Reduced Activation Ferritic Martensitic Steel Research Articles

Creep deformation behaviour of thermomechanically treated India specific reduced activation ferritic martensitic steel

In the present work, the influence of thermomechanical treatment (TMT) on microstructural evolution and creep deformation behaviour of India specific reduced activation ferritic martensitic (INRAFM) steel was investigated. Firstly, the TMT was carried out on INRAFM steel by reducing its thickness to 25% at warm rolling temperature (973 K), followed by tempering for 90 min at 1033 K. The resultant microstructural characteristics were compared with those attained in normalized and tempered (N+T) condition. TMT treated steel shows a refinement in lath structure with an increment in population of finer M23C6 precipitates along the hierarchal PAG/packet/block boundaries and MX carbonitrides in the intralath region. After TMT, creep tests were performed on the steel over the stress range of 200–300 MPa at temperature of 823 K. The rupture time of the steel significantly increased upon TMT, due to refinement in microstructural features that enhanced the creep resistance. Dependence of minimum creep rate on applied stress obeys Norton's power law and the stress exponent values indicated the deformation mechanism as dislocation creep. Modified Monkman-Grant relation shows a good relation between rupture time/strain to failure with minimum creep rate. Thereafter, the effect of TMT on the transient, secondary and tertiary creep behaviour was also examined and presented in this work. The TMT treated steel shows lower rate of exhaustion, higher threshold stress and prolonged tertiary region.

The neglected activation of tantalum in reduced activation materials

Due to the expression of “reduced activation”, the severe radioactivity of tantalum (Ta) after neutron irradiation in early decay years is easily neglected. In the presented study, after an irradiation of only 0.3 dpa by fission neutron and followed by a cooling stage of 305 days, the RAFM steel specimen with a weight of 0.8 g possessed drastic radioactivity of 4 mSv/hour at a measuring distance of 20 cm. It was ascertained that the high radioactivity was generated by 182Ta. The severe radioactivity seriously increased the danger of material experiments, even though the original Ta addition of the steel was only 0.13 wt%. Similar deteriorative radioactivity of Ta can be generated in both fission and fusion neutron irradiations. We appeal to seriously deliberate the radioactivity of Ta and control the Ta content in the RAFM steel.

A novel design of transitional layer structure between reduced activation ferritic martensitic steels and tungsten for plasma facing materials

Plasma-facing components (PFCs) are among the most critical gaps for fusion energy to establish technical and economic feasibility. Tungsten as a first wall/blanket material in PFCs requires to be integrating with reduced activation ferritic martensitic (RAFM) steels as a structural component. Currently, major drawbacks are the requirement of brazing, the formation of a brittle interface, and a large difference between the coefficients of thermal expansion of tungsten and steel. Here, a novel transitional multilayer structure was designed and investigated to join tungsten and RAFM steels using three interlayers. The composition of each interlayer was selected based on computational thermodynamics and diffusion kinetics to ensure a body-centered cubic (bcc) single-phase structure and prevent the formation of a brittle intermetallic phase region in the temperature range of 600–1150 °C. Although the transitional layer structure was designed for additive manufacturing, spark plasma sintering (SPS) as proof of concept was used to bond the individual layers. Interfaces were investigated using scanning and transmission electron microscopy methods but no layered intermetallic phase was observed. Nanoindentation maps across the interface suggest major hardness differences at the interface between tungsten and the vanadium interlayer, as well as the interface between RAFM steel and the FeCrAl interlayer.

Correction: Low-Cycle Fatigue Behavior of Reduced Activation Ferritic-Martensitic Steel at Elevated Temperatures

Correction: Low-Cycle Fatigue Behavior of Reduced Activation Ferritic-Martensitic Steel at Elevated Temperatures

An update on the latest and most relevant improvements achieved in the European helium-cooled ceramic breeder TBM design

The European TBM (Test Blanket Module) program under development will test two alternative concepts for tritium generation and extraction in ITER: a helium cooled / ceramic pebble breeder concept and a water cooled / lithium lead breeder concept.The former implements Advanced Ceramic Breeder pebbles that act as tritium breeding material, and Beryllium pebbles as neutron multiplier. A Reduced Activation Ferritic Martensitic steel, the EUROFER-97 (X10CrWVTa9-1), is used as structural material and pressurized Helium technology is implemented for efficient heat extraction (300-500°C, 8 MPa).The design of the TBM box involves challenges that must be tackled with a holistic approach to simultaneously consider a variety of aspects. Some of these challenges include: the thermal performance and the structural integrity of the TBM under ITER normal operations and accidental conditions, the implementation of pressurized equipment regulation(s), the convergence with manufacturing activities being developed in parallel and the high congestion in the reduced TBM volume –especially when considering instrumentation integration– without affecting too much the tritium breeding capability. It is also key to develop advanced methodologies based on design by analysis to virtually test alternative design solutions in an agile manner. These should be linked to the applicable nuclear design code (RCC-MRx) to consolidate design choices. In this way, a traceable evolution of the TBM design process can be built where alternatives are tested, compared and either discarded or selected in a consistent and comprehensive framework.This paper presents an overview of the very significant progress achieved in the design of the helium cooled concept TBM over the last two years, a period that corresponds to the first steps of the preliminary design phase towards the Preliminary Design Review (PDR) gate. This includes a significant change in the design of the back manifold area and the connection with the rest of the TBM-Set that solves long-standing historical problems inherited from pre-conceptual design stage. An update on the latest and most relevant methodological developments achieved is also provided in this paper, since these have been key to progress with the design evolutions in a sound and traceable manner. To end, the paper also presents some examples of the instrumentation definition and its complex integration within and outside the TBM box.

Preliminary study on Hot Isostatic Pressing diffusion bonding of Fe–Cr–Al and CLF–1 steel for preparation of tritium permeation barrier

In the internal cooling channel forming of the blanket component for fusion reactor, the combination of Fe-Cr-Al ferritic steel with RAFM steel (reduced activation ferritic/martensitic steel) can form alumina film and then be served as tritium permeation barrier (TPB), which can inhibit the penetration of tritium into structural materials. To realize reliable joining between the two materials, Hot Isostatic Pressing (HIP) diffusion bonding, which is considered as the most promising fabrication technique for blanket component was used. The microstructure and mechanical properties of the joints were analyzed, and the results showed that Fe-Cr-Al/CLF-1 steel interfaces generated some large-size AlN, and the interfacial microvoids are not totally eliminated. While adding Ni interlayer between Fe-Cr-Al/CLF-1 steel could significantly reduce the size of AlN and promote interfacial microvoids closure. Fine NiAl and Ni3Al phases, as the strengthening phases in Nickel-based superalloys, were formed due to atom interdiffusion across the interfaces. Fe-Cr-Al/Ni/CLF-1 joints exhibited average tensile strength of 636 and 529 MPa, and elongation rate of 3% and 15% when tested at 25 and 300 °C, respectively. Microhardness test demonstrated Fe-Cr-Al/Ni/CLF-1 steel joints could be enhanced by precipitation-hardening of NixAly phases. These preliminary studies have laid a foundation for the preparation of TPB in the cooling channel of blanket components.

Effect of applied stress on the ductility of RAFM steel during long-term exposure in flowing Pb-17Li

The effect of applied stress on the ductility of reduced activation ferritic martensitic (RAFM) steel during long-term exposure in flowing Pb-17Li in the temperature interval 300–400 ℃ was investigated using tensile tests. The ductility of RAFM steels at a stress of 250 MPa decreases appreciably after exposure in following Pb-17Li at 400 ℃ for 1500 h. And a slight ductility loss occurred for the samples tested at 300 ℃, 350 ℃ and 400 ℃ for 500 h. Coarsening of martensitic laths occurred in RAFM steels owing to the thermal aging with a 250 MPa stress for 1500 h in liquid Pb-17Li. Due to the poor ductility of coarse laths, micro-cracks originated on the surface of necking zone during tensile test, which facilitates the penetration of liquid metal. With sufficient liquid Pb-17Li penetrated into cracks continuously, the surface tension of the newly formed fracture surface is declined by liquid metal in the crack tip. The fracture mode was altered from uniform to unstable crack propagation thus leading to a reduction of ductility. There was no depletion of Fe and Cr or Pb infiltration into the substrate for RAFM steels after exposure. The fracture mechanisms were proposed in terms of supply of liquid metal and micro-cracks caused by coarse laths. Accordingly, the reduction of ductility for RAFM steels after exposure in liquid Pb-17Li was the result of combined effect of applied stress and liquid metal.

Effect of austenitizing temperature on microstructure and mechanical properties evaluation of microalloyed low-carbon RAFM steel

The microstructure and mechanical properties of microalloyed low-carbon reduced activation ferritic/martensitic (RAFM) steel after austenitization at 950 °C, 1000 °C and 1050 °C for 0.5 h and tempering at 750 °C for 1.5 h were investigated using scanning electron microscopy, transmission electron microscopy, electron back scattering diffraction, X-ray diffraction, tensile tests, and impact tests. The grain size and lath width obviously increase because more M23C6 carbides dissolve after austenitization at 1050 °C. Besides, more fine MX particles would precipitate and result in higher dislocation density. Furthermore, the strengthening mechanisms at room temperature were also systematically discussed, the predominant strengthening mechanism changes from grain boundary strengthening and dislocation strengthening to dislocation strengthening when the austenitizing temperature increases from 950 °C to 1000 °C and 1050 °C. After taking the microstructure, strength, ductility, and toughness into thorough consideration, the optimum austenitizing temperature for microalloyed low-carbon RAFM steel is about 1000 °C.

A comprehensive atomistic investigation on the cascade induced helium bubble motion in bcc iron for neutron irradiated RAFM steels

Cascade induced by the primary knock-on atom could significantly influence the evolution of helium bubbles in RAFM steels. In this work, the helium bubble migration and helium bubbles coalescence under cascade overlapping, i.e., when the helium bubble was overlapped by the thermal spike of one cascade, in bcc Fe was investigated by molecular dynamics simulation. The influence of He-to-vacancy (He/V) ratio as well as cascade energy, temperature, bubble size and bubbles distance were investigated based on the statistical results of 4700 cases. A method which could quantitatively estimate the overlapping extent between the cascade and helium bubbles was put forward to confirm that the cascade did overlap with the helium bubble in all cases. It was found that under cascade overlapping, the helium bubble tends to migrate toward the thermal-vacancy-rich center. The mechanisms of migration change from vacancy diffusion to He atoms migration with the increase of the He/V ratio, and the corresponding migration distance could decrease one order of magnitude. The cascade induced coalescence of two close helium bubbles is associated with the synergy effect of migration, dissociation and joint strain field of the two bubbles. Especially, it was found that the joint strain field of two close helium bubbles could significantly promote the bubbles’ coalescence by fostering the two bubbles’ directional migration towards each other under cascade overlapping. It was also found that the stable He/V ratio of helium bubble under cascade overlapping is around 0.5.

Electromagnetic load evaluation of K-DEMO divertor for MD and VDE plasma disruption scenarios

Electromagnetic (EM) force caused by the instability of plasma operation is one of the most significant external loads in the DEMO, as well as the extremely high heat flux (more than 10 MW/m2). Abrupt disruption of plasma current of several MA causes thermal quench and current quench sequentially. This paper presents the EM force on the divertor calculated by considering the Poloidal field variation (PFV), halo current (HC), and toroidal field variation (TFV) generated during the quenching process for the K-DEMO. The K-DEMO with the fusion power of 2,200 MW has a major radius of 6.8 m and a plasma current of 12 MA in steady-state operation. The K-DEMO divertor basically employs a water-cooled tungsten monoblock concept with a reduced activation ferritic martensitic RAFM steel heat sink. EM force was calculated using ANSYS Mechanical/Emag solver. For the convenience of calculation, the existing inconvenient GUI environment based on MAPDL was changed to ANSYS Workbench and improved. To evaluate the EM forces generated in the K-DEMO divertor when various disruption scenarios occur, three scenarios were selected for major disruption (MD) and vertical displacement event (VDE), respectively. The variation of PFV, HC, and TFV in the disruption are significant input data for EM analysis. The major disruption data were converted from ITER disruption scenarios by using the in-house code since the plasma data for the disruption scenarios has not been developed for the K-DEMO. This study presents the preliminary evaluation of EM force for the K-DEMO and improved calculation platform although the disruption data derived from ITER are used. The EM forces generated by the selected six scenarios were calculated and compared. In the MD_DW_slow_fast and VDE_DW_slow_fast scenarios, the maximum EM force was 4.8 MN on the K-DEMO divertor module along the vertical direction, and the HC influence was dominant.

Corrosiveness of Li2TiO3 ceramic tritium breeder to structural RAFM steel: Effects on microstructure and thermal conductivity

Structural steel will be in long-term contact with ceramic tritium breeder at a high temperature in a solid tritium breeding blanket. The maximum temperature in the tritium breeding zone would reach nearly 1173 K. At this high temperature, a large amount of lithium oxide would evaporate and deposit on the structural materials, causing the overall corrosion of the structural materials. Corrosiveness of Li2TiO3 ceramic tritium breeder to reduced activation ferritic-martensitic steel CLF-1 was investigated in this study. The experiment was carried out at 1123 K under He + 0.1 vol% H2 atmosphere for up to 400 h. The formation of non-uniform corrosion layers with a Fe-rich outer layer and a Cr-rich inner layer were found on the surface of CLF-1 steel. Secondary ion mass spectrometry (SIMS) test at contact and non-contact areas were performed to demonstrate evaporation of lithium compound. The corrosion layers were characterized by scanning electron microscopy (SEM) and energy dispersive X-ray (EDX) spectroscopy. The non-contact area with tritium breeding pebbles were corroded indicating a large amount of lithium oxide on the non-contact zone, resulting in serious overall corrosion of structural materials. LiFe5O8 was found as the main phase on the surface of the corroded samples by X-ray diffraction (XRD). The thermal conductivity of LiFe5O8 (∼0.8 W/m/K, 300 K) calculated by first principles phonon Boltzmann equation (PBE) is significantly lower than the original CLF-1 (∼23 W/m/K, 300 K). Almost no change of thermal conductivity for experimental results was found after corrosion, which is due to non-uniform corrosion layer presenting thinner layer at some areas, as well as the hole, crack and peeling of corrosion layer indicated by SEM results. However, in the long-term service of blanket structural material, it is predicted that a much thicker corrosion layer may form on the surface of RAFM steels thus bring a risk to the heat transfer of blanket.

Effect of tantalum addition on the mechanical properties of tungsten/tantalum composite thin films

Tungsten (W) when used as a plasma-facing material (PFM) in a tokamak will face harsh neutron radiation and high-temperature fluctuations. Due to its ductile-to-brittle transition temperature (DBTT), W behaves as a brittle material in lower temperatures and thus helps in propagating cracks easily resulting in dust formation and failure of material. In this study, an attempt has been made to improve the ductility, hardness, and toughness of W thin films by adding tantalum (Ta) to it, resulting in composite thin films. W/Ta composite thin films deposited on reduced-activation ferritic martensitic steel (RAFM) steel by RF magnetron sputtering method are compared with W thin films deposited on the same substrate. Taguchi orthogonal array is used to get the best properties for the composite thin films and nanoindentation technique is used to evaluate the mechanical properties of developed thin films. With the least hardness of 7.19 Gpa, composite film hardness is 452% lesser than the maximum hardness of (39.54 Gpa) W films. The reduction in hardness for composite thin films has not depleted the toughness as the Ef/Es ratio is higher than cracking threshold of 1.3. The maximum plastic energy absorbed for yield by composite films during nanoindentation is (1710 mN nm) 14% higher than W thin film (1500 mN nm), indicating the composite film is a ductile and tougher than the W film.

Microstructure and Mechanical Properties Evaluation of Microalloyed Low‐Carbon Reduced Activation Ferritic/Martensitic Steel

A novel method for reducing the amount of M23C6 carbides, increasing the amount of MX carbonitrides, and decreasing the coarsening rate of M23C6 carbides is put forward to improve the high‐temperature properties of reduced activation ferritic/martensitic (RAFM) steel. Scanning electron microscopy, transmission electron microscopy, X‐ray diffraction, tensile tests, and impact tests are used to systematically investigate the microstructure evolution and mechanical properties of RAFM (RAFM‐0.1C) steel, low‐carbon RAFM (RAFM‐0.04C) steel, and minor boron and nitrogen microalloyed low‐carbon RAFM (RAFM‐CBN) steel. Some δ ferrite develops when C decreases from 0.1 to 0.04 wt% and subsequently disappears after 0.01 wt% N addition. The amount and size of M23C6 carbides and dislocation density decrease with the decrease of C (RAFM‐0.04C), while the amount of MX carbonitrides is 2‐3 times, and dislocation density is ≈2 times higher than that of RAFM‐0.04C steel after 0.01 wt% N addition (RAFM‐CBN). Compared with RAFM‐0.1C steel, the yield strength at room temperature and 550 °C slightly decreases for RAFM‐0.04C steel and considerably increases for RAFM‐CBN steel. The contributions of microalloy on strengthening mechanisms of RAFM steel at room temperature are also systematically revealed by combining the experimental and theoretical data.

Effect of irradiation temperature on radiation hardening of CLF-1 steel

A kind of reduced-activation ferritic-martensitic steel, CLF-1, was developed by the Southwestern Institute of Physics, China, as a candidate structural material for fusion and fourth-generation fission reactors. This work reports the results from the nanoindentation and transmission electron microscopy study performed on this steel after irradiation with 2.75 MeV Fe11+ ions up to 1 and 10 dpa at temperatures of 300 - 500 °C. An increase in hardness was observed in the damaged layer of samples irradiated at 400 °C, while the samples irradiated at 500 °C exhibited a softening of the damaged layer. The hardening can be primarily, although not exclusively, attributed to the increase of <100>-type dislocation loops observed at the increasing irradiation temperature. Nano-sized cavities around the dislocation loops and along dislocation lines were also observed, explaining the lattice swelling after irradiation at 500°C. The research results provide a valuable reference for the development of irradiation-hardening-resistant Fe-Cr alloys for advanced nuclear energy systems.

Initial Neutronics Investigation of a Chlorine Salt–Based Breeder Blanket

Tritium breeding blankets within D-T-fueled fusion reactors contain lithium compounds and typically require neutron multiplier materials to achieve a tritium breeding ratio (TBR) consistent with self-sustaining operation. Liquid breeder blankets have some advantages over solid blankets, and previous blanket studies have investigated liquid metal as well as liquid salt–based blankets. Liquid salts have reduced magnetohydrodynamic effects as compared to liquid metals, but typically have a lower TBR. Recently, advanced fission reactor concepts have considered chloride-based salts in their design, and there is a significant amount of research work occurring to study these salts. Chloride salts have previously been considered for fusion reactors, but studies have typically found lower breeding ratios than for fluoride salts, such as 2(LiF)-BeF2 (flibe) so they have not been further developed. In this work, we use a one-dimensional cylindrical radiation transport model of a conceptual tokamak fusion reactor to investigate the neutronics feasibility of using a chloride salt–based blanket that uses chlorine enriched in 37Cl, which has both a low capture cross section and a substantial (n,2n) cross section. It is found that chloride salts (LiCl mixed with BeCl2 and/or PbCl2) can potentially achieve a ~3% to 5% higher TBR than fluoride molten salts, notably flibe, in the absence of a solid multiplier. Including a solid multiplier, however, does narrow this advantage, with TBRs estimated within ~1% of flibe with a 2-cm Be multiplier. Chloride salts can also reach lower melting points than flibe, potentially improving the scope for the use of reduced activation ferritic-martensitic steel as a structural material. There is substantial uncertainty in the calculations driven by limited thermochemical data for the Cl salts, plus cross-section uncertainties. The production of 36Cl through 35Cl(n,g) and 37Cl(n,2n) has the potential to challenge the waste disposal rating of the blanket. Calculations indicate that, while this is not an immediate showstopper, this case depends upon the exact waste disposal rating criteria used for 36Cl. Further work could reduce these uncertainties with improved thermochemical data, higher-fidelity modeling for downselected salts, and more refined waste disposal calculations and regulatory guidance. Finally, it must be recognized that, as for all molten salts, corrosion and chemistry can present appreciable technical challenges that require further assessment in developing a practical blanket concept, and also that the enrichment of chlorine presents an additional technical and supply chain challenge.

Enhanced Mechanical Properties of CLAM by Zirconium Alloying and Thermo-Mechanical Processing

In this study, we present the effects of 0.004~0.098 wt% Zr and thermo-mechanical processing (TMP) on the microstructure and mechanical properties of the China RAFM steel, CLAM, as a feasibility study for improving mechanical properties. The inclusions in ingots were characterized using optical microscope (OM) and scanning electron microscope (SEM), which could be classified as fine simple particles and large complex particles. The complexity of the alloy’s inclusion composition increases with the increasing Zr concentration. The higher the Zr content, the more complex the composition of inclusions in the alloy. The average diameter of inclusions in 0.004Zr steel was the smallest, which was 0.79 μm and the volume fraction was 0.018%. The highest yield strength, tensile strength, elongation, and impact energy of 0.004Zr alloy at room temperature were 548.3 MPa, 679.4 MPa, 25.7%, and 253.9 J. The structure of the TMPed steels was all tempered martensite. With the increase in tempering temperature, the yield and tensile strength of the experimental steel gradually decreased, while the elongation and impact energy gradually increased. The 0.004ZrD and 0.004ZrH alloys had the best yield strength and impact energy, which were 597.9 and 611.8 MPa and 225.9 and 243.3 J, respectively. In addition, the alloys showed good thermal stability during the aging at 600 °C for 1500 h. It was discovered that TMP is a simple and practical industrial technique that could successfully enhance the mechanical properties of CLAM steel without sacrificing impact toughness.

Effect of pre-irradiation defects on helium trapping and diffusion in RAFM steel

In order to analyze the effect of pre-irradiation defects on helium trapping and diffusion behaviors in detail by slow positron beam, Si ion with high energies (300 keV and 6 MeV) and 100 eV helium plasma were sequentially introduced into reduced activation ferritic/martensitic steel. Massive microscopic defects were observed in Si ion irradiated specimens. Low energy helium (∼100 eV) atoms could diffuse into much deeper region in metals without displacement damage. In the diffusion process of helium atoms, they would be trapped by the Si-ion pre-irradiation defects. The formation of He-vacancy complexes would affect the annihilation of positrons with the electrons in pre-irradiation defects and decrease the S parameters. The shallow region defects induced by 300 keV irradiation have more obvious advantage of hindering the diffusion of helium over the deep defects than that in 6 MeV irradiation. Thus, the pre-introduced defects can trap effectively helium atoms and hinder their diffusion.

Towards better understanding the question of helium-driven swelling in RAFM-ODS alloys

We irradiated three 9Cr alloys: Fe9Cr model alloy, reduced-activation ferritic/martensitic (RAFM), and advanced oxide dispersion strengthened ferritic (RAFM-ODS) alloys, with 100keV He ions at 723K to the peaked concentration of 6500 appm (low helium level) and 65000 appm (high helium level), respectively. The question of helium-driven swelling in Fe9Cr, RAFM, and RAFM-ODS alloys was investigated by multiple characterizations. Homogeneous helium bubbles, main in the form of VmHen (n < m), nucleated on different sinks, producing swelling behavior and dominating the microstructures of irradiated Fe9Cr, RAFM, and RAFM-ODS steels. The consequent swelling rate values in Fe9Cr, RAFM, and RAFM-ODS were ∼ 0.15% (1.33%), ∼ 0.058% (0.83%), and ∼ 0.048% (0.73%) in the low (high) levels, respectively. The RAFM-ODS steel exhibits better tolerance to helium bubble swelling and superior grain size stability than Fe9Cr and RAFM, which is manifested by the smaller average sizes and the lower number densities of bubbles. It can be attributed that the oxides in RAFM-ODS steel could provide more effective helium trapping sites that sequester the helium into smaller bubbles and away from the precipitate-matrix interfaces and grain boundaries. In addition, the existence of high density of oxide particles could also delay desorption peaks of helium atoms from RAFM-ODS steel compared to RAFM and Fe9Cr.

Hydrogen isotope plasma-driven permeation through RAFM steel: isotope exchange and helium irradiation effect

Hydrogen isotope exchange in China low activation martensitic (CLAM) steel was investigated by plasma-driven permeation (PDP) measurements in the temperature range of ∼670–920 K. It was found that as H was introduced into D plasma, the permeation flux of HD increased while that of D2 decreased. The total D permeation flux, however, remained almost unchanged. Such a result allowed the analysis of isotope replacement and showed that the reduction of D2 is balanced by the rise of HD. In addition, the formation of HD was found to be closely related to the H and D content in plasma. Hydrogen isotopic effects on H and D plasma dissociation/ionization rate and on incident flux were evaluated by measuring plasma parameters including electron temperature and electron density using a triple Langmuir probe. The correlation between isotope effects on ionization rate in the plasma and isotope effects in the permeation flux was discussed. He plasma pre-irradiation and seeding during D PDP were also conducted to investigate the He effects on D diffusion. The total D permeation flux was found to be reduced by He seeding. Data analysis of Langmuir probe showed that ionization rate of D plasma was influenced by He seeding. Besides, under simultaneous D + He irradiation, He atoms were preferentially trapped at the defects resulting in a shallower D concentration gradient that described the observed decrease in D permeation rate. He plasma pre-irradiation was found to reduce D permeation flux with a much slower breakthrough to reach steady-state. The amount of D trapped at He-induced defects was derived quantitatively from the transient permeation curves. It is believed that He bubbles formed at the near surface act as trapping sites of D atoms, leading to a less effective diffusion coefficient and an enhanced retention in the material.

Large Component Simulation of an Inboard Sector of a Dual-Coolant Lead-Lithium Blanket

Advances in high-performance computing now enable the engineering evaluation of large blanket components from a systems perspective at the beginning of a normal design cycle. As an example, we discuss the computational fluid dynamics (CFD) involved in characterizing the thermal performance of an entire 22.5 toroidal sector of a dual-coolant lead-lithium blanket with particular attention to the integrated manifolding and flow distributions. The inboard sector is roughly 7 m tall, has 321 first-wall helium-cooled channels, and five parallel PbLi breeder channels complete with insulating SiC flow channel inserts. A finite volume model was developed with various degrees of mesh refinement from 36 to 187 million cells to perform the flow calculations on disparate fluids with conjugate heat transfer in the solid. Steady-state CFD calculations were performed using a realizable k-ε turbulence model. Simplifications include a uniform applied surface heat flux and a one-dimensional radial volumetric neutron heating profile. Constant material properties were used for the F82H reduced-activation ferritic martensitic (RAFM) steel walls, SiC flow channel inserts, and the PbLi breeder. Ideal gas behavior was assumed for the helium, which includes compressibility. Helium mass flows of 54 kg/s at 8 MPa and PbLi flows of 3 kg/s at 101 kPa supply the sector, both at an inlet temperature of 350°C. This early model is not optimized; however, it reveals important features not obvious in a conglomeration of smaller independent models. For example, all the manifolding was included to evaluate flow distributions throughout the full component. Submodels were only used to obtain the convective heat transfer coefficients (HTCs) inside the helium first-wall channels equipped with enhancement vanes that allow the first wall to handle 0.8 MW/m2 of plasma heat flux with the aim of keeping bulk RAFM steel temperatures near the creep-fatigue limit of 550°C. Average HTCs obtained from these detailed submodels were then used in the large model to predict thermal performance without incurring the meshing overhead from the thousands of internal vanes. Although much progress was made, initial results indicate that more must be done to further reduce hot spots on the first wall, minimize pressure drops, and provide optimal flow distributions.