Irradiation hardening in unalloyed and ODS molybdenum during low dose neutron irradiation at 300 °C and 600 °C

Unalloyed molybdenum and oxide dispersion strengthened (ODS) molybdenum were irradiated at 300 °C and 600 °C in HFIR to neutron fluences of 0.2, 2.1, and 24.3 × 10²⁴ n/m² (E > 0.1 MeV). The size and number density of voids and loops as well as the measured irradiation hardening and electrical resistivity were found to increase sub-linearly with fluence. This supports the idea that the formation of the extended defects that produce irradiation hardening in molybdenum is the result of a nucleation and growth process rather than the formation of sessile defects directly from the displacement damage cascades. This conclusion is further supported by molecular dynamics (MD) simulations of cascade damage. The unalloyed molybdenum had a low impurity interstitial content with less irradiation hardening and lower change in electrical resistivity than is observed for ODS Mo. This result suggests that high-purity can result in slightly improved resistance to irradiation embrittlement in molybdenum at low fluences.     

Microstructural analysis of an HT9 fuel assembly duct irradiated in FFTF to 155 dpa at 443 °C

The majority of data on the irradiation response of ferritic/martensitic steels has been derived from simple free-standing specimens irradiated in experimental assemblies under well-defined and near-constant conditions, while components of long-lived fuel assemblies are more complex in shape and will experience progressive changes in environmental conditions. To explore whether the resistance of HT9 to void swelling is maintained under more realistic operating conditions, the radiation-induced microstructure of an HT9 ferritic/martensitic hexagonal duct was examined following a six-year irradiation of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). The calculated irradiation exposure and average operating temperature of the duct at the location examined were ∼155 dpa at ∼443 °C. It was found that dislocation networks were predominantly composed of (a/2)<1 1 1> Burgers vectors. Surprisingly, for such a large irradiation dose, type a<1 0 0> interstitial loops were observed. Additionally, a high density of precipitation occurred. These two microstructural characteristics may have contributed to the rather low swelling level of 0.3%.     

Creep rupture of a 9Cr1MoNbV steel at 500 °C: Base metal and welded joint

ASME Grade 91 steel base metal and a similar weld were tested under creep at 500 °C for rupture time up to 18,000 h. Creep failure of cross-weld specimens occurs in the weld metal at this temperature. No significant microstructural changes were observed after creep. Analysis of creep deformation of smooth creep bars, welded joints and slightly notched bars indicated an apparent creep stress exponent of 19. For the creep conditions considered, failure of the material can be explained by the viscoplastic instability of the specimens without significant damage development. This allowed to develop a simple analysis for time to failure prediction.     

Nanoscale characterization of HT-9 exposed to lead bismuth eutectic at 550 °C for 3000 h

Fast reactors and targets in spallation neutron sources may use lead bismuth eutectic (LBE) as a coolant. Its physical and chemical properties and irradiation properties make it a safe and high performance coolant in radiation environments. However, LBE is a corrosive medium for most steels. In the present study, the atomic force microscopy, magnetic force microscopy, conductive atomic force microscopy, surface potential microscopy, and scanning electron analysis with energy dispersive X-ray spectroscopy were performed to get a better understanding of the corrosion and oxidation mechanism of the HT-9 stainless steel in an LBE environment. What was believed in the past to be simply a double oxide layer structure was revealed here to be more complicated. It is found that the inner most oxide layer maintains the grain structure of what used to be the bulk steel material while the outer oxide layer possessed a columnar structure. The EDS line scans and the conductive and magnetic properties measured using the scanning probe techniques give us the local properties of the formed oxide layers. This leads to a more detailed view of the oxide layers formed on HT-9 in LBE.     

Tensile behaviour of 9Cr-1Mo tempered martensitic steels irradiated up to 20 dpa in a spallation environment

Tensile specimens of 9Cr-1Mo (EM10) and mod 9Cr-1Mo (T91) martensitic steels in the normalized and tempered metallurgical conditions were irradiated with high energy protons and neutrons up to 20 dpa at average temperatures up to about 360 °C. Tensile tests were carried out at room temperature and 250 °C and a few samples were tested at 350 °C. The fracture surfaces of selected specimens were characterized by Scanning Electron Microscopy (SEM). While all irradiated specimens displayed at room temperature considerable hardening and loss of ductility, those irradiated to doses above approximately 16 dpa exhibited a fully brittle behaviour and the SEM observations revealed significant amounts of intergranular fracture. Helium accumulation, up to about 0.18 at.% in the specimens irradiated to 20 dpa, is believed to be one of the main factors which triggered the brittle behaviour and intergranular fracture mode. One EM10 and one T91 specimen irradiated to 20 dpa were annealed at 700 °C for 1 h following irradiation and subsequently tensile tested. In both cases, a remarkable recovery of ductility and strain-hardening capacity was observed after annealing, while the strength remained significantly above that of the unirradiated material.     

Corrosion experiments of steels in flowing Pb at 500 °C and in flowing LBE at 450 °C

In accelerator driven systems (ADS), as well as in the next Generation IV reactors, one of the concerned issues is the material compatibility and corrosion in liquid Pb, which is considered a candidate coolant. Liquid metal corrosion of the structural materials can proceed via different processes: species dissolution and penetration of liquid metal along grain boundaries and metal. The occurrence of these corrosion phenomenon depend on the experimental parameters, such as temperature, thermal gradients, solid and liquid metal compositions, velocity of the liquid metal and oxygen activity in Pb. One possible technique to prevent any corrosive attack by the liquid metals is the in situ passivation of the containment steels. This technique is achieved through an active control and monitoring of the dissolved oxygen concentration. This paper summarizes the data gathered from the CHEOPE III loop, where passivation of T91 and AISI 316L steels is tested in pure Pb at 500 °C were carried out, comparing them with preliminary corrosion data, in LBE, gathered from the LECOR loop.     

Effects of Fe and Cr on corrosion behavior of ZrFeCr alloys in 500 °C steam

A study of the corrosion behaviors of ZrFeCr alloy and the influence of microstructure on corrosion resistance are described by X-ray diffraction and scanning electron microscope in this paper. The results show that several ZrFeCr alloys exhibit protective behavior throughout the test and oxide growth is stable and protective. The best alloy has the composition Zr1.0Fe0.6Cr. Fitting of the weight gain curves for the protective oxide alloys in the region of protective behavior, it showed nearly cubic behavior for the most protective alloys. The Zr1.0Fe0.6Cr has the more laves Zr(Fe,Cr)₂ precipitate in matrix and it has the better corrosion resistance. The Zr0.2Fe0.1Cr has little precipitate, the biggest hydrogen absorption and the worst corrosion resistance. The number of precipitates and the amount of hydrogen absorption in Zr alloy plays an important role on corrosion resistance behaviors in 500 °C/10.3 MPa steam.     

Creep-to-rupture tests of T91 steel in flowing Pb-Bi eutectic melt at 550 °C

Uniaxial creep-to-rupture tests were performed on T91 in air and in flowing lead-bismuth eutectic melts. Compared to specimens tested in air, the specimens tested in liquid-metal show: (i) strain and strain rate increase up to a factor of about 50 (strain rate); (ii) time-to-rupture decrease; (iii) rapid transition into the third creep stage at high stress (above 180 MPa). The analysis of the test results revealed several important surface phenomena, which lead to different behavior of the specimens tested in lead-bismuth eutectic melts compared to those tested in air. Under high stress, and therefore high strain, the crack propagation process is mostly controlled by the reduction of the surface energy due to Pb and Bi adsorption on the steel surface. Under low stress (140 and 160 MPa) and low strain, this process is delayed due to the competing mechanism of healing the oxide scale cracks.     

Low cycle fatigue of irradiated and unirradiated Eurofer97 steel at 300 °C

Low cycle fatigue results are reported for unirradiated and irradiated reduced activation ferritic martensitic steel Eurofer97. The neutron irradiation experiment (irradiation at 300 °C to a nominal dose of 2.5 dpa) has been performed in the High Flux Reactor, Petten, the Netherlands. Post-irradiation low cycle fatigue tests have been performed in air at 300 °C at a total strain range of 0.6%, 1.0% and 1.4%. Neutron irradiation at 300 °C resulting in irradiation hardening is found to be beneficial for fatigue life at low strain amplitudes and to be adverse at high strain amplitudes. No effect of the different technological product forms on the fatigue life in Eurofer97 is observed, and fatigue behavior of Eurofer97 steel is found to be similar to that of F82H steel.     

Corrosion behavior of CLAM steel weldment in flowing liquid Pb-17Li at 480 °C

CLAM (China Low Activation Martensitic) steel is considered as one of the candidate structural materials in liquid LiPb blanket concepts. Welding is one of the essential technologies for its practical application, CLAM steel weldment shows a great difference with base metal due to the effect of welding thermal cycle. In order to investigate the corrosion behavior and mechanism of CLAM weldments in liquid Pb-17Li, the experiments were performed by exposing the TIG weldment samples in flowing LiPb at 480 °C. The weight loss test of exposed specimens show that in 500 h, 1000 h dynamic conditions, corrosion resistance of CLAM steel weldment is poor, SEM analysis shows that the thicker martensite lath in weld area lead to higher corrosion amount, EDS results show that the influence of corrosion on surface elements is small, and surface corrosion is even, EDX analysis shows that the penetration of Pb-17Li does not exist in the joint. With the increasing of exposure time, the corrosion rate decreases. Metallographic analysis shows that the presence of Cr has great influence on the corrosion resistance of the steel matrix. The area short of Cr in thick martensite lath of CLAM steel weldment is easily corroded. After a series of theoretical and experimental analysis, a basic presumably corrosion behavior model is established, which makes contributions to the in-depth understanding of the corrosion mechanism of CLAM weldments.     

Radiation stability of ZrN under 2.6 MeV proton irradiation

Zirconium nitride is a promising alternative material for the use as an inert matrix for transuranic fuel, but the knowledge of the radiation tolerance of ZrN is very limited. We have studied the radiation stability of ZrN using a 2.6 MeV proton beam at 800 °C. The irradiated microstructure and hardening were investigated and compared with annealed samples. A high density of nano-sized defects was observed in samples irradiated to doses of 0.35 and 0.75 dpa. Some defects were identified as vacancy-type pyramidal dislocation loops using lattice resolution imaging and Fourier-filter image processing. A very slight lattice expansion was noted for the sample with a dose of 0.75 dpa. Hardening effects were found for samples irradiated to both 0.35 and 0.75 dpa using Knoop indentation.     

Thermo-mechanical characteristics of concrete at elevated temperatures up to 310 °C

Both, the normal strength concretes (NSC) and high strength concretes (HSC) have been used in structures which may be exposed to elevated temperatures. Concretes have also been used in the construction of radiation shielding structures. Data on the behaviour of concrete at high temperature is of prime concern in predicting the constructions and safety of buildings in response to certain accidents or particular service conditions. Prediction of mechanical behaviour, thermo-mechanical deformations and moisture migration in non-uniformly heated concrete is important for safe operation of concrete containment.This paper presents the results of an experimental investigation carried out to predict the behaviour of concrete intended for nuclear applications. For this purpose, normal concrete having compressive strength of 40 MPa was designed using limestone aggregates. Cylindrical specimens (110 mm × 22 mm) were made and subjected to heating-cooling cycles at 110, 210 and 310 °C. Measurements were taken for thermal gradient, mass loss, deformations, residual mechanical properties, thermal conductivity, and porosity. This investigation developed some important data on the properties of concrete exposed to elevated temperatures up to 310 °C. Comparisons and interesting conclusions were drawn about the thermal stability at high temperature and the residual mechanical properties of the tested concrete.     

Mechanical properties of steel fiber reinforced reactive powder concrete following exposure to high temperature reaching 800 °C

This study investigates the stress-strain relation of RPC in quasi-static loading after an elevated temperature. The cylinder specimens of RPC with φ 50 mm × 100 mm are examined at the room temperature and after 200-800 °C. Experimental results indicate that the residual compressive strength of RPC after heating from 200-300 °C increases more than that at room temperature, but, significantly decreases when the temperature exceeds 300 °C. The residual peak strains of RPC also initially increase up to 400-500 °C, then decrease gradually beyond 500 °C. Meanwhile, Young's modulus diminishes with an increasing temperature. Based on the regression analysis results, this study also develops regression formulae to estimate the mechanical properties of RPC after an elevated temperature, thus providing a valuable reference for industrial applications and design.     

Tensile properties of EC316LN irradiated in SINQ to 20 dpa

The wide application of 316-type austenitic stainless steels in existing spallation targets requires a comprehensive understanding of their behavior in spallation irradiation environments. In the present study, EC316LN specimens were irradiated in SINQ targets to doses between 3 and 17.3 dpa at temperatures between about 80 °C and 390 °C. Tensile tests were conducted at room and irradiation temperatures. The results demonstrate that the irradiation induced significant hardening and embrittlement in the specimens. The irradiation hardening and embrittlement effects show a trend of saturation at doses above about 10 dpa. Although the ductility was greatly reduced, all specimens broke with strong necking, which indicates a ductile fracture mode.     

380 keV proton irradiation effects on photoluminescence of Eu-doped GaN

The effect of 380 keV proton irradiation on the photoluminescence (PL) properties has been investigated for undoped and Eu-doped GaN. As the proton irradiation exceeds , a drastic decrease of PL intensity of the near band-edge emission of undoped GaN was observed. On the other hand, for Eu-doped GaN, the PL emission corresponding to the ⁵D₀→⁷F₂ transition in Eu³⁺ kept the initial PL intensity after the proton irradiation up to . Present results, together with our previous report on electron irradiation results, suggest that Eu-doped GaN is a strong candidate for light emitting devices in high irradiation environment.     

Fracture toughness of the hydrogen charged EUROFER 97 RAFM steel at room temperature and 120 °C

The static fracture toughness of EUROFER 97 reduced activation ferritic-martensitic steel was investigated in presence of higher content of hydrogen. The hydrogen effect is shown during fracture toughness testing both of base and weld metals at room temperature and at 120 °C. At the room temperature testing the J_0.2 integral values will decrease depending on hydrogen content in the range of 2-4 wppm. The same hydrogen content of 2 wppm manifests itself by an uneven level of hydrogen embrittlement for base metal and weld metal. This corresponds to a different J_0.2 integral value and a different mechanism of fracture mode. At the hydrogen content of 4 wppm more evident decrease of J_0.2 was observed for both metals. At 120 °C hydrogen decreases J_0.2 integral in base metal at a limited scale only in comparison to weld metal. At room temperature and hydrogen content of about 4 wppm the base metal specimen exhibits inter-granular fracture and trans-granular cleavage on practically the whole crack surface. The weld metal fracture has shown inter-granular and trans-granular mechanism with ductile and dimple rupture.     

Enhanced susceptibility of CaF2(1 1 1) to adsorption due to ion irradiation

We have investigated morphological changes of freshly cleaved CaF₂(1 1 1) single crystal surfaces before and after ion irradiation. We show that with or without irradiation the surface undergoes serious changes within minutes after the cleavage if the samples are exposed to ambient conditions. This is most likely due to the adsorption of water and could be avoided only if working under clean ultra-high-vacuum conditions. Ion-induced modifications on this surface seem to act as centers for an increased rate of adsorption so that any quantitative numbers obtained by atomic force microscopy in such experiments have to be treated with caution.     

Temperature dependence of lattice disorder in Ar-irradiated (1 0 0), (1 1 0) and (1 1 1) MgO single crystals

To better appreciate dynamic annealing processes in ion irradiated MgO single crystals of three low-index crystallographic orientations, lattice damage variation with irradiation temperature was investigated. Irradiations were performed with 100 keV Ar ions to a fluence of 1 × 10¹⁵ Ar/cm² in a temperature interval from −150 to 1100 °C. Rutherford backscattering spectroscopy combined with ion channeling was used to analyze lattice damage. Damage recovery with increasing irradiation temperature proceeded via two characteristic stages separated by 200 °C. Strong radiation damage anisotropy was observed at temperatures below 200 °C, with (1 1 0) MgO being the most radiation damage tolerant. Above 200 °C damage recovery was isotropic and almost complete recovery was reached at 1100 °C. We attributed this orientation dependence to a variation of dynamic annealing mechanisms with irradiation temperature.     

Nano-cavities observed in a 316SS PWR flux thimble tube irradiated to 33 and 70 dpa

The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290 °C and 70 dpa at 315 °C were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during post-irradiation slow strain rate testing in PWR water conditions.     

Transmission electron microscopy of oxide development on 9Cr ODS steel in supercritical water

Oxide layers formed on 9Cr oxide dispersion strengthened ferritic steel alloys during exposure to 600 °C supercritical water for 2- and 4-weeks were examined using cross-sectional transmission electron microscopy. A focused ion beam insitu lift-out technique was used to produce site-specific samples with electron transparent areas up to 8 μm by 10 μm. The oxide layers consist of several sub-layers: an Fe-rich outer oxide, a Cr-rich inner oxide, and a diffusion layer, extending beyond the oxide front into the metal. An evolution of the oxide layer structure is seen between 2 and 4 weeks, resulting in the development of a band of Cr₂O₃ at the diffusion layer/metal interface from the previously existing continuous mixture of FeCr₂O₄ ‘fingers’ and bcc metal. It is believed that transport in this Cr₂O₃ layer at the diffusion layer/metal interface becomes the rate-limiting step for oxide advancement, since this change in oxide structure also corresponds to a decrease in corrosion rate.