Behaviour of chromium steels in liquid Pb-55.5Bi with changing oxygen content and temperature

The behaviour of protective oxide layers on P122 steel and its welds and of ODS steel in liquid Pb_44.5Bi_55.5 (LBE) is examined under conditions of changing temperatures and oxygen concentrations. P122 (12Cr) and its welded joints are exposed to LBE at 550 °C for 4000 h with oxygen concentrations of 10⁻⁶ and 10⁻⁸ wt% (p(O₂) = 8.1 × 10⁻²³ bar and 5.2 × 10⁻²⁷ bar) which change every 800 h. It is found that like in case of constant oxygen concentration of 10⁻⁶ wt% a protective spinel layer (Fe(Fe_1−xCr_x)₂O₄) was maintained on P122 and also on its welded joint. Two experiments with exposure times of 4800 h are conducted on ODS steel, both with temperatures changing from 550 to 650 °C and back every 800 h, one experiment with 10⁻⁶ the other with 10⁻⁸ wt% oxygen in LBE. Both experiments show strong local dissolution attack after 4800 h which is in agreement with the behaviour of ODS in LBE at a constant temperature of 650 °C. However, dissolution attack is less in LBE with 10⁻⁸ wt% oxygen (p(O₂) = 3.0 × 10⁻²⁵ bar).     

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

Mechanical and thermo-physical properties of refractory metal alloys and mechanically alloyed (MA)-oxide dispersion strengthened (ODS) steels are reviewed and their potential for use in space nuclear reactors is examined. Preferable refractory alloys for use in liquid metal and gas-cooled space reactors include Nb-1%Zr, PWC-11, Mo-TZM, Mo-xRe where x varies from 7% to 44.5%, T-111 and ASTAR-811C. These alloys are heavy, difficult to fabricate, and are not readily available. The advantages of the MA-ODS alloys are: (a) their strength at high temperatures (>1000 K), which decreases slower with temperature than those of niobium and molybdenum alloys; (b) relatively lightweight and less expensive; (c) low swelling and no embrittlement with exposure to high-energy neutrons (>0.1 MeV) up to 10²⁷ n/m²; and (d) high resistance to oxidation and nitration. The few data available on compatibility of MA-ODS alloys with alkali liquid metals up to 1100 K are encouraging, however, additional tests at typical temperatures (1000-1400 K) in space nuclear reactors are needed. The anisotropy of MA-ODS alloys when cold worked, and particularly rolled into tubes, should not hinder their use in space nuclear power systems, in which operation pressure is either near atmospheric or as high as 2 MPa, but joints weldability is an issue.     

Determining the shear properties of the PyC/SiC interface for a model TRISO fuel

The fracture behavior of TRISO-coated fuel particles is dependent on the shear strength of the interface between the inner pyrolytic carbon (PyC) and silicon carbide coatings. This study evaluates the interfacial shear properties and the crack extension mechanism for TRISO-coated model tubes using a push-out technique. The interfacial debond shear strength was found to increase with increasing sample thickness and finally approached a constant value. The intrinsic interfacial debond shear strength of ∼280 MPa was estimated. After the layer is debonded, the applied load is primarily transferred by interfacial friction. A non-linear shear-lag model predicts that the residual clamping stress at the interface is ∼350 MPa, and the coefficient of friction is ∼0.23, yielding a frictional stress of ∼80 MPa. These relatively high values are attributed to the interfacial roughness. Of importance in these findings is that this unusually high interfacial strength could allow significant loads to be transferred between the inner PyC and SiC in application, potentially leading to failure of the SiC layer.     

Experimental studies of mechanical properties of solid zirconium hydrides

Zirconium alloy Zr-2.5Nb has been hydrided to ZrH_x (x = 1.15-2.0), and studied using microhardness and unconfined and confined compression techniques. At room temperature, results on Young’s modulus and yield strength of solid hydrides show that these mechanical properties remain about the same as the original zirconium alloy for hydrogen compositions up to about ZrH_1.5. The levels of these properties start to drop when δ hydride becomes the major phase and reaches a minimum for the ε hydride phase. Between room temperature and 300 °C, Young’s modulus of solid hydrides decreases with temperature at about the same rate as it does for the original zirconium alloy.     

Effect of the irradiation-annealing-irradiation cycle on the mechanical properties of pure copper and copper alloy

The report presents the results of two irradiation experiments. The first experiment was carried out in the SM-2 reactor with the aim to study the effect of single annealing after irradiation on mechanical properties of pure Cu and GlidCopAl25IG alloy. The aim of the second experiment performed in the RBT-6 reactor was to investigate the effect of the irradiation-annealing-irradiation (IAI) cycle. Pure Cu and GlidCopAl25IG alloy specimens were irradiated in the SM-2 and RBT-6 reactors to ?10⁻³, 10⁻² and 10⁻¹ dpa at T_irr=80 °C. The investigations performed revealed that IAI cycles do not cause an accumulation of embrittlement of pure copper and GlidCopAl25IG alloy in the cycles. The experiments lead to the conclusion that the regime of intermediate annealing produces the structure in the material (relatively low density of SFT), sufficiently insensitive to subsequent irradiation (at a low dose level ?10⁻² dpa).     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

Post irradiation plastic properties of F82H derived from the instrumented tensile tests

F82H (Fe-8Cr-2W) and its variant doped with 2%Ni were irradiated up to 20 dpa at 300 °C in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Post irradiation tensile testing was performed at room temperature. During testing, the images of the specimens including the necked region were continuously recorded. Tests on cold worked material were also carried out for comparison. From the load-displacement curves and the strain distributions obtains from the images, flow stress levels and strain hardening behavior was evaluated. A preliminary constitutive equation for the plastic deformation of irradiated F82H is presented. The results suggest that the irradiation mainly causes defect-induced hardening while it did not strongly affect strain hardening at the same flow stress level for F82H irradiated at 300 °C. The strain hardening of Ni doped specimens was, however, strongly affected by irradiation. Results provide basics to determine allowable stress levels at temperatures below 400 °C.     

The preparation of fine-grain doped graphite and its properties

Fine-grain doped graphite was prepared by the ball-milling dispersion method for the first time. Such composite has not only high thermal conductivity and excellent bending strength (116 MPa), but also better oxidation resistance at elevated temperature and outgassing properties than those of composite doped with normal size carbides. Correlations between microstructure and properties of such composites are also discussed in detail.     

The neutron irradiation effect on mechanical properties of HIP joint material

Dispersion strengthened copper (DSCu) and stainless steel are the candidate material for the heat sink and the structural material of the ITER shielding blanket and these materials are joined by hot isostatic pressing (HIP). In this study, the neutron irradiation effect on mechanical properties of HIP joint material was examined by tensile and impact tests using specimens with irradiation damage of about 1.5 dpa. The results of tensile tests show that tensile strength of HIP joint material was about the same as that of DSCu base material, and this trend did not change after neutron irradiation. On the other hand, the impact value of HIP joint material was smaller than that of DSCu base material because of the diffusion of main elements at joint boundary. It was shown that embrittlement by the neutron irradiation effect is smaller than that of the effect by HIP joint.     

Natural sintering of carbonate powder for use as a carbon-14 immobilization matrix

Carbon-14 generated during fuel irradiation in the reactor is later partially released as ¹⁴CO₂ during reprocessing. It must not be released directly into the environment because of its long half-life and rapid assimilation by living organisms. The waste volume generated by conditioning depends not only on the type of fuel but especially on the selected conditioning matrix. We propose that a sintered carbonate be used directly as an alternative to cement with the main objective of providing a significant waste volume reduction while ensuring durability equal to or better than that of cement. Many carbonates are difficult to sinter naturally because of their relatively low thermal decomposition temperature. They also exhibit low hardness and are generally friable. We show that a binary carbonate BaCa(CO₃)₂ yields a sintered carbonate with satisfactory mechanical properties and acceptable water solubility. The carbonate fabrication, pelletizing and sintering conditions are examined and discussed.     

Effect of radial hydrides on the axial and hoop mechanical properties of Zircaloy-4 cladding

The effect of radial hydrides on the mechanical properties of stress-relief annealed Zircaloy-4 cladding was studied. Specimens were firstly hydrided to different target hydrogen levels between 100 and 600 wt ppm and then thermally cycled in an autoclave under a constant hoop stress to form radial hydrides by a hydride reorientation process. The effect of radial hydrides on the axial properties of the cladding was insignificant. On the other hand, the cladding ductility measurements decreased as its radial hydride content increased when the specimen was tested in plane strain tension. A reference hydrogen concentration for radial hydrides in the cladding was defined for assessing the fuel cladding integrity based on a criterion of the tensile strength 600 MPa. The reference hydrogen concentration increased with the specimen (bulk) hydrogen concentration to a maximum of ∼90 wt ppm at the bulk concentration ∼300 wt ppm H and then decreased towards higher concentrations.     

Strain rate dependence of impact properties of sintered 316L stainless steel

This paper uses a material testing system (MTS) and a compressive split-Hopkinson bar to investigate the impact behaviour of sintered 316L stainless steel at strain rates ranging from 10⁻³ s⁻¹ to 7.5 × 10³ s⁻¹. It is found that the true stress, the rate of work hardening and the strain rate sensitivity vary significantly as the strain rate increases. The flow behaviour of the sintered 316L stainless steel can be accurately predicted using a constitutive law based on Gurson’s yield criterion and the flow rule proposed by Khan, Huang and Liang (KHL). Microstructural observations reveal that the degree of localized grain deformation increases, but the pore density and the grain size decrease, with increasing strain rate. Adiabatic shear bands associated with cracking are developed at strain rates higher than 5.6 × 10³ s⁻¹. The fracture surfaces exhibit ductile dimples. The depth and density of these dimples decrease with increasing strain rate.     

Mechanical property and microstructural change by thermal aging of SCS14A cast duplex stainless steel

The aging behavior, especially saturation, of JIS SCS14A cast duplex stainless steels was investigated on the basis of the mechanical properties and microstructural changes during accelerated aging at 350 °C and 400 °C. The aging behavior of the materials mainly proceeds via two stages. During the first stage, the generation and concentration of the iron-rich and chromium-enriched phase in ferrite occurs by phase decomposition. The first stage corresponds to aging times of up to 3000 h at 400 °C. During the first stage, the ferrite hardness achieved is approximately 600 VHN, and the Charpy impact energy is almost saturated. During the second stage, the precipitated chromium-enriched phase aggregates and coarsens, and the G phase precipitation also occurs. The second stage corresponds to the aging times range of 3000-30 000 h at 400 °C. During the second stage, the ferrite hardness achieved is about 800 VHN; however, further hardening exceeding 600 VHN does not influence the Charpy impact energy.     

Mechanical behaviour of the T91 martensitic steel under monotonic and cyclic loadings in liquid metals

The paper deals with the mechanical properties in liquid metals of the T91 martensitic steel, a candidate material for the window of an accelerating driven system (ADS). Two main questions are examined, the risk of liquid metal embrittlement and the accelerated fatigue damage by a liquid metal. It is found that the transition from ductile to brittle behaviour induced by a liquid metal is possible as a result of a decrease in surface energy caused by the adsorbed liquid metal. The embrittlement can occur only with a hard microstructure and a nucleation of very sharp defects inside the liquid metal. Under cycling straining, the fatigue resistance of the standard T91 steel is decreased by a factor of about 2 in the liquid metal as compared to air. It is proposed that short crack growth is promoted by the liquid metal which weakens the microstructural grain boundary barriers and skip the microcrack coalescence stage.     

Design feasibility study of a divertor component reinforced with fibrous metal matrix composite laminate

Fibrous metal matrix composites possess advanced mechanical properties compared to conventional alloys. It is expected that the application of these composites to a divertor component will enhance the structural reliability. A possible design concept would be a system consisting of tungsten armour, copper composite interlayer and copper heat sink where the composite interlayer is locally inserted into the highly stressed domain near the bond interface. For assessment of the design feasibility of the composite divertor concept, a non-linear multi-scale finite element analysis was performed. To this end, a micro-mechanics algorithm was implemented into a finite element code. A reactor-relevant heat flux load was assumed. Focus was placed on the evolution of stress state, plastic deformation and ductile damage on both macro- and microscopic scales. The structural response of the component and the micro-scale stress evolution of the composite laminate were investigated.     

Investigation of physico-mechanical properties of ceramic oxide kernels for nuclear applications

In the present work, two different classes of oxide kernels were investigated: unirradiated thoria, urania and (Th,U)O₂ fuel kernels and low-density Al₂O₃ kernels for the incorporation of minor actinides. The physical mechanism of oxide kernel failure under uniaxial compression was investigated. A new method for determining the physico-mechanical properties of kernels has been developed and the parameters P_S and δ, characterising the level of stress required for destruction of the material structure and the brittleness of the investigated materials, respectively, were evaluated and discussed. It was shown that the value of P_S is analogous to traditional characteristics of material such as microhardness Hv. The `quantisation' effect was revealed in the kernel crushing strength and deformation distributions. The physico-mechanical properties of ceramic kernels (average particle size, microstructure, phase state, density, P_S and δ) were investigated and comparative analysis of different kernel types was performed. Additionally, the impact of annealing time on the properties of low-density Al₂O₃ kernels was examined.     

The influence of roughness on tribological properties of nuclear grade graphite

The influence of surface roughness on tribological properties of graphite IG-11 was investigated on a standard SRV tester. The experimental condition was selected as: 30 N normal load, room temperature and a 10 Hz frequency with different strokes. The experiments environments included helium and air. Five types of roughness were studied in the experiments. The experiments revealed that the surface roughness greatly affected the graphite friction behavior. When the friction surface was smooth, the friction coefficient was high because of intensive adhesion accompanied by many pits at the friction surface. When the friction surface was rough, the adhesion was very poor, but the wear was excessive and generated many graphite particles at the friction surface. These particles can separate the friction surfaces, which reduced the friction action between them. For very rough specimens, the friction coefficient decreased with sliding velocity at about 0.004 m/s and then increases gradually.     

Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.     

Effect of contact conditions on embrittlement of T91 steel by lead-bismuth

The T91 martensitic steel is a candidate structural material for the liquid lead-bismuth eutectic (LBE) MEGAPIE spallation target. This paper first reviews some results on Liquid Metal Embrittlement (LME) of martensitic steels by liquid metals. It appears that LME of steels can occur provided a few criteria are fulfilled simultaneously. Intimate contact between liquid metal and solid metal is the first one. Usually, it is impossible to avoid the oxide film formation on the steel surface even after short exposure to air. This explains the difficulty arising when one would like to determine the susceptibility to LME of T91 steel whilst put into contact with lead-bismuth. Later, we report on different methods of surface preparation in order to remove the oxide layer on the T91 steel (PVD, soft soldering fluxes) and the resulting susceptibility to LME.     

In-reactor uniaxial tensile testing of pure copper at a constant strain rate at 90 °C

Recently, annealed specimens of pure copper have been tensile tested in a fission reactor at a damage rate of 6 × 10⁻⁸ dpa/s with a constant strain rate of 1.3 × 10⁻⁷ s⁻¹. The specimen temperature during the test was about 90 °C. The stress response was continuously recorded as a function of irradiation time (i.e. displacement dose and strain). The experiment lasted for 308 h. During the dynamic in-reactor test, the specimen deformed and hardened homogeneously without showing any sign of yield drop and plastic instability. However, the specimen yielded a uniform elongation of only about 12%. The preliminary results are briefly described and discussed in the present note.