Mechanism of dynamic strain aging and characterization of its effect on the low-cycle fatigue behavior in type 316L stainless steel

Low-cycle fatigue tests were carried out in air in a wide temperature range from 20 to 650 °C with strain rates of 3.2 × 10⁻⁵–1 × 10⁻² s⁻¹ for type 316L stainless steel to investigate dynamic strain aging (DSA) effect on the fatigue resistance. The regime of DSA was evaluated using the anomalies associated with DSA and was in the temperature range of 250–550 °C at a strain rate of 1 × 10⁻⁴ s⁻¹, in 250–600 °C at 1 × 10⁻³ s⁻¹, and in 250–650 °C at 1 × 10⁻² s⁻¹. The activation energies for each type of serration were about 0.57–0.74 times those for lattice diffusion indicating that a mechanism other than lattice diffusion is involved. It seems to be reasonable to infer that DSA is caused by the pipe diffusion of solute atoms through the dislocation core. Dynamic strain aging reduced the crack initiation and propagation life by way of multiple crack initiation, which comes from the DSA-induced inhomogeneity of deformation, and rapid crack propagation due to the DSA-induced hardening, respectively.     

Corrosion studies in support of lead-bismuth cooled FBRs

The performance of structural materials in lead or lead-bismuth eutectic (LBE) systems is evaluated. The materials evaluated included several US steels (austenitic steel [316L], carbon steels [F-22, Fe-Si], ferritic/martensitic steels [HT-9 and 410]), and several experimental Fe-Si-Cr alloys that were expected to demonstrate corrosion resistance. The materials were exposed in either a dynamic corrosion cell for periods from 100 to 1,000 h at temperatures of 400, 500, 600 and 700°C, depending on material and exposure location. Weight change and optical SEM or X-ray analysis of the specimen were used to characterize oxide film thickness, corrosion depth, microstructure, and composition changes. The tests conducted with stainless steels (410, 316L and HT-9) produced mass transfer of elements (e.g., Ni and Cr) into the LBE, resulting in degradation of the material. With Fe-Si alloys a Si rich layer (as SiO₂) is formed on the surface during exposure to LBE from the selective dissolution of Fe.     

Influence of the surface conditions on permeation in the deuterium–MANET system

The condition of the surfaces is of crucial importance for the deuterium permeation through materials. In this work a study of the surface constants for the adsorption (σk₁) and release (σk₂) of deuterium under different surface conditions on the martensitic steel DIN 1.4914 (MANET) has been carried out. The growth of an oxide surface layer (Cr₂O₃) of about 25–30 nm in a MANET sample, heat treated in an oxidizing environment, compared to the bare MANET that have a ‘natural' oxide of about 5 nm has provoked a reduction of both the permeation rate and the recombination coefficient (about 3 orders of magnitude). In addition, the permeation governing process has changed from diffusion-limited to surface-limited. The measurements of the permeation rate of deuterium were performed by a gas-phase permeation technique over the temperature range 574–746 K and for deuterium driving pressures in the range from 3 to 10⁵ Pa.     

Helium dilution effect on hydrogen permeation in 316L stainless steel and nickel-base heat-resistant alloys

Effects of inert-gas dilution on hydrogen permeation have been investigated in 316L stainless steel, Inconel 600, Inconel 750, Nimonic 80A and Hastelloy X at 1173 K and 1073 K, by employing a gas-flow system. We used gas mixtures of hydrogen and helium, whose hydrogen concentration ranged from 10^?5 to 10^?1. For the steady-state permeation, the dilution of hydrogen caused no anomalous effects and the permeation rate conformed to Sieverts' law. However, for the transient state, the hydrogen permeation was retarded by the dilution with helium. The retardation effect is discussed in terms of an adsorption model and explained by a decrease in sticking probability at the alloy surface with the dissociative adsorption of hydrogen.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

Corrosion behaviour of aluminized martensitic and austenitic steels in liquid Pb-Bi

The Pb-Bi liquid alloy is under consideration as a spallation target material in the hybrid systems due to its suitable nuclear and physical properties. In order to limit the risks of corrosion of the structural elements in contact with the liquid Pb-Bi, protection by means of aluminized coatings was investigated for 316L austenitic steel and T91 martensitic steel. For both steels, no damages were observed after immersions in static Pb-Bi up to 500 °C for low oxygen concentrations and long durations. However, at 600 °C in the same conditions, a non-uniform degradation of the coatings was observed. Only coated 316L was tested in dynamic conditions. The results were generally satisfying for temperatures from 350 to 600 °C and for fluid velocities up to 2.3 m s⁻¹. However, in both the IPPE loops and the CICLAD device, some localized damage of the coatings, attributed to erosion, was observed.     

Removal of polonium from stainless steel surface contaminated by neutron irradiated lead-bismuth eutectic

Lead-bismuth eutectic (LBE) has good characteristics for the coolant and/or target of various nuclear systems, but it also has a problem of polonium contamination. In the study, baking experiment was performed to remove polonium contamination on type 316 stainless steel plate that was originated from neutron irradiated LBE. The contaminated type 316 stainless steel plate was baked in a vacuum condition at various temperatures from 200°C to 600°C. In the previous preliminary study, the effect of short time baking was investigated. In the study, the effect of long time baking was investigated. The detail of the experimental method was also described. The result of long time baking experiment showed that the baking method was effective for removal of polonium from stainless steel surface contaminated by neutron irradiated LBE, if the baking was performed at 500°C and higher in a vacuum condition. The obtained result was consistent to the previous preliminary study.     

Observation of micro-cavities in nickel during He and D irradiation and post-irradiation annealing

Changes in sizes and morphology of small cavities in nickel irradiated by 25 keV helium ions and 20 keV deuterons were investigated during irradiation and on annealing after irradiation by means of transmission electron microscopy. In the early stage of He⁺ irradiations at 600 and 700° C, roundish cubes appeared, gradually changed to octahedra. and, then, by the truncation of apexes, finally reached cubo-octahedra. Nucleation and growth behavior of cubic cavities in D⁺ irradiated nickel was different from the case of He⁺ irradiation. On annealing of the He⁺ irradiated specimen, only octahedral cavities showed marked growth, finally changing to roundish cubes at 750° C. Cavities of roundish cubes and cubo-octahedra did not grow nor change their shapes remarkably by the annealing. The cubic cavities formed by D⁺ irradiation at 360° C showed gradual shrinkage on annealing at 600° C and disappeared at 625° C. The changes of cavities during irradiation and on annealing were interpreted by the effect of the internal gas pressure.     

Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 °C and 500 °C.During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 °C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 °C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.     

Defect and void evolution in oxide dispersion strengthened ferritic steels under 3.2 MeV Fe ion irradiation with simultaneous helium injection

In an attempt to explore the potential of oxide dispersion strengthened (ODS) ferritic steels for fission and fusion structural materials applications, a set of ODS steels with varying oxide particle dispersion were irradiated at 650°C, using 3.2 MeV Fe⁺ and 330 keV He⁺ ions simultaneously. The void formation mechanisms in these ODS steels were studied by juxtaposing the response of a 9Cr–2WVTa ferritic/martensitic steel and solution annealed AISI 316LN austenitic stainless steel under the same irradiation conditions. The results showed that void formation was suppressed progressively by introducing and retaining a higher dislocation density and finer precipitate particles. Theoretical analyses suggest that the delayed onset of void formation in ODS steels stems from the enhanced point defect recombination in the high density dislocation microstructure, lower dislocation bias due to oxide particle pinning, and a very fine dispersion of helium bubbles caused by trapping helium atoms at the particle–matrix interfaces.     

Tracer diffusion of Co into SUS-304

A piece of stainless steel 304 was thermal annealed either in a ⁶⁰Co solution at 280°C, or in a He atmosphere at 300 or 600°C after being smeared with a ⁶⁰Co solution. The depth profile of ⁶⁰Co in SUS was determined with these samples, together with those sampled from the water cleaning system in the Japan Power Demonstration Reactor and those exposed to a high-temperature water ( ˜ 280°C) circulating OWL-loop installed in the Japan Material Testing Reactor, in order to determine the apparent values of the volume and grain boundary diffusion constants. The results have been compared with those of the previous measurements on Ni, Cr, and Fe above 500°C. A consistency has been noticed between both results at 600°C, while the present values at 280 and 300°C are much larger than those expected from the extrapolation of the previous measurements. Although this inconsistency is ascribed to the contribution of the dislocation pipe diffusion at such a low temperature, an attempt has been made to apply the present values to estimate how deep ⁶⁰Co would penetrate SUS—one of the main constructing materials of the primary cooling system—in the life time of a power reactor.     

Segregation effects on the corrosion behaviour of a phosphorus-doped AISI type 304L stainless steel

The effect of grain boundary (GB) segregation on intergranular stress corrosion cracking (IGSCC) in hot water environments at 150°C and 250°C was studied in a P-doped AISI type 304L stainless steel. The extent of segregation was measured by an exposure test in boiling 5 N HNO₃ + 8g/L K₂Cr₂O₇ solution as well as by a potentiostatic etch test at +1325 mV (SHE) in 5 N H₂SO₄ solution. Although GB segregation was detected in all the aged specimens, IGSCC was shown by only the specimens aged for 550°C/1000 h. The results suggest that it is the GB chromium depletion, rather than the segregation of phosphorus at the GBs, that controls IGSCC of stainless steels in the hot water environments studied.     

Corrosion of stainless steels in lead-bismuth eutectic up to 600 °C

An experimental program has been carried out to understand the differences in the corrosion behaviour between different stainless steels: the austenitic steels 304L and 316L, the martensitic steels F82Hmod, T91 and EM10, and the low alloy steel P22. The influence of oxygen level in Pb-Bi, temperature and exposure time is studied. At 600 °C, the martensitic steels and the P22 steel exhibit thick oxide scales that grow with time, following a linear law for the wet environment and a parabolic law for the dry one. The austenitic stainless steels show a better corrosion behaviour, especially AISI 304L. Under reducing conditions, the steels exhibit dissolution, more severe for the austenitic stainless steels. At 450 °C, all the materials show an acceptable behaviour provided a sufficient oxygen level in the Pb-Bi. At reducing conditions, the martensitic steels and the P22 steel have a good corrosion resistance, while the austenitic steels exhibit already dissolution at the longer exposures.     

Deuterium permeation of amorphous alumina coating on 316L prepared by MOCVD

The deuterium permeation behavior of the alumina coating on 316L stainless steel prepared by metal organic chemical vapor deposition (MOCVD) was investigated. The alumina coating was also characterized by X-ray photoelectron spectroscopy (XPS), X-ray diffraction (XRD) and scanning electron microscope (SEM). It was found that the as-prepared coating consisted of amorphous alumina. This alumina coating had a dense, crack-free and homogeneous morphology. Although the alumina coating was amorphous, effective suppression of deuterium permeation was demonstrated. The deuterium permeability of the alumina coating was 51–60 times less than that of the 316L stainless steel and 153–335 times less than that of the referred low activation martensitic steels at 860–960 K.     

SiC coating as hydrogen permeation reduction and oxidation resistance for nuclear fuel cladding

Silicon carbide (SiC) coating is one of the countermeasures for the prevention of oxidation and hydrogen embrittlement of fuel claddings because SiC has high resistance of oxidation and hydrogen permeation. Hydrogen permeation and oxidation experiments for the cladding materials with SiC coatings were conducted in unirradiated conditions. The sputtering method was employed to make SiC coatings. In the hydrogen permeation experiment, 316 type of stainless steel (SS316) was used as a base material of the coating. SS316 with SiC coatings showed hydrogen permeation reduction by one order of magnitude. In the oxidation experiments, Zircaloy 4 (Zry-4) and SS316 were used as base materials of the coatings. The weight gain of the Zry-4 specimens with a SiC coating decreased by about one-fifth compared to the uncoated ones at 750 °C and 1200°C. This phenomenon was observed for SS316 at 750 °C as well. The peel-off of the coating was observed in several experiments, and it is considered that the peel-off was caused by the difference of the thermal expansions between coatings and base materials. Thicker coatings showed better oxidation resistance, but thinner coatings showed more tolerance of peel-off.     

Results of steel corrosion tests in flowing liquid Pb/Bi at 420-600 °C after 2000 h

Corrosion tests were carried out on austenitic AISI 316L and 1.4970 steels and on MANET steel up to 2000 h of exposure to flowing (up to 2 m/s) Pb/Bi. The concentration of oxygen in the liquid alloy was controlled at 10⁻⁶ wt%. Specimens consisted of tube and rod sections in original state and after alloying of Al into the surface. After 2000 h of exposure at 420 and 550 °C the specimen surfaces were covered with an intact oxide layer which provided a good protection against corrosion attack of the liquid Pb/Bi alloy. After the same time corrosion attack at 600 °C was severe at the original AISI 316L steel specimens. The alloyed specimens containing FeAl on the surface of the alloyed layer still maintained an intact oxide layer with good corrosion protection up to 600 °C.     

Medium temperature carbon dioxide gas turbine reactor

A carbon dioxide (CO₂) gas turbine reactor with a partial pre-cooling cycle attains comparable cycle efficiencies of 45.8% at medium temperature of 650 °C and pressure of 7 MPa with a typical helium (He) gas turbine reactor of GT-MHR (47.7%) at high temperature of 850 °C. This higher efficiency is ascribed to: reduced compression work around the critical point of CO₂; and consideration of variation in CO₂ specific heat at constant pressure, C_p, with pressure and temperature into cycle configuration. Lowering temperature to 650 °C provides flexibility in choosing materials and eases maintenance through the lower diffusion leak rate of fission products from coated particle fuel by about two orders of magnitude. At medium temperature of 650 °C, less expensive corrosion resistant materials such as type 316 stainless steel are applicable and their performance in CO₂ have been proven during extensive operation in AGRs. In the previous study, the CO₂ cycle gas turbomachinery weight was estimated to be about one-fifth compared with He cycles. The proposed medium temperature CO₂ gas turbine reactor is expected to be an alternative solution to current high-temperature He gas turbine reactors.     

Effect of chemical potential of carbon on phase transformation and corrosion of JLF-1 steel in a static lithium

The corrosion and the phase transformation of RAFM (Reduced Activation Ferritic/Martensitic) steel, JLF-1(Fe-9Cr-2W-0.1C), in static lithium (Li) were investigated. The specimens were exposed to static Li at 600 °C for 250 h. The carbon potential in Li was controlled by the carbide formation on inner surface of the crucible, which was made of Mo and Nb. These materials were expected to form stable carbide, and worked as carbon trap in Li, which resulted in the decrease in the carbon potential of Li. The effect of low carbon potential on the corrosion and the phase transformation was characterized by comparison with the test performed in the crucible made of Fe-9Cr and SUS316L (18Cr-12Ni), which has no effect as carbon traps. The low carbon potential caused by Nb and Mo caused the dissolution of carbon from the surface of JLF-1 specimen to Li. The depth of the phase transformation observed after the test in Nb crucible was deeper than that tested in Mo crucible. This was because the stability of the Nb carbide was larger than that of Mo carbide causing larger driving force for the dissolution of carbides from the surface.     

Behavior of steels in flowing liquid PbBi eutectic alloy at 420-600 °C after 4000-7200 h

This paper presents the results of steel exposure up to 7200 h in flowing LBE at elevated temperatures and is a follow-up paper of that with results of an exposure of up to 2000 h. The examined AISI 316 L, 1.4970 austenitic and MANET 10Cr martensitic steels are suitable as a structural material in LBE (liquid eutectic Pb₄₅Bi₅₅) up to 550 °C, if 10⁻⁶ wt% of oxygen is dissolved in the LBE. The martensitic steel develops a thick magnetite and spinel layer while the austenites have thin spinel surface layers at 420 °C and thick oxide scales like the martensitic steel at 550 °C. The oxide scales protect the steels from dissolution attack by LBE during the whole test period of 7200 h. Oxide scales that spall off are replaced by new protective ones. At 600 °C severe attack occurs already after 2000 and 4000 h of exposure. Steels with 8-15 wt% Al alloyed into the surface suffer no corrosion attack at all experimental temperatures and exposure times.     

The formation of tritium permeation barriers by CVD

The effectiveness as permeation barriers of the following CVD coatings have been investigated: TiC (1 to 2 μm in thickness); a bi-layer of TiN on TiC (3 μm total thickness) and CVD A1₂O₃ on a TiN/TiC bi-layer. The substrate materials were TZM (a Mo alloy) and 316L stainless steel in the form of discs of diameter 48 mm and thickness 0.1 or 1 mm. Permeation measurements were performed in the temperature range 515–742 K using deuterium at pressures in the range 1–50 kPa. CVD layers were shown to form reasonably effective permeation barriers. At a temperature of 673 K TiC is around 6000 times less permeable to deuterium than 316L stainless steel.