Danny J. Edwards

Influence of irradiation temperature and dose gradients on the microstructural evolution in neutron-irradiated 316SS

A cold worked 316SS baffle bolt was extracted from the Tihange pressurized water reactor and sectioned at three different positions. The temperature and dose at the 1-mm bolt head position were 593 K and 19.5 dpa respectively, whereas at two shank positions the temperature and dose was 616 K and 12.2 dpa at the 25-mm position and 606 K and 7.5 dpa at the 55-mm position. Microstructural characterization revealed that small faulted dislocation loops and cavities were visible at each position, but the cavities were most prominent at the two shank positions. Measurable swelling exists in the shank portions of this particular bolt, and accompanying this swelling is the retention of very high levels of hydrogen absorbed from the environment. The observation of cavities in the CW 316SS at temperatures and doses relevant to LWR conditions has important implications for pressurized water reactors since SA 304SS plates surround the bolts, a steel that usually swells earlier due to its lower incubation period for swelling.

Evolution of fine-scale defects in stainless steels neutron-irradiated at 275 C

Six austenitic stainless steel heats (three heats each of 304SS and 316SS) neutron-irradiated at 275 °C from 0.6 to 13.3 dpa have been carefully characterized by TEM and their hardness measured as a function of dose. The characterization revealed that the microstructure is dominated by a very high density of small Frank loops present in sizes as small as 1 nm and perhaps lower, which could be of both vacancy and interstitial-type. Frank loop density saturated at the lowest doses characterized, whereas the Frank loop size distributions changed with increasing dose from an initially narrow, symmetric shape to a broader, asymmetric shape. Although substantial hardening is caused by the small defects, a simple correlation between hardness changes and density and size of defects does not exist. These results indicate that radiation-induced segregation to the Frank loops could play a role in both defect evolution and hardening response.

Temperature and dose dependencies of microstructure and hardness of neutron irradiated OFHC copper

Abstract Tensile specimens of pure oxygen free high conductivity (OFHC) copper were irradiated with fission neutrons between 320 and 723 K to fluences in the range 5 × 10 21 to 1.5 × 10 24 n/m 2 ( E > 1 MeV) with a flux of 2.5 × 10 17 n/m 2 s. Irradiated specimens were investigated by transmission electron microscopy (TEM) and quantitative determinations were made of defect clusters and cavities. The dose dependence of tensile properties of specimens irradiated at 320 K was determined at 295 K. Hardness measurements were made at 295 K on specimens irradiated at different temperatures and doses. Microstructures of tensile tested specimens were also investigated by TEM. Results show that the increase in cluster density and hardening nearly saturate at a dose of ∼ 0.3 dpa. Irradiations at 320 K cause a drastic decrease in the uniform elongation already at ≌ 0.1 dpa. It is suggested that the irradiation-induced increase in the initial yield stress and a drastic decrease in the ability of copper to deform plastically in a homogeneous fashion are caused by a substantial reduction in the ability of grown-in dislocations to act as efficient dislocation sources.

Effect of neutron irradiation and post-irradiation annealing on microstructure and mechanical properties of OFHC-copper

Specimens of oxygen-free high conductivity (OFHC) copper were irradiated in the DR-3 reactor at Riso at 100 °C to doses in the range 0.01–0.3 dpa (NRT). Some of the specimens were tensile tested in the as-irradiated condition at 100 °C whereas others were given a post-irradiation annealing treatment at 300 °C for 50 h and subsequently tested at 100 °C. The microstructure of specimens was characterized in the as-irradiated as well as irradiated and annealed conditions both before and after tensile deformation. While the interstitial loop microstructure coarsens with irradiation dose, no significant changes were observed in the population of stacking fault tetrahedra (SFT). The post-irradiation annealing leads to only a partial recovery and the level of recovery depends on the irradiation dose level. However, the post-irradiation annealing eliminates the yield drop and reinstates enough uniform elongation to render the material useful again. These results are discussed in terms of the cascade-induced source hardening (CISH) model.

Effects of neutron irradiation on mechanical properties and microstructures of dispersion and precipitation hardened copper alloys

Tensile specimens of CuAl2O3, CuCrZr and CuNiBe alloys were irradiated with fission neutrons to fluences of 5 × 1022, 5 × 1023and1 × 1024n/m2 (E > 1MeV) at 47°C. Tensile properties and Vickers hardness were determined at 22°C. Microstructures of irradiated as well as unirradiated specimens were examined using a transmission electron microscope and the fractured surfaces were investigated in a scanning electron microscope. The most significant effect of irradiation is a drastic decrease in the ductility of copper alloys at a dose level as low as 0.2 dpa. The loss of ductility appears to be related to the intrinsic hardness of the grain interior and not to the grain boundary embrittlement. It is suggested that the irradiation-induced hardening and the lack of uniform elongation may be understood in terms of difficulty in dislocation generation due to pinning of grown-in dislocation by defect clusters (loops) and/or impurity atoms.

Austenitic stainless steels and high strength copper alloys for fusion components

Abstract An austenitic stainless steel (316LN), an oxide-dispersion-strengthened copper alloy (GlidCop Al25), and a precipitation-hardened copper alloy (Cu–Cr–Zr) are the primary structural materials for the ITER first wall/blanket and divertor systems. While there is a long experience of operating 316LN stainless steel in nuclear environments, there is no prior experience with the copper alloys in neutron environments. The ITER first wall (FW) consists of a stainless steel shield with a copper alloy heat sink bonded by hot isostatic pressing (HIP). The introduction of bi-layer structural material represents a new materials engineering challenge; the behavior of the bi-layer is determined by the properties of the individual components and by the nature of the bond interface. The development of the radiation damage microstructure in both classes of materials is summarized and the effects of radiation on deformation and fracture behavior are considered. The initial data on the mechanical testing of bi-layers indicate that the effectiveness of GlidCop Al25 as a FW heat sink material is compromised by its strongly anisotropic fracture toughness and poor resistance to crack growth in a direction parallel to the bi-layer interface.

Effects of heat treatments and neutron irradiation on microstructures and physical and mechanical properties of copper alloys

Abstract Tensile specimens of CuAl 2 O 3 , CuCrZr and CuNiBe alloys were given different heat treatments and then irradiated with fission neutrons at 250°C to a dose level of ≅ 0.3 dpa. Both unirradiated and irradiated specimens were tensile tested at 250°C. The microstructure and electrical resistivity were determined in the unirradiated as well as irradiated conditions. The post-deformation microstructure and fracture surfaces were also investigated. The main effect of the bonding thermal cycle heat treatment was a slight decreased in the strength of CuCrZr and CuNiBe alloys. The strength of CuAl-25, on the other hand, remained almost unaltered. The post-irradiation tests at 250°C showed a severe loss of ductility in the case of CuNiBe alloy. The irradiated CuAl-25 and CuCrZr specimens, on the other hand, exhibited a reasonable amount of uniform elongation. The results are briefly discussed in terms of thermal and irradiation stability of precipitates and particles and irradiation-induced segregation, precipitation and recovery of dislocation microstructure.

Microstructure and mechanical behaviour of TZM and Mo-5% Re alloys irradiated with fission neutrons

Abstract The response of microstructural and mechanical properties of TZM and Mo-5% Re alloys to neutron irradiation are reported. Irradiations were carried out at five temperatures between 323 and 723 K to a dose level of ∼ 0.16 dpa. The resulting microstructures consisted of a high density of small loops in both alloys with the addition of characteristic loop rafting in TZM. There was a marked sensitivity of microstructure to temperature. Void formation was found in both TZM and Mo-5% Re after the 623 and 723 K irradiations. The irradiations induced a significant increase in Vickers hardness in all cases but with no obvious dependence on irradiation temperature. Even at the present low neutron dose, large increases in tensile strength and a drastic decrease in ductility were observed. These results are discussed in terms of source hardening of the matrix within grains rather than obstacle hardening. Although the fracture processes are concentrated at the grain boundaries, these are not due to grain boundary embrittlement.

Low-temperature radiation embrittlement of copper alloys

The effect of low-temperature (T{sub irr} less than 0.3. T{sub melt}) irradiation on the tensile properties of copper and precipitation-hardened (PH) and dispersion-strengthened (DS) copper alloys was investigated. Samples were irradiated with fission neutrons at temperatures of 80 to 200{degrees}C to doses of 0.6 to 5 dpa. Irradiation at temperatures <150{degrees}C resulted in significant hardening and accompanying embrittlement in all of the materials. By comparing the present results with literature data, it is concluded that severe radiation embrittlement occurs in copper alloys irradiated at temperatures {le}I00{degrees}C for doses above {approximately} 0.01 to 0.1 dpa. On the other hand, irradiation at temperatures above 150{degrees}C causes only moderate embrittlement for doses up to {approximately}5 dpa. It is recommended that the minimum operating temperature for copper alloys intended for structural applications in fusion energy systems should be 150{degrees}C, unless uniform elongations

The effect of neutron spectrum on the mechanical and physical properties of pure copper and copper alloys

The electrical resistivity and tensile properties of copper and oxide dispersion strengthened (DS) copper alloys have been measured before and after fission neutron irradiation to damage levels of 0.5 to 5 displacements per atom (dps) at {approximately}100 to 400{degrees}C. Some of the specimens were irradiated inside a 1.5 mm Cd shroud in order to reduce the thermal neutron flux. The electrical resistivity data could be separated into two components, a solid transmutation component {Delta}{rho}{sub tr} which was proportional to thermal neutron fluence and a radiation defect component {Delta}{rho}{sub rd} which was independent of the displacement dose. The saturation value for {Delta}{rho}{sub rd} was {approximately}1.2 nanohm-meters for pure copper and {approximately}1.6 nanohm-meters for the DS copper alloys irradiated at 100{degrees}C in positions with a fast-to-thermal neutron flux ratio of 5. Considerable radiation hardening was observed in all specimens at irradiation temperatures below 200{degrees}C. The yield strength was relatively insensitive to neutron spectrum in specimens strengthened by dispersoids or cold- working. 17 refs., 7 figs., 1 tab.