Hardening and Recovery of Neutron-Irradiated Pressure Vessel Steel (ASTM A533B)

Specimens of ASTM A533B steel were studied to gain information on the annealing process following irradiation, through measurements of internal friction and of hardness.

The specimens were quenched from 900°C and tempered at 650°C, then irradiated in the JMTR reactor at 65°–75°C to a neutron dose of 1.4–1.7×10²⁰ n/cm² (E _n >1MeV).

Peaks were observed on the internal friction curves from unirradiated specimens. These peaks disappeared upon irradiation, but reappeared with annealing treatment at 150°C.

Radiation-anneal hardening was observed at 250°C. The recovery of radiation hardening begins at a temperature between 250° and 350°C, but is not completed even at 550°C.     

Radiation damage in silicon carbide and graphite for fusion reactor first wall applications

Silicon carbide and graphite materials were exposed to fast neutron fluences of 2 × 10²³ to 2 × 10²⁴n/m² (E > 1 MeV) and a study was made of changes in fracture strength, Weibull modulus and electrical resistivity. Silicon carbide (Norton NC-430) exhibits a decrease in fracture strength (25%) at the higher fluence if the temperature is kept at 298 K, while at 1473 K the decrease in fracture strength is only 10% indicative of recovery due to thermal annealing. The fracture strength of the graphite (POCO AXF-5Q) tested at 298 K increases rapidly by ~20% after 2 × 10²³n/m² and remains constant at higher fluence. Analyses of the data using the Weibull weakest link model were given, in addition to annealing and swelling results.     

Effect of irradiation on the deformation of copper and copper-aluminium alloys

The effect of low-temperature irradiation (<70°C) on the deformation of pure copper, Cu-0.09 wt% Al alloy, and Cu-0.19 wt% Al alloy was investigated. Tensile specimens were prepared from single crystals of various orientations, and were exposed to a neutron fluence of 6.85 ×10¹⁹ n/cm²(E>1 MeV) at a temperature less than 70°C. All irradiated and unirradiated specimens were pulled. A large increase in critical shear stress due to irradiation was observed; the increase was smaller in Cu-0.19 wt% Al alloy than in pure copper and Cu-0.09 wt% Al alloy. Ultimate shear stress and shear strain were less influenced by irradiation. Yield points were observed in all irradiated specimens. The yield drop was large in irradiated pure copper, and decreased with the increase of aluminium content in copper-aluminium alloys. All unirradiated specimens showed a high work-hardening coefficient (n) in the beginning of the deformation, followed by a lower value. By irradiation, the first value drastically decreased, while the second remained nearly constant. Shear stress and shear strain were influenced by crystal orientation.     

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.     

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 fast-neutron irradiation on deformation twinning in zirconium deformed at 77 K

The effect of neutron irradiation on twinning deformation is investigated at 77 K using unirradiated and to a fluence of 4.1 × 10¹⁹n/cm² irradiated zirconium. The slip line density N_s, the twin volume fraction V_t and the twin density N_t for both specimens were determined as a function of strain by means of optical microscopy. The twins were determined using the diffraction profiles of {101&#x0304;2} twins at the early stages, as well as, {112&#x0304;2} and {112&#x0304;4} twins at the latter stages of deformation. It is shown that the nucleation of twins was affected more strongly than the growth by neutron irradiation.     

Anisotropic growth of reactor fuel pin cladding

Anisotropic growth of 316 stainless steel reactor fuel pin cladding was found to occur after irradiation in the Experimental Breeder Reactor-II (EBR-II). Pressurized tube specimens were irradiated to a peak fluence of 10²³n/cm² (E >0.1 MeV) at temperature ranging from 430°C to approximately 590°C. Growth was observed in both the annealed and 20% cold worked conditions and was found to decrease with increasing hoop stress. The anisotropic growth is more pronounced in the cold worked condition. The growth is attributed to a preferred orientation of Burgers vectors in the preirradiated cold worked dislocation structure.     

Effects of neutron irradiation on polymer matrix composites at 5 K and at room temperature: II. Degradation of mechanical properties

Neutron irradiation with a low flux of accompanying γ-rays in the Intense Pulsed Neutron Source was carried out at 5 K and at room temperature on four kinds of polymer matrix composites (filler: E-glass or carbon fiber cloth; matrix: epoxy or polyimide resin). The specimen irradiated at 5 K was warmed up to room temperature before the mechanical test was performed at 77 K. The Young's modulus of these composites scarcely decreases even when a total neutron fluence is 3.0 × 10¹⁸n/cm² (2.1 × 10¹⁸n/cm² for E > 0.1 MeV) for the 5 K irradiation and 1.6 × 10¹⁹n/cm² (8.0 f 10¹⁸ n/cm² for E > 0.1 MeV) for the room-temperature irradiation. The ultimate strength, however, decreases significantly at this neutron fluence for all the composites except the carbon/epoxy composite whose initial strength is comparatively low. Comparison of this result with that obtained for ⁶⁰Co γ-ray irradiation demonstrates that the radiation sensitivity of the glass/epoxy and glass/polyimide composites is 1.8–2.6 times higher towards neutrons than γ-rays. As to the irradiation temperature of 5 K and room temperature, no significant influence on the degradation efficiency of the composite strength is observed under the present conditions of mechanical test.     

Irradiation creep of previously irradiated 304 stainless steel in a fast flux environment

Results of a fast flux neutron irradiation experiment designed to investigate the effects of high levels of prior irradiation (to 10²³ n/cm², E > 0.1 MeV) on the irradiation creep of type 304 stainless steel at 700 K (800°F) are reported. The steady state creep coefficient is found to increase by a factor of 7 as the specimen fluence increases from 3 to 10 × 10²²n/cm², (E > 0.1 MeV). A non-linear dependence of strain on stress is exhibited. The results of this study are compared with previously reported stress relaxation data and with predictions of a swelling enhanced irradiation creep model.     

Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

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.     

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.     

Annealing of neutron damage in graphite irradiated and stored at room temperature

Nimonic PE16, a gamma-prime Ni₃(Al,Ti) precipitate-strengthened alloy under consideration for fast reactor structural applications, has been neutron irradiated in three heat treatment conditions: solution treated, aged, and overaged. After irradiation at 600° C to 5.4 × 10²²n/cm² (E > 0.1 MeV), or 27 dpa, specimens were characterized for gamma-prime precipitate stability by transmission electron microscopy. The precipitate microstructures after irradiation reflected the influence of the preirradiation heat treatment; and indeed the precipitate particles present prior to irradiation remained stable. However, additional precipitation occurred during irradiation in each of the specimens examined. The in-reactor gamma-prime precipitation process decorated such microstructural features as voids, dislocations and carbide precipitates. Examples were found in the solution-treated condition where gamma prime in the form of Archimedes' screws had precipitated on climbing screw dislocations. The precipitation behavior observed is compared with predictions from existing models. It is concluded that models for solute diffusion to point-defect sinks and for Ostwald coarsening can account for the observations, but that the models for precipitate stability controlled by cascade dissolution during neutron irradiation do not.     

The effect of irradiation and post-irradiation annealing on the yield stress and microstructure of vanadium-carbon alloys

Vanadium and vanadium-carbon alloys containing 0.15, 0.3 and 1 at% carbon were irradiated in JMTR with fast neutrons (E_n > 1 MeV) at 773 K to a dose of 5 × 10²⁴n/m². Tensile test and microstructural observations were carried out after irradiation and post-irradiation annealing. All of the. specimens showed radiation hardening. The irradiation produced voids, dislocations and radiation-induced quasi-carbides, which were formed by the agglomeration of vacancies and carbon atoms. The radiation anneal hardening in the alloys occurred at 873 K. The void number densities in the alloys had a peak at 873 K while the quasi-carbides decomposed at the same temperature. Therefore, invisible voids existing in the as-irradiated condition would grow by absorbing the vacancies, which were released in the process of decomposition of the quasicarbides during annealing, and the increase of the visible voids would effectively contribute to the radiation anneal hardening of these alloys.     

Void Swelling of Solution-Annealed Type 316 Stainless Steel under Long Time Proton Irradiation

Solution-annealed type 316 stainless steel was irradiated by 150 keV proton to a dose of about 6 dpa at the irradiation temperature ranging 450–700°C. To examine the effect of aging during irradiation, the present proton irradiation was carried out for about 25 h at a low dose rate of 7×10–^?5dpa/s. The specimens without He preinjection showed much smaller void swelling than those preinjected with He to the content of 10 at.ppm. Similarly to the case of neutron irradiations, the void swelling in the He preinjected specimens showed the temperature dependence with double peaks, and the peak swelling temperatures were about 550 and 650°C. In these specimens with He preinjection. void number density decreased and average void diameter increased with the increase of irradiation temperature in the range of 450–600°C, but these trends were reversed between 600 and 650°C. The volume of the grain boudary M₂₃C6 precipitates increased with the increase of irradiation temperature from 600 to 700°C, and it was concluded that the decrease of soluble carbon due to the precipitation of M₂₃C6 caused the second swelling peak at 650°C.     

Dose and Temperature Dependence of Void Swelling in Electron Irradiated Stainless Steel

In a high voltage electron microscope, solution treated Type 316 stainless steel was electron-irradiated at temperatures in the range of 370–630°C to a dose of about 30 dpa. The swelling (ΔV/V) induced by the irradiation beyond about 5 dpa is well described by an empirical equation, ΔV/V=A(dpa) ⁿ , under constant void and dislocation densities. With increasing irradiation temperature, the fluence exponent n increases and the pre-exponent term A decreases. At 550°C irradiation, the fluence exponent takes the value of 1.5 due to the diffusion-limited void growth. The value of n larger than 1.5 at higher temperature (>550°C) is attributable to the surface reaction-limited void growth. The smaller value of n for the low temperature (?500°C) irradiation appears to arise from the dislocation assisted vacancy diffusion. The peak swelling temperature of the specimen irradiated to 30 dpa is about 570°C, which shifts to a higher temperature with increase in electron dose.     

Void-swelling in 20% cold-worked FV548 austenitic stainless steel irradiated with 22 MeV C2+ ions or 46.5 MeV Ni6+ ions

A study has been made of the void-swelling behaviour of 20% cold-worked FV548 steel irradiated with 22 MeV C²⁺ or 46.5 MeV Ni⁶⁺ ions in the Harwell Variable Energy Cyclotron after room-temperature pre-injection with 10 ppm helium. The dose dependence of void-swelling under 46.5 MeV Ni⁶⁺ irradiation at 500 and 600°C, and the temperature dependence of void-swelling of specimens irradiated with 22 MeV C²⁺ ions in the range 500–650°C have been established. In addition the temperature dependence of void-swelling in specimens aged for 1000 h at 650° C before irradiation has also been studied. Subsidiary annealing experiments have demonstrated the high recovery resistance of the 20% cold-worked structure in FV548, compared with types 316 and 321 steels.The void-swelling behaviour is interpreted in terms of the balance between the dislocation and NbC point defect sink strengths and the observed resistance to recovery of the cold-worked structure in FV548.     

Neutron Irradiation Effects on Thermally Activated Process of Tensile Properties of Titanium

This paper deals with the relationship between mechanical properties and irradiation, effects for titanium irradiated to fast neutron fluxes. The neutron fluences applied are 6.9×10¹⁸, 8.6 × 10¹⁸ and 3.0 × 10¹⁹ n/cm². Tensile deformation is carried out over the temperature range of 77–about 600°K retaining the strain rate constant on one hand and changing the strain rate by a factor of about 5 and 10 on the other.

The fluence (φ) dependence of the yield stress at room temperature for an athermal component of the stress, σμ is greater than that for a thermal component σ* which does not change remarkably after irradiation. Their increments Δσ, Δμ and Δ σ* are proportional toσ ^1/3, σ^1/2σ^1/4 and, respectively.

The relationship between activation volume V* and effective shear stress τ* is investigated for both the unirradiated and irradiated specimens. In terms of the τ*/τ*₀ analysis (τ*_o is the value of τ* at T = 0°K), V* shows a tendency to decrease with increase in neutron fluence.

Irradiation defects observable by electron microscopy seem to be related to the athermal activation stress (σ_u) and those too small to be observed by electron microscopy to the thermal activation stress. The yield stress in the thermal activation can be given by Conrad's formula. The activation energy H₀ shows a constant value of about 1.8 eV irrespective of the neutron fluence applied. This value is 0.3–0.4eV higher than that for unirradiated specimens.     

Effect of Neutron-Irradiation on Thermal Conductivity,Electric Resistivity and Thermal Expansion of Boron Carbide

Boron carbide possessing porosities of 91 and 77%T.D. were irradiated at about 300°C to 6.0x10¹⁸～ 3.3×10¹⁹ n/cm² in helium atmosphere. The thermal conductivity, the electric resistivity and the linear thermal expansion were determined. Upon irradiation to 3.3×10¹⁹ n/cm², cracking occurred in all of the 91%T.D. specimens and in two of the five 77%T.D. specimens. The thermal conductivity changed with irradiation more markedly in the case of the 77%T.D. specimen, which also increased its specific electric resistivity by a factor of 10⁴, while that of the 91%T.D. specimen increased by less than 10-fold. This significant difference in thermal and electric conductivity change by irradiation is attributed to the difference in microstructure. It was observed that the recovery of radiation damage began at about 300°C when the specimens were subjected to heating and the changes observed in respect of thermal conductivity, electric resistivity and linear thermal expansion.