12Cr-1MoVW Steels Research Articles

Irradiation-induced impurity segregation and ductile-to-brittle transition temperature shift in high chromium ferritic/martensitic steels

A model is presented to predict irradiation-induced impurity segregation and its contribution to the ductile-to-brittle transition temperature (DBTT) shift in high chromium ferritic steels. The hardening contribution (dislocation loops, voids and precipitates) is also considered in this study. The predicted results are compared with the experimental DBTT shifts data for irradiated 9Cr1MoVNb and 12Cr1MoVW steels with different grain sizes.

Heat treatment effects on impact toughness of 9Cr–1MoVNb and 12Cr–1MoVW steels irradiated to 100 dpa

Plates of 9Cr–1MoVNb and 12Cr–1MoVW steels were given four different heat treatments: two normalizing treatments were used and for each normalizing treatment two tempers were used. Miniature Charpy specimens from each heat treatment were irradiated to ≈20 dpa at 365°C and to ≈100 dpa at 420°C in the Fast Flux Test Facility (FFTF). In previous work, the same steels were irradiated in FFTF to 4–5 dpa at 365°C and 35–36 dpa at 420°C. The tests indicated that prior austenite grain size, which was varied by the different normalizing treatments, affected the impact behavior of the 9Cr–1MoVNb but not the 12Cr–1MoVW. Tempering had relatively little effect on the impact behavior of both steels. Conclusions are presented on how heat treatment can be used to optimize impact properties.

Embrittlement of CrMo steels after low fluence irradiation in HFIR

Subsize Charpy impact specimens of 9Cr1MoVNb (modified 9Cr1Mo) and 12Cr1MoVW (Sandvik HT9) steels and 12Cr1MoVW with 2% Ni (12Cr1MoVW2Ni) were irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400°C to damage levels up to 2.5 dpa. The objective was to study the effect of the simultaneous formation of displacement damage and transmutation helium on impact toughness. Displacement damage was produced by fast neutrons, and helium was formed by the reaction of 58Ni with thermal neutrons in the mixed-neutron spectrum of HFIR. Despite the low fluence relative to previous irradiations of these steels, significant increases in the ductile-brittle transition temperature (DBTT) occurred. The 12Cr1MoVW2Ni steel irradiated at 400°C had the largest increase in DBTT and displayed indications of intergranular fracture. A mechanism is proposed to explain how helium can affect the fracture behavior of this latter steel in the present tests, and how it affected all three steels in previous experiments, where the steels were irradiated to higher fluences.

Effects of temperature on the cyclic deformation behaviour and microstructural changes of 12Cr-1MoVW martensitic stainless steel

Strain-controlled uniaxial push-pull fatigue tests were carried out with cylindrical specimens of 12Cr-1MoVW steel from room temperature up to 600°C. Specimens were tested at a total strain range of 2% at room temperature, 300°C, 400°C, 500°C and 600°C, respectively, and were also tested at 500°C for a total strain range of 1·6%, 1% and 0·6%. The fatigue test results showed that the cyclic deformation behaviour of this steel was strongly dependent on the test temperatures. TEM studies revealed that cellular dislocation structure comprised the major microstructure and the subgrain size increased with increasing test temperatures. Carbides tended to redistribute along the subgrain boundaries in the specimens subjected to low cycle fatigue at high temperatures. The undeformed section of each specimen exhibited a completely different microstructure from that in the gauge section in all the temperatures tested. The lath martensite structures were not changed obviously when HT-9 steel cyclic deformed at lower strain range.

Heat treatment effects on toughness of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated at 365°C

The 9Cr-1MoVNb and 12Cr-1MoVW steels were austenitized at 1040 and 1100°C to produce different prior austenite grain sizes, after which they were given different tempering treatments (1 h at 760°C or 2.5 h at 780°C). Subsize Charpy impact specimens from these materials were irradiated at 365°C up to 5 dpa. For 9Cr-1MoVNb steel in the unirradiated condition, the smaller the prior austenite grain size and the higher the tempering temperature, the lower the ductile-brittle transition temperature (DBTT). Regardless of the DBTT in the unirradiated condition, however, the DBTT shift for 9Cr-1MoVNb steel due to irradiation was the same for all heat treatments. This means heat treatment can be used to ensure a lower DBTT before and after irradiation. The 12Cr-1MoVW steel showed little effect of heat treatment on DBTT in the unirradiated condition, and the shift in DBTT was relatively constant. Thus, it appears that heat treatment cannot be used to reduce the effect of irradiation on DBTT for this steel.

Dissolution kinetics of steels exposed in lead-lithium and lithium environments

An analysis of weight loss kinetics was made for type 316 stainless steel and Fe12Cr1MoVW steel in thermally convective Fe17 at.% Li (500°C) and lithium (600°C). For both steels, the kinetics are similar in lithium and Pb17Li if the comparison is made at a temperature where solubility-driven mass transport dominates. At the temperatures of this study, a model based on paralinear reaction kinetics accurately described the weight loss curves for type 316 stainless steel in both liquid metals, while the ferritic steel rapidly achieved weight losses that were linear with time. The differences between the kinetic behavior of the two steels can thus be interpreted in terms of the preferential dissolution reactions and localized attack observed for the austenitic alloy.

Small specimen test techniques for the evaluation of toughness degradation

The small punch (SP) test was applied to neutron irradiated TEM disks to obtain fracture related mechanical properties (DBTT, J IC) of low activation ferritic steels of 8Cr-2WV and 8Cr-2WVTa and conventional 12Cr-1MoVW steel. The SP test using 3 mm diameter TEM disks is confirmed to be useful, especially for J IC evaluation. DBTT determination is also thought to be possible with a sufficient number of specimens.

Embrittlement of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated in HFIR

Charpy impact specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels and these steels with 2% Ni (9Cr-1MoVNb-2Ni and 12Cr-1MoVW-2Ni) were irradiated in the High Flux Isotope Reactor (HFIR) at 300°C up to 34 dpa and at 400°C to 42 dpa. Nickel was added to study the simultaneous effect of displacement damage and transmutation helium on toughness. Displacement damage is produced by fast neutrons and helium is produced by the reaction of 58Ni with thermal neutrons in the mixed spectrum of HFIR. Irradiation caused large increases in the ductile-brittle transition temperature (DBTT) of all four steels. The saturation in the shift in DBTT with fluence that has been observed in such steels irradiated in fast reactors does not apply to specimens irradiated in HFIR. Shifts in DBTT of over 300°C were observed at 400°C for 9Cr-1MoVW-2Ni and 12Cr-1MoVW-2Ni; shifts of over 200°C were observed for the standard steels. The shifts are the largest ever recorded for this type of steel and are attributed to the high helium concentrations (up to 400 appm).

Effect of irradiation in HFIR on tensile properties of Cr-Mo steels

Tensile specimens of tempered martensitic 9Cr-1MoVNb and 12Cr-1MoVW steels with and without additions of 2% Ni were irradiated in the High Flux Isotope Reactor (HFIR) at 400 and 600°C to displacement-damage levels of up to 72 dpa. Nickel was added to the steels to produce transmutation helium, and up to 700 appm He was produced in the steels with 2% Ni. Irradiation at 400°C caused an increase in strength and a decrease in ductility, while irradiation at 600°C caused a decrease in strength and an increase in ductility. Results at 400°C were similar to data from previous tests on steel irradiated to ~ 11 dpa (up to 110 appm He), where it was concluded that helium caused a strength increase in addition to the hardening caused by displacement damage alone. The present results indicate that the helium-enhanced strengthening at 400°C saturates with increasing helium concentration, similar to the saturation (at a lower strength level) with displacement-damage dose found in the absence of helium.

Tensile properties of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated to 23 dpa at 390 to 550 ° C

Normalized-and-tempered 9Cr-1MoVNb and 12Cr-1MoVW steels were irradiated in the Experimental Breeder Reactor II (EBR-II) at 390, 450, 500, and 550°C to displacement damage levels of up to 23 to 25 dpa. Tensile tests were conducted at the irradiation temperatures on three types of specimens: irradiated specimens, normalized-and-tempered specimens, and specimens thermally aged 5000 h at the irradiation temperatures. Observations from these tests were compared with results on these same materials irradiated in EBR-II at the same temperatures to 9 to 13 dpa. Irradiation of specimens at 390°C to 23 to 25 dpa caused an increase in strength and a decrease in ductility relative to the aged and unaged control specimens; however, there was little change from the specimens irradiated to 9 to 13 dpa. After irradiation at 450, 500, and 550 ° C, there was little difference between the irradiated specimens, the normalized-and-tempered specimens, and the specimens aged for 5000 h. Results were interpreted in terms of the precipitate and dislocation microstructures developed during normalizing and tempering, thermal aging, and irradiation.

Impact behavior of 9Cr-1MoVNb and 12Cr-1MoVW steels irradiated in HFIR

Charpy impact specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels were irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400°C to displacement-damage levels as high as 42 dpa. For both steels, irradiation caused large increases in the ductile-brittle transition temperature (DBTT) and a decrease in the upper-shelf energy (USE). The increase in DBTT and the decrease in USE were greater at 400°C than at 300°C. At 400°C, irradiation to ~ 40 dpa produced shifts in DBTT of 204 and 242°C for the 9Cr-1MoVNb and 12Cr-1MoVW steels, respectively. These are the largest shifts ever observed for these steels and are attributed to the high helium concentration generated during irradiation in HFIR.

Impact behavior of Cr-W steels

Chromium-molybdenum ferritic (martensitic) steels are leading candidates for the structural components for fusion reactors. However, irradiation of steels containing molybdenum or niobium in a fusion environment will produce long-lived radioactive isotopes that will lead to difficult waste disposal problems. To alleviate the waste disposal problem, ferritic steels are being developed that are analogous to conventional Cr-Mo steels, but with molybdenum replaced by tungsten and niobium replaced by tantalum. Experimental steels containing 0.1% C, 2 1/4 to 12% Cr, 0 to 2% W, 0 or 0.25% V, and 0 or 0.07% Ta were produced. Charpy impact properties were determined. A 5Cr-2W-0.25V steel and a 9Cr-2W-0.25V-0.07Ta steel had the best impact properties. The impact properties of these two steels as well as those of a 9Cr-2W-0.25V and a 12Cr-2W-0.25V steel were as good or better than the properties of similarly heat treated conventional 9Cr-1MoVNb and 12Cr-1MoVW steels.

Fluence and helium effects on the tensile properties of ferritic steels at low temperatures

Tensile specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels with up to 2% Ni and 2 1 4 Cr-1Mo steel were tested at room temperature after irradiation at ~ 50 ° C to displacement-damage levels of up to 25 dpa in the High-Flux Isotope Reactor (HFIR). Nickel was added to the ferritic steels to produce helium by a two-step (n, α) reaction with thermal neutrons during irradiation in the mixed neutron spectrum of HFIR. Up to 327 appm He was produced in the steels with 2% Ni. Irradiation caused an increase in the strength of all the steels. With an increase in fluence, there was a decrease in the rate of strength increase, but the strength was still increasing after 20 to 25 dpa. Strength increases were accompanied by a loss of ductility, although the ductility appeared to go through a minimum and was greater at the highest fluences than at intermediate fluences. The results were interpreted to mean that the transmutation helium that was generated during irradiation caused part of the strength increase.

Experimental and analytical investigations of mass transport processes of 12cr-1 movw steel in thermally-convected lithium systems

Experimental data on corrosion and mass transport in lithium/12Cr-1 MoVW steel were obtained from two thermal convection loops; one operated from 360 to 505 °C for 3040 h and the other from 525 to 655 °C for 2510 h. The experimental effort was supported by analytical investigations of possible mechanisms of corrosion and mass transport. It was found that mass transfer is not a simple function of temperature and alloy component solubility. Above 580 °C mass transfer appears to be dominated by alloy solubility via the temperature gradient. Between 450 and 580 °C, mass transfer appears to be related to surface reactions involving dissolved carbon and nitrogen in lithium with chromium, and carbides on the steel surface. The corrosion rates, as interperted from weight-change as uniform dissolution, from this work, are significantly lower than those adopted in recent blanket design studies.

Impact behavior of 9-Cr and 12-Cr ferritic steels after low-temperature irradiation

Abstract Miniature Charpy impact specimens of 9Cr-1MoVNb and 12Cr-1MoVW steels and these steels with 1 and 2% Ni were irradiated in the High-Flux Isotope Reactor (HFIR) at 50°C to displacement damage levels of up to 9 dpa. Nickel was added to study the effect of transmutation helium. Irradiation caused an increase in the ductile-brittle transition temperature (DBTT). The 9Cr-1MoVNb steels, with and without nickel, showed a larger shift than the 12Cr-1MoVW steels, with and without nickel. The results indicated that helium also increased the DBTT. The same steels were previously irradiated at higher temperatures. From the present and past tests, the effect of irradiation temperature on the DBTT behavior can be evaluated. For the 9Cr-1MoVNb steel, there is a continuous decrease in the magnitude of the DBTT increase up to an irradiation temperature of about 400°C, after which the shift drops rapidly to zero at about 450°C. The DBTT of the 12Cr-1MoVW steel shows a maximum increase at an irradiation temperature of about 400°C and less of an increase at either higher or lower irradiation temperatures.

Reduced-activation ferritic steels: A comparison with Cr-Mo steels

“Reduced-activation” ferritic steels are being developed to replace the Cr-Mo steels presently being considered for fusion-reactor applications. To meet the reduced-activation criteria, the steels cannot contain Mo, Nb, Ni. Cu, and N. Tungsten is proposed as a substitute for molybdenum, and niobium is replaced by tantalum and an increased use of vanadium. An initial series of Cr-W steels have been produced that are analogous to the Cr-Mo steels. Microstructural examination, tempering studies, tensile tests, and Charpy impact tests indicate that three Cr-W steels with 9 to 12% Cr are comparable to the 9Cr-1MoVNb and 12Cr-1MoVW steels, which are presently the primary candidate ferritic steels for fusion applications.

Effect of nickel content of 9Cr-1MoVNb and 12Cr-1MoVW steels on the aging and irradiation response of impact properties

Irradiating nickel-containing alloys in a mixed-spectrum reactor can simulate both transmutation helium and displacement damage expected in a fusion reactor first wall. Impact properties of 9Cr-1MoVNb and 12Cr-1MoVW steels doped with nickel were determined in the as-heat-treated, thermally aged, and irradiated conditions to determine if nickeldoping affects the behavior. The irradiation was carried out in a fast-spectrum reactor which produces only an insignificant amount of helium during irradiation, thereby evaluating the effect of nickel alone. Only limited property changes resulted from thermal aging or irradiation to 12 dpa at 450 to 550°C. Irradiation of the 12Cr-1MoVW steel at 390°C produced severe degradation of impact properties. Nickel additions affected the unirradiated material properties, but subsequent radiationinduced changes were similar. The results indicate that nickel doping and subsequent irradiation in a mixed spectrum reactor is a viable method for simulating irradiation effects in a fusion reactor first wall.

Tensile behavior of irradiated 12Cr-1MoVW steel

The effect of irradiation on the tensile properties of 12Cr-1MoVW steel given two different normalized-and-tempered heat treatments was determined for specimens irradiated in the Experimental Breeder Reactor-II (EBR-II) at 390. 450, 500 and 550°C to ~13 dpa. Tensile tests were conducted at the irradiation temperature and room temperature on the irradiated specimens, as-heat-treated specimens, and as-heat-treated specimens thermally aged 5000 h at the irradiation temperature. Irradiation at 390°C increased the 0.2% yield stress and ultimate tensile strength above the strength of the unaged and aged controls for both heat-treated conditions. The effect was greater for the steel heat treated with the shorter austenitization and tempering times. The increased strength was accompanied by a slight decrease in ductility. After irradiation at 450, 500 and 550°C the yield stress, ultimate tensile strength, and ductility were similar to the as-heat-treated and the thermally aged controls. Strength changes were discussed in terms of microstructural changes observed in other studies.