M.L. Grossbeck

Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios

Abstract Effects of helium on mechanical properties of irradiated structural materials are reviewed. In particular, variations in response to the ratio of helium to displacement damage serve as the focus. Ductility in creep and tensile tests is emphasized. A variety of early work has led to the current concentration on helium effects for fusion reactor materials applications. A battery of techniques has been developed by which the helium to displacement ratio can be varied. Our main discussion is devoted to the techniques of spectral tailoring and isotopic alloying currently of interest for mixed-spectrum reactors. Theoretical models of physical mechanisms by which helium interacts with displacement damage have been developed in terms of hardening to dislocation motion and grain boundary cavitation. Austenitic stainless steels, ferritic/martensitic steels and vanadium alloys are considered. In each case, work at low strain rates, where the main problems may lie, at the helium to displacement ratios appropriate to fusion reactor materials is lacking. Recent experimental evidence suggests that both in-reactor and high helium results may differ substantially from post-irradiation or low helium results. It is suggested that work in these areas is especially needed.

An assessment of tensile, irradiation creep, creep rupture, and fatigue behavior in austenitic stainless steels with emphasis on spectral effects

Abstract A review of mechanical properties of austenitic stainless steels is made to assess their behavior in fusion reactors. Since the first walls of fusion devices are expected to range in temperature from 100 to over 500°C, behavior over a wide range of temperatures is reviewed. For tensile properties, the neutron spectrum has little effect on strength, but appears to influence plastic deformation and ductility. The effect is apparent at temperatures above 400°C where ductility begins to increase. Ductility appears to be higher for alloys irradiated in fast reactors where little helium is produced, but even for fast reactor irradiation, ductility begins to drop above 700°C. This behavior, which is enhanced by higher helium concentrations, is believed to be a result of helium embrittlement. Irradiation creep is very weakly dependent on temperature, with nearly constant creep rates from room temperature to half the melting point. There are not yet sufficient data to determine the effect, if any, of helium on irradiation creep. However, creep rupture is exacerbated by helium since bubble formation at grain boundaries is the primary mechanism of in-reactor failure, especially at high temperatures. Fatigue is also dependent upon high temperature helium embrittlement. High temperatures, high helium concentrations, and low strain rates enhance reduction in fatigue life. At temperatures as high as 550°C, there is no apparent effect of helium even at concentrations of 500 appm. at strain rates of 10−3 s−1 but either a decrease in strain rate by a factor of 100 or an increase in helium concentration by a factor of 6 produces a degradation in fatigue life. In general, helium affects mechanical properties of austenitic stainless steels; however, it is mostly a high temperature phenomenon. Caution must be exercised at temperatures below 250°C, where uniform tensile elongation is extremely low in irradiated alloys. Low-temperature embrittlement must be investigated further to determine if incremental hardening by helium contributes.

Impurity effects on reduced-activation ferritic steels developed for fusion applications

Abstract Reduced-activation steels are being developed for fusion applications by restricting alloying elements that produce long-lived radioactive isotopes when irradiated in the fusion neutron environment. Another source of long-lived isotopes is the impurities in the steel. To examine this, three heats of reduced-activation martensitic steel were analyzed by inductively coupled plasma mass spectrometry for low-level impurities that compromise the reduced-activation characteristics: a 5-ton heat of modified F82H (F82H-Mod) for which an effort was made during production to reduce detrimental impurities, a 1-ton heat of JLF-1, and an 18-kg heat of ORNL 9Cr–2WVTa. Specimens from commercial heats of modified 9Cr–1Mo and Sandvik HT9 were also analyzed. The objective was to determine the difference in the impurity levels in the F82H-Mod and steels for which less effort was used to ensure purity. Silver, molybdenum, and niobium were found to be the tramp impurities of most importance. The F82H-Mod had the lowest levels, but in some cases the levels were not much different from the other heats. The impurity levels in the F82H-Mod produced with present technology did not achieve the low-activation limits for either shallow land burial or recycling. The results indicate the progress that has been made and what still must be done before the reduced-activation criteria can be achieved.

Fracture toughness and tensile behavior of ferritic-martensitic steels irradiated at low temperatures

Abstract Disk compact tension and sheet tensile specimens of the ferritic-martensitic steels F82H and Sandvik HT-9 were irradiated in the High Flux Isotope Reactor (HFIR) at 90°C and 250°C to neutron doses of 1.5–2.5 dpa. For both steels, radiation hardening was accompanied by a reduction in strain hardening capacity (SHC). When combined with other literature data it is apparent that severe loss of SHC occurs in F82H for irradiation temperatures below ∼400°C and in HT-9 for irradiation temperatures below ∼250°C. For both alloys, severe loss of SHC does not correlate with brittle behavior during fracture toughness testing.

Effects of low temperature neutron irradiation on deformation behavior of austenitic stainless steels

Abstract Two experiments have been conducted to quantify the effects of neutron irradiation on the deformation and fracture behavior of solution annealed austenitic stainless steels irradiated to doses ranging from 3 to 19 dpa at temperatures from 60 to 400°C. For all alloys, yield strength increases rapidly with dose in the 60–300°C regime. Radiation hardening is accompanied by changes in the flow properties with the appearance of an initial yield drop and a significant reduction in strain hardening capacity. The magnitude of the changes is dependent upon both neutron dose and irradiation temperature, with reductions in strain hardening capacity occurring most rapidly in the range 250–350°C. It is shown that for neutron doses up to about 3 dpa, although the changes in deformation mode reduce the fracture thoughness, the toughness remains satisfactorily high.

Modeling of strengthening mechanisms in irradiated fusion reactor first wall alloys

Abstract Tensile property evaluations of austenitic stainless steels irradiated at 60 to 400°C in the ORR have been conducted. The neutron spectrum was tailored to achieve the appropriate helium/dpa ratio for austenitic stainless steels in a fusion reactor. The components of the alloy microstructure considered to contribute to strengthening are black dot defect clusters, Frank loops, network dislocations, voids, bubbles, and precipitates. Accepted expressions for hardening by defect interactions were employed in the calculation. It was found that the strengthening could be accounted for by the observed defects at temperatures below 330°C. However, strength was underpredicted at the higher temperatures. Since the voids and precipitates were observed to nearly always occur in coupled pairs, they were considered to be one large defect contributing to hardening through the Orowan bowing mechanism. This mechanism alone could not account for the strengthening observed. However, radiation-induced segregation of alloy components to the bubbles could account for the observed hardening, although defects below the limit of detectability are also believed to make a contribution.

High-temperature fatigue life of type 316 stainless steel containing irradiation induced helium

Abstract Specimens of 20%-cold-worked AISI type 316 stainless steel were irradiated in the High Flux Isotope Reactor (HFIR) at 550°C to a maximum damage level of 15 dpa and a transmutation produced helium level of 820 at. ppm. Fully reversed strain controlled fatigue tests were performed in a vacuum at 550°C. No significant effect of the irradiation on low-cycle fatigue life was observed; however, the strain range of the 10 7 cycle endurance limit decreased from 0.35 to 0.30%. The relation between total strain range and number of cycles to failure was found to be ΔE t = 0.02 N f −0.12 + N f −0.6 for N f 7 cycles.

Irradiation creep in type 316 stainless steel and us PCA with fusion reactor He/dpa levels☆

Abstract Irradiation creep was investigated in Type 316 stainless steel (316 SS) and US Fusion Program PCA using a tailored spectrum of the Oak Ridge Research Reactor in order to achieve a He/dpa value characteristic of a fusion reactor first wall. Pressurized tubes with stresses of 20 to 470 MPa were irradiated at temperatures of 330, 400, 500, and 600°C. It was found that irradiation creep was independent of temperature in this range and varied linearly with stress at low stresses, but the stress exponent increased to 1.3 and 1.8 for 316 SS and PCA, respectively, at higher stresses. Specimens of PCA irradiated in the ORR and having helium levels up to 200 appm experienced a 3 to 10 times higher creep rate than similar specimens irradiated in the FFTF and having helium levels below 20 appm. The higher creep rates are attributed to either a lower flux or the presence of helium. A mechanism involving interstitial helium-enhanced climb is proposed.

Influence of details of reactor history on microstructural development during neutron irradiation

Abstract Microstructurally-oriented irradiation experiments are shown in this paper to be strongly dependent on details of reactor history that frequently are not brought to the experimenters attention. In some cases, these details can dominate the experiment so as to produce very misleading results. To aid in the design and interpretation of microstructurally-oriented experiments, a number of studies are reviewed to highlight history effects and then guidelines are presented to minimize the impact of reactor history in new experiments.

Stress-strain relations of irradiated stainless steels below 673 K☆

Abstract Most specimens, irrespective of thermo-mechanical treatment, exhibited proof stress levels of above 800 MPa and uniform elongations below 1% after irradiation in the High Flux Isotope Reactor (HFIR). Only the solution annealed specimens irradiated at a low temperature of 328 K showed uniform elongations larger than 5% and proof stresses smaller than 800 MPa. Irradiation in the High Flux Reactor (HFR) caused more hardening than did irradiation in the HFIR. Ductility loss and change in work hardening characteristics by HFR irradiation were evaluated from reduction of area values. Residual ductility was revealed to be larger than 0.5 in natural strain, and the irradiation was estimated to have a small effect on work hardening characteristics and on fracture stress. The ductility of the irradiated alloys was found to be about 58% of that for the unirradiated alloys, as has been previously reported for irradiation in the HFIR. It was also demonstrated that true stress-strain relations, except for the fracture conditions, could be represented by Swifts type constitutive equation.