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.     

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.     

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.     

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.     

Sodium fast reactor evaluation: Core materials

In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.     

SANS investigation of a neutron-irradiated Fe-9 at%Cr alloy

Available experimental results indicate that the addition of Cr to Fe and steels significantly influences the response of Fe-Cr alloys and ferritic/martensitic high-Cr steels to neutron irradiation. A level of 9 at%Cr is of particular interest because this composition is close to the boundary of the Fe-Cr miscibility gap. Furthermore, it corresponds to the composition of several candidate steels for application in nuclear technology. However, experimental evidence has been incomplete so far. The reported study by means of small-angle neutron scattering is devoted to the effect of neutron irradiation at 300 °C up to fluences of 0.6 and 1.5 dpa on the microstructure of an Fe-9 at%Cr alloy. We have observed a pronounced irradiation-induced increase of scattering cross-sections for both magnetic and nuclear scattering. Bimodal size distributions of irradiation-induced defect-solute clusters have been reconstructed. The restrictions on the composition of these clusters have been discussed in terms of the scattering contrast. We have found that vacancy clusters and α′-particles alone cannot explain the full set of experimental findings. The remaining inconsistency can be solved by taking into account a contribution of impurity carbon.     

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.     

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.     

Irradiation Performance of Fast Reactor MOX Fuel Assemblies Irradiated to High Burnups

Mixed oxide fuel assemblies (MFA-1 and MFA-2 assemblies) were irradiated in the fast flux test facility to evaluate the irradiation performance of fast reactor core fuels at high burnups and high fast neutron fluences. The MFA-1 and MFA-2 assemblies achieved respective peak pellet burnups of 147 and 162GWd/t, and resisted to respective peak fast neutron fluences (E > 0:1 MeV) of 21:4 _ 1026 and 23:8 _ 1026 n/m2, without any indication of fuel pin breaching. Structural components of these assemblies were made of modified type 316 stainless steel and 15Cr-20Ni base advanced austenitic stainless steel. Postirradiation examinations of these assemblies revealed dimensional changes of fuel pins and assembly ducts due to irradiation-induced void swelling and irradiation creep, and fuel cladding local oval distortions due to bundle-duct interaction (BDI). The swelling resistance of 15Cr-20Ni base advanced austenitic stainless steel fuel pin cladding was almost the same as that of the modified type 316 stainless steel cladding, while the assembly duct of the former material had a slightly higher swelling resistance than that of the latter material. Analyses of fuel pin bundle deformations indicated that these assemblies likely mitigate BDI mainly by fuel pin bowings and cladding oval distortions.     

Nanoindentation on ion irradiated steels

Radiation induced mechanical property changes can cause major difficulties in designing systems operating in a radiation environment. Investigating these mechanical property changes in an irradiation environment is a costly and time consuming activity. Ion beam accelerator experiments have the advantage of allowing relatively fast and inexpensive materials irradiations without activating the sample but do in general not allow large beam penetration depth into the sample. In this study, the ferritic/martensitic steel HT-9 was processed and heat treated to produce one specimen with a large grained ferritic microstructure and further heat treated to form a second specimen with a fine tempered martensitic lath structure and exposed to an ion beam and tested after irradiation using nanoindentation to investigate the irradiation induced changes in mechanical properties. It is shown that the HT-9 in the ferritic heat treatment is more susceptible to irradiation hardening than HT-9 after the tempered martensitic heat treatment. Also at an irradiation temperature above 550 °C no detectable hardness increase due to irradiation was detected. The results are also compared to data from the literature gained from the fast flux test facility.     

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.     

Temperature dependence of lattice disorder in Ar-irradiated (1 0 0), (1 1 0) and (1 1 1) MgO single crystals

To better appreciate dynamic annealing processes in ion irradiated MgO single crystals of three low-index crystallographic orientations, lattice damage variation with irradiation temperature was investigated. Irradiations were performed with 100 keV Ar ions to a fluence of 1 × 10¹⁵ Ar/cm² in a temperature interval from −150 to 1100 °C. Rutherford backscattering spectroscopy combined with ion channeling was used to analyze lattice damage. Damage recovery with increasing irradiation temperature proceeded via two characteristic stages separated by 200 °C. Strong radiation damage anisotropy was observed at temperatures below 200 °C, with (1 1 0) MgO being the most radiation damage tolerant. Above 200 °C damage recovery was isotropic and almost complete recovery was reached at 1100 °C. We attributed this orientation dependence to a variation of dynamic annealing mechanisms with irradiation temperature.     

Response of nanoclusters in a 9Cr ODS steel to 1 dpa, 525 °C proton irradiation

Ferritic-martensitic (F/M) alloys are expected to play an important role as cladding or structural components in Generation IV and other advanced nuclear systems operating in the temperature range 350-700 °C and to doses up to 200 displacements per atom (dpa). Oxide dispersion strengthened (ODS) F/M steels have been developed to operate at higher temperatures than traditional F/M steels. These steels contain nanometer-sized Y-Ti-O nanoclusters for additional strengthening. A proton irradiation to 1 dpa at 525 °C has been performed on a 9Cr ODS steel to determine the nanocluster stability at low dose. The evolution of the nanocluster population and the composition at the nanocluster-matrix interface were studied using electron microscopy and atom probe tomography. The data from this study are contrasted to those from a previous study on heavy-ion irradiated 9Cr ODS steel.     

Mechanical properties of high-nickel alloys EP-753 and РЕ-16 after neutron irradiation to 54 dpa at 400-650 °С

Short-term mechanical properties and void swelling were investigated for high-nickel alloys РЕ-16 and three compositional variants of Russian alloy EP-753 and in various starting conditions after side-by-side irradiation in the BN-350 fast reactor at 400, 500, 600 and 650 °С to 54 dpa. For both alloys irradiation resulted in significant hardening and ductility reduction dependent on their chemical composition and initial heat treatment. At test temperatures equal to the irradiation values both alloys exhibited a high level of strength and satisfactory ductility. In the test temperature range of 550-650 °С the phenomenon of high-temperature irradiation embrittlement was observed.     

CHF experiment and CFD analysis in a 2 × 3 rod bundle with mixing vane

In this study, the CHF enhancement using various mixing vanes is evaluated and the flow characteristics are investigated through the CHF experiments and CFD analysis.CHF tests were performed using 2 × 2 and 2 × 3 rod bundles and with R-134a as the working fluid. The test section geometry was identical to that of commercial PWR fuel assembly not including the heated length (1.125 m) and number of fuel rods. From the CHF tests, it was found that the CHF enhancement using mixing vanes under higher mass flux (1400 kg/m² s) and lower pressure (15 bar) conditions is larger than the CHF enhancements under other conditions. Among the mixing vanes used in this study, the swirl vane showed the best performance under relatively low pressure (15 bar) and mass flux (300-1000 kg/m² s) conditions and the hybrid vane performed best near the PWR operating conditions.The detailed flow characteristics were also investigated by CFD analysis using the same conditions as the CHF tests. To calculate the subcooled boiling flow, the wall partitioning model was applied to the wall boundary and various two-phase parameters were also considered. The reliability of the CFD analysis in the boiling analysis was confirmed by comparing the average void fractions of the analysis and the experiments: the results agreed well. From the CFD analysis, the void fraction flattening as a result of the lateral velocity induced by the mixing vane was observed. By the lateral motion of the liquid, the void fraction in the near wall was decreased and that of the core region was increased resulting in the void fraction flattening. The decrease of the void fraction in the near wall region promoted liquid supply to the wall and consequently the CHF increased. For the quantification of the void flatness, an index was developed and the applicability of the index in the CHF assessment was confirmed.     

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.     

Irradiation hardening in unalloyed and ODS molybdenum during low dose neutron irradiation at 300 °C and 600 °C

Unalloyed molybdenum and oxide dispersion strengthened (ODS) molybdenum were irradiated at 300 °C and 600 °C in HFIR to neutron fluences of 0.2, 2.1, and 24.3 × 10²⁴ n/m² (E > 0.1 MeV). The size and number density of voids and loops as well as the measured irradiation hardening and electrical resistivity were found to increase sub-linearly with fluence. This supports the idea that the formation of the extended defects that produce irradiation hardening in molybdenum is the result of a nucleation and growth process rather than the formation of sessile defects directly from the displacement damage cascades. This conclusion is further supported by molecular dynamics (MD) simulations of cascade damage. The unalloyed molybdenum had a low impurity interstitial content with less irradiation hardening and lower change in electrical resistivity than is observed for ODS Mo. This result suggests that high-purity can result in slightly improved resistance to irradiation embrittlement in molybdenum at low fluences.     

Study of 200 MeV Ag ion induced amorphisation in nickel ferrite thin films

Thin films of nickel ferrite of thickness ∼100 and 150 nm were deposited by pulsed laser deposition. The films were irradiated with a 200 MeV Ag¹⁵⁺ beam of three fluences 1 × 10¹², 2 × 10¹² and 4 × 10¹² ions/cm². X-ray diffraction showed a decrease in the intensity of peaks indicating progressive amorphisation with increased irradiation fluence. Fourier transform infra-red and Raman spectra of pristine and irradiated films were also recorded which showed a degradation of the crystallinity of the samples after irradiation. The damage cross section of the infra-red bands was determined. It was found that the two bands at 557 and 614 cm⁻¹ did not show similar behaviour with fluence.     

Transmission electron microscopy investigation of irradiated U-7 wt%Mo dispersion fuel

The microstructural evolution of atomised U-7 wt%Mo alloy fuel under irradiation was investigated by transmission electron microscopy on material from the experimental fuel plates used in the FUTURE irradiation. The interaction layer that forms between the U(Mo) particles and the Al matrix is assumed to become amorphous under irradiation and as such cannot retain the fission gas in stable bubbles. As a consequence, gas filled voids are generated between the interaction layer and the matrix, causing the fuel plate to pillow and finally fail. The present analysis confirms the assumption that the U(Mo)-Al interaction layer is completely amorphous after irradiation. The Al matrix and the individual U(Mo) particles, with their cellular substructure, have retained their crystallinity. It was furthermore observed that the fission gas generated in the U(Mo) particles has formed a bubble superlattice, which is coherent with the U(Mo) lattice. Bubbles of roughly 1-2 nm size have formed a 3-dimensional lattice with a lattice spacing of 6-7 nm.     

Change in magnetic properties of MgB2 thin films after irradiation by 200 MeV Au ion beam

The SHI irradiation induced effects on magnetic properties of MgB₂ thin films are reported. The films having thickness 300-400 nm, prepared by hybrid physical chemical vapor deposition (HPCVD) were irradiated by 200 MeV Au ion beam (S_e ∼ 23 keV/nm) at the fluence 1 × 10¹² ion/cm². Interestingly, increase in the transition temperature T_c from 35.1 K to 36 K resulted after irradiation. Substantial enhancement of critical current density after irradiation was also observed because of the pinning provided by the defects created due to irradiation. The change in surface morphology due to irradiation is also studied.