Compatibility of ferritic-martensitic steel T91 welds with liquid lead-bismuth eutectic: Comparison between TIG and EB welds

The 9 wt.% chromium ferritic-martensitic steel T91 is being considered as candidate structural material for a future experimental accelerator driven system (XT-ADS). This material and its welded connections would need to be used in contact with liquid lead-bismuth eutectic (LBE), under high irradiation doses. Both unirradiated tungsten inert gas (TIG) and electron beam (EB) welds of T91 have been examined by means of metallography, scanning electron microscopy (SEM-EDX), transmission electron microscopy (TEM), Vickers hardness measurements and tensile testing in both gas and liquid lead-bismuth environment. The TIG weld was commercially produced and post weld heat treated by a certified welding company while the post weld heat treatment of the experimental EB weld was optimized in terms of the Vickers hardness profile across the welded joint. The mechanical properties of the T91 TIG and EB welds in contact with LBE have been examined using slow strain rate tensile testing (SSRT) in LBE at 350 °C. All welds showed good mechanical behaviour in gas environment but total elongation was strongly reduced due to liquid metal embrittlement (LME) when tested in liquid lead-bismuth eutectic environment. The reduction in total elongation due to LME was larger for the commercially TIG welded joint than for the EB welded joint.     

Liquid metal embrittlement of an austenitic 316L type and a ferritic-martensitic T91 type steel by mercury

The susceptibility to liquid metal embrittlement (LME) of 316L and T91 steels by mercury has been studied at room temperature. A dedicated experimental device using center crack tension (CCT) specimens was built. We developed a specimen preparation procedure that must be rigorously applied in order to investigate the embrittling effect of Hg. The high strength ferritic-martensitic steel of type T91 is embrittled by Hg at room temperature over a large range of crosshead speeds, between 6.67 × 10⁻⁷ and 6.67 × 10⁻³ m s⁻¹. More surprisingly, the austenitic steel of type 316L is also embrittled by Hg between 1.67 × 10⁻⁸ and 2.5 × 10⁻⁴ m s⁻¹. The fracture of the T91 and 316L CCT specimens in contact with Hg occurs by shear band decohesion over the above-mentioned range of crosshead speeds.     

Fracture toughness assessment of ferritic-martensitic steel in liquid lead-bismuth eutectic

The presence of micro-cracks at the surface of a ferritic-martensitic steel is known to favour its embrittlement by liquid metals and thus decrease the mechanical properties of the structural materials. Unfortunately, conventional fracture mechanics methods cannot be applied to tests in liquid metal environment due to the opaque and conducting nature of the LBE. Therefore new methods based on the normalization technique for assessment of plain strain fracture toughness in LBE were examined. This paper discusses the assessment of the plain strain fracture toughness of T91 steel in liquid lead bismuth environment at 473 K, tested at a displacement rate of 0.25 mm min⁻¹ and makes the comparison with results obtained in air at the same temperature and displacement rate. Although there is a decrease of the fracture toughness by 20-30% when tested in LBE, the toughness of the T91 steel remains sufficient under the tested conditions.     

Corrosion behavior of steels in flowing lead-bismuth under abnormal conditions

The project IP EUROTRANS, domain DEMETRA, is primary focused on the study of the technology of the interaction between steels and heavy liquid metals. The characterization of the metal response to sudden changes, simulating accidental conditions in liquid lead-bismuth eutectic was carried out. This paper reports the results of two hot-spot simulations with two different oxygen concentrations (10⁻⁸ wt%, 10⁻⁶ wt%). Each experiment was divided in two main periods: the initial, long period at the standard operating temperature 550 °C; the second, short period, at higher temperature, 650 °C. The damage that occurs on the austenitic steel AISI 316L and the ferritic-martensitic steel T91 was investigated. The amount of damage for both steels was higher at lower oxygen contents and the short, hot spot simulation, markedly affected the T91. At higher oxygen content the amount of damage decreased. A few, localized pits, were observed; however, there was no visible increment in the amount of damage after the hot spot simulation.     

Liquid metal corrosion/erosion investigations of structure materials in lead cooled systems: Part 1

In future lead cooled nuclear power systems the heavy liquid-metal pump will be placed in the hot temperature region of the reactor. This combines corrosion problematic at elevated temperatures with erosion at the impeller of the pump. Several steels designed for conventional mechanical high loaded pumps and the SiSiC have been tested in oxygen containing stagnant lead (10⁻⁶ and 10⁻⁸ wt%) at 550 and 600 °C for 2000 and 4000 h in the COSTA-facilities. Only SiSiC shows no influence by the liquid metal. No dissolution attack occurs at cast iron steels but inner oxidation takes place. NORILOY shows no dissolution attack. All other steels are attacked by liquid lead at one of the conditions. To evaluate the erosion-corrosion attack a new test facility allowing velocities of the lead of up to 20 m/s was designed and constructed. With a CFD-code the behaviour and flow velocity of the lead are simulated.     

On the kinetics of LME for the ferritic-martensitic steel T91 immersed in liquid PbBi eutectic

Use of heavy liquid metals (HLM), such as the Lead-Bismuth eutectic (LBE), is foreseen for targets and coolants in the newly designed accelerator driven systems (ADS). However, most of the structural materials suffer severe damage when in contact with the HLM. In particular, the simultaneous presence of load and environment can induce a phenomenon named Liquid Metal Embrittlement (LME) in the reference material (ferritic-martensitic steel T91). This phenomenon could give rise to high crack growth rates and the abrupt collapse of the structural material.In this study the effect of LBE on the crack growth rate of the T91 steel was evaluated. The effect of pre-exposure to flowing LBE was considered and assessed in terms of crack growth rates and fracture surface appearance.It was concluded that if conditions for wetting were reached during exposure in LBE, e.g. by dissolution or rupture of protective oxide layers, LME was likely to occur under stress, at 300 °C, regardless of the LBE chemistry.     

Irradiation embrittlement characterization of the EUROFER 97 material

The paper summarizes original results of irradiation embrittlement study of EUROFER 97 material that has been proposed as one candidate of structural materials for future fusion energy systems and GEN IV.Test specimens were manufactured from base metal as well as from weld metal and tested in initial unirradiated condition and also after neutron irradiation.Irradiation embrittlement was characterized by testing of toughness properties at transition temperature region - static fracture toughness and dynamic fracture toughness properties, all in sub-size three-point bend specimens (27 × 4 × 3 mm³). Testing and evaluation was performed in accordance with ASTM and ESIS standards, fracture toughness K_JC and K_Jd data were also evaluated with the “Master curve” approach. Moreover, J-R dependencies were determined and analyzed.The paper compares unirradiated and irradiated properties as well as changes in transition temperature shifts of these material parameters. Discussion about the correlation between static and dynamic properties is also given.Results from irradiation of EUROFER 97 show that this steel - base metal as well as weld metal - is suitable as a structural material for reactor pressure vessels of innovative nuclear systems - fusion energy systems and GEN IV. Transition temperature shifts after neutron irradiation by 2.5 dpa dose show a good agreement in the case of EUROFER 97 base material for both static and dynamic fracture toughness tests. From the results it can be concluded that there is a low sensitivity of weld metal to neutron irradiation embrittlement in comparison with EUROFER 97 base metal.     

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

A comparison between pearlitic 2CrMoV steel (WWER-440) and 9% Cr based ferritic-martensitic steels (EUROFER 97 and LA12TaLC) is presented as regards irradiation induced ductile-brittle transition temperature shifts. For neutron doses of 1.5-2 dpa and irradiation temperatures around 300 °C the transition temperature shifts for WWER-440 steel and EUROFER 97 welds are comparable. In the temperature range 350-500 °C the radiation embrittlement levels of both steels are low. Moreover, post-irradiation annealing is proposed as a promising method to predict results of high temperature irradiation embrittlement from existing lower temperature irradiation embrittlement data.     

Susceptibility to LME of 316L and T91 steels by LBE: Effect of strain rate

The effect of liquid lead-bismuth eutectic on 316L and T91 steels at 160 °C has been studied as a function of strain rate, using a centre cracked in tension specimen adapted for the study of crack propagation. Brittle fracture, characterized by elongated river cracks on all the fracture surfaces, indicates that T91 is sensitive to the embrittlement by LBE. This embrittlement effect is very pronounced at low deformation rate (∼10⁻⁵ mm s⁻¹). A ductile-brittle transition is observed in the high strain rate range investigated. In the transitory regime, there is a competition between the growth of dimples and brittle cracking induced by the liquid metal. Ductility recovery is complete at the highest investigated deformation rate. The mechanical properties of the 316L steel are not clearly affected by the presence of LBE, in spite of a modification in the plastic deformation mode which strongly affects fracture surfaces.     

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.     

Mechanical properties of Eurofer 97 in Pb-16Li and irradiation effect

To be used in a fusion reactor, structural materials, and in particular steels, has to be selected and optimised in their composition to achieve a reduction in the long-term radioactive waste. A reduction in the long-term radioactive inventory could be reached substituting elements like molybdenum, niobium and nickel with other ones like tantalum and tungsten which have the same functions as alloying elements and, if irradiated, do not produce long lived radioisotopes. The martensitic steel belonging to the family of 8-9% Cr Eurofer 97 is considered the reference structural steel for fusion application. However, only few information are available about its mechanical properties in the liquid eutectic alloy Pb-16%Li. Particularly, the problem of liquid metal embrittlement (LME) has not been studied in detail and the effect of neutron irradiation on LME has not been investigated at all so far. This work presents the results obtained irradiating tensile specimens of Eurofer 97 up to 5.9 dpa in lead lithium. Tensile tests of samples have been performed out of pile in the same alloy at the same temperature at which irradiation was carried out.     

Nano-indentation measurement of oxide layers formed in LBE on F/M steels

Ferritic/martensitic (F/M) steels (T91, HT-9, EP 823) are candidate materials for future liquid lead or lead bismuth eutectic (LBE) cooled nuclear reactors. To understand the corrosion of these materials in LBE, samples of each material were exposed at 535 °C for 600 h and 200 h at an oxygen content of 10⁻⁶ wt%. After the corrosion tests, the samples were analyzed using SEM, WDX and nano-indentation in cross section. Multi-layered oxide scales were found on the sample surfaces. The compositions of these oxide layers are not entirely in agreement with the literature. The nano-indentation results showed that the E-modulus and hardness of the oxide layers are significantly lower than the values for dense bulk oxide materials. It is assumed that the low values stem from high porosity in the oxide layers. Comparison with in-air oxidized steels show that the E-modulus decreases with increasing oxide layer thickness.     

Corrosion experiments and materials developed for the Japanese HLM systems

The static corrosion tests in lead-bismuth eutectic (LBE) were conducted from 450 °C to 600 °C to understand corrosion behavior and develop corrosion resistant materials for heavy liquid metal systems. While increase of Cr content in steels enhances corrosion resistance in LBE, the effect approaches a constant value above 12 wt% of Cr. Corrosion depth in LBE increases with increasing temperature and corrosion attack becomes severe above 550 °C even under the condition of high oxygen concentration. Nickel dissolution and Pb-Bi penetration occur in 316SS and JPCA above 550 °C under the condition of high oxygen concentration. When oxygen concentration decreases below the level of Fe oxide formation, corrosion attack on these steels also becomes violent due to dissolution of various elements and grain boundary corrosion. Whereas additions of 1.5 wt% Si to T91 and 2.5 wt% Si to 316SS improve corrosion resistance, the effect is insufficient taking fluctuation of oxygen concentration in LBE into consideration. Furthermore, addition of 1.5 wt% Si to T91 causes rise in DBTT. A new coating method using Al, Ti and Fe powders produces corrosion resistant coating layers on 316SS. The coating layers containing 6-8 wt% Al exhibit good corrosion resistance at 550 °C for 3000 h in LBE containing 10⁻⁶-10⁻⁴ wt% of oxygen.     

Fine structure behaviour of VVER-1000 RPV materials under irradiation

Changes in the fine structure and mechanical properties of the base metal (BM) and weld metal (WM) of VVER-1000 pressure vessels during accumulation of neutron dose in the range of fluences ∼(3.2-15) × 10²³ m⁻² (E > 0.5 MeV) at 290 °C are studied using methods of transmission electron microscopy, fractographic analysis, and Auger electron spectroscopy. A correlation was found between the changes of mechanical properties and the micro- and nano-structures of the studied steels. Accumulation of neutron dose considerably raises the strength characteristics and transition temperature of VVER-1000 pressure vessel steels. The rate of changes in the mechanical properties of the weld metal is significantly higher than that of the base metal. The slower growth of strength characteristics and transition temperature shift of the base metal under irradiation as compared with the weld metal is due to the slower growth of the density of radiation defects and radiation-induced precipitates. The level of intergranular embrittlement under irradiation in the weld metal is not higher then in the base metal in spite of the higher content of nickel.     

T91 cladding tubes with and without modified FeCrAlY coatings exposed in LBE at different flow, stress and temperature conditions

Corrosion tests of 2000 h duration are conducted on tubes consisting of the steel T91 in liquid metal loops containing eutectic lead-bismuth melt with 10⁻⁶ wt% oxygen in solution. The experiments include tests at temperatures of 480-600° C, at liquid metal flow velocities of 1, 2 and 3 m/s and under mechanical stress due to an internal pressure of 15 MPa. The surface of tubes exposed to 600 °C and to different flow velocities are coated with a FeCrAlY alloy to examine its suitability as a protective coating for high loaded parts like cladding tubes. The coating was remelted by an electron pulse of GESA to homogenize the coating and improve its bonding to the bulk material. In all of the tests no liquid metal attack was observed. As received steel specimens developed multilayer oxide scales of a thickness increasing with temperature and internal pressure, while coated tubes had a thin protective alumina scale. Flow velocities above 2 m/s permanently removed formed magnetite at 550 °C. No influence of the flow velocity was observed for the coated surfaces which keep their stable thin alumina scale. The internal pressure of 15 MPa caused a strain of 0.7% in the tube wall, which obviously increases iron diffusion and enhances magnetite formation.     

Compatibility of Ti3SiC2 with liquid Pb and PbBi containing oxygen

Ti₃SiC₂ is considered as a possible pump material for heavy liquid metal (HLM) cooled nuclear systems because it resists oxygen and has high temperature strength. Therefore the material was tested in stagnant lead and lead bismuth containing oxygen at temperatures between 550 and 750 °C up to 4000 h in the Karlsruher COSTA facilities. Two different oxygen concentrations (10⁻⁶ and 10⁻⁸ wt%) were chosen. During the exposure a thin TiO_x-layer was formed at the surface of the specimens. No dissolution attack was observed under these conditions.     

Influence of Zn as a spallation product on the behaviour of martensitic steel T91 and austenitic steel 316L in liquid Pb-Bi

The liquid Pb-Bi alloy is proposed as material for the spallation target in hybrid systems. During the spallation process, several chemical elements are produced in the target which could generate specific liquid metal embrittlement phenomena. Among these species, zinc is known as an element which can promote LME (liquid metal embrittlement). Corrosion tests were carried out in liquid Pb-Bi in isothermal static conditions without and with 80 wppm of zinc at 150 °C, 350 °C and 600 °C up to 6000 h. No modification of the corrosion kinetics of T91 martensitic and 316L austenitic steels was observed for either unstressed or U-bend specimens with zinc in Pb-Bi. Moreover, no sign of embrittlement was observed for any of the samples with and without zinc.     

Corrosion behaviour of stainless steels in flowing LBE at low and high oxygen concentration

The corrosion behaviours of austenitic steel AISI 316L and martensitic steel T91 were investigated in flowing lead-bismuth eutectic (LBE) at 400 °C. The tests were performed in the LECOR and CHEOPE III loops, which stood for the low oxygen concentration and high oxygen concentration in LBE, respectively. The results obtained shows that steels were affected by dissolution at the condition of low oxygen concentration (C_[O2] = 10⁻⁸-10⁻¹⁰ wt%) and were oxidized at the condition of high oxygen concentration (C_[O2] = 10⁻⁵-10⁻⁶ wt%). The oxide layers detected are able to protect the steels from dissolution in LBE. Under the test condition adopted, the austenitic steel behaved more resistant to corrosion induced by LBE than the martensitic steel.     

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.     

Evolution of the nanostructure of VVER-1000 RPV materials under neutron irradiation and post irradiation annealing

A high nickel VVER-1000 (15Kh2NMFAA) base metal (1.34 wt% Ni, 0.47% Mn, 0.29% Si and 0.05% Cu), and a high nickel (12Kh2N2MAA) weld metal (1.77 wt% Ni, 0.74% Mn, 0.26% Si and 0.07% Cu) have been characterized by atom probe tomography to determine the changes in the microstructure during neutron irradiation to high fluences. The base metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 14.9 × 10²³ m⁻² (E > 0.5 MeV), and the weld metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 11.5 × 10²³ m⁻² (E > 0.5 MeV). High number densities of ∼2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals. No significant copper enrichment was associated with these nanoclusters and no copper-enriched precipitates were observed. The number densities of these nanoclusters correlate with the shifts in the ΔT_{41 J} ductile-to-brittle transition temperature. These nanoclusters were present after a post irradiation anneal of 2 h at 450 °C, but had dissolved into the matrix after 24 h at 450 °C. Phosphorus, nickel, silicon and to a lesser extent manganese were found to be segregated to the dislocations.