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.     

Effects of Temperature,Strain Rate,and Specimen Orientation on Localized Plastic Deformation of Irradiated Zircaloy-2

Both transverse and longitudinal Zircaloy-2 specimens irradiated up to 1.2 × 10²⁰ n/cm² (E> 1 MeV) were tested in tension with strain rates ranging 1.1 × 10^-4~1.1 × 10^-2 s^-1 in the temperature range 200~400°C. Detailed observations of the specimen wall surface and microstructure were also made on samples deformed to various amounts of plastic strain, with a projector and an optical microscope.

It was found that localized plastic deformation bands occurred in the temperature range approximately 280~330°C during straining to the ultimate tensile stress. Results also showed that the strain rate dependence of tensile properties, particularly the strain to the ultimate tensile stress, was associated with changes in the number and width of the localized deformation band with strain rates at a temperature of 300°C at which localized bands occurred. From a break of the straight line tracing the true stress-true plastic strain relationship, it was established that the onset stress and strain of the localized deformation band could be estimated.

The effect of specimen orientation on localized deformation band was also discussed on the basis of differences in the onset stress and strain between the transverse and longitudinal specimens.     

Elevated-temperature tensile properties of irradiated 9 Cr-1 MoVNb steel

Normalized-and-tempered 9 Cr-1 MoVNb steel tensile specimens were irradiated in the Experimental Breeder Reactor-11 (EBR-11) at 390, 450, 500, and 550°C to ～2.1 and 2.5 × 10²⁶ neutrons/m² (> 0.1 MeV), which produced displacement damage levels of ～10 and 12 dpa, respectively. Tensile tests were conducted at the irradiation temperature and at room temperature. In addition to the irradiated specimens, as-heat-treated specimens and as-heat-treated specimens thermally aged at the irradiation for 5000 h were also tested.Thermal aging had no effect on the unirradiated tensile properties. Irradiation at 390°C increased the 0.2% yield stress and the ultimate tensile strength above those of the unirradiated control specimens. The ductility decreased slightly. After irradiation at 450, 500, and 550°C, the tensile properties were essentially the same as the unirradiated values. The hardening at 390°C was attributed to the dislocation and precipitate structure formed during the irradiation. The lack of hardening at 450°C and higher correlates with an absence of an irradiation-induced damage structure.     

Effect of Neutron Irradiation on Deformation Behavior of Zirconium

The effect of neutron irradiation on the tensile deformation behavior of zirconium was examined at room temperature at various strain rates ranging of 2.2×10^?4~2.2× 10^?2 sec^?1. The microstructure of the deformed specimens was observed by transmission electron microscopy. It was established that neutron irradiation diminishes the uniform elongation and the strain hardening rate, and hastens the onset of plastic instability. These phenomena are attributed to inhomogeneous deformation in the dislocation channels in the irradiated and deformed zirconium.

From the relation between strain rate and tensile properties (yield stress, ultimate tensile stress, uniform elongation and strain hardening rate), it was established that in unirradiated zirconium deformation is controlled by slip at strain rates below 6×10^?3 sec^?1, while above this threshold, twinning as well as slip contribute to deformation.

Neutron irradiation markedly inhibits deformation twinning in zirconium at room temperature. At 77 K, on the other hand, deformation by twinning is more prominent in irradiated specimens. The mechanism of twinning inhibition due to neutron irradiation is discussed.     

High temperature deformation and fracture behaviour of 316L stainless steel under high strain rate loading

The high temperature deformation and fracture behaviour of 316L stainless steel under high strain rate loading conditions are investigated by means of a split Hopkinson pressure bar. Impact tests are performed at strain rates ranging from 1 × 10³ s^?1 to 5 × 10³ s^?1 and temperatures between 25 °C and 800 °C. The experimental results indicate that the flow response and fracture characteristics of 316L stainless steel are significantly dependent on the strain rate and temperature. The fracture analysis results indicate that the 316L specimens fail predominantly as the result of intensive localised shearing. Furthermore, it is shown that the flow localisation effect leads to the formation of adiabatic shear bands. The fracture surfaces of the deformed 316L specimens are characterised by a dimple-like structure with knobby features. The knobby features are thought to be the result of a rise in the local temperature to a value greater than the melting point.     

The effects of warm work on some mechanical and fabrication properties of uranium

Uranium plates have been warm rolled at 250°C to reductions of up to 90%. Subsequent tensile tests at ambient temperature have shown a progressive increase in fracture stress and ductility with increased deformation. No evidence of anistropy in tensile properties was observed for material unidirectionally rolled to 80% reduction at 250°C. Cupping tests have been performed on 0.048 in (1.2 mm) uranium sheet warm rolled up to 90% reduction. Tests were made at R.T. and 200° C on both the as-rolled and the annealed conditions. Values of $\text{R}$, the average width strain to thickness strain ratio, were determined during the tensile tests. No correlation was observed between the determined limiting drawing ratios and $\text{R}$ values. However, good cupping behaviour was shown to be a characteristic of material with a high tensile ductility.     

Mechanical properties of molybdenum neutron-irradiated at a high temperature

Molybdenum specimens prepared by two processes, powder-metallurgy (PM) and electron-beam melting (EB), were irradiated to a fast neutron fluence of 2.74 × 10²⁴n/m² (E_n? 1 MeV) at about 600°C (873 K), and their mechanical properties were studied in detail. It was shown that the degree of irradiation embrittlement in EB-Mo was smaller than that in PM-Mo, which might be caused by stronger grain-boundaries and probably smaller irradiation-hardening in the former. From the relation between the recovery of ductility and microstructural changes in post-irradiation annealed PM-Mo at 800 (1073 K), 1000 (1273 K) and 1200°C (1473 K), it was concluded that the recovery resulted from a decrease of irradiation hardening due to a rearrangement and a disappearance of depleted-zones, dislocation-loops and voids in order with increasing annealing temperature. An anomalous mode of fracture was observed in as-irradiated specimens, which consisted of inhomogeneous deformation, then brittle fracture not at the center but at the root of the deformation neck. This mode was observed in a narrow temperature range near the DBTT. A possible mechanism is discussed.     

The irradiation creep of nickel and AISI 321 stainless steel during 4 MeV proton bombardment

An apparatus has been developed to study the creep of thin metal specimens under tensile stress during bombardment by 4 MeV protons from the Harwell Van de Graaff Accelerator. The specimen is held in a helium atmosphere and the proton beam reaches it through a thin metal window at the end of the accelerator beam line. The proton beam passes through the thin (25 μm) specimen, losing ~1.5 MeV in the process (most of which contributes to heating the specimen) and creating almost uniform radiation damage at the rate of $\text{(1–10) × 10}{}^{\mathrm{?7}}$ displacements per atom per second (dpa s^?1). The specimen temperature is monitored by infra-red pyrometry and controlled to ± 0.2°C by additional DC heating via the infra-red pyrometer output to compensate for ion beam fluctuations. The irradiation creep strain of the specimen is continuously measured with a sensitivity of $\text{5 × 10}{}^{\mathrm{?6}}$ by a linear variable differential transformer. Irradiation times up to about 100h with reasonable beam stability are possible. Results are presented of the irradiation creep behaviour of pure Ni and both solution treated and cold-worked AISI 321 stainless steel bombarded in the temperature range 400–600°C under tensile stresses in the range 20–250 MPa.     

Effect of Neutron Irradiation on Mechanical Properties of Al-Li Alloys

An increase in yield stress at room temperature was observed in Al-0.6W/0 Li alloy irradiated to thermal neutron doses of 2.9 × 10¹⁹ to 7.2 × 10¹⁹ cm^?2. The hardening of as-irradiated specimens is accompanied with yield point followed by jerky yield-elongation in the stress-strain curve. The radiation hardening could not be annealed out by heating for 30 min at temperatures up to 350°C, whereas the yield-elongation disappeared gradually with increasing heating temperature in the l mm diam. specimens; with the 2 mm diam. specimens the yield-elongation still remained even after post-irradiation heating for 30 min at 350°C. Strengthening accompanied by jerky yield-elongation is considered to be due to He atom clusters precipitated along the dislocation. The hardening observed in the specimens heat-treated after irradiation at temperatures above 250°C is caused by randomly distributed gas bubbles.

In heavily cold-worked Al-0.6%W/o Li specimens, recovery of work hardening occurred during neutron irradiation to 4.2 × 10¹⁹ cm^?2. Hardening due to gas bubbles was also observed in the cold-worked specimens. In Al-2.7W/0 Li alloy, an increase in yield stress took place in the specimens irradiated to 4.2 × 10¹⁹ cm^?2 and heated for 30 min at temperatures of 155° to 260°C. The hardening is thought to be due to re-precipitation of β-phase resolved during the neutron irradiation.     

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.     

Self-Limiting Power Excursion Characteristics of Light Water Reactor, (IV)

A strain gage was used for the measurement, of fuel cladding strain generated during pulse operation tests on the Hitachi Training Reactor. In the analysis of the measured strain, two kinds of correction were called for: (1) the fadiation effect on the strain gage and lead wires, and (2) the temperature effect due to the lag of the gage filament temperature behind the true fuel cladding temperature. The experimental axial strain after the two corrections were applied was 781 × 10^?6 cm/cm for the hottest fuel rod in the pulse operation test with an inserted reactivity of 1.20%δk/k. This maximum strain corresponded to 2,169 kg/cm² of thermal stress and 111 cal/cm²·sec of heat flux. These results were obtained under the condition of maximum temperature in the fuel center of 1,200°C and a fuel cladding temperature of 140°C. When the axial strain was calculated with consideration given to the gap or contact conductance between the fuel and its cladding, a reasonable agreement was obtained between the calculation and the experimental results.     

Irradiation-creep in a materials testing reactor

Helical springs made from titanium, zirconium, Nimonic PE 16 alloy and two austenitic stainless steels namely Firth-Vickers FV 548 and a steel with a composition within the AISI 316 specification have been irradiated in the Dounreay Materials Testing Reactor, at temperatures between 85 °C and 100 °C and to neutron fluences up to about 4 × 10²⁴ m^?2 (> 1 MeV), whilst loaded in tension. Irradiation-creep was observed in all the materials studied and initial strain rates/unit stressneutron ranged from 1 to 1.9 × 10^?35 m⁴/N for Nimonic PE 16 and zirconium respectively. Data obtained from an earlier experiment are re-presented and compared with the present results.Springs which received no heat-treatment after coiling unwound during irradiation at rates which were independent of the supported loads. The phenomenon is attributed to the relaxation of internal stresses (produced during the manufacture of the springs) with an irradiation-creep constant which is an increasing function of prior cold-work.A mechanism of irradiation-creep is proposed which involves the re-arrangement of the dislocation network in a crystal as the dislocations climb by absorbing interstitials produced by irradiation.     

Effects of Dissolved Hydrogen and Strain Rate on IASCC Behavior in Highly Irradiated Stainless Steels

IASCC behavior in cold-worked SUS316 stainless steels irradiated to 35 dpa was examined using slow strain rate tensile testing at a strain rate of 6.7 × 10^?8/s in 320°C simulated PWR primary water while varying the dissolved hydrogen (DH) concentration from 0 to 2.8 ppm. The results were compared with those previously obtained at a higher strain rate using specimens of different sizes and with those of the previous interrupted experiment. The initiation and propagation of IASCC enhanced with increasing DH concentration and lower strain rate. The IASCC initiation stress decreased to almost half of the yield strength at high DH. Accompanying slow tensile tests in an argon gas environment showed that a lower strain rate did not change in the initiation stress that exceeded the yield strength, but enhanced the propagation of intergranular cracking.     

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.     

The effects of large concentrations of helium on the mechanical properties of neutron-irradiated stainless steel

Tensile and creep properties have been determined on specimens of type 316 stainless steel irradiated in the High Flux Isotope Reactor in the range 380 to 785°C. Irradiation of type 316 in this reactor partially simulates fusion reactor irradiation, with displacement damage levels up to 120 dpa and helium contents up to 6000 appm achieved in two years. Samples irradiated in the annealed condition to about 100 dpa and 4000 appm helium showed an increased yield strength between 350 and 600°C and, except at 350°C, a reduced ultimate tensile strength compared with values for the unirradiated material. Samples irradiated in the 20%-cold-worked condition showed decreases in both yield and ultimate tensile strengths at all test temperatures. The irradiated samples of both annealed and cold-worked material exhibited little strain hardening, and total elongations were small and became zero,for tests at 650° C. Tensile tests at 575°C and creep-rupture tests at 550°C showed strong effects of fluence on strength and ductility for helium contents above about 30 appm. Optical metallography showed extensive carbide precipitation at all temperatures and precipitation of a second phase, believed to be sigma, at the higher temperatures.     

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 effects of gas transport properties on blister formation in He+-implanted glass

Blister formation in He⁺-implanted glasses is correlated with the measured helium gas diffusivity. A scries of glasses with diffusivities from ～3 × 10^?7 to ～5 × 10^?12cm²sec^?1 was implanted under nearly identical conditions with 150 keV He⁺ ions at a flux of 15 μA cm^?2 and a nominal sample temperature of 110°C. Glasses with D less than ～1 × 10^?9cm²sec^?1 were fully blistered, whereas those with D greater than ～3 × 10^?8cm²sec^?1 showed no surface deformation. Glasses with diffusivities between ～3 × 10^?8 and ～1 × 10^?9cm²sec^?1 had local regions with low density coverage of relatively large blisters. The critical concentration of implanted helium, estimated by comparing experimental data with results from a simple theoretical model, is ^{～1 × 1019 cm?3}, consistent with high pressure solubility measurements. Reemission data at low fluence are qualitatively in agreement with analytical calculations. Implications for CTR technology are discussed.     

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.