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In part I of this helium-effects study on ferritic/martensitic steels, results were presented on tensile and Charpy impact properties of 9Cr–1MoVNb (modified 9Cr–1Mo) and 12Cr–1MoVW (Sandvik HT9) steels and these steels containing 2% Ni after irradiation in the High Flux Isotope Reactor (HFIR) to 10–12 dpa at 300 and 400 °C and in the Fast Flux Test Facility (FFTF) to 15 dpa at 393 °C. The results indicated that helium caused an increment of hardening above irradiation hardening produced in the absence of helium. In addition to helium-effects studies on ferritic/martensitic steels using nickel doping, studies have also been conducted over the years using boron doping, ion implantation, and spallation neutron sources. In these previous investigations, observations of hardening and embrittlement were made that were attributed to helium. In this paper, the new results and those from previous helium-effects studies are reviewed and analyzed.

Tensile and Charpy specimens of 9Cr–1MoVNb (modified 9Cr–1Mo) and 12Cr–1MoVW (Sandvik HT9) steels and these steels doped with 2% Ni were irradiated at 300 and 400 °C in the High Flux Isotope Reactor (HFIR) up to ≈12 dpa and at 393 °C in the Fast Flux Test Facility (FFTF) to ≈15 dpa. In HFIR, a mixed-spectrum reactor, ( n , α ) reactions of thermal neutrons with 58 Ni produce helium in the steels. Little helium is produced during irradiation in FFTF. After HFIR irradiation, the yield stress of all steels increased, with the largest increases occurring for nickel-doped steels. The ductile–brittle transition temperature (DBTT) increased up to two times and 1.7 times more in steels with 2% Ni than in those without the nickel addition after HFIR irradiation at 300 and 400 °C, respectively. Much smaller differences occurred between these steels after irradiation in FFTF. The DBTT increases for steels with 2% Ni after HFIR irradiation were 2–4 times greater than after FFTF irradiation. Results indicated there was hardening due to helium in addition to hardening by displacement damage and irradiation-induced precipitation.

The dependence of helium production on radiation hardening and -embrittlement has been examined in a reduced-activation martensitic F82H steel (8Cr–2W–0.2V–0.04Ta–0.1C) irradiated at 250 °C to 2.3 dpa. In this study, 10 B and 11 B-doped specimens were irradiated to minimize the errors from the effect of B on mechanical properties by comparing the results. The specimens used were 10 B-doped, 10 B + 11 B-doped and 11 B-doped F82H steels. The total amounts of doping boron were about 60 mass ppm. The range of helium concentration produced in the specimens was from about 5 to about 330 appm. Tensile and fracture toughness tests were performed after neutron irradiation. 50 MeV-He 2+ irradiation was also performed to implant about 85 appm He atoms at 120 °C by AVF cyclotron to 0.03 dpa, and small punch testing was performed to obtain ductile-to-brittle transition temperatures (DBTT). Radiation hardening of the neutron-irradiated specimens increased slightly with increasing helium production. The 100 MPa m 1/2 DBTT for the F82H + 11 B, F82H + 10 B + 11 B, and F82H + 10 B specimens were 40, 110, and 155 °C, respectively. The shifts of DBTT due to helium production were evaluated as about 70 °C by 190 appm He and 115 °C by 330 appm He. In cyclotron experiment using standard F82H, a similar DBTT shift due to He was measured. These results suggest that helium production can increase the DBTT.

F82H (Fe–8Cr–2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load–displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 °C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 °C.

The effects of irradiation on the Charpy impact properties of reduced-activation ferritic/martensitic steels were investigated on a microstructural basis. It was previously reported that the ductile–brittle transition temperature (DBTT) of F82H-IEA and its heat treatment variant increased by about 130 K after irradiation at 573 K up to 5 dpa. Moreover, the shifts in ORNL9Cr–2WVTa and JLF-1 steels were much smaller, and the differences could not be interpreted as an effect of irradiation hardening. The precipitation behavior of the irradiated steels was examined by weight analysis and X-ray diffraction analysis on extraction residues, and SEM/EDS analysis was performed on extraction replica samples and fracture surfaces. These analyses suggested that the difference in the extent of DBTT shift could be explained by (1) smaller irradiation hardening at low test temperatures caused by irradiation-induced lath structure recovery (in JLF-1), and (2) the fracture stress increase caused by the irradiation-induced over-solution of Ta (in ORNL9Cr–2WVTa).

The effects of irradiation up to 20 dpa on the Charpy impact properties of reduced-activation ferritic/martensitic steels (RAFs) were investigated. The ductile-brittle transition temperature (DBTT)...

The synergistic effect of displacement damage and hydrogen or helium atoms on microstructures in F82H steel irradiated at 250–400 °C to 2.8–51 dpa in HFIR has been examined using isotopes of 54 Fe or 10 B . Hydrogen atoms increased slightly the formation of dislocation loops and changed the Burgers vector for some parts of dislocation loops, and they also affected on the formation of cavity at 250 °C to 2.8 dpa. Helium atoms also influenced them at around 300 °C, and the effect of helium atoms was enhanced at 400 °C. Furthermore, the relations between microstructures and radiation-hardening or ductile to brittle transition temperature (DBTT) shift in F82H steel were discussed. The cause of the shift increase of DBTT is thought to be due to the hardening of dislocation loops and the formation of α′-precipitates on dislocation loops.

An alloy of F82H prepared using the isotope 54Fe in order to encourage H and He production in a fission reactor has been irradiated in the HFIR JP20 experiment at three temperatures to 7 dpa as TEM disks. Irradiated disks were shear punch tested, examined by TEM, analyzed for He and H content, and compared with previous results in order to quantify irradiation hardening due to transmutation-induced H and He. Hardening due to irradiation is found following irradiation at 300 and 400 °C, that is intermediate between that at lower and higher dose, but hardening is negligible following irradiation at 500 °C. Microstructural examinations show typical behavior of irradiation as a function of irradiation temperature, with moderate swelling after 400 °C irradiation but few bubbles after irradiation at 300 °C. Correlations of change in hardening with He and H content show little indication of transmutation-induced hardening, but measured H levels do not agree with predictions and therefore H production and analysis requires further study.

In the absence of a 14 MeV neutron source, the effect of helium on structural materials for fusion must be simulated using fission reactors. Helium effects in ferritic/martensitic steels have been studied by adding nickel and boron and irradiating in a mixed-spectrum reactor. Although the nickel- and boron-doping techniques have limitations and difficulties to estimate helium effects on the ferritic/martensitic steels, past irradiation experiments using these techniques have demonstrated similar effects on the swelling and Charpy impact properties that are indicative of a helium effect. Although both techniques have disadvantages, it should be possible to plan experiments using the nickel- and boron-doping techniques to develop an understanding of the effects of helium on mechanical properties.

Miniature tensile and Charpy specimens of four ferritic/martensitic steels were irradiated at 300°C and 400°C in the high flux isotope reactor (HFIR) to a maximum dose of ≈12 dpa. The steels were standard F82H (F82H-Std), a modified F82H (F82H-Mod), ORNL 9Cr–2WVTa, and 9Cr–2WVTa–2Ni, the 9Cr–2WVTa containing 2% Ni to produce helium by (n,α) reactions with thermal neutrons. More helium was produced in the F82H-Std than the F82H-Mod because of the presence of boron. Irradiation embrittlement in the form of an increase in the ductile–brittle transition temperature (ΔDBTT) and a decrease in the upper-shelf energy (USE) occurred for all the steels. The two F82H steels had similar ΔDBTTs after irradiation at 300°C, but after irradiation at 400°C, the ΔDBTT for F82H-Std was less than for F82H-Mod. Under these irradiation conditions, little effect of the extra helium in the F82H-Std could be discerned. Less embrittlement was observed for 9Cr–2WVTa steel irradiated at 400°C than for the two F82H steels. The 9Cr–2WVTa–2Ni steel with ≈115 appm He had a larger ΔDBTT than the 9Cr–2WVTa with ≈5 appm He, indicating a possible helium effect.

Reduced-activation ferritic/martensitic steel, F82H (8Cr–2W–0.2V–0.04Ta–0.1C), and variants doped with isotopically tailored boron were irradiated at 673 K up to 51 dpa in the high flux isotope reactor (HFIR). The concentrations of 10B in these alloys were 4, 62, and 325 appm during HFIR irradiation which resulted in the production of 4, 62 and 325 appm He, respectively. After irradiation, transmission electron microscopy (TEM) was carried out. The number density of cavities increased and the average diameter of cavities decreased with increasing amounts of 10B. The number density decreased and the average diameter increased with increasing displacement damage. Swelling increased as a function of displacement damage and He concentration.

F-82H alloys isotopically enriched in 54 Fe up to 86% were irradiated in the high flux isotope reactor (HFIR) to determine the accelerated production of helium and hydrogen due to isotopic effects. Results are compared to calculations using isotopic helium production cross-sections from ENDF/B-VI or GNASH and measured neutron spectra. Helium measurements demonstrated an accelerated helium (appm)/dpa ratio of 2.3 after a 1.25-year irradiation, an increase of a factor of 4.3 over natural iron. The accelerated helium production is due to higher helium production cross-sections for 54 Fe and 55 Fe. Alloys doped with 55 Fe could achieve helium/dpa ratios up to about 20, well above the fusion reactor ratio of 10. Hydrogen measurements were performed using a newly developed quadrupole mass spectrometer system at PNNL. Calculations predict that hydrogen production will be accelerated by about a factor of 13 over natural iron. However, measurements show that most of this hydrogen is not retained in the samples.

The effects of helium production and heat treatment on the swelling of F82H steel irradiated in the HFIR to 51 dpa have been investigated using 10B, 58Ni and 60Ni-doped specimens. The swelling of tempered F82H-std and F82H doped with 10B irradiated at 400°C ranged from 0.52% to 1.2%, while the swelling of the non-tempered F82H doped with 58Ni or 60Ni was less than 0.02%. At 300°C the swelling in all steels was insignificant. In the F82H + Ni, a high number of density carbides formed in the matrix at these temperatures. The production of helium atoms enhanced the swelling of the F82H steel. However, the non-tempered treatment for the F82H + Ni suppressed remarkably the swelling. The cause of low swelling in the F82H + Ni may be due to the occurrence of the high density of carbides acting as sinks or the decrease of mobility of vacancies interacting with carbon atoms in matrix.

The effects of neutron spectrum on tensile properties of the low-activation martensitic steel F82H (8Cr–2WVTa) was examined using a thermal neutron shield to tailor the neutron spectrum for steels irradiated in the high flux isotope reactor (HFIR). The yield stresses of spectrally tailored specimens irradiated in HFIR to 5 dpa at 300°C and 500°C are on trend lines obtained from unshielded irradiation in HFIR. No significant effect of the neutron spectrum on tensile properties could be detected.

Microstructural evolution of welded austenitic stainless steel irradiated with mixed-spectrum neutrons was examined by transmission electron microscopy (TEM). TEM disks were obtained from electron-beam (EB) welded plates of JPCA, which is a Ti-midified austenitic stainless steel. Specimens were irradiated in HFIR up to 17 dpa at 670 and 770 K, and the estimated helium concentration was around 1100 appm. Cavities formed at 670 K irradiation were very small (

The austenitic stainless steel (Fe–0.06C–0.5Si–1.8Mn–14Cr–16Ni–2Mo–0.24Ti) was irradiated in a triple ion facility and the High Flux Isotope Reactor. The materials used were in the solution annealed (SA) and 15% cold-worked (CW) condition. TEM and tensile specimens were irradiated to a dose level of 30 and 10 dpa at 200°C. Some of the specimens were annealed after the irradiation at 500°C for 8 h in a vacuum. Microhardness tests were carried out on the surface of the TEM disks at room temperature. Tensile tests were carried out at 200°C in a vacuum with strain rate of about 1×10−3 s−1. The microhardness of both SA and CW increased by ion irradiation and then decreased by annealing. The yield strengths of the neutron irradiated SA and CW decreased to 610 and 650 MPa by annealing, respectively. The strain to necking of the irradiated CW recovered from 0.7% to 7.6%. The fracture mode remained ductile in each case.

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.

Vanadium, V1Ni, V10Ti and V10Ti1Ni (at%) were irradiated in HFIR to doses ranging from 18 to 30 dpa and temperatures between 300 and 600°C. Since the irradiation was conducted in a highly thermalized neutron spectrum without shielding against thermal neutrons, significant levels of chromium (15–22%) were formed by transmutation. The addition of such large chromium levels caused strong embrittlement. At higher irradiation temperatures radiation-induced segregation of transmutant Cr and solute Ti at specimen surfaces caused strong increases in the density of the alloy. The resultant shrinkage, possibly compounded by thermal cycling, led to cracks developing at all intersections of grain boundaries with the specimen surface. This caused specimens irradiated at 500°C or below to often fail during retrieval from the reactor, as well as during electropolishing and other handling operations. At 600°C, the cracking and embrittlement processes are so severe that only a fine dust, composed mostly of individual grains or chunks of grains, was found in the irradiation capsule.

Tensile and Charpy impact properties of neutron irradiated F82H (Fe8Cr2WVTa) with and without boron have been investigated to obtain the basic irradiation response on mechanical properties in low damage regime less than 1 dpa at the temperature ranging from 300° to 590°C. Boron-doped steel was used for the helium effect due to (n, α) reaction. Typical irradiation hardening was observed at 300°C. The irradiation above 520°C did not reveal increase in yield stress, but the specimen irradiated at 590°C showed some reduction in elongation in room temperature tensile testing. Slight difference in the tensile properties between boron-doped and boron-free were observed at 590°C. No changes in ductile brittle transition temperature (DBTT) occurred at a temperature between 335° and 460°C by Charpy impact testing.

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.

Type 316 stainless steel from the core of the experimental fast breeder reactor (FBR) JOYO was examined by the slow strain rate tensile (SSRT) test in pure, oxygenated-water and air and by the electrochemical potentiokinetic reactivation (EPR) test to evaluate a susceptibility to the irradiation assisted stress corrosion cracking (IASCC) and the radiation-in-duced segregation (RIS). The solution annealed and 20% cold-worked materials had been irradiated at 425°C to a neutron fluence of 8.3 × 1026n/m2 (〉 0.1 MeV) which is equivalent to 40 displacement per atom (dpa). Intergranular cracking was induced by the SSRT in water at 200 and 300°C, but was not observed on specimen tested in water at 60°C and in air at 300°C. This indicates that irradiation increased a susceptibility to stress corrosion cracking (SCO in water. After the EPR test, grain boundary etching was observed in addition to grain face etching. This suggests Cr depletion may have occurred both at grain boundary and at defect clusters during the irradiation. The results are compared with the behavior of similar materials irradiated with different neutron spectrum.

The release of 133Xe from (Th,U)O2 was studied at a low fission density by using a post-irradiation technique. The uranium concentrations of the specimens ranged from 0.15 to 20 mol% U02. Heating curves of the release gave almost the same pattern in shape, while the total release, which was the combination of an in-pile release and a post-irradiation release up to 1000°C, increased with uranium concentration except for the nominally pure ThO2. Effects of preparation conditions of specimens such as atmosphere, temperature and stoichiometry were also studied and found to be minor. Possible release mechanisms were discussed.

The lattice parameter change of Th 0.937 U 0.063 O 2 fuel kernels for high temperature gas cooled reactors (HTGR) was studied after irradiation from 1.1 × 10 15 to 4.4 × 10 17 f/cm 3 using a micro-beam X-ray diffractometer. It was found that the lattice parameter increases initially in a linear relationship with the increasing fission dose and then reaches a saturation value ( Δ a a ⋍9.4 × 10 −4 ) at about 1 × 10 17 f / cm 3 . The damage volume and the fission-induced defect number per fission were estimated using a damage saturation function for the fission dose dependence of the lattice parameter changes. The isochronal annealing results indicated that the recovery was roughly characterized by two stages. The isothermal annealing results showed that the recovery process seems to obey a second order kinetics with the activation energy depending on the fraction of recovery.

The polished hemispheres of ThO2 kernels were irradiated by fission fragments, which recoiled from the covered Al-U alloy foils. After irradiation, the lattice parameters of the kernels were measured by microbeam X-ray diffractometer. It was found that the lattice parameter increased initially in a linear relationship with increasing fission fragment dose and reached a saturation value at doses higher than about 1×1016 fission fragments⋅cm-3. The saturation value was smaller for the fine-grained specimen than for the coarsegrained. The recovery began at about 400°C for both the specimens and lasted to higher temperature for the coarse-grained.

The polished hemispheres of ThO2 kernels were irradiated by fission fragments, which recoiled from the covered Al-U alloy foils. After irradiation, the lattice parameters of the kernels were measured by micro-beam X-ray diffractometer. It was found that the lattice parameter increased with increasing fission fragment dose and reached a saturation value at doses higher than about 5×1016 fission fragments-cm-3. With increasing grain sizes from 4.0 to 19.3 μm, the saturation values of lattice parameter changes increased from approximately 0.07 to 0.18%. The recovery began at about 400°C in all specimens and the temperature of the complete recovery had a tendency to become higher with increasing grain size. These results lead to a conclusion that the grain boundaries act as a sink of the defects created by fission fragments.

In-pile release of 85mKr, 87Kr, 88Kr, 133Xe, 135Xe and 138Xe from natural graphite was measured over a temperature range of 100 to 950°C and that of 133I and 135I was estimated. The effects of fission rate and fission density on the fission gas release were examined. The β-decay process of iodine within and outside the graphite was observed to control the gross release of xenon. Possible mechanisms of an anomalous sink of 88Kr in the plot of log R B versus log λ were discussed. A new model for knock-out release was proposed on the basis of its temperature dependence found. These results were compared with those of post-irradiation experiments and explained by a mechanism whereby fission gas is trapped in defects created in graphite by fission and β-decay energy, and released through annealing of the defects.