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.     

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.     

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.     

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.     

Electrochemical behaviour of gadolinium ion in molten LiCl-KCl eutectic

This work presents the electrochemical study of GdCl₃ in the molten LiCl-KCl eutectic in the temperature range 723-823 K. Transient electrochemical techniques such as cyclic voltammetry and chronopotentiometry, on an inert metallic tungsten working electrode, have been used in order to investigate the reduction mechanism and transport parameters. This study shows that Gd³⁺ ions are reduced to Gd metal by a single step mechanism with exchange of three electrons. Diffusion coefficient of GdCl₃ ions was determined at various temperatures, at 723 K the value is D = 0.88 10⁻⁵ cm² s⁻¹. Apparent standard reduction potential of the redox couple Gd³⁺/Gd has been determined by the open-circuit chronopotentiometry technique at several temperatures. Also the Gibbs free energy of GdCl₃ formation was determined and compared with thermodynamic data for pure compounds in the supercooled state in order to estimate the activity coefficient of Gd³⁺ in the molten LiCl-KCl eutectic.     

Be AMS measurements at low energies (E < 1 MeV)

The longlived isotope ¹⁰Be is of great importance in earth sciences for dating applications, reconstruction of the solar activity or in climate research. Routine AMS measurements with BeO samples are performed on accelerators with a terminal voltage above 2 MV. Applying the degrader foil technique for boron suppression, first tests with BeO samples on the 0.6 MV ETH/PSI machine were limited by background to a ¹⁰Be/⁹Be ratio of 10⁻¹³. The background was identified as ⁹Be which reaches the detector by scattering processes. By applying an additional magnetic mass filter to the high energy mass spectrometer the background was effectively removed. A ¹⁰Be/⁹Be background ratio of <5 × 10⁻¹⁵ was achieved. The overall efficiency (detected ¹⁰Be compared to BeO⁻ injected into the accelerator) was 7-8%.     

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between −160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between −170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.     

Study of thermodynamic properties of Np-Al alloys in molten LiCl-KCl eutectic

Pyrochemical methods are investigated worldwide within the framework of Partitioning and Transmutation concepts for spent nuclear fuel reprocessing. Electroseparation techniques in a molten LiCl-KCl are being developed in ITU to recover all actinides from a mixture with fission products. During the process, actinides are selectively electrochemically reduced on a solid aluminium cathode, forming solid actinide-aluminium alloys. This work is focused on the thermodynamic properties of Np-Al alloys in a temperature range of 400-550 °C and on the characterisation of the structure and chemical composition of deposits obtained by electrodeposition of Np on solid Al electrodes in a LiCl-KCl-NpCl₃ melt. Cyclic voltammetry and open circuit chronopotentiometry have been used to examine the electrochemical behaviour of Np on inert W and reactive Al electrodes. Gibbs energies, enthalpy and entropy of formation and standard electrode potentials of Np-Al alloys were evaluated and compared with ab initio calculations. Galvanostatic electrolyses at 450 °C were carried out to recover Np onto Al plates and the solid surface deposits were characterised by XRD and SEM-EDX analyses. Stable and dense deposits consisting of NpAl₃ and NpAl₄ alloys were identified. In addition, the conversion of NpO₂ to NpCl₃ is described, using chlorination of the oxide in a molten salt media by pure chlorine gas.     

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.     

Immobilisation of I by encapsulation in tin by hot-pressing at 200 °C

The volatility of iodine-129 and its soluble nature in anionic form makes it very difficult to incorporate in ceramic or glassy solids for the purpose of long-term immobilisation. Thus encapsulation in a low-melting metal such as tin is an attractive option, and we describe experiments in which we have hot-pressed AgI-bearing alumina beads surrounded by tin powder at 200 °C.     

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%.     

Leach testing at 50 °C of α-doped SON68 glass alteration gels

Gels formed by altering α-doped (Np, Pu, Am) SON68 glass at 300 °C were leached during 1 year at 44 cm⁻¹ and 50 °C under oxidizing conditions (Eh_/NHE ≈ +150 mV) and under reducing conditions (Eh_/NHE ≈ −250 mV). After 3 days of leaching the gel dissolution was highly incongruent. The gel dissolution rate calculated from the silicon concentrations was 4.4 × 10⁻⁵ g m⁻² d⁻¹, except for the Am-doped gel, for which the rate was two times higher. During leaching, Np is weakly retained in the gel (35% under oxidizing conditions and 50% under reducing conditions) whereas Pu and Am are strongly retained (over 90%). The three lanthanides La, Ce, and Nd exhibit exactly the same leaching behavior, but different from that of actinides. Speciation and complexation calculations for neodymium showed that its solubility could be controlled by Nd(OH)₃ for periods beyond 3 months. Conversely, no simple chemical compound appears to control the solubility of the actinides.     

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.     

Microstructural characterization of U-7Mo/Al-Si alloy matrix dispersion fuel plates fabricated at 500 °C

The starting microstructure of a dispersion fuel plate will impact the overall performance of the plate during irradiation. To improve the understanding of the as-fabricated microstructures of U-Mo dispersion fuel plates, particularly the interaction layers that can form between the fuel particles and the matrix, scanning electron microscopy (SEM) and transmission electron microscopy (TEM) analyses have been performed on samples from depleted U-7Mo (U-7Mo) dispersion fuel plates with either Al-2 wt.% Si(Al-2Si) or AA4043 alloy matrix. It was observed that in the thick interaction layers, U(Al, Si)3 and U6Mo4Al43 were present, and in the thin interaction layers, (U, Mo) (Al, Si)3, U(Al, Si)4, U3Si3Al2, U3Si5, and possibly USi-type phases were observed. The U3Si3Al2 phase contained some Mo. Based on the results of this investigation, the time that a dispersion fuel plate is exposed to a relatively high temperature during fabrication will impact the nature of the interaction layers around the fuel particles. Uniformly thin, Si-rich layers will develop around the U-7Mo particles for shorter exposure times, and thicker, Si-depleted layers will develop for the longer exposure times.     

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.     

Structural recovery of self-irradiated natural and Pu-doped zircon in an acidic solution at 175 °C

We have investigated the aqueous stability of self-irradiated natural and synthetic ²³⁸Pu-doped zircon (4.7 wt% of ²³⁸Pu) in an acidic solution at 175 °C. Both zircon samples have suffered a similar degree of self-irradiation damage, as given by their degree of amorphization. X-ray diffraction measurements revealed that during the hydrothermal treatment only the disordered crystalline remnants recovered in the natural zircon, whereas in the ²³⁸Pu-doped zircon the amorphous phase strongly recrystallized. Such a different alteration behavior of natural and Pu-doped zircon is discussed in terms of two fundamentally different alteration mechanisms. Our results demonstrate that further experimental studies with Pu-doped zircon are required before any reliable prediction about the long-term aqueous stability of an actinide waste form based on zircon can be made.     

Isotropic pyrocarbon deposited at 1250 °C by means of thermal gradient chemical vapor deposition

Isotropic pyrocarbon was fabricated at 1250 °C via thermal gradient chemical vapor deposition. The density of the deposit was measured using Archimedes methods and the microstructure was examined by polarized light microscope and scanning electron microscope. The results showed that the density of the deposit was about 1.85 g/cm³ and it was constituted of spherical pyrocarbon grains which was about 0.6-0.8 μm in diameter and compacted densely and there are some microcracks in the deposit.     

Low cycle fatigue tests of surface modified T91 steel in 10 wt% oxygen containing Pb45Bi55 at 550 °C

Low cycle fatigue tests in air and LBE containing 10⁻⁶ wt% dissolved oxygen were conducted with T91 steel at 550 °C. T91 was employed in two modifications, one in the as-received state, and the other after alloying FeCrAlY into the surface by pulsed electron beam treatment (GESA process). Tests were carried out with symmetrical cycling (R = −1) with a frequency of 0.5 Hz and a total elongation Δε_t/2 between 0.3% and 2%. No influence from LBE on fatigue could be detected. Results in air and LBE showed similar behaviour. Additionally, no difference was observed between surface treated and none treated T91 specimens.     

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.     

Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °C proton irradiation

Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350-700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y-Ti-O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster-matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.