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The liquid lead-lithium (Pb–17Li) blanket has many applications in fusion reactors due to its good tritium breeding performance, high heat transfer efficiency and safety. The compatibility of liquid Pb–17Li alloy with the structural material of blanket under magnetic field is one of the concerns. In this study, corrosion experiments China low activation martensitic (CLAM) steel and 316L steel were carried out in a forced convection Pb–17Li loop under 1.0 T magnetic field at 480 °C for 1000 h. The corrosion results on 316L steel showed the characteristic with a superficial porous layer resulted from selective leaching of high-soluble alloy elements and subsequent phase transformation from austenitic matrix to ferritic phase. Then the porous layers were eroded by high-velocity jet fluid. The main corrosion mechanism of CLAM steel was selective dissolution-base corrosion attack on the microstructure boundary regions and exclusively on high residual stress areas. CLAM steel performed a better corrosion resistance than that of 316L steel. The high Ni dissolution rate and the erosion of corroded layers are the main causes for the severe corrosion of 316L steel.

Highlights • Mechanism of protrusion formation on steel surface in flowing Pb-17Li. • Small-scale, mid-scale and large-scale corrosion-erosion behaviors in flowing Pb-17Li. • Precipitation behaviors of Fe, Cr and Ni in a nonisothermal Pb-17Li system. Abstract Corrosion tests were performed to study corrosion–erosion behaviors of reduced activation ferritic/martensitic steel JLF-1 and mass transfer behaviors of metal elements in a nonisothermal Pb-17Li flow. The tests were performed using isothermal and nonisothermal mixing pots at approximately 773 K for up to 1000 h. The test results indicate that the steel surface corroded according to the dissolution of steel compositions (i.e., Fe and Cr) in the flowing Pb-17Li. The corroded surface then revealed a granular structure with small pebble-shaped protrusions. The protrusions were formed by preferential corrosion along the boundaries of microstructures on the steel surface. Small-scale corrosion-erosion was caused by the removal of the small protrusions from the steel surface in flowing Pb-17Li. There was also large-scale corrosion–erosion according to the destruction of the steel surface after the surface was deeply corroded. Metal elements, such as Fe, Cr and Ni, were dissolved from the steel in the flowing Pb-17Li. They were transferred to a low-temperature region with the Pb-17Li flow and finally precipitated as round-shaped particles on metal wires installed in the low-temperature region of the mixing pot. [ABSTRACT FROM AUTHOR]

• Ductility of RAFM steels at a stress of 250 MPa decreased appreciably after exposure in Pb-17Li at 400 ℃ for 1500 h. • Coarsening of martensitic laths occurred in RAFM steel during long-term exposure in Pb-17Li with a stress. • A mechanism of fracture for RAFM steel under combined effect of stress and Pb-17Li is proposed. The effect of applied stress on the ductility of reduced activation ferritic martensitic (RAFM) steel during long-term exposure in flowing Pb-17Li in the temperature interval 300–400 ℃ was investigated using tensile tests. The ductility of RAFM steels at a stress of 250 MPa decreases appreciably after exposure in following Pb-17Li at 400 ℃ for 1500 h. And a slight ductility loss occurred for the samples tested at 300 ℃, 350 ℃ and 400 ℃ for 500 h. Coarsening of martensitic laths occurred in RAFM steels owing to the thermal aging with a 250 MPa stress for 1500 h in liquid Pb-17Li. Due to the poor ductility of coarse laths, micro-cracks originated on the surface of necking zone during tensile test, which facilitates the penetration of liquid metal. With sufficient liquid Pb-17Li penetrated into cracks continuously, the surface tension of the newly formed fracture surface is declined by liquid metal in the crack tip. The fracture mode was altered from uniform to unstable crack propagation thus leading to a reduction of ductility. There was no depletion of Fe and Cr or Pb infiltration into the substrate for RAFM steels after exposure. The fracture mechanisms were proposed in terms of supply of liquid metal and micro-cracks caused by coarse laths. Accordingly, the reduction of ductility for RAFM steels after exposure in liquid Pb-17Li was the result of combined effect of applied stress and liquid metal. [ABSTRACT FROM AUTHOR]

Diffusion characteristics of transition metals – nickel (Ni) and iron (Fe) diffusion in the molten lead–17lithium (Pb–17Li) were investigated experimentally under the argon atmosphere in the temperature range from 300 °C to 600 °C by using the long capillary method. The axial concentration distributions of Ni and Fe in Pb–17Li specimens were measured by ICP-MS, and the diffusion coefficients of Ni and Fe were determined based on the Fick's second law. The Arrhenius correlations of diffusion coefficient of Ni and Fe were obtained as D Ni =7.37 × 10 −5 exp (−4.66 × 10 3 / RT ) (300 °C ≤ T ≤ 450 °C) [cm 2 /s], D Fe =3.21 × 10 −4 exp (−2.64 × 10 4 / RT ) (450 °C ≤ T ≤ 600 °C) [cm 2 /s]. It is found that the diffusion coefficient of Ni is one order magnitude bigger than that of Fe in Pb–17Li. The diffusion coefficient of Fe in Pb–17Li was approximately in the same order as that in lead–bismuth eutectic reported by authors. [ABSTRACT FROM AUTHOR]

A rotating disc type corrosion test facility has been developed to study the compatibility of structural steels in lead–lithium eutectic at selected temperatures and flow velocities. The compatibility of an Indian Reduced Activation Ferritic–Martensitic Steel (IN RAFMS) disc rotating at 360 revolutions per minute (RPM) in Pb–17Li was studied at 773 K after an exposure of 3000 h. Results indicated that weight loss occurred at a rate of 1.02 μg/cm 2 h for a total duration of 3000 h. The corrosion mechanism was governed by dissolution of alloying elements from the sample surface and penetration of Pb–17Li into the IN RAFMS matrix. The microstructure of the parent material got altered up to a depth of 30 μm from the exposed surface although effects of Pb–17Li penetration were noted up to 100 μm. The depth of Pb–17Li attack increased with increase in linear flow velocities along the radial direction of the IN RAFMS disc. [ABSTRACT FROM AUTHOR]

Structural and elemental analyses of the RAFM steel (EUROFER 97) interface with flowing Pb–17Li eutectic (velocity 5 cm/s at 550 °C, 1000 h) under the action of a strong magnetic field (B = 1.7 T) were performed using optical microscopy, SEM, confocal microscopy, precision micro-hardness methods, SIMS and point or line-scan EDX analyses. The results show that the magnetic field induces a faster crushing of martensite into the grains, a deeper dissolution of grain boundaries, an enhancement of the Fe and Cr mass transfer and a fast detachment of corrosion layers due to MHD effects. [ABSTRACT FROM AUTHOR]

Corrosion of modified 9Cr-1Mo steel (P91) and Indian Reduced Activation Ferritic Martensitic Steel (IN-RAFMS) was studied in static lead-17 atom% lithium eutectic for 355 h at 773 K. Exposure to molten Pb-17Li led to loss of weight in P91 material whereas IN-RAFMS gained weight which was possibly due to longer incubation period and post-exposure surface oxidation. Depletion of Fe and Cr occurred from a thin surface layer in both P91 and IN-RAFMS although higher oxygen buildup on the surface of IN-RAFMS justified its weight gain. [ABSTRACT FROM AUTHOR]

A rotating disc type corrosion test facility has been developed to study the compatibility of structural steels in lead-lithium eutectic at selected temperatures and flow velocities. The compatibility of a modified 9Cr-1Mo steel (P91) disc in Pb-17Li was studied at 773 K and 700 revolutions per minute during an exposure of 2,000 h. Results indicated that weight loss occurred at a rate of 4.2 µg/cm h for a total duration of 2,000 h. The corrosion mechanism was governed by dissolution of iron and chromium from the sample surface into molten Pb-17Li. The depth of dissolution of major alloying elements and selective removal of the Fe and Cr depleted surface layers depended upon the variation of linear flow velocities along the radial direction. [ABSTRACT FROM AUTHOR]

Abstract: An experimental study was performed to investigate hydrogen transport under conditions of gas bubbling into the liquid breeders of Li, Pb–17Li and Flinak. Gas of H2 or high-purity Ar at 0.1MPa was injected into the liquid breeders using an I-shape nozzle. Then, the H2 partial pressure in the cover gas under transition state was measured by a solid electrolyte sensor. Variations of the partial pressure in the exhaust gas were compared with a numerical model to describe hydrogen transfer through the bubble–liquid interface. The experimental rates of hydrogen transport in the liquid breeders were correlated in terms of a mass-transfer coefficient for the hydrogen transport under gas bubbling. The overall mass-transfer coefficient for H2 charging was compared with that for Ar purging. The coefficient for the recovery was about 10 times lower than that of the charging in the respective cases of the Pb–17Li and the Flinak. [Copyright &y& Elsevier]

Abstract: Formation of droplet of liquid Li–17Pb released from a nozzle into vacuum was studied for the evaluation of the feasibility as a tritium extraction process. Size of droplets formed from the nozzles was estimated by theoretical and experimental methods. For the theoretical estimation, the non-dimensional comparison of the physical bulk property of liquid Pb–17Li with water (H2O) at ambient temperature was applied. It was found to be reasonable to apply the Plateau-Rayleigh-Instability theory for the droplet size formula of the fluid Pb–17Li for the nozzle diameter 0.4mm–1.0mm, temperature 400°C–500°C, at initial velocity of 3m/s. The experimental results of the droplet size showed good agreement with the theory. This device was used for the parametric study of extraction of deuterium during their free fall in vacuum. The scaling of the device suggests the engineering feasibility of the process. [Copyright &y& Elsevier]

Abstract: 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 500h, 1000h 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. [Copyright &y& Elsevier]

Abstract: The dual-coolant lead-lithium, or DCLL, blanket concept is of strong interest in the US fusion technology program. In the DCLL blanket, the flow channel insert (FCI) is a critical component. FCIs must have low electrical and thermal conductivity and be compatible with lead–lithium eutectic alloy (Pb–17Li) at elevated temperatures. FCIs must retain structural integrity and desirable properties even under irradiation and large temperature gradients during operation. FCIs must not fail in such a way that Pb–17Li enters the FCI and changes its electrical or thermal conductivity significantly. Another important issue for the DCLL is the development of a suitable tritium extraction from the Pb–17Li to achieve low tritium partial pressure, thus facilitating decisive tritium control. In this paper, the state of DCLL development in the US is presented including recent design modifications and results from recent R&D efforts. Such R&D includes the progress on development and property quantification of SiC/SiC composites and SiC foams as candidate FCI materials; Pb–17Li material capability and infiltration studies; simulations of MHD Pb–17Li flow characteristics and of resultant temperature distributions; and the analysis of FCI stress states based on these thermal loads. In addition, tritium extraction from Pb–17Li based on a vacuum permeator concept is shown to have the potential to achieve the desired tritium control. A discussion of DCLL optimization and unresolved DCLL issues and future R&D needs is also presented. [Copyright &y& Elsevier]

The corrosion behavior of structural materials in liquid lead alloys like Pb-17Li as a liquid tritium breeder for future fusion reactors and Pb-55Bi (lead-bismuth eutectic [LBE]) as an envisaged spallation target and/or coolant for accelerator-driven systems (ADS) and lead-cooled Gen IV reactors is a key issue in the road map for developing new energy systems within the next decades. Ferritic-martensitic steels of the 8 wt% to 10 wt% Cr type are considered as structural materials for use in liquid Pb-17Li at temperatures up to 550°C. Due to the specific physical chemistry of Pb-17Li, dissolution corrosion is the major corrosion mechanism of iron-based alloys. No oxide formation on steel surfaces can occur because of the very low oxygen potential. Therefore, coatings could/must be applied to minimize corrosion effects at medium (500°C) and higher operational temperatures (650°C) of advanced fusion blanket concepts. Ferritic-martensitic and austenitic CrNi steels are considered potential structural materials for ADS and Gen IV reactor applications. Nevertheless, the high nickel solubility in LBE limits the use of unprotected austenitic steels to temperatures of about 550°C. Because of the much higher oxygen potential in LBE, a totally different strategy of minimizing corrosion must be adopted. By using active oxygen control, a desired oxygen activity will be adjusted in the liquid metal via gas phase equilibrium, which enables the controlled formation of oxide layers on the steel surfaces. These in situ-formed oxides can act as corrosion barriers and make use of high-nickel alloys possible, even at temperatures around 600°C. [ABSTRACT FROM AUTHOR]

Since several years ENEA is involved in experimental activities concerning the interaction between molten lithium lead alloy, in eutectic composition, and pressurised water in conditions relevant for DEMO fusion reactor, in order to predict the behaviour of a WCLL blanket module in case of an in blanket LOCA. In this ambit LIFUS 5 apparatus was designed and constructed to carry out the experimental campaign on water large leaks into liquid Pb–17Li, with the final aim to provide data for the validation of the mathematical modelling of the related phenomena. After two tests performed to qualify all mechanical components and the data acquisition system, LIFUS 5 was operated at the ENEA C.R. Brasimone for tests nos. 3 and 4. Water was injected into the reaction tank at a pressure of 155 bar with two different values of sub-cooling. The initial liquid metal temperature was fixed to 330 °C. The first pressure peak due to the water vaporization and jet expansion was clearly recognized together with the subsequent pressure increase due to further water injection and hydrogen generation. In the performed experiments the maximum pressure peak, as detected in both reaction and expansion vessels, never overcame the value of the injected water. Moreover, in these two tests a significant temperature increase in the reaction vessel occurred, strictly connected to the amount of injected water. [Copyright &y& Elsevier]

Within the European Fusion Technology Programme, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decades, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be tested in ITER-FEAT. The present paper is focused on the study of the WCLL-TBM nuclear response in ITER-FEAT, being specifically oriented to the investigation of the local effects due to the typical C-shaped tubes of the breeder zone, since they could play a pivotal role in the module-relevant thermo–mechanical design. A 3D heterogeneous model of the WCLL-TBM, realistically simulating its new lay out and taking into account 9% Cr martensitic steel as reference structural material, has been set-up. A particular attention has been paid to the simulation of the characteristic ‘C’ shape of the breeder zone double walled tubes, which have been realistically reproduced. The WCLL-TBM model has been inserted into an existing ITER-FEAT 3D semi-heterogeneous model accounting for a proper D–T neutron source. Analyses have been performed by means of MCNP-4C code running on a cluster of four workstations through the implementation of a parallel virtual machine. For each analysis a large number of histories (>10.000.000) have been simulated, obtaining statistical uncertainties on the results lower than 3%. The main features of the WCLL-TBM nuclear response have been determined focusing the attention on power deposition density, material damage through DPA and He and H production rate, daily tritium production and tritium production rate radial distribution in the module. The obtained results are herewith presented and critically discussed. [Copyright &y& Elsevier]

The interaction between pressurised water and liquid Pb–17Li is a topical issue for the WCLL blanket concept and needs to be studied in detail because of the potentially significant effect on the reliability and safety of the blanket. Particularly, it seems important from the design point of view to deeply investigate the consequences of coolant micro-leaks in the blanket module generated by micro-cracks with a size in the order of 10−3 mm2. This kind of interaction has been studied during an extensive experimental campaign on the RELA loops in the ENEA Research Centre at Brasimone. The results showed a reduction in the liquid metal flow-rate in the circuit and a deterioration of the heat exchange properties between breeder and coolant due to the formation and growth of solid reaction products generated by the chemical reaction between lithium and water. The molar ratio between the recovered hydrogen recovered and water injected was in the range 0.3–0.5, confirming that, in the test conditions, the main solid reaction product is lithium hydroxide (LiOH). The possible impact of these results on the TBM-ITER design is also presented and discussed. [Copyright &y& Elsevier]

The advanced dual coolant (A-DC) blanket is one of the EU advanced concepts to be investigated in the frame of the long-term power plant conceptual study (PPCS). Its basic concept—following the ARIES-ST concept—is based on the use of helium-cooled ferritic steel structure, the self-cooled Pb–17Li breeding zone, and SiC/SiC flow channel inserts. The latter serves as electrical and thermal insulators and therefore minimize the pressure losses and enable a relatively high Pb–17Li exit temperature leading to a high thermal efficiency. The present work on PPCS is drawn extensively on the preparatory study on plant availability (PPA) carried out in 1999 where a maximum neutron wall load of 5 MW/m2 (corresponding maximum surface heat load of 0.9 MW/m2) was given in the reference case of the A-DC blanket. In the following stage of PPCS the A-DC blanket is normalized and adapted to a typical size of commercial reactors (e.g. 1500 MWe) which requires iterative calculations between the blanket layout and the system code analysis. The status of the work with some idea improvements is reported. [Copyright &y& Elsevier]