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.     

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.     

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.     

In-reactor deformation of solution annealed type 304L stainless steel

Over the past six years at EBR-II, a great deal of information has been obtained on the in-reactor behaviour of solution annealed-Type 304L stainless steel. This information consists of the following: (1) Irradiation induced swelling results in the form of immersion density and transmission electron microscope (TEM) measurements on unstressed material that extends over a temperature range of 395° to 530°C and a neutron fluence range of 1.8 to 9.3 × 10²² n/cm² (E > 0.1 MeV). (2) Irradiation induced creep results from helium pressurized capsules irradiated at a temperature of 415°C. The hoop stress range covered in the experiment was 0 to 27.3 ksi, and the peak neutron fluence obtained to date is 7 × 10²² n/cm² (E > 0.1 MeV). (3) Residual stress measurements (slit tube technique) with complementary TEM gradient studies on stressed and unstressed capsules. (4) Comparative swelling studies of stressed cladding material and unstressed capsule material from encapsulated EBR-II driver fuel experiments over wide ranges of temperature and neutron fluences. The deformation information derived from the four above studies represent an extensive data base from which to obtain an understanding of the in-reactor deformation of austenitic stainless steel. It is the purpose of this paper to review our information on the in-reactor deformation of solution annealed Type 304L stainless steel.     

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 × 10²⁰ n/cm²($\text{E > 1}$ MeV) and 2.03 × 10²¹ n/cm² ($\text{E > 1}$ 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 ($\text{E > 1}$ 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.     

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.     

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.     

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.     

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.     

Irradiation creep of previously irradiated 304 stainless steel in a fast flux environment

Results of a fast flux neutron irradiation experiment designed to investigate the effects of high levels of prior irradiation (to 10²³ n/cm², E > 0.1 MeV) on the irradiation creep of type 304 stainless steel at 700 K (800°F) are reported. The steady state creep coefficient is found to increase by a factor of 7 as the specimen fluence increases from 3 to 10 × 10²²n/cm², (E > 0.1 MeV). A non-linear dependence of strain on stress is exhibited. The results of this study are compared with previously reported stress relaxation data and with predictions of a swelling enhanced irradiation creep model.     

Effects of neutron irradiation on polymer matrix composites at 5 K and at room temperature: II. Degradation of mechanical properties

Neutron irradiation with a low flux of accompanying γ-rays in the Intense Pulsed Neutron Source was carried out at 5 K and at room temperature on four kinds of polymer matrix composites (filler: E-glass or carbon fiber cloth; matrix: epoxy or polyimide resin). The specimen irradiated at 5 K was warmed up to room temperature before the mechanical test was performed at 77 K. The Young's modulus of these composites scarcely decreases even when a total neutron fluence is 3.0 × 10¹⁸n/cm² (2.1 × 10¹⁸n/cm² for E > 0.1 MeV) for the 5 K irradiation and 1.6 × 10¹⁹n/cm² (8.0 f 10¹⁸ n/cm² for E > 0.1 MeV) for the room-temperature irradiation. The ultimate strength, however, decreases significantly at this neutron fluence for all the composites except the carbon/epoxy composite whose initial strength is comparatively low. Comparison of this result with that obtained for ⁶⁰Co γ-ray irradiation demonstrates that the radiation sensitivity of the glass/epoxy and glass/polyimide composites is 1.8–2.6 times higher towards neutrons than γ-rays. As to the irradiation temperature of 5 K and room temperature, no significant influence on the degradation efficiency of the composite strength is observed under the present conditions of mechanical test.     

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.     

The relation between stress-relaxation and creep for some zirconium alloys during neutron irradiation

In-reactor stress-relaxation tests on beam specimens of several zirconium alloys have been performed at 566 K in a fast neutron flux (E>1 MeV) of 2 × 10¹⁷ n/m² · s. The stress-relaxation behaviour is characterized by an initial rapid decrease in the unrelaxed stress ratio followed by a slower steady-state decrease which can be expressed by the relation,

$\text{ln}\text{(σ}\text{σ}{}_0\text{) = ? (AE)t +}\text{ln}\text{D}$

, where $\text{σ}\text{σ}{}_0$ is the unrelaxed stress at a given time t as a fraction of the initial stress σ₀, A is a flux-, material-,temperature-dependent constant, E is Young's modulus, and D is given by $\text{σ}{}_{\mathrm{p}}\text{σ}{}_0$ where σ_p approximates the initial stress decrease.The linear dependence of In $\text{σ}\text{σ}{}_0$ on time and the independence of the stress-relaxation behaviour on the initial stress implies that the creep rate in the steady-state period can be given by the expression, $\text{ε}\text{?}\text{=}\text{A}\text{σ}$. Good agreement was obtained between creep rates derived from stress-relaxation tests on experimental pressure tube materials and creep rates derived from diameter measurements of pressure tubes for identical temperature, flux and stress conditions.     

Ferrite formation in neutron-irradiated type 304L stainless steel

Austenitic (γ) to ferrite (α) transformation was observed using transmission electron microscopy in type 304L stainless steel that had been irradiated at ~500°C to fast-neutron (E > 0.1 MeV) fluences greater than ~ 3 × 10²²n/cm². Previous studies on similar unirradiated stainless steels found no such transformation, indicating that the γ → αtransformation was irradiation-induced. The α phase appeared to nucleate on stacking faults, indicating that the presence of large Frank loops was the critical step in the transformation. After an entire grain of austenite had transformed, the only remaining γ phase existed as shells around voids. Coincidence of rapid swelling behavior with γ → α transformation indicated that the two were related, perhaps by reaction of both phenomena to the effects of irradiation and temperature on microchemical segregation. A volume expansion of about 2.5% was found to be associated with the transformation. Inferences are drawn relating this behavior in type 304L steel to the effects of radiation on other reactor structural materials such as type 316 stainless steel, which is also a metastable austenitic composition.     

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.     

Anisotropic growth of reactor fuel pin cladding

Anisotropic growth of 316 stainless steel reactor fuel pin cladding was found to occur after irradiation in the Experimental Breeder Reactor-II (EBR-II). Pressurized tube specimens were irradiated to a peak fluence of 10²³n/cm² (E >0.1 MeV) at temperature ranging from 430°C to approximately 590°C. Growth was observed in both the annealed and 20% cold worked conditions and was found to decrease with increasing hoop stress. The anisotropic growth is more pronounced in the cold worked condition. The growth is attributed to a preferred orientation of Burgers vectors in the preirradiated cold worked dislocation structure.     

Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

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.