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.     

Subject index

In 1965 eight surveillance subassemblies were placed in row 12 of the EBR-II sodium-cooled fast breeder reactor with an irradiation temperature near the sodium-inlet temperature of 371°C. At the same time, two other surveillance subassemblies were placed in the primary storage basket, which receives minimal neutron exposure but is immersed in primary sodium and experiences a temperature of 371°C. Each of the subassemblies contained 18 preloaded springs made of Inconel X750. Springs from four of the in-core subassemblies and one subassembly from the storage basket have been evaluated to determine irradiation-enhanced deformation rates to neutron exposures of 4.2 dpa.It was found that the creep coefficient derived from the stress relaxation measurements on Inconel X750 springs was 1.0 × 10^?12 (Pa-dpa)^?1 for springs irradiated up to 4.2 dpa (3751 d) at an in-reactor temperature of 371°C. The relaxation behavior was adequately described by a creep law that was linear in neutron fluence and applied stress. Springs encapsulated in helium showed identical in-reactor relaxation rates to springs exposed to the flowing primary sodium. The creep coefficient derived from the present work on Inconel X750 springs was shown to be the same as the creep coefficients determined from various austenitic stainless steel alloys.     

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.     

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.     

Irradiation creep relaxation of void swelling-driven stresses

Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 °C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 °C and 504 °C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are analyzed using a recently proposed multiaxial creep-swelling model.     

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.     

Crack Initiation Mechanism in Non-ductile Cracking of Irradiated 304L Stainless Steels under BWR Water Environment

The deformation behavior and initiation mechanisms of intergranular (IG) and transgranular (TG) cracks in irradiated 304L stainless steel were studied by slow-strain-rate tensile tests in inert gas and simulated BWR water environments, followed by fractographic and microstructural examinations. Neutron irradiation was made in test reactors to fluences of up to 6.2x10²⁰ n/cm² (E>1 MeV). Intergranular cracking occurred in water above a critical neutron fluence of around 1 × 10²⁰ n/cm², based on the results of the SSRT tests and SEM fractography. That critical fluence is mechanistically supported by irradiated, deformed microstructures exhibiting dislocation channeling at that fluence, while radiation-induced Cr depletion at the grain boundaries was minor. Transgranular cracking of the irradiated material occurred in water below the critical fluence, initiating in the non-uniformly strained surface region of the test bar in the later stages of plastic deformation. The initiation of TG cracking is hypothesized to be related to a high density of deformation twins. Intergranular cracking is proposed to have initiated where localized slip bands terminated at grain boundaries, while TG cracking is inferred to have initiated at deformation twin boundaries. High stress and strain concentrations at grain/twin boundaries would be the common cause of non-ductile crack initiation.     

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.     

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.     

The effects of low dose rate irradiation and thermal aging on reactor structural alloys

As part of the EBR-II reactor materials surveillance program, test samples of fifteen different alloys were placed into EBR-II in 1965. The surveillance (SURV) program was intended to determine property changes in reactor structural materials caused by irradiation and thermal aging. In this work, the effect of low dose rate (approximately 2 × 10⁻⁸ dpa/s) irradiation at 380–410°C and long term thermal aging at 371°C on the properties of 20% cold worked 304 stainless steel, 420 stainless steel, Inconel X750, 304/308 stainless weld material, and 17-4 PH steel are evaluated. Doses of up to 6.8 dpa and thermal aging to 2994 days did not significantly affect the density of these alloys. The strength of 304 SS, X750, 17-4 PH, and 304/308 weld material increased with irradiation. In contrast, the strength of 420 stainless steel decreased with irradiation. Irradiation decreased the impact energy in both Inconel X750 and 17-4 PH steel. Thermal aging decreased the impact energy in 17-4 PH steel and increased the impact energy in Inconel X750. Tensile property comparisons of 304 SURV samples with 304 samples irradiated in EBR-II at a higher dose rate show that the higher dose rate samples had greater increases in strength and greater losses in ductility.     

The effect of dose rate on the response of austenitic stainless steels to neutron radiation

Depending on reactor design and component location, austenitic stainless steels may experience significantly different irradiation dose rates in the same reactor. Understanding the effect of dose rate on radiation performance is important to predicting component lifetime. This study examined the effect of dose rate on swelling, grain boundary segregation, and tensile properties in austenitic stainless steels through the examination of components retrieved from the Experimental Breeder Reactor-II (EBR-II) following its shutdown. Annealed 304 stainless steel, stress-relieved 304 stainless steel, 12% cold-worked 316 stainless steel, and 20% cold-worked 316 stainless steel were irradiated over a dose range of 1–56 dpa at temperatures from 371 to 440 °C and dose rates from 0.5 to 5.8 × 10⁻⁷ dpa/s. Density and tensile properties were measured for 304 and 316 stainless steel. Changes in grain boundary composition were examined for 304 stainless steel. Swelling appears to increase at lower dose rates in both 304 and 316 stainless steel, although the effect was not always statistically significant. Grain boundary segregation also appears to increase at lower dose rate in 304 stainless steel. For the range of dose rates examined, no measurable dose rate effect on tensile properties was noted for any of the steels.     

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.     

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.     

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.     

Residual stress behavior in fast neutron irradiated SA AISI 304l stainless steel cylindrical tubing

Residual stress measurements were made on solution-annealed (SA) AISI 304L stainless steel (SS) irradiated in EBR-II over a temperature range from 402 to 524°C by axially slitting short sections of tubing. The data were analyzed by using SA AISI 304 SS physical properties and SA AISI 304L SS swelling and irradiation creep empirical equations to calculate the slit width change (δ) versus fluence (φt) curve. At temperatures equal to and above 445°C, δ versus φt calculations indicate that the stress effect on swelling is sufficiently large to reduce the swelling rate temperature gradient, and consequently the on-power stress gradient, to zero. This behavior is confirmed by void volume gradient measurements. At lower temperatures, δ versus φt calculations indicate that stress affected swelling is smaller and does not relax the swelling rate temperature gradient. Void volume gradient measurements confirm the presence of a swelling gradient. Calculations of the δ versus φt curve were made with four different empirical swelling equation fluence dependencies, and the best agreement with the δ versus φt data was obtained with a power form type swelling equation. The equations fit the immersion density data ($\text{ΔV}\text{V}{}_0\text{versus}\text{φt}$) within experimental scatter, but predict significantly different δ versus φt behavior. These results show that the slit tube results are very sensitive to the empirical swelling equation form.     

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.     

In-reactor investigations of the creep of structural materials

Conclusions 1. A series of in-reactor tests was performed on a sample used to study radiation creep in 00X16H15M3B steel, XHM1 chrome-nickel alloy, the zirconium based alloys é110 and é635, and the vanadium-based alloy BTX8. The radiation creep modulus (in units of Pa⁻¹·(displacements/atom)⁻¹ equals 1.7·10⁻¹¹ for 00X16H15M3B steel, 4.6·10⁻¹¹ for XHM alloy with fluence up to 2.3·10²⁰ cm⁻² and 1.6·10⁻¹¹ for a fluence above 1·10²¹ cm⁻², (4.6–4.9)·10⁻¹¹ for é110 alloy, and 1.8·10⁻¹¹ for é635 alloy. For the alloy BTX8, at stresses below half the yield point and t=450°C, the modulus equals 3.3·10⁻¹² Pa⁻¹·(displacements/atom)⁻¹. At a higher stress, the deformation rate of the alloy increases progressively. 2. In the investigation of the temperature dependence of in-reactor creep of the alloy é110, it was found that at 350–370°C and higher, the thermal creep makes the predominant contribution to deformation. In the experimental range 370–455°C, the thermal activation energy of in-reactor creep was determined to be 36 ± 8 kcal/(g·atom). At temperatures below 350°C the creep of the alloy é110 is a temperature-independent radiation-stimulated process. 3. In the case of tests of zirconium alloys, a previously unobserved phenomenon of periodic rapid deformation of the material against the background of creep at stresses even well below the yield point of the irradiated material was discovered. The effect was manifested at a temperature of about 230°C. As the temperature increases up to 290°C and higher, no plastic movements are observed. Translated from Atomnaya énergiya, Vol. 80, No. 5, pp. 386–391, May, 1996.     

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.     

Residual stress measurements in irradiated solution-annealed type 304 stainless steel tubing

Immersion density and residual stress measurements were made on solution-annealed type 304 stainless steel capsule tubing irradiated up to fluence levels of 9 × 10²² n/cm² (E > 0.1 MeV). The measured residual stress is dependent on the competition between differential swelling which builds up differential stresses, and irradiation creep which relaxes these stresses. The measurements were analyzed using a bilinear swelling equation formulated with swelling data from the same heat of material. The temperatures and fluence levels of the swelling and slit tube data were each calculated with the same computer code. At high fluence, when swelling was in the steady-state region, the effective irradiation creep rate increased by a factor of about three. Further analysis was made assuming that stress-enhanced swelling and swelling-enhanced irradiation creep were the enhanced relaxation mechanisms.     

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.