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.     

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.     

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.     

The temperature dependence of nickel-ion damage in nickel

The temperature dependence of void and dislocation structures was studied in high-purity nickel irradiated with 2.8 MeV ⁵⁸Ni⁺ ions to a displacement density of 13 displacements per atom (dpa) at a displacement rate of $\text{7 × 10}{}^{\mathrm{?2}}\text{dpa/sec}$ over the temperature range 325 to 625°C. Dislocation loops, with no significant concentrations of voids, were observed in specimens irradiated at 475°C and below. Specimens irradiated between 525 and 725°C contained both voids and dislocations. The maximum swelling was measured as 1.2% at 625°C. Analysis of the data by theoretical models for void nucleation and growth indicated that the swelling in the present experiment was principally limited by void growth at low temperatures and by void nucleation at high temperatures. The data were also compared with previously reported neutron and nickel-ion irradiation results.     

Design,operation, and initial results from a series of graphite creep irradiation experiments

A series of irradiation tests was designed to evaluate the creep characteristics at elevated temperatures and to high fast fluences of various graphites of interest to HTGR designers. The series encompasses the irradiation of 28 specimens, each 15.24 mm (0.6 in.) diam × 25.4 mm (1 in.) long, at 900°C to incremental exposures of $\text{1,2,4, and 8 × 10}{}^{21}\text{neutrons/cm}{}^2$ ($\text{E > 0.18}\text{MeV}$); 28 similar specimens at 600°C to the same exposures; and 28 other similar specimens at 1250°C under the same conditions. A compressive stress of 13.79 MPa (2000 psi) is applied to 20 of the specimens in each test by means of a metal bellows expanded by gas pressure against the specimen columns. Eight of the stacked specimens in each test are stressed to 20.68 MPa (3000 psi) by a reduction in diameter. Special features of the capsules are described which include (1) movable center-line thermocouples which measure the temperature profile along the axis of the capsule, (2) linear variable differential transformer type load cells to monitor the applied load, and (3) computerized temperature control designed to achieve accurate longitudinal temperatures over the 0.508 m (20 in.) length of the test specimen columns. The operational characteristics of the first capsule, which is currently being irradiated and is one of a total of twelve in the series to be irradiated over a four-year period, are presented. The performance of the Oak Ridge Research Reactor Data Acquisition and Control System (ORRDACS) is briefly described.     

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.     

Examination of the plastic properties of zircaloy-4 at elevated temperatures in air atmosphere

Results are reported about the examination of the influence of test temperatures and sample anisotropy upon the yield stress, the strain rate sensitivity and the ductility of Zry-4 sheet-type tensile specimens. Most of the investigations were performed in air atmosphere in the temperature range between 750 and 1050°C at cross-head speeds ranging over three orders of magnitude. Superplastic deformation behaviour was observed at temperatures between 860 and 1000°C for a strain rate of the order of 10^?4 s^?1. For samples deformed under these conditions in air atmosphere the strain rate sensitivity index $\text{m}{}_{\mathrm{a}}$ has shown to depend sensitively upon strain. Although no influence of anisotropy upon the yield stress was observed above 550°C, in the superplastic region $\text{m}{}_{\mathrm{a}}$ revealed a strong dependence upon sample orientation. For higher temperatures even in air the total elongation increased with temperature up to 850°C. At the highest temperature due to severe take-up of oxygen the samples failed brittle. The apparent activation energy for superplastic flow agrees very well with the energy of grain boundary diffusion in α-Zr.     

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.     

Neutron irradiation-induced voids in β-silicon carbide

Pyrolytic β-silicon carbide was irradiated at temperatures between 625°C and 1500°C to neutron fluences up to 12 × 10²¹ n/cm² ($\text{E > 0.18 MeV}$). Density changes were measured and the samples were examined by transmission electron microscopy. Irradiation below 1000°C created small Frank dislocation loops on {111} planes. Irradiation at 1250°C and 1500°C produced tetrahedral voids which caused continuing expansion of the samples. Void sizes increased with increasing fluence and with increasing irradiation temperature, while void concentrations decreased with increasing temperature. One-hour post-irradiation anneals at 1700°C to 2100°C reduced the void concentration and total void volume while increasing the maximum void size.     

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 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.     

Irradiation effects of boron carbide used as control rod elements in fast breeder reactors

The paper describes radiation effects on 8₄C pellets used as control rod elements in the Enrico Fermi Fast Breeder Reactor. Pellet swelling (Δ$\text{V/V}$) caused by irradiation was less than 1% in which crystal lattice swelling was less than 20%. Many microcracks, a main cause of pellet swelling, appeared in the irradiated pellets. The production of microcracks was related to graphite precipitation in the pellets before irradiation. Open pores which did not exist in the unirradiated pellets were formed in the irradiated ones. In a unit cell of B₄C, the α-axis elongated by 0.025 Å and the $\text{c-}\text{axis}$ shrank by 0.07 Å by irradiation. Moreover, we found three recovery stages which were from room temperature to 400°C, from 400 to 750°C and from 850 to 1100°C. The recovery mechanisms in the irradiated pellets are discussed in terms of the helium behavior.     

Neutron damage and fission product gas release in the initial burst region of uranium monocarbides

Thermal neutron damage and fission product gas (¹³³ Xe) release in a burst region of uranium monocarbides were studied. After neutron irradiation, the electrical resistivity was measured from room temperature to 800° C. Three recovery stages were revealed in the resistivity of UC irradiated to $\text{4.0 × 10}{}^{16}$ nvt. The lattice parameter of UC with the same irradiation also showed three stages of recovery up to 1050°C. The initial burst of Xe from UC was studied in a dose range between $\text{1.6 × 10}{}^{15}$ and $\text{2.9 × 10}{}^{18}$ nvt. The burst occurred in three steps for lightly irradiated specimens, while there were two steps of the burst in heavily irradiated specimens. The activation energies for each burst step were calculated. From the results obtained here, we concluded that the burst was correlated with the recovery of damage in the neutron-irradiated UC.     

Post-irradiation ductility loss of FeNibase precipitation hardenable alloys

Several γ′- and γ′/γ″-strengthened Fe-Ni-base superalloys have shown near-zero ductility after neutron irradiation to fluences of $\text{～ 4 × 10}{}^{22}\text{n/cm}{}^2$, $\text{E > 0.1 MeV}$, at 500 to 650°C. The ductility loss is most pronounced in solution-treated or in solution-treated and aged specimens tested at 110°C above the irradiation test temperature. Failed specimens exhibit brittle intergranular fractures. Microstructural examination of the embrittled specimens showed that continuous or semi-continuous coatings of γ′ formed at grain boundaries during the irradiation. In some cases, the grain boundaries were also decorated by small bubbles, thought to be transmutation-induced helium or contained trace elements such as sulphur and phosphorus. All of these grain boundary alterations are attributed to radiation-induced solute segregation. Microanalyses of the γ′ coatings indicate that Ni, Al, Si, Ti and Nb had segregated to grain boundary sinks in irradiated FeNiCr based alloys. Nonequilibrium segregation of helium and trace impurities is also considered likely. The role of radiation-induced segregation in the embrittlement phenomenon is consistent with the observation that introducing a high density of dislocation sinks by cold-working reduces γ′ formation at the grain boundaries and reduces the ductility loss. The embrittlement is attributed to concurrent strengthening within grains by irradiation-induced γ′ precipitation and brittle cleavage failure of grain boundary precipitates.     

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.     

Design of an irradiation device for the determination of the in-pile creep behaviour of zircaloy cladding tubes under internal and external over-pressure,in FRG-2

The dimensional stability of fuel rods in light water reactors is influenced by the creep strength of the Zircaloy (Zry) cladding. The Gesellschaft für Kernenergieverwertung in Schiffbau und Schiffahrt m.b.H and Kraftwerk Union AG are jointly carrying out irradiation experiments in FRG-2, to determine the effect of neutron irradiation on the creep behaviour of Zry cladding. In these capsule experiments specimens can be tested in a helium environment at temperatures from 280 to 400°C in a fast neutron flux of approx. $\text{5 × 10}{}^{13}\text{/}\text{cm}{}^2\text{s}$ under biaxial tensile and compressive stresses from 70 N/mm² to 150 N/mm². In the paper the test equipment, the experimental techniques and the initial results are described.     

Einführung in die Werkstoffwissenschaft (in German): ed. W. Schatt (VEB Deutscher Verlag für Grundstoffindustrie,Leipzig, 1972, DM 35)

Copper samples were irradiated with fast neutrons at temperatures in the range 220–550 °C and at instantaneous fluxes in the range 2 × 10¹³-3 × 10¹⁴ n/cm².sec > 0.1 MeV. The maximum swelling was observed at 0.45 $\text{T}{}_{\mathrm{f}}$ for an instantaneous flux of 3 × 10¹⁴ n/cm²-sec. A fourfold reduction of the instantaneous flux, at constant dose, displaces the maximum to lower temperatures and slightly increases its magnitude. Cold work before irradiation does not appear to have a significant effect on swelling. Alloying with solutes which lower the stacking-fault energy appears to displace the domains of swelling towards lower temperatures for a fixed instantaneous flux and towards lower flux for a fixed temperature.