

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.     

Comparison of 16 MeV proton, 14 MeV neutron and fission neutron damage in copper

The irradiation damage structures produced in high-purity copper by a fluence of $\text{3 × 10}{}^{16}$ particles/cm² of 16 MeV protons, 14 MeV neutrons, and fission neutrons ($\text{E > 1 MeV}$) were studied by transmission electron microscopy. The damage consists of vacancy-and-interstitial clusters or sessile Frank dislocation loops oriented on {111} planes of the copper matrix, and ranges in size from 25 Å (lower limit of resolution) to 200Å in diameter. p]The size-density distributions of the clusters in the 14 MeV neutron and 16 MeV proton irradiated samples were virtually identical, and the average size of the clusters in these two groups of specimens was substantially larger than was the case for those in the fission-neutron-irradiated copper.     

Swelling of type-316 stainless steel at high fluences in EBR-II

A test to measure swelling induced by fast neutron irradiation in unstressed specimens of type-316 stainless steel has completed irradiation in the EBR-II reactor. Results are reported and discussed which describe the swelling as a function of neutron fluence, temperature of irradiation and extent of cold work in the alloy. Density determinations showed swellings of up to 15% $\text{ΔV}\text{V}{}_{\mathrm{f}}$ for 20% cold worked type-316 stainless steel at a neutron fluence level of 1.4 × 10²³n/cm², E > 0.1 MeV (70 dpa). The peak swelling temperature range was 550°C–600°C regardless of the extent of cold working. Increasing the cold work level reduced the swelling and tended to broaden the swelling temperature peak. Transmission electron microscopy (TEM) investigations showed that cold working had reduced the average void sizes compared to those observed in the solution annealed material.     

Neutron irradiation effects on the mechanical properties of organic composite materials

Neutron irradiations with low γ-ray flux in the Intense Pulsed Neutron Source were carried out on four kinds of cloth-filled organic composites (filler: E-glass or carbon fiber; matrix: epoxy or polyimide resin) and a unidirectional alumina fiber/epoxy composite. These composites were examined with regard to the mechanical properties at room temperature. Following irradiation at room temperature, the Young's (tensile) modulus of these composites remains practically unchanged up to a total neutron fluence of $\text{5.0 × 10}{}^{18}\text{n/cm}{}^2$ ($\text{1.4 × 10}{}^{18}\text{n/cm}{}^2$ for $\text{E > 0.1 MeV}$). The shear modulus and the ultimate strength, on the other hand, decrease significantly at this neutron fluence for the glass/epoxy and glass/polyimide composites, whereas for the other composites both properties do not degrade. This result is most likely ascribed to the radiation damage at fiber/matrix interface due to recoil particles produced by a ${}^{10}\text{B(n,α)}{}^7$ Li reaction in the boron-containing E-glass fibers. Only for the E-glass fiber composites, in fact, the fracture propagation energy is appreciably increased by irradiation, while for the other composites the propagation energy is scarcely changed, thus confirming the significant contribution due to the ¹⁰B reaction. As to the 5 K irradiation, degradation of the present composites was not observed up to a total neutron fluence of $\text{1.0 × 10}{}^{18}\text{n/cm}{}^2$ ($\text{7.0 × 10}{}^{17}\text{n/cm}{}^2$ for $\text{E > 0.1}\text{MeV}$) when tested at room temperature.     

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.     

Annealing behavior of neutron irradiated β-SiC

The post-irradiation annealing behavior of β-SiC for use as a monitor of irradiation temperature is discussed. Powder and rods of polycrystalline β-SiC were irradiated to $\text{1.5 × 10}{}^{17}$ to $\text{5.0 × 10}{}^{19}\text{n/cm}{}^2\text{(E > 0.18 MeV)}$ at temperatures between 290 and 500°C. The estimated temperatures deduced from the changes in lattice constant and specific electric resistivity during progressive annealing, and from thermal expansion measurement by high-temperature X-ray diffraction agreed with values determined by means of a thermocouple. Thermal expansion measurement in a conventional dilatometer resulted in an over estimate of the irradiation temperature, and further improvement of this method is required for experimental application.     

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.     

The characteristics of 15 MeV and fission neutron damage in niobium

Displacement damage by 15 MeV (d-Be source) and fission neutrons at 30°C in high purity niobium single crystals has been studied by transmission electron microscopy. The fluence of the 15 MeV neutrons was 1.8 × 10¹⁷n/cm² and the fluence of the fission neutrons (5 × 10¹⁷ n/cm²) was chosen so that samples from both types of irradiations had approximately the same damage energy. In both 15 MeV and fission neutron irradiated specimens, the loops were observed to be about $\text{2}\text{3}$ interstitial and $\text{1}\text{3}$ vacancy type. The analysis of Burgers vectors of the dislocation loops showed that more than $\text{2}\text{3}$ of the loops were perfect $\text{a}\text{2}\text{〈111〉}$ and that the rest were $\text{a}\text{2}\text{〈110〉}$ faulted. It is concluded that at equal damage energies, the detailed nature of the damage is the same for 15 MeV and fission neutron irradiated niobium.     

Effects of stress on swelling in 316 stainless steel

Stress was found to increase the magnitude of irradiation-induced swelling in 316 stainless steel. Measurement of the densities of pressurized tube specimens, irradiated at temperatures of ~ 430–475°C to peak fluences of $\text{~ 9 × 10}{}^{22}\text{n/cm}{}^2$ ($\text{E > 0.1 MeV}$) in EBR-II, has indicated increased swelling in both the annealed and 20% cold worked conditions of this alloy. Swelling in the annealed specimens was observed to increase linearly with hoop stress up to ~ 20 ksi (130 MPa), whereupon further increases in stress resulted in reduced swelling. Swelling in the cold worked material was linear with stress up to levels of ~ 28 ksi (193 MPa).     

Neutron irradiation-induced voids in β-silicon carbide

Pyrolytic β-silicon carbide was irradiated at temperatures between 625°C and 1500°C to neutron fluences up to 12 × 10²¹ n/cm² ($\text{E > 0.18 MeV}$). Density changes were measured and the samples were examined by transmission electron microscopy. Irradiation below 1000°C created small Frank dislocation loops on {111} planes. Irradiation at 1250°C and 1500°C produced tetrahedral voids which caused continuing expansion of the samples. Void sizes increased with increasing fluence and with increasing irradiation temperature, while void concentrations decreased with increasing temperature. One-hour post-irradiation anneals at 1700°C to 2100°C reduced the void concentration and total void volume while increasing the maximum void size.     

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.     

Einführung in die Werkstoffwissenschaft (in German): ed. W. Schatt (VEB Deutscher Verlag für Grundstoffindustrie,Leipzig, 1972, DM 35)

Copper samples were irradiated with fast neutrons at temperatures in the range 220–550 °C and at instantaneous fluxes in the range 2 × 10¹³-3 × 10¹⁴ n/cm².sec > 0.1 MeV. The maximum swelling was observed at 0.45 $\text{T}{}_{\mathrm{f}}$ for an instantaneous flux of 3 × 10¹⁴ n/cm²-sec. A fourfold reduction of the instantaneous flux, at constant dose, displaces the maximum to lower temperatures and slightly increases its magnitude. Cold work before irradiation does not appear to have a significant effect on swelling. Alloying with solutes which lower the stacking-fault energy appears to displace the domains of swelling towards lower temperatures for a fixed instantaneous flux and towards lower flux for a fixed temperature.