Kinetics of carbon transfer governing compatibility between uranium carbide and stainless steel bonded with sodium

The purpose of the present study was undertaken to provide information on the kinetics of carbon transfer governing the degree of compatibility between carbide fuel and austenitic stainless steel, by comparing the experimental data with calculated results. Hyperstoichiometric uranium carbide containing ${}^{\mathrm{14C}}\text{(6.0 wt\%)}$ and stainless steel (AISI 304) specimens were enclosed in a nickel capsule together with sodium, and then heated to 750°C according to the pre-determined schedule. With respect to the degree of carburization of the stainless steel using the radioactive carbide, the activity in the steel decreased exponentially with depth and increased with heating time. The kinetics of carbon transfer was well interpreted by assuming that carbon atoms leached out into sodium from $\text{UC}{}_2$ platelets in the carbide and then diffused into the stainless steel along its grain boundary.     

Corrosion of various ferritic steels in an isothermal sodium loop system

Ferritic chromium-molybdenum steels with chromium contents of 1 wt% up to 12 wt% have been exposed for 8370 h to flowing sodium at 550°C. The oxygen content of the sodium was 6–7 ppm by weight. Weight measurements, carbon analyses and metallographic examinations were carried out. The low chromium steels show weight loss and decarburisation. The high chromium steels show weight gain and carburisation. The crossover point is at about 5 wt% Cr. The composition at the utmost surface (<10 μm) of the various steels tends to about 8 wt% chromium, about 2 wt% nickel and 0.02–0.09 wt% carbon. Sodium chromite crystals were present on the steels with a chromium content of 5 wt% or more. At the exposed surface of the $\text{2}\text{1}\text{4}$ wt% chromium steel sodium chromite crystals were found locally.     

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.     

Carbon and nitrogen transfer in Fe-9 Cr-Mo ferritic steels exposed to a sodium environment

Kinetics of carburization/decarburization of five commercial and two high-purity Fe-9 Cr-1 to 2.5 Mo feiritic steels have been studied in a sodium environment at temperatures between 773 and 973 K. Carbon concentration-distance profiles were obtained as a function of sodium-exposure time, temperature, and carbon in sodium and the carburization/ decarburization rate constants were evaluated. The results show that the Fe-9 Cr-Mo steels are more resistant to carbon transfer than the low-alloy Fe-$\text{2}\text{1}\text{4}$ Cr-1 Mo steel. The conditions of temperature and carbon concentration in sodium for carburization or decarburization of Fe—9 Cr—Mo steels are quite similar to those for stainless steels. However, the extent of carbon transfer in Fe-9 Cr—Mo steels is lower than that of the stainless steels. The composition and carbide structure of the steel had a significant effect on the carburization/decarburization behavior. Fe-9 Cr-Mo steels that decarburize to very low carbon concentrations either contain M₂X phase or have M₆C as the only stable carbide.     

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.     

Elevated-temperature tensile properties of irradiated 214 Cr-1 Mo steel

The effect of neutron irradiation on the tensile properties of normalized-and-tempered $\text{2}\text{1}\text{4}$ Cr-1 Mo steel was determined for specimens irradiated in Experimental Breeder Reactor II (EBR-II) at 390 to 550°C. Two types of unirradiated control specimens were tested: as-heat-treated specimens and as-heat-treated specimens aged for 5000 h at the irradiation temperatures. Irradiation to approximately 9 dpa at 390° C increased the strength and decreased the ductility compared to the control specimens. Softening occurred in samples irradiated and tested at temperatures of 450, 500, and 550 °C; the amount of softening increased with increasing temperature. The tensile results were explained in terms of the displacement damage caused by the irradiation and changes in carbide precipitates that occur during elevated-temperature exposure.     

Compatibility between several heat resistant alloys and sintered Li2O in static helium gas environment

The reaction of sintered Li₂O discs with several commercial heat resistant alloys has been investigated under the conditions of 3.3 × 10⁴Pa ($\text{1}\text{3}\text{atm}$) static He gas atmosphere in the temperature range of 500 and 750° C. Reaction products were identified by X-ray diffraction analysis to be two phases of Li₅FeO₄ and LiCrO₂. The former was dominant below 650° C and the latter was dominant above 650° C. The activation energies of the reaction were determined by the Arrhenius plots for weight changes and sub-scale thickness measurements. The reactivity of the four Fe-Ni-Cr alloys decreased according to the order of Incoloy 800, 316 SS, Hastelloy X-R and Inconel 600. Grain boundary penetration was observed above 500° C for Incoloy 800, 550° C for 316 SS and 600° C for Inconel 600. There was no grain boundary penetration in Hastelloy X-R.     

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.     

Thermal conductivity of neutron-irradiated pyrolytic β-silicon carbide

Samples of pyrolytic β-silicon carbide deposited at 1400 °C (grain size $\text{～ 1 μ}\text{m}$) and at 1750 °C (grain size $\text{～ 3 μ}\text{m}$) were irradiated with fast neutrons to 2.7-7.7 × 10²¹ n/cm² ($\text{E > 0.18 MeV}$) at 550 °–1100 °C. Irradiation reduced the room-temperature thermal conductivity from $\text{～0.15}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ to $\text{～ 0.02}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ after irradiation at 550 °C and to $\text{～ 0.05}\text{cal/cm}\text{·}\text{sec}\text{· °}\text{C}$ for an irradiation temperature of 1100 °C. The thermal conductivity of unirradiated samples decreased with increasing measurement temperature, while that of the irradiated samples was much less temperature dependent. No difference in behaviour was found between the samples with $\text{～ 1 μ}\text{m}$ grain size and the samples with $\text{～ 3 μ}\text{m}$ grain size.     

The influence of helium distribution and grain size on high-temperature embrittlement of stainless steel

In order to study the high-temperature embrittlement related to helium from the viewpoint of the distribution of helium and the grain size, helium ($\text{7.5 × 10}{}^{\mathrm{?6}}$ atomic fraction) was injected into stainless steel treated in various ways, using a cyclotron. Helium retarded the tendency towards recrystallization of stainless steel, and the grain size of the specimen recrystallized after injection was clearly smaller. The high-temperature embrittlement related to helium in stainless steel was appreciable above 650°C, and could be classified in terms of the mean grain diameter and the distribution of helium.     

Irradiation induced creep experiments in the BR2 reactor using the resonant cavity method

A 16Cr16NiNb stainless steel (steel 1.4981) will be used as wrapper material for SNR 300. Therefore, some in-pile creep tests have been performed with this material in the temperature range 420–700°C. The main objective of this programme was to see, whether the creep rates of steel 1.4981 followed at low fluences ($\text{? 2.5 × 10}{}^{22}\text{n/cm}{}^2$, $\text{E > 0.1}\text{MeV}$) the same rules as for other austenitic stainless steels. The experiments were performed in the BR2 reactor at Mol/Belgium, using creep rigs which were developed and manufactured by CEN Grenoble. The creep strains were measured by the resonant cavity method. The paper describes the main characteristics of the creep capsules, and reports on the performance of those types of rigs. Finally, the experimental results are presented and discussed.     

Nickel ion bombardment of type 304 stainless steel: Comparison with fast reactor swelling data

Annealed Type 304 stainless steel containing 15 atomic ppm of helium has been bombarded with 5 MeV nickel ions at 525°C to 700°C. A pronounced swelling peak occurs at 625°C, compared to a swelling peak temperature of about 475°C in reactor. TEM measurements of void swelling at 625°C as a function of ion dose show a swelling of almost 40% at 124 dpa without evidence of saturation. Measurements of gross swelling of the ion-bombarded material by a new step-height method provide information that is in good agreement with TEM data, and can be extended to larger swellings. The step-height results indicate a swelling of over 90% at 290 dpa at 625°C. The ion-produced swelling agrees well with in-reactor data when the two are compared at the respective peak swelling temperatures, and the void concentrations and average void diameters are comparable for the two cases. The high ion dose results are used to guide extrapolation of reactor data to higher fluences, leading to the following predictions for swellings at the peak swelling temperature in reactor: 18% swelling at $\text{1× 10}{}^{23}\text{n/cm}{}^2$ (fast), 50% at $\text{2 × 10}{}^{23}$, and 80% at $\text{3 × 10}{}^{23}$.     

Effect of cyclotron-injected helium on void-formation characteristics of type 304 stainless steel

Samples of Type 304 stainless steel were injected with helium by cyclotron bombardment to concentrations ranging between $\text{1.1 × 10}{}^{\mathrm{?7}}$ and $\text{1 × 10}{}^{\mathrm{?4}}$ ppma. Following cyclotron injection, the samples were given a variety of heat treatments prior to insertion in EBR-II for irradiation at 450 °C to a total dose of $\text{1 × 10}{}^{21}\text{n/cm}{}^2$. Samples that were not heat treated or that were annealed at 650 °C following cyclotron injection formed few voids and dislocation loops after EBR-II irradiation. This behavior is apparently due to the precipitate clusters that were formed during the helium injection. These precipitates were analyzed by electron microscopic techniques and found to have spherically symmetric strain fields that were of interstitial character. Samples that were annealed at 760 °C following cyclotron injection formed a larger number density of both voids and dislocation loops than did the control sample after EBR-II irradiation. The void volume also exceeded that of the control. Clustering of the dislocation loop population near grain boundaries and precipitate particles was observed in the control and low helium concentration samples.     

The effect of cold working on creep rupture properties for helium-injected austenitic stainless steel

In order to study the effect of helium on the high temperature embrittlement of stainless steel, helium $\text{(7.5 × 10}{}^{\mathrm{?6}}\text{atomic fraction)}$ was injected into cold-worked stainless steels by using a cyclotron. At 650°C, it appeared that the reduction in creep-rupture strength due to helium was larger as cold-working was increased, but a loss of rupture elongation was less for a particular degree of cold-working. The 10% cold-worked material showed particularly good creep-rupture properties in the presence of helium. The loss of ductility was more pronounced in the creep test than in the tensile test.     

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 minor constituents on swelling in stainless steel

The small quantities of solute interstitial elements in stainless steel (C, N and possibly Si) reduce the swelling under neutron irradiation ($\text{～ 2 × 10}{}^{22}\text{neutrons/cm}{}^2$) by more than an order of magnitude between 500 and 600° C over high purity material. The solute interstitials reduce both the numbers and sizes of irradiation-caused voids. Current swelling models ignore — of necessity — this gross effect. Several possible mechanisms are suggested to account for the effect.     

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).     

The effect of carbon on 2.25 Cr-1 Mo steel: (I). Microstructure and tensile properties

The microstucture, hardness, and the tensile properties of 2.25 Cr-1 Mo steel with 0.009, 0.030, 0.120, and 0.135 wt % C were determined on steels in the annealed (furnace-cooled from 927°C), normalized (air-cooled from 927°C), and normalized-and-tempered conditions. As annealed, the microstructure was primarily proeutectoid ferrite with spherical carbides and pearlite, the amounts increasing with increasing carbon content. During normalization ($\text{7}\text{8}$ in rods or 1 in plates were heat-treated), granular bainite formed: 1 to 2 and 15 to 20% bainite (remainder proeutectoid ferrite) for the 0.009 and 0.030 wt % C steels, respectively; the 0.120 and 0.135 wt % C steels were entirely bainite. On tempering, carbides precipitated. In all heat-treated conditions, there was little difference in the room temperature hardness of the 0.009 and 0.030 wt % C steels and between the 0.120 and 0.135 wt % C steels. Tensile tests from 25 to 565°C indicated that strength depends on microstructure, which is determined by carbon content.     

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.