Compatibility of CLAM steel weldments with static LiPb alloy at 550 °C

CLAM steel is considered as a structural material to be used in the Test Blanket Module as a barrier or blanket adjacent to liquid LiPb in fusion reactors. In this paper, CLAM steel is welded by tungsten inert gas (TIG) welding, and the compatibility of the weldment with liquid LiPb is tested. Specimens were corroded in static liquid LiPb, with corrosion times of 500 h and 1000 h, at 550 °C, and the corresponding weight losses are 0.272 mg/cm² and 0.403 mg/cm² respectively. Also the corrosion rate decreases with increased corrosion time. In the as-welded condition, corrosion resistance of the weld zone is higher than that of the HAZ (Heat Affected Zone). Likely, thick martensite lath and large residual stresses at the welding zone result in higher corrosion rates. The compatibility of CLAM steel weld joints with high temperature liquid LiPb can be improved to some extent through a post-weld tempering process. The surface of the as-welded CLAM steel is uniformly corroded and the concentration of Cr on the surface decreases by about 50% after corrosion. Penetration of LiPb into the matrix is observed for neither the as-welded nor the as-tempered conditions. Influenced by thick martensite lath and large residual stresses, the welded area, especially the weld zone, is easily corroded, therefore it is of primary importance to protect the welded area in the solid blanket of the fusion reactor.     

Mechanical properties and microstructural investigations of TIG welded 40 mm and 60 mm thick SS 316L samples for fusion reactor vacuum vessel applications

The development of advanced fusion reactors like DEMO will have various challenges in materials and fabrication. The vacuum vessel is important part of the fusion reactor. The double walled design for vacuum vessel with thicker stainless steel material (40–60 mm) has been proposed in the advanced fusion reactors like ITER. Different welding techniques will have to be used for such vacuum vessel development. The required mechanical, structural and other properties of stainless steels have to be maintained in these joining processes of components of various shapes and sizes in the form of plates, ribs, shells, etc. The present paper reports characterization of welding joints of SS316L plates with higher thicknesses like 40 mm and 60 mm, prepared using multi-pass Tungsten Inert Gas (TIG) welding process. The weld quality has been evaluated with non-destructive tests by X-ray radiography and ultrasonic methods. The mechanical properties like tensile, bend tests, Vickers hardness and impact fracture tests have been carried out for the weld samples. Tensile property test results indicate sound weld joints with efficiencies over 100%. Hardening was observed in the weld zone in non-uniform manner. Macro and microstructure studies have been carried out for Base Metal (BM), Heat Affected Zone (HAZ) and Weld Zone (WZ). Scanning Electron Microscopy (SEM) analysis carried out for the impact fractured specimens show ductile fracture. The microstructural study and ferrite number data indicate the presence of high content of delta ferrite in the weld zone as compared to the delta ferrite in base metal.     

Helium-implanted CLAM steel and evolutionary behavior of defects investigated by positron-annihilation spectroscopy

China Low Activation Martensitic (CLAM) steel has been chosen as the primary candidate structural material for the first wall/blanket for fusion reactor. The excessive helium irradiation induced damage of CLAM steel at high temperatures and the evolution of defects were investigated in this paper. The samples were homogeneously implanted with 1e + 17 ions/cm² and 100 keV of helium at room temperature, 473, 673, and 873 K. Irradiation induced damage of CLAM steel and the annealing behavior of defects were probed by slow positron beam Doppler broadening technique. Helium implantation produced a large number of vacancy-type defects which bound with helium and formed helium–vacancy complexes. Target atoms’ displacement capacity was strengthened with rising irradiation temperatures, so the S parameter increased with increasing irradiation temperatures, and helium–vacancy complexes were main defects after helium implantation at damage layers. Helium bubbles would be unstable and the desorption of helium bubbles would promote the density of defects above 673 K. By analyzing the curves of S–W and annealing tests of irradiated specimen, it suggested that there werenot only one type of defect in damage layers. Though helium–vacancy complexes were primary defects after helium implanted, introducing excessive helium might also generated other point defects or dislocation loops in the material.     

Vacancy-type defects and hardness of helium implanted CLAM steel studied by positron-annihilation spectroscopy and nano-indentation technique

China Low Activation Martensitic (CLAM) steel was implanted with helium up to 1e + 16/cm² at 300–873 K using 140 keV helium ions. Vacancy-type defects induced by implantation were investigated with positron beam Doppler broadening technique, and then nano-hardness measurements were performed to investigate helium-induced hardening effect. He implantation produced a large number of vacancy-type defects in CLAM steel, and the concentration of vacancy-type defects decreased with increasing temperature. Vacancy–helium complexes were main defects at different temperatures. Irradiation induced hardening was observed at all irradiation temperatures, and the peak value of hardness was at 473 K. The result suggested that both vacancy–helium complexes and helium bubbles had contribution to irradiation induced hardening. The decomposition and annihilation of irradiation-induced defects became more and more significant with increasing temperature, which induced the increment of hardness became more and more small.     

Progress in compatibility experiments on lithium–lead with candidate structural materials for fusion in China

Liquid LiPb eutectic is one of the promising candidate tritium breeder materials for fusion reactors. This paper presents the progress in compatibility experiments with liquid LiPb achieved up to now in China for some candidate structural materials. The results showed that CLAM steel had good compatibility with flowing LiPb at 480 °C with the velocity of 0.08 m/s after 5000 h in DRAGON-I loop. On the other hand, after exposed in static LiPb at 700 °C for 500 h in a SiC crucible, the W and Mo specimens suffered much more weight loss compared with Nb specimen, and a thin reaction product layer was visible on the surface of all the three refractory metals. Preliminary analysis on SiC_f/SiC composite specimens indicated that there was no penetration of LiPb and no reaction products on the surface with CVD SiC coating, which showed SiCf/SiC composite were stable and compatible with static LiPb under 700 °C after 500 h exposure.     

Diffusion bonding beryllium to Reduced Activation Ferritic Martensitic steel: Development of processes and techniques

Beryllium was successfully bonded to a Reduced Activation Ferritic Martensitic (RAFM) steel with a maximum strength of 150 MPa in tension and 168 MPa in shear. These strengths were achieved using Hot Isostatic Pressing (HIP), at temperatures between 700 °C and 750 °C for 2 h and under a pressure of 103 MPa. To obtain these strengths, 10 μm of titanium and 20 μm of copper were deposited on the beryllium substrate prior to HIP bonding. The copper film acted a bonding aid to the RAFM steel, while the titanium acted as a diffusion barrier between the copper and the beryllium, suppressing the formation of brittle intermetallics that are known to compromise mechanical performance. Slow cooling from the peak HIP temperature along with an imposed hold time at 450 °C further enhanced the final mechanical strength of the bond.     

Weldability of neutron irradiated austenitic stainless steels

Degradation of weldability in neutron irradiated austenitic stainless steel is an important issue to be addressed in the planning of proactive maintenance of light water reactor core internals. In this work, samples selected from reactor internal components which had been irradiated to fluence from 8.5 × 10²² to 1.4 × 10²⁶ n/m² (E > 1 MeV) corresponding to helium content from 0.11 to 103 appm, respectively, were subjected to tungsten inert gas arc (TIG) welding with heat input ranged 0.6–16 kJ/cm. The weld defects were characterized by penetrant test and cross-sectional metallography. The integrity of the weld was better when there were less helium and at lower heat input. Tensile properties of weld joint containing 0.6 appm of helium fulfilled the requirement for unirradiated base metal. Repeated thermal cycles were found to be very hazardous. The results showed the combination of material helium content and weld heat input where materials can be welded with little concern to invite cracking. Also, the importance of using properly selected welding procedures to minimize thermal cycling was recognized.     

Progress of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO [1] has been developed at Karlsruhe Institute of Technology (KIT) since a couple of years with the goal of reaching a heat flux of 10 MW/m² anticipated for DEMO. The reference concept HEMJ (He-cooled modular divertor with multiple-jet cooling) is based on the use of small cooling fingers – each composed of a tungsten tile brazed to a tungsten alloy thimble – as well as on impingement jet cooling with helium at 10 MPa, 600 °C. The cooling fingers are connected to the main structure of ODS Eurofer steel by brazing in combination with a mechanical interlock. This paper reports progress to date of the design accompanying R&Ds, i.e. primarily the fabrication technology and HHF experiments. For the latter a combined helium loop and electron beam facility (200 kW, 40 keV) at Efremov Institute, St. Petersburg, Russia, has been used. This facility enables mock-up testing at a nominal helium inlet temperature of 600 °C, a pressure of 10 MPa, and a maximal pressure head of 0.5 MPa. HHF test results till now confirm well the divertor design performance. In the recent test series in early 2010 the first breakthrough was achieved when a mock-up has survived over 1000 cycles at 10 MW/m² unscathed.     

Verification test results of a cutting technique for the ITER blanket cooling pipes

For replacement of the first wall (FW) of the international thermonuclear experimental reactor (ITER), cutting and welding tools for the cooling pipes must be able to access a pipe from the surface side of the FW and cut/weld the pipe from the inside the cooling pipe (inner diameter: 42.72 mm, thickness: 2.77 mm). The cutting tool for the pipe end is required to cut a flat plate circularly from the surface side of the FW (cutting diameter: approximately 44 mm, plate thickness: 5 mm). To determine the specifications for both the tools and the blanket hydraulic connections, the ITER Organization (IO) and the Japan Domestic Agency (JADA) conducted research and development activities regarding the FW replacement. This paper describes the current status of the development of cutting tools for the cooling pipe connection.     

Numerical simulation of a blanket cooling system for fusion reactor based on PWR conditions

The simulations of a blanket cooling system were presented to address the choice of cooling channel geometry and coolant input data which are related to blanket engineering implementation. This work was performed using computer aided design (CAD) and computational fluid dynamics (CFD) technology. Simulations were carried out for the blanket module with a size of 0.6 m × 0.45 m in toroidal plane, and the nuclear heat was applied on the cooling system at P_n (neutron wall load) of 5 MW/m². The structure factors and input data of hydraulics were investigated to explore the optimal parameters to match the PWR condition. It was found that the inlet velocity of first wall (FW) channel should be within the range of 2.48–3.34 m/s. As a result, the temperature rise (TR) of the coolant in the FW channel would be 24–25 K. This leads to the remaining space for TR within the range of 15 K in the piping circuits. It also indicated that the FW plays an important role in TR (reaches 60% of the whole cooling system) due to its high level of P_n and heat flux in the zones. It was predicted that the nuclear heat inside blanket module could be removed completely by the piping circuits with an acceptable pipe bore and the related input data. Finally, a possible design range of cooling parameters was proposed in view of engineering feasibility and blanket neutronics design.     

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5 MW/m²) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V = 1–3 m/s, P = 2–3 MPa, T = 110–170 °C). The heat flux was gradually varied in the range of 1–10 MW/m² until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2 m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3–8 MW/m²) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.     

Effect of hot rolling process on the mechanical and microstructural property of the 9Cr–1Mo steel

The effect of hot rolling on the mechanical and microstructural property has been investigated to simulate the effect of hot extrusion during the manufacturing process of the fuel cladding for sodium cooled fast reactors (SFRs). Hot rolling of modified 9Cr–1Mo steel was carried out either at 1050 °C or 950 °C upon cooling after normalizing. Continuous annealing right after the hot rolling at 950 °C for 1 h has been carried out followed by the mechanical testing and microstructural analysis. The results showed that hot rolling without any annealing or tempering treatment leaves residual stress so that it leads to the abrupt increase of material strength that would affect cladding formability. Continuous annealing right after the hot rolling process can alleviate residual stress without decreasing too much of material strength. Hot rolling either at 1050 °C or 950 °C increases the number density of the remained precipitate which leads to the precipitation hardening. Introduction of continuous annealing results in an increase in the fraction of secondary V-rich MX precipitate that leads to an increase in the stability at high temperature mechanical property.     

Fabrication of W/FMS joint mock-ups using a hot isostatic pressing

Blocks of tungsten and ferritic–martensitic steel (FMS) were joined without any interfacial defects or cracks. For the joining, two times of a hot isostatic pressing (HIP) were performed. The first HIP (900 °C, 100 MPa, 1.5 h) facilitates the diffusion bonding between W and FMS. The second HIP (750 °C, 70 MPa, 2 h) corresponds to a tempering process to retain the mechanical properties of the FMS. As an interlayer material, titanium foil that can mitigate the thermal expansion difference between W and FMS was used. In addition, a molybdenum foil was inserted to prevent an unwanted bonding of W to a canning material. The lateral cracks in W plates, which were usually observed in the case of a conventional HIP process, were not observed when the molybdenum separator was used. W/FMS joint mock-ups with a dimension of 50 mm × 50 mm × 32 mm (T) were successfully fabricated. The shear strength of the joints was 89 MPa on average.     

Design of the dummy coil for magnet power supply

It is necessary to test it on a dummy coil, before using a magnet power supply (MPS) to energize a Poloidal Field (PF) coil in the Korea Superconducting Tokamak Advanced Research (KSTAR) device. The dummy coil should accept the same large current from the MPS as the PF coil and be within the capability of the utilities located at the KSTAR site. Therefore a coil design based on the characteristics of the MPS and other restrictive conditions needed to be made. There are three requirements to be met in the design: an electrical requirement, a structural requirement, and a water cooling requirement. The electrical requirement was that the coil should have an inductance of 40 mH. For the structural requirement, the material should be non magnetic. The coil support structure and water cooling manifold were made of SUS 304. The water cooling requirement was that there should be sufficient flow rate so that the temperature rise ΔT should not exceed 12 °C for operation at 12.5 kA for 5 min. Square cross-section hollow conductor with dimensions of 38.1 mm × 38.1 mm was used with a 25.4 mm center hole for cooling water. However, as a result of tests, it was found that the electrical and structural requirements were satisfied but that the water cooling was over designed. It is imperative that the verification will be redone for a test with 12.5 kA for 5 min.     

Irradiation creep relaxation of void swelling-driven stresses

Swelling-driven-creep test specimens are used to measure the compressive stresses that develop due to constraint of irradiation void swelling. These specimens use a previously non-irradiated 20% CW Type 316 stainless steel holder to axially restrain two Type 304 stainless steel tubular specimens that were previously irradiated in the US Experimental Breeder Reactor (EBR-II) at 490 °C. One specimen was previously irradiated to fluence levels in the void nucleation regime (9 dpa) and the other in the quasi-steady void growth regime (28 dpa). A lift-off compliance measurement technique was used post-irradiation to determine compressive stresses developed during reirradiation of the two specimen assemblies in Row 7 of EBR-II at temperatures of 547 °C and 504 °C, respectively, to additional damage levels each of about 5 dpa. Results obtained on the higher fluence swelling-driven-creep specimen show that compressive stress due to constraint of swelling retards void swelling to a degree that is consistent with active load uniaxial compression specimens that were irradiated as part of a previously reported multiaxial in-reactor creep experiment. Swelling results obtained on the lower fluence swelling-driven creep specimen show a much larger effect of compressive stress in reducing swelling, demonstrating that the larger effect of stress on swelling is on void nucleation as compared to void growth. Test results are analyzed using a recently proposed multiaxial creep-swelling model.     

Fracture toughness of China low activation martensitic (CLAM) steel at room temperature

The fracture toughness (J_IC) of China low activation martensitic (CLAM) steel was tested at room temperature through the compact tension specimen, the result is 417.9 kJ/m², which is similar to the JLF-1 at same experimental conditions. The microstructural observation of the fracture surface shows that the fracture mode is a typical ductile fracture. Meanwhile, the fracture toughness is also calculated on the basis of the fractal dimension and the calculated result is 454.6 kJ/m², which is consistent well with the experimental result. This method could be used to estimate the fracture toughness of materials by analyzing of the fracture surface.     

Tritium permeation in EUROFER in EXOTIC and LIBRETTO irradiation experiments

The reduced activation martensitic steel (RAFM) EUROFER is foreseen as a structural material in test breeder module (TBM) in ITER and breeder blanket in DEMO design. In a number of irradiation experiments conducted in high flux reactor (HFR) in Petten EUROFER was used as a containment wall of the breeder material, through which tritium permeation was monitored on line. Thus in EXOTIC-9/1 (EXtraction Of Tritium In Ceramics) experiment where Li₂TiO₃ pebbles were the breeder material, EUROFER was irradiated up to 1.3 dpa at 340–580 °C. In LIBRETTO experiments (LIBRETTO-4/1, -4/2 and -5) the breeder material was lead lithium eutectic which was in direct contact with the EUROFER containment wall. The neutron damage in steel achieved in the LIBRETTO experiments varied from 2 to 3.5 dpa. The irradiation temperature was 350 °C (LIBRETTO-4/1), 550 °C (LIBRETTO-4/2), and 300–500 °C (LIBRETTO-5).Tritium permeability was studied by varying the irradiation temperature and hydrogen concentration in the purge gas. From the analysis of the temperature transients performed in all four experiments yielded the tritium diffusion coefficients were derived, which appear to be factor ten lower than the literature data obtained in the gas driven permeation experiments.     

Creep-to-rupture of the steel P92 at 650 °C in oxygen-controlled stagnant lead in comparison to air

Creep-to-rupture experiments were performed on 9%-Cr ferritic–martensitic steel P92 in the CRISLA facility. The specimens of P92 were examined at 650 °C and static tensile stress between 75 and 325 MPa in both stagnant lead with 10^?6 mass% dissolved oxygen and air. The steel showed an insignificant difference in time-to-rupture, t_R, and ductile fracture in both environments at >100 MPa, corresponding to t_R < 3,442 h. At 75 MPa in Pb (t_R = 13,090 h), the steel, however, featured purely brittle fracture pointing to liquid metal embrittlement. Structural changes in the steel and surface oxidation in the different environments were studied using metallographic techniques. The Laves phase that forms during thermal aging at 650 °C was found along prior austenite grain boundaries and martensite laths already after relatively short testing time, along with chromium carbides that are already present in the as-received condition of the steel.     

Phase diagram of the ZrO2–FeO system

The results on the ZrO₂–FeO system studies in a neutral atmosphere are presented. The refined eutectic point has been found to correspond to a ZrO₂ concentration of 10.3 ± 0.6 mol% at 1332 ± 5 °C. The ultimate solubility of iron oxide in zirconia has been determined in a broad temperature range, taking into account the ZrO₂ polymorphism. A phase diagram of the pseudobinary system in question has been constructed.     

High temperature deformation and fracture behaviour of 316L stainless steel under high strain rate loading

The high temperature deformation and fracture behaviour of 316L stainless steel under high strain rate loading conditions are investigated by means of a split Hopkinson pressure bar. Impact tests are performed at strain rates ranging from 1 × 10³ s^?1 to 5 × 10³ s^?1 and temperatures between 25 °C and 800 °C. The experimental results indicate that the flow response and fracture characteristics of 316L stainless steel are significantly dependent on the strain rate and temperature. The fracture analysis results indicate that the 316L specimens fail predominantly as the result of intensive localised shearing. Furthermore, it is shown that the flow localisation effect leads to the formation of adiabatic shear bands. The fracture surfaces of the deformed 316L specimens are characterised by a dimple-like structure with knobby features. The knobby features are thought to be the result of a rise in the local temperature to a value greater than the melting point.