Irradiation of structural materials in contact with lead bismuth eutectic in the high flux reactor

In the framework of the materials domain DEMETRA in the European Transmutation research and development project EUROTRANS, irradiation experiment IBIS has been performed in the High Flux Reactor in Petten. The objective was to investigate the synergystic effects of irradiation and lead bismuth eutectic exposure on the mechanical properties of structural materials and welds. In this experiment ferritic martensitic 9 Cr steel, austenitic 316L stainless steel and their welds have been irradiated for 250 Full Power Days up to a dose level of 2 dpa. Irradiation temperatures have been kept constant at 300 °C and 500 °C.During the post-irradiation test phase, tensile tests performed on the specimens irradiated at 300 °C have shown that the irradiation hardening of ferritic martensitic 9 Cr steel at 1.3 dpa is 254 MPa, which is in line with the irradiation hardening obtained for ferritic martensitic Eurofer97 steel investigated in the fusion program. This result indicates that no LBE interaction at this irradiation temperature is present. A visual inspection is performed on the specimens irradiated in contact with LBE at 500 °C and have shown blackening on the surface of the specimens and remains of LBE that makes a special cleaning procedure necessary before post-irradiation mechanical testing.     

Embrittlement of irradiated ferritic/martensitic steels in the absence of irradiation hardening

Irradiation damage caused by neutron irradiation below 425-450 °C of 9-12% Cr ferritic/martensitic steels produces microstructural defects that cause an increase in yield stress. This irradiation hardening causes embrittlement observed in a Charpy impact test as an increase in the ductile-brittle transition temperature. Little or no change in strength is observed in steels irradiated above 425-450 °C. Therefore, the general conclusion has been that no embrittlement occurs above these temperatures. In a recent study, significant embrittlement was observed in F82H steel irradiated at 500 °C to 5 and 20 dpa without any change in strength. Earlier studies on several conventional steels also showed embrittlement effects above the irradiation-hardening temperature regime. Indications are that this embrittlement is caused by irradiation-accelerated or irradiation-induced precipitation. Observations of embrittlement in the absence of irradiation hardening that were previously reported in the literature have been examined and analyzed with computational thermodynamics calculations to illuminate and understand the effect.     

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.     

Tensile properties of ferritic/martensitic steels irradiated in STIP-I

Specimens of ferritic/martensitic (FM) steels T91, F82H, Optimax-A and the electron beam weld (EBW) of F82H were irradiated in the Swiss spallation neutron source (SINQ) Target-3 in a temperature range of 90-370 °C to displacement doses between 3 and 12 dpa. Tensile tests were performed at room temperature and the irradiation temperatures. The tensile test results demonstrated that the irradiation hardening increased with dose up to about 10 dpa. Meanwhile, the uniform elongation decreased to less than 1%, while the total elongation remained greater than 5%, except for an F82H specimen of 9.8 dpa tested at room temperature, which failed in elastic deformation regime. At higher doses of 11-12 dpa, the ductility of some specimens recovered, which could be due to the annealing effect of a short period of high temperature excursion. The results do not show significant differences in tensile properties for the different FM steels in the present irradiation conditions.     

Mechanical properties of irradiated ODS-EUROFER and nanocluster strengthened 14YWT

Irradiations to 1.5 dpa at 300-750 °C were conducted to investigate the changes in mechanical properties of an advanced nanocluster strengthened ferritic alloy, designated 14YWT, and an oxide dispersion strengthened ferritic alloy ODS-EUROFER. Two non-dispersion strengthened variants, 14WT and EUROFER 97, were also irradiated and tested. Tensile results show 14YWT has very high tensile strengths and experienced some radiation-induced hardening, with an increase in room temperature yield strength of 125 MPa after irradiation, while results for ODS-EUROFER show a 275 MPa increase following irradiation. Master curve fracture toughness analysis show 14YWT has a cryogenic T_o reference temperatures before and after irradiation of about −188 and −176 °C, respectively, and upper-shelf K_JIc values between 175 and 225 MPa√m. The favorable fracture toughness properties and resistance to radiation-induced changes in mechanical properties observed for 14YWT are attributed to a fine grain structure and high number density of Y-Ti-O nanoclusters.     

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.     

Tensile behaviour of 9Cr-1Mo tempered martensitic steels irradiated up to 20 dpa in a spallation environment

Tensile specimens of 9Cr-1Mo (EM10) and mod 9Cr-1Mo (T91) martensitic steels in the normalized and tempered metallurgical conditions were irradiated with high energy protons and neutrons up to 20 dpa at average temperatures up to about 360 °C. Tensile tests were carried out at room temperature and 250 °C and a few samples were tested at 350 °C. The fracture surfaces of selected specimens were characterized by Scanning Electron Microscopy (SEM). While all irradiated specimens displayed at room temperature considerable hardening and loss of ductility, those irradiated to doses above approximately 16 dpa exhibited a fully brittle behaviour and the SEM observations revealed significant amounts of intergranular fracture. Helium accumulation, up to about 0.18 at.% in the specimens irradiated to 20 dpa, is believed to be one of the main factors which triggered the brittle behaviour and intergranular fracture mode. One EM10 and one T91 specimen irradiated to 20 dpa were annealed at 700 °C for 1 h following irradiation and subsequently tensile tested. In both cases, a remarkable recovery of ductility and strain-hardening capacity was observed after annealing, while the strength remained significantly above that of the unirradiated material.     

Core materials development for the fuel cycle R&D program

The Fuel Cycle Research and Development program is investigating methods of burning minor actinides in a transmutation fuel. One of the challenges of achieving this goal is to develop fuels capable of reaching extreme burnup levels (e.g. 40%). To achieve such high burnup levels’ fast reactor core materials (cladding and duct) must be able to withstand very high doses (>300 dpa design goal) while in contact with the coolant and the fuel. Thus, these materials must withstand radiation effects that promote low temperature embrittlement, radiation induced segregation, high temperature helium embrittlement, swelling, accelerated creep, corrosion with the coolant, and chemical interaction with the fuel (FCCI).To develop and qualify materials to a total fluence greater than 200 dpa requires development of advanced alloys and irradiations in fast reactors to test these alloys. Test specimens of ferritic/martensitic alloys (T91/HT-9) previously irradiated in the FFTF reactor up to 210 dpa at a temperature range of 350-750 °C are presently being tested. This includes analysis of a duct made of HT-9 after irradiation to a total dose of 155 dpa at temperatures from 370 to 510 °C. Compact tension, charpy and tensile specimens have been machined from this duct and mechanical testing as well as SANS and Mossbauer spectroscopy are currently being performed. Initial results from compression testing and Charpy testing reveal a strong increase in yield stress (∼400 MPa) and a large increase in DBTT (up to 230 °C) for specimens irradiated at 383 °C to a dose of 28 dpa. Less hardening and a smaller increase in DBTT was observed for specimens irradiated at higher temperatures up to 500 °C. Advanced radiation tolerant materials are also being developed to enable the desired extreme fuel burnup levels. Specifically, coatings are being developed to minimize FCCI, and research is underway to fabricate large heats of radiation tolerant oxide dispersion steels with homogeneous oxide dispersions.     

Void swelling in a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions at elevated temperatures

In this work the void swelling behavior of a 9Cr ferritic/martensitic steel irradiated with energetic Ne-ions is studied. Specimens of Grade 92 steel (a 9%Cr ferritic/martensitic steel) were subjected to an irradiation of ²⁰Ne-ions (with 122 MeV) to successively increasing damage levels of 1, 5 and 10 dpa at a damage peak at 440 and 570 °C, respectively. And another specimen was irradiated at a temperature ramp condition (high flux condition) with the temperature increasing from 440 up to 630 °C during the irradiation. Cross-sectional microstructures were investigated with a transmission electron microscopy (TEM). A high concentration of cavities was observed in the peak damage region in the Grade 92 steel irradiated to 5 dpa, and higher doses. The concentration and mean size of the cavities showed a strong dependence on the dose and irradiation temperature. Enhanced growth of the cavities at the grain boundaries, especially at the grain boundary junctions, was observed. The void swelling behavior in similar 9Cr steels irradiated at different conditions are discussed by using a classic void formation theory.     

Post-irradiation mechanical properties of three EUROFER97 joints

Within the development of reduced activation ferritic martensitic (RAFM) steels as prominent structural materials for future fusion reactors, EUROFER97 has recently emerged in Europe as the reference material for the DEMO design. In order to characterise the in-service performance of EUROFER97 as structural material, it is important to assess the properties of welded joints, particularly under irradiation. In the present paper, three EUROFER97 joints (two diffusion welds and one TIG weld) have been irradiated in the BR2 reactor of SCK-CEN at 300 °C up to 1.8 dpa and subsequently characterised for tensile, impact and fracture toughness properties. Comparisons of the results are provided with base EUROFER97 (both unirradiated and irradiated under similar conditions) and, where available, with properties measured on the joints in the unirradiated condition. The post-irradiation mechanical behaviour of both diffusion joints (“laboratory” and “mock-up”) appears similar to that of the base material; therefore, diffusion joining looks a very promising technique. On the other hand, the properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region.     

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%.     

Neutron/proton irradiation and He effects on the microstructure and mechanical properties of ferritic/martensitic steels T91 and EM10

In the European EUROTRANS/DEMETRA program, the synergistic effect of radiation damage and helium on microstructure and mechanical properties of two 9Cr 1Mo ferritic/martensitic (FM) steels T91 and EM10 was evaluated after irradiation in SINQ targets. In addition, the helium induced effect was investigated using helium implanted specimens. The results demonstrate that helium can induce significant embrittlement effect in FM steels as shown by the tremendous increase in ductile-to-brittle transition temperature, the great reduction in ductility and fracture toughness at >∼15 dpa and 1000 appm He and the occurrence of intergranular fracture mode. Further, high-density helium bubbles can produce pronounced hardening effect.     

High temperature irradiation effects in selected Generation IV structural alloys

In the Generation IV Materials Program cross-cutting task, irradiation and testing were carried out to address the issue of high temperature irradiation effects with selected current and potential candidate metallic alloys. The materials tested were (1) a high-nickel iron-base alloy (Alloy 800H); (2) a nickel-base alloy (Alloy 617); (3) two advanced nano-structured ferritic alloys (designated 14YWT and 14WT); and (4) a commercial ferritic-martensitic steel (annealed 9Cr-1MoV). Small tensile specimens were irradiated in rabbit capsules in the High-Flux Isotope Reactor at temperatures from about 550 to 700 °C and to irradiation doses in the range 1.2-1.6 dpa. The Alloy 800H and Alloy 617 exhibited significant hardening after irradiation at 580 °C; some hardening occurred at 660 °C as well, but the 800H showed extremely low tensile elongations when tested at 700 °C. Notably, the grain boundary engineered 800H exhibited even greater hardening at 580 °C and retained a high amount of ductility. Irradiation effects on the two nano-structured ferritic alloys and the annealed 9Cr-1MoV were relatively slight at this low dose.     

Tensile properties of EC316LN irradiated in SINQ to 20 dpa

The wide application of 316-type austenitic stainless steels in existing spallation targets requires a comprehensive understanding of their behavior in spallation irradiation environments. In the present study, EC316LN specimens were irradiated in SINQ targets to doses between 3 and 17.3 dpa at temperatures between about 80 °C and 390 °C. Tensile tests were conducted at room and irradiation temperatures. The results demonstrate that the irradiation induced significant hardening and embrittlement in the specimens. The irradiation hardening and embrittlement effects show a trend of saturation at doses above about 10 dpa. Although the ductility was greatly reduced, all specimens broke with strong necking, which indicates a ductile fracture mode.     

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.     

The effect of neutron irradiation on mechanical properties of YAG laser weldments using previously irradiated material

The objective of this study is to make clear the effect of neutron irradiation on mechanical properties of laser weldments using irradiated material. This estimation is necessary for the application to joining coolant piping of the ITER blanket. Irradiation testing was performed at Japan Material Testing Reactor (JMTR). On the irradiation condition for weldments using irradiated material, fast neutron fluence was 1.4 × 10²⁴ n/m², which corresponds to a displacement damage rate of 0.26 displacement per atom (dpa) and irradiation temperature 200 °C. The results of this study show that tensile properties of all weldments changed into that of base material by the effect of neutron irradiation. The results of hardness tests show that irradiation hardening at an irradiation damage dose of 0.3 dpa is almost same as that at irradiation damage 0.6 dpa. It is concluded that irradiated weldments using irradiated material were moved toward irradiated base material on tensile and hardness properties up to 0.6 dpa. On the other hand, tensile properties of base material were changed by the effect of neutron irradiation up to about 0.3 dpa, and with much less change from 0.3 dpa to 0.6 dpa. It is inferred that the effect of neutron irradiation of SS316LN-IG almost saturated up to 0.3 dpa.     

Liquid Pb-Bi embrittlement effects on the T91 steel after different heat treatments

The present work aims to investigate the susceptibility of ferritic/martensitic steels of different strength to the embrittlement of liquid Pb-Bi eutectic (LBE). Slow strain rate tensile (SSRT) tests on specimens of the T91 steel in three tempering conditions at 500, 600 and 760 °C were conducted in Ar and in LBE at temperatures between 150 and 500 °C. For the specimens tempered at 760 °C (the normal tempering condition) the susceptibility of the steel to LBE embrittlement appeared at temperatures between 300 and 450 °C. With increasing the strength of specimens by lowering the tempering temperature, specimens tempered at 600 and 500 °C demonstrated more pronounced embrittlement effects, reflected by wider and deeper ‘ductility-troughs’. The results suggest that ferritic/martensitic steels with higher strength are more susceptible to LBE embrittlement. The LBE embrittlement effects can be attributed to the decrease of fracture stress resulted from the ‘weakening inter-atomic bond’ by LBE contacting at crack tips.     

Effect of Cr on the mechanical properties and microstructure of Fe-Cr model alloys after n-irradiation

High-chromium ferritic-martensitic steels are candidate structural materials for high-temperature applications in fusion reactors and accelerator driven systems (ADS). Cr concentration has been shown to be a key parameter which needs to be optimized in order to guarantee the best corrosion and swelling resistance, together with the minimum embrittlement. The behavior of Fe-Cr model alloys with different Cr concentrations (0, 2.5, 5, 9 and 12 wt%Cr) has been studied. Tensile tests have been performed in order to characterize the flow properties in the temperature range from −160 °C to 300 °C. The trend of the yield strength with temperature shows that the strain hardening is the same for all materials at low temperatures, even though they have different microstructures. The same materials have been neutron-irradiated at 300 °C in the BR2 reactor of SCK·CEN, up to three different doses (0.06, 0.6 and 1.5 dpa). The results obtained so far indicate that even at these low doses, the Cr content affects the hardening behavior of Fe-Cr binary alloys. Using the Orowan mechanism, the TEM observed microstructure provides an explanation of the obtained hardening but only at the very low dose, 0.06 dpa. At higher doses, other hardening mechanisms are needed.     

Liquid metal embrittlement susceptibility of ferritic-martensitic steel in liquid lead alloys

The susceptibility of the ferritic-martensitic steels T91 and EUROFER97 to liquid metal embrittlement (LME) in lead alloys has been examined under various conditions. T91, which is currently the most promising candidate material for the high temperature components of the future accelerator driven system (ADS) was tested in liquid lead bismuth eutectic (LBE), whereas the reduced activation steel, EUROFER97 which is under consideration to be the structural steel for fusion reactors was tested in liquid lead lithium eutectic. These steels, similar in microstructure and mechanical properties in the unirradiated condition were tested for their susceptibility to LME as function of temperature (150-450 °C) and strain rate (1 × 10⁻³-1 × 10⁻⁶ s⁻¹). Also, the influence of pre-exposure and surface stress concentrators was evaluated for both steels in, respectively, liquid PbBi and PbLi environment. To assess the LME effect, results of the tests in liquid metal environment are compared with tests in air or inert gas environment. Although both unirradiated and irradiated smooth ferritic-martensitic steels do not show any or little deterioration of mechanical properties in liquid lead alloy environment compared to their mechanical properties in gas as function of temperature and strain rate, pre-exposure or the presence of surface stress concentrators does lead to a significant decrease in total elongation for certain test conditions depending on the type of liquid metal environment. The results are discussed in terms of wetting enhanced by liquid metal corrosion or crack initiation processes.     

Microstructural stability of an HT-9 fuel assembly duct irradiated in FFTF

To explore whether the known resistance of fully tempered HT-9 to neutron-induced phase instability and void swelling are maintained under realistic time-dependent reactor operating conditions, the radiation-induced microstructure of an HT-9 ferritic/martensitic hexagonal duct was examined following a 6-year irradiation campaign of a fuel assembly in the Fast Flux Test Reactor Facility (FFTF). Microscopy examination was conducted on specimens irradiated to 4 dpa at 505 °C, 28 dpa at 384 °C and 155 dpa at 443 °C where quoted temperatures are the average operating temperatures over the lifetime of the duct.The dislocation and phase microstructure were observed to remain relatively unchanged at 4 dpa at 505 °C, but significant microstructural changes were observed to have occurred at 28 and 155 dpa and 384 and 443 °C respectively. At these doses the microstructures have experienced precipitation and formation of interstitial loops. In addition, void swelling had occurred at 155 dpa with an average swelling of ∼0.3%, although some local areas swelled as much as 1.2%. In general it appears that this alloy retains its swelling resistance under typical reactor operation conditions up to 155 dpa.