HT-9 Steel Research Articles

Long-term thermal aging effects in ferritic-martensitic steel HT9

The ferritic-martensitic steel HT9 is a candidate material for fuel cladding and core components in advanced nuclear reactors, such as sodium-cooled fast reactors, thanks to their high temperature mechanical properties and low susceptibility to irradiation induced swelling phenomena. However, thermal stability and elevated temperature microstructural evolution in these alloys may impact their long-term behavior and reliability. In this work, the effects of thermal aging on the microstructural and mechanical properties of HT9 have been investigated through complementary electron microscopy, synchrotron X-ray diffraction, microhardness, and thermodynamic modeling. Plates of HT9 were aged up to 50 kh at relevant sodium-cooled fast reactor operational temperatures (360 °C - 700 °C). Trends in microstructure as a function of aging time and temperature were apparent from qualitative and quantitative analysis. These observations were further supported by thermodynamic modeling of the bulk and precipitate phases. Specific phases observed include BCC Fe, FCC M23C6, HCP and FCC MX phase and Laves M2X phase. Through the application of our multi-scale and multi-modal approach, clear information on the aging mechanism of HT9 was obtained, allowing for a more informed prediction, and understanding of the long-term behavior, performance and thermal stability of ferritic-martensitic alloys exposed to elevated temperatures.

Thermomechanical Processing for Improved Mechanical Properties of HT9 Steels.

Thermomechanical processing (TMP) of ferritic-martensitic (FM) steels, such as HT9 (Fe-12Cr-1MoWV) steels, involves normalizing, quenching, and tempering to create a microstructure of fine ferritic/martensitic laths with carbide precipitates. HT9 steels are used in fast reactor core components due to their high-temperature strength and resistance to irradiation damage. However, traditional TMP methods for these steels often result in performance limitations under irradiation, including embrittlement at low temperatures (<~430 °C), insufficient strength and toughness at higher temperatures (>500 °C), and void swelling after high-dose irradiation (>200 dpa). This research aimed to enhance both fracture toughness and strength at high temperatures by creating a quenched and tempered martensitic structure with ultrafine laths and precipitates through rapid quenching and unconventional tempering. Mechanical testing revealed significant variations in strength and fracture toughness depending on the processing route, particularly the tempering conditions. Tailored TMP approaches, combining rapid quenching with limited tempering, elevated strength to levels comparable to nano-oxide strengthened ferritic alloys while preserving fracture toughness. For optimal properties in high-Cr steels for future reactor applications, this study recommends a modified tempering treatment, i.e., post-quench annealing at 500 °C or 600 °C for 1 h, possibly followed by a brief tempering at a slightly higher temperature.

General corrosion behavior and mechanism of ferritic/martensitic steel with and without TaWVCr coating in static oxygen-saturated lead bismuth eutectic at 550 °C

Here in present work, the ferritic/martensitic (F/M) steel HT9 and refractory TaWVCr high entropy alloy coated HT9 were tested in a static oxygen-saturated lead bismuth eutectic (LBE) at 550 °C through corrosion tests and slow strain rate tension tests to investigate the corrosion behavior and mechanism. The fracture pattern of HT9 and TaWVCr/HT9 both show the ductile fracture. Comprehensive microstructural characterizations from atomic-scale examinations to micro-scale fracture analyses were performed to illustrate the underlying mechanisms of the liquid LBE corrosion. It was revealed that the corrosion of HT9 is related to the surface oxide scale cracking, which propagates into the matrix but terminated in the Cr2O3 enriched zone at the crack tip. On the contrary, the stress concentration caused by local degradation of the TaWVCr coating accelerates the oxidation and dissolution of the coating in LBE, leading to the formation of a duplex oxide layer underneath the coating, which allows cracks initiate in the magnetite layer and annihilate in the chromite layer. The micrometer-scale penetrative oxide layer was found within the matrix, due to the inward diffusion of O through the diffusion channels such as dislocation channels of the matrix and the dissolution channels of the coating and this can accelerate the oxidation corrosion of the matrix.

Comparison of hardening and microstructures of ferritic/martensitic steels irradiated with fast neutrons and dual ions

Ferritic/martensitic steels T91 and HT9 were irradiated with neutrons (BOR-60 reactor) and dual ions (9 MeV Fe3+ and 3.42 MeV energy degraded He2+) from 369 to 520 °C and damage levels of 16.6 to 72 dpa to quantify the possibility of using ion irradiation to simulate neutron irradiation in terms of microstructures and mechanical properties. Nanoindentation testing was performed to obtain the bulk equivalent hardness of the dual-ion irradiated samples. For the neutron irradiated samples, both nanoindentation and Vickers hardness testing were conducted. Transmission Electron Microscopy (TEM) characterizations of the cavities, dislocation loops and precipitates were conducted to account for the strengthening contribution of each microstructure element. The good agreement between the microstructure-predicted (dispersed barrier hardening) and measured strength of the irradiated specimens demonstrated the accuracy of the strengthening model and the nanoindentation tests. The comparison of mechanical property and microstructure changes in ion and neutron irradiated structural materials indicated that ion irradiation replicated many neutron irradiation features. However, a single 70 °C temperature shift is insufficient to match all complex microstructures of neutron vs. ion irradiation over the irradiation temperature range of 369–520 °C.

Direct and pulse electroplating effects on the diffusion barrier property of plasma-nitrided Cr coatings on HT9 steel

The introduction of a Cr and chromium nitride multi-coating barrier between the nuclear metallic fuel and the cladding is a potential candidate to mitigate fuel cladding chemical interaction (FCCI). Cr coatings were electroplated on HT9 disks using direct and pulse currents followed by plasma nitriding for synthesizing additional nitride coating. The pulse current induced Cr coatings revealed severe microstructural changes from crack-free to the abundance of cracks during the plasma nitriding due to phase transformation during the plasma nitriding. The phase transformation of hexagonal CrH to cubic Cr occurred at a temperature as low as 110 °C resulting in the liberation of hydrogen pores agglomerated at Cr grain boundaries and internal stress. Finally, the diffusion barrier property of the pulse current induced Cr coatings was significantly reduced, while the plasma-nitrided Cr coating produced by a direct current exhibited a crack-closure effect, and an additional nitride layer successfully hindered the elemental diffusion of Ce and Nd at 650 °C for 25 h. This current study points to the critical effect of current waveforms on the final barrier property of Cr-chromium nitride coatings for nuclear cladding applications.

Microchemical evolution of irradiated additive-manufactured HT9

The microstructural responses under 5 MeV Fe2+ single-ion-beam irradiation of three conditions of additive-manufactured (AM) HT9 steel using a powder-based directed energy deposition (DED) technique with and without postbuild heat treatments were investigated. Besides the observed dislocation loop formation and the absence of cavities at the irradiation condition of 50 dpa at 460 °C, Ni/Si/Mn-rich precipitates are found to form in all three conditions of AM-HT9, whereas Cu-rich clusters that arise from Cu uptake from the DED process are only observed in the heat-treated conditions, and not in the as-built (ASB) condition. Coprecipitation of the Cu- and Ni/Si/Mn-rich clusters occur near defect sinks such as line dislocations and grain boundaries in the heat-treated AM-HT9. The variation in microchemical evolution can be directly linked to the starting sink strength of the 3 AM-HT9 conditions, and the ASB condition with higher sink strength suppressed the responses observed in the postbuild heat-treated specimens.

The influence of nitrogen and nitrides on the structure and properties of proton irradiated ferritic/martensitic steel

The 12Cr1MoWV (wt%) ferritic/martensitic steel HT9 is a candidate material for fuel cladding in advanced nuclear reactors, such as the Versatile Test Reactor currently under development. As such, understanding the relationship between microstructure and mechanical properties in the context of irradiation environments for these steels is critical. N content, and more specifically interstitial N, has been hypothesized to be detrimental to irradiated properties at lower temperatures (less than 0.3Tm) to a total of 6 dpa; however, in this work at a dose of 1 dpa the irradiated microstructure was improved with added N, leading to less irradiation hardening. Three variants of HT9 were irradiated with 1.5 MeV protons to a dose of 1 dpa at 300 ˚C. The HT9 variants included Low (10 ppm), Mid (190 ppm), and High (440 ppm) N alloys that were otherwise nearly identical. Changing the N content had a variety of effects on the irradiated defect structures. As N content increased, the average dislocation loop diameter decreased, while the number density of loops increased. Additionally, extensive Ni clustering was observed on dislocations and interfaces. The Mid and High N specimens exhibited significantly less hardening (ΔHV≅100) relative to the Low N specimen (ΔHV≅160). The decrease in hardening is attributed to vanadium carbonitride acting as a sink for Ni clusters that would otherwise form on dislocations. Under the irradiation conditions used, these results suggest increasing the N content in HT9 may have a desirable effect on the irradiated structure and properties at the dose studied, as well as the swelling resistance at higher doses. In other words, N content appears to be a powerful tool for tailoring the self-interstitial atom cluster mobility in F/M steels for different temperature and dose applications.

Synthesis of plasma-nitrided Cr coatings on HT9 steel for advanced chemical barrier property in a nuclear cladding application

Chromium–chromium nitride multi-coating barriers were synthesized on HT9 steel disks using the electroplating and radiofrequency (RF) plasma nitriding techniques, respectively, to significantly mitigate fuel–cladding chemical interactions in sodium fast reactors. It was revealed that plasma nitriding not only synthesized an additional chromium nitride layer at the outer surface of the Cr coating but also had a significant crack-closure effect on hard Cr coatings. Further, Cr coatings nitrided for durations ranging from 0 to 72 h were investigated to observe crack-closure stages; notably, nitriding for 72 h at 600 °C and an RF power of 180 W resulted in the elimination of most visible interior and penetration cracks. Moreover, an additional 0.75–1 μm thick nitride layer at the outer surface of the Cr coating included columnar grains of CrN and Cr2N phases aligned with the growth direction. Diffusion couple tests of Ce–Nd alloys at 650 °C for 25 h revealed that the 72-h-nitrided Cr coating possessed an excellent barrier property without any visible inter-diffusion, in contrast with the untreated Cr coating, for which a substantial inter-diffusion area of up to 17,500 μm2 was observed.

Phase stability and microstructural evolution in neutron-irradiated ferritic-martensitic steel HT9

Ferritic Martensitic (F/M) steel HT9 specimens were irradiated in the Advanced Test Reactor up to 4.16 dpa in three temperature ranges (roughly from 300 to 600 °C). The post-irradiation microstructure, including dislocation structure , precipitation and radiation-induced segregation (RIS) was characterized using analytical scanning / transmission electron microscopy (S/TEM), and atom probe tomography (APT). Irradiation hardening was measured using nanoindentation . The results reveal a distinctive pattern of dislocation and precipitate evolution at high temperature, around 600 °C, where various defects and precipitates formed in the low dose regime followed by a recovering process with increasing dose. Dislocation loops formed in all temperature ranges, and the growth of dislocation loops is unconstrained above certain critical temperature, contributing to the increasing dislocation density even prior to doses of 0.5 dpa at 600 °C. Ni/Mn/Si clusters were identified in all temperature ranges and the compositions of these clusters converged to G phase stoichiometrically. Significant coarsening of G phase particles was observed at 600 °C, accompanied by the formation of G phase on grain boundaries. α ’ precipitates were only found in the medium and low temperature ranges (below 500 °C). The number density and volume fraction were higher in the low temperature specimens, while larger particles were observed in the medium temperature range. RIS of Cr, Ni, Mn, Si, P was identified at dislocation lines , grain boundaries and phase boundaries, and the temperature dependence is consistent with previous studies. The RIS of Cr to the existing VN particles was confirmed by APT and may accelerate the transition of VN to Cr-rich nitrides . The irradiation hardening contribution from dislocation loops, dislocation lines, G phase and α ’ phase was parsed based on a linear dispersed barrier hardening model. The results suggest that most irradiation hardening at high temperature is due to increasing dislocation density with dose.

Nano-mechanical property assessment of a neutron-irradiated HT-9 steel cladding and a fuel-cladding chemical interaction region of a uranium–10 wt% zirconium nuclear fuel

Nano-mechanical property assessment of a neutron-irradiated HT-9 steel cladding and a fuel-cladding chemical interaction region of a uranium–10 wt% zirconium nuclear fuel

Limitations of Thermal Stability Analysis via In-Situ TEM/Heating Experiments.

This work highlights some limitations of thermal stability analysis via in-situ transmission electron microscopy (TEM)-annealing experiments on ultrafine and nanocrystalline materials. We provide two examples, one on nanocrystalline pure copper and one on nanocrystalline HT-9 steel, where in-situ TEM-annealing experiments are compared to bulk material annealing experiments. The in-situ TEM and bulk annealing experiments demonstrated different results on pure copper but similar output in the HT-9 steel. The work entails discussion of the results based on literature theoretical concepts, and expound on the inevitability of comparing in-situ TEM annealing experimental results to bulk annealing when used for material thermal stability assessment.

Stable, Ductile and Strong Ultrafine HT-9 Steels via Large Strain Machining.

Beyond the current commercial materials, refining the grain size is among the proposed strategies to manufacture resilient materials for industrial applications demanding high resistance to severe environments. Here, large strain machining (LSM) was used to manufacture nanostructured HT-9 steel with enhanced thermal stability, mechanical properties, and ductility. Nanocrystalline HT-9 steels with different aspect rations are achieved. In-situ transmission electron microscopy annealing experiments demonstrated that the nanocrystalline grains have excellent thermal stability up to 700 °C with no additional elemental segregation on the grain boundaries other than the initial carbides, attributing the thermal stability of the LSM materials to the low dislocation densities and strains in the final microstructure. Nano-indentation and micro-tensile testing performed on the LSM material pre- and post-annealing demonstrated the possibility of tuning the material’s strength and ductility. The results expound on the possibility of manufacturing controlled nanocrystalline materials via a scalable and cost-effective method, albeit with additional fundamental understanding of the resultant morphology dependence on the LSM conditions.

High-throughput nitride and interstitial nitrogen analysis in ferritic/martensitic steels via time-of-flight secondary ion mass spectrometry

The following work focuses on characterizing nitrogen in several modified alloys of the precipitation strengthened ferritic/martensitic steel HT9 (12Cr-1MoWV (wt%)). Alloy HT9 is a candidate material for fuel rod cladding and ductwork in the core of the Versatile Test Reactor and other next-generation nuclear fast reactors. Nitrogen content has been shown to have a significant effect on irradiated properties in alloy HT9. To explore that theory, a more comprehensive assessment of the location of nitrogen must be performed; however, locating nano-precipitates, such as nitrides and light interstitial elements like nitrogen in steel, has proven difficult. To begin answering questions surrounding the location of nitrogen, Time-of-Flight Secondary Ion Mass Spectrometry (ToF-SIMS) was employed as a primary analysis technique, together with extensive transmission electron microscopy, internal friction testing, and thermodynamic simulations, providing verification and context to trends found in the ToF-SIMS data. For the first time, ToF-SIMS was demonstrated to reliably measure relative differences in the vanadium carbonitride precipitate volume fraction across several chemistries and heat treatments of ferritic/martensitic steels. Additionally, associating matrix elements (iron and chromium) to nitrogen-containing ion clusters significantly reduced the ‘matrix effect that biases nitrogen detection towards nitride precipitates, providing a potential, high-throughput method for the measurement of relative interstitial nitrogen content in steels with ToF-SIMS. • ToF-SIMS provided a new high-throughput analysis method for microalloy precipitate and microstructure design. • ToF-SIMS was employed for semi-quantitative, bulk detection and measurement of nitrides and interstitial nitrogen in steel. • S/TEM provided a quantitative comparison of nitride content, while internal friction testing provided a semi-quantitative comparison of interstitial N.

Feasibility of Sodium-Cooled Breed-and-Burn Reactor with Rotational Fuel Shuffling

The objective of the present study is to show that it is feasible to establish the breed-and-burn (B&B) mode of operation with rotational fuel shuffling in the S-PRISM core based on neutronic and thermal-hydraulic analyses. The results quantified the impact of major core design choices on the criticality of a core that uses sodium as a coolant material and HT9 steel as structural material. The design variables examined include the binary metallic fuel U-Zr with different weight percentages of zirconium as well as different core heights and fuel rod pitch-to-diameter ratios (P/Ds) in the fuel assembly. We found that a core using the binary metallic fuel U-Zr with 2 wt% zirconium, with a core height of 200 cm, a P/D of 1.086, and a core power of 400 MW(thermal), could overcome some major design constraints. It was also found that with shuffling intervals of 1125 to 1250 days, the core with rotational fuel shuffling was critical in the equilibrium state, and the possible average discharged burnup was from 274.8 to 305.3 GWd/ton HM. Reactor characteristics such as neutron flux and power profile were almost stable during the equilibrium cycle. A steady-state thermal-hydraulic analysis was performed for the hottest channel in the core. It revealed that both the fuel and cladding maximum temperatures were less than the melting point of the fuel and the chemical interaction temperature of the HT-9, respectively. The mixed coolant outlet temperature was somewhat below the temperature usually observed in sodium-cooled fast reactors. Thus, it appears that the S-PRISM can be principally designed to have a B&B core.

Investigation of hardening mechanisms and size effects in proton-irradiated HT-9 steels

Ferritic/martensitic steels, such as HT-9, are known for their complex microstructural features and mechanical properties. In this paper, in-situ micro-tensile tests and traditional fractography methods were utilized to study the fracture behavior of proton-irradiated HT-9 steels. First, to evaluate the viability of micro-tensile tests for nuclear material qualification process, meso‑tensile tests on as-received HT-9 steels were performed. Fracture mechanisms of unirradiated HT-9 steels at both length scales were compared and underlying mechanisms discussed. The direct comparison of micro- and meso‑scale data shows a distinctive size effect demonstrated by the increase in yield stress (YS). Upon completion of initial assessment, specimens were irradiated with 4 MeV+ protons to three fluences, all of which were lower than 0.01 displacements per atom (dpa). As expected, the YS increases with irradiation. However, at 7 × 10−3 dpa, the reversal of the trend was observed, and the YS exhibited sharp decline. We demonstrate that at lower length scales, grain structure has a more profound impact on the mechanical properties of irradiated materials, which provides information needed to fill in the gap in current understanding of the HT-9 fracture at different length scales.

Helium bubble nucleation and growth in alloy HT9 through the use of in situ TEM: Sequential he-implantation and heavy-ion irradiation versus dual-beam irradiation

The formation of He bubbles in Ferritic/Martensitic steel HT9 is investigated through the use of in situ Transmission Electron Microscopy coupled with He implantation and heavy ion irradiation. Of particular interest is the effect of increasing He appm/dpa ratio on the formation and growth of the bubbles, as well as the effect of the sequential order of ion irradiation i.e. He-pre-implantation followed by heavy-ion irradiation versus true dual-beam irradiation. The role of He is discussed.

Enhancing the performance of a long-life modified CANDLE fast reactor by using an enriched 208Pb as coolant

The investigation of the utilization of enriched 208Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched 15N), 208Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, 208Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of 208Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of 208Pb is the highest among Na, natPb, and Bi. The small inelastic scattering cross-section of 208Pb can harden the neutron energy spectrum. Therefore, 208Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of 239Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched 208Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched 208Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched 208Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched 208Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched 208Pb can enhance the performance of a long-life Modified CANDLE fast reactor.

Effect of pulse current and coating thickness on the microstructure and FCCI resistance of electroplated chromium on HT9 steel cladding

In this study, we examined the interrelations among electroplating parameters, such as pulse current and electroplating duration; structural properties (such as crystalline structure, grain size, micro-cracking, and coating thickness); and physical properties (such as hardness and Fuel and Cladding Chemical Interaction (FCCI) barrier properties) of chromium barriers. Chromium coatings were electroplated onto HT9 steel disks using direct current and pulse current (PC) for different durations to investigate the influence of PC and coating thickness on the resulting microstructure and FCCI resistance. PC and a coating thicker than 20 μm were found to be effective for improving the barrier properties of Cr coatings. Scanning electron microscopy and electron backscatter diffraction analyses revealed sound microstructures without micro-cracking and twofold grain structures including large and small grains near the outer surface and HT9 substrate of the pulse-plated Cr coatings thicker than 20 μm, respectively. These resulted in reduced interdiffusion area and interdiffusion depth in the microstructure of the diffusion-coupled specimens. Heat-treatment of Cr coatings at 650 °C for times ranging from 50 to 1000 h revealed that the grain size increases with increasing heat-treatment time. The measured Vickers hardness of Cr coatings decreased with increasing grain size and resulted in only 42.6% hardness of as-electroplated Cr coatings after 1000 h of heat-treatment.

Postoperation Dose Rate Estimates for the Very-Small, Long-Life, Modular Reactor

Abstract The very-small, long-life, modular (VSLLIM) nuclear reactor generates 1.0–10 MWth, for ∼92 and 5.8 full power years (FPY), respectively, without refueling. This factory fabricated, assembled, and sealed reactor is cooled by natural circulation of in-vessel liquid sodium (Na). It offers redundant control and passive operation and decay heat removal features. The VSLLIM reactor, together with other plant components for generating electricity using open Brayton cycle, is deployable on a portable platform. Alternatively, multiple units could be installed at a site and use high thermal efficiency Rankine or supercritical CO2 cycles for electricity generation. Extensive analyses are performed to estimate postoperation external biological dose rate and the storage time required on-site for this dose rate to decrease to or below the U.S. federal transportation limit (0.2 rem/h or 2 mSv/h) for safe handling, removal, and replacement with a new unit loaded with fresh fuel. Results show that gamma photons emission from the uranium nitride (UN) fuel and the activated HT-9 steel cladding for the UN fuel rods, the core structure, and the reactor primary vessel, is a major contributor to the radiological source term, and the external biological dose rate estimates. After operating at 10 MWth for 5.8 FPY, using 6 cm thick lead shielding of the reactor guard vessel, decreases the postoperation external biological dose rate to be in compliance with the federal limit, only after∼ 103 days of postoperation on-site storage.

Effects of Silicon Content and Tempering Temperature on the Microstructural Evolution and Mechanical Properties of HT-9 Steels.

Two kinds of experimental ferritic/martensitic steels (HT-9) with different Si contents were designed for the fourth-generation advanced nuclear reactor cladding material. The effects of Si content and tempering temperature on microstructural evolution and mechanical properties of these HT-9 steel were studied. The microstructure of experimental steels after quenching and tempering were characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM), and transmission electron microscopy (TEM); the mechanical properties were investigated by means of tensile test, Charpy impact test, and hardness test. The microscopic mechanism of how the microstructural evolution influences mechanical properties was also discussed. Both XRD and TEM results showed that no residual austenite was detected after heat treatment. The results of mechanical tests showed that the yield strength, tensile strength, and plasticity of the experimental steels with 0.42% (% in mass) Si are higher than that with 0.19% Si, whereas hardness and toughness did not change much; when tempered at 760 °C, the strength and hardness of the experimental steels decreased slightly compared with those tempered at 710 °C, whereas plasticity and toughness increased. Further analysis showed that after quenching at 1050 °C for 1 h and tempering at 760 °C for 1.5 h, the comprehensive mechanical properties of the 0.42% Si experimental steel are the best compared with other experimental steels.