Low Activation Ferritic Steel Research Articles

Study on Deuterium Permeation Behavior of Palladium Films Prepared by Magnetron Sputtering Method

Pre-depositing a Pd film is crucial for accurately acquiring the hydrogen permeability of metal materials, as it permits the production of ultra-pure hydrogen. However, the microstructure of Pd film and its effect on the hydrogen isotope permeation behavior of substrate materials have been neglected. In this study, Pd films were deposited on China Low-activation Ferritic (CLF−1) steel by magnetron sputtering. The effect of sputtering pressure on the microstructure and deuterium permeation behavior of Pd films at temperatures of 550−650 °C is presented. SEM results demonstrated that the films had a columnar crystal structure with a thickness of 0.6 ± 0.2 μm. The gas-driven permeation results revealed that the deuterium ion current intensity of the coated CLF−1 sample was at least three times lower than that of uncoated CLF−1 steel, which was influenced by the combined effect of oxygen and surface cracks. Oxygen could not be excluded from the films at a sputtering pressure of 10−3 Pa order of magnitude. It was also found that the films cracked during deuterium permeation experiments, which affected the deuterium permeation behavior. Films with large surface coverage and small grain sizes exhibited better cracking resistance. Our study provides promising insights into the hydrogen permeability of Pd films.

Experimental investigation on microstructure and mechanical properties of joining stainless steel 316LN to Low Activation Ferritic Martensitic steel (LAFM) using activated flux TIG welding

ABSTRACT One of the most significant requirements for joining dissimilar materials is to produce a minimum number of deeper-penetration welds and thereby avoid flaws. The Activated Flux TIG welding process improves penetration depth in a single pass, reducing the number of weldments. The dissimilar weld between LAFM – Stainless Steel 316LN also possesses the same requirement for a particular application. The present study aims to investigate the structure-property relationship and mechanical properties of dissimilar weldments of LAFM – SS 316LN using conventional TIG and Activated TIG welding using TiO2 and Co3O4 oxide fluxes. The microstructure study of conventional TIG weld reveals the fine lath martensitic structure as compared to weld prepared with A-TIG using TiO2 flux due to difference in the cooling rate. Also, the carbide formation in conventional TIG and A-TIG using TiO2 confirms the presence of delta ferrite But, the same observation was not confirmed with A-TIG weldments prepared using Co3O4. Similarly, the hardness results also support the microstructure study. The higher transverse tensile strength of weld indicates that it is stronger than base metal. The lower impact energy in case of Activated Flux TIG welding process was reported due to presence of oxide flux.

Characterization of Corrosion Behavior of CLF-1 in Liquid Lithium Using Calibration-Free Laser-Induced Breakdown Spectroscopy in Depth Profile Analysis.

It is important to get fast and quantitative compositional depth profiles for the boundary layer of the corroded specimen in order to understand the corrosion process and mechanism due to liquid lithium induced corrosion problems to structural material of fusion reactors. In this work, calibration-free laser-induced breakdown spectroscopy (CF-LIBS) is introduced to investigate the compatibility of CLF-1(China low-activation Ferritic steel) exposed in liquid lithium at 500 °C for 500 h. The results show that CF-LIBS constitutes an effective technique to observe the corrosion layer of specimens which are non-uniform and the elements of matrix show gradient distribution from the boundary to the inner layer. The concentration was 82–95 wt.% Fe, 5–12 wt.% Cr, 0.45–0.85 wt.% Mn, 1.6–1.1 wt.% W, 0.11–0.16 wt.% V, and <0.2 wt.% Li along the longitudinal corrosion depth for the corrode CLF-1. The results reveal the quantitative elemental variation trend of CLF-1 in the lithium corrosion process and indicate that the CF-LIBS approach can be applied to the analysis of composition in multi-element materials.

Trace tritium removal in China reduced activation ferritic-martensitic steels through thermal desorption with/without hydrogen isotope exchange

Effective removal of deuterium and tritium accumulated in fusion reactor materials is crucial for practical and safe fusion energy. In this paper, investigations on tritium removal in China Low Activation Martensitic (CLAM) steel and China Low-activation Ferritic (CLF-1) steel through thermal desorption with/without hydrogen isotope exchange have been performed. The results indicate that the tritium radioactivity in CLF-1 steel is higher than that of CLAM steel. The tritium removal efficiency is both less than 50% at 473 K, while significantly improved after introduction of hydrogen, 82.24% for CLAM steel and 70.34% for CLF-1 steel respectively. With the increase of temperature, the tritium removal efficiency can be further improved, however, hydrogen isotope exchange can slightly enhance tritium release at 673 K. In consequence, tritium retained in both steels can be removed effectively at relatively moderate temperature (<673 K) due to the high tritium diffusivity, especially for CLAM steel.

Surface studies on the characterization of RAFM steel after deuterium plasma exposure

Previous works have identified the potential for preferential sputtering of low/middle-Z elements in the reduced-activation ferritic-martensitic (RAFM) steels, leaving behind a W-enriched surface, which could then potentially reduce further sputtering of Fe and other elements in the steel. Thus, it is of vital importance to observe the morphology and concentration of the W-enriched surface. In this work, we detail the near-surface analysis of China low-activation ferritic (CLF-1) steel which has been exposed to deuterium plasma (with an impinging ion energy of 150 eV per D, incident fluence of 7.2 × 1024 D/m2, and specimen temperature of 470 K). Transmission electron microscopy and Rutherford backscattering spectroscopy tests were applied with particular focus on assessing the structure of the near-surface region. Results revealed that a shallow W-enriched surface consisting of W, Fe, Cr and O with a thickness of approximately 15 nm formed in the near-surface region of CLF-1 steel after deuterium-plasma exposure. In this W-enriched surface, the areal density and concentration of W were approximately 2.0 × 1015 atoms/cm2 and 2.0 at.%, respectively, evaluated by simulation with SIMNRA. Moreover, most of the W crystalline grains were scattered on the outside edge of the accumulated layer.

Diffusion of hydrogen in China Reduced Activation Ferritic-martensitic steels at low temperatures

Reduced Activation Ferritic/Martensitic (RAFM) steels have been considered as the most promising candidate structural materials for a fusion reactor. In this paper, the hydrogen diffusion behavior of two types of RAFM steels, called China Low Activation Martensitic (CLAM) steel and China Low-activation Ferritic (CLF-1) steel, have been studied by electrochemical technique over the temperature range of 293–323 K. Before hydrogen permeation, both sides of the CLAM and CLF-1 steels were coated with a thin layer of uniform and dense palladium (Pd) membrane by electro-plating method. The results showed that the diffusion coefficients of hydrogen through CLAM and CLF-1 steels were submitted to the relationships of D(cm2/s) = 3.04 × 10−6exp[-4.72(kJ⋅mol-1)/RT] and D(cm2/s) = 1.82 × 10-5exp[-10.07(kJ⋅mol-1)/RT], which were far higher than that of some other austenitic stainless steels, such as 316 S.S. This indicates that if the RAFM steels as the structural material for fusion reactor are free from hydrogen environment, the hydrogen isotopes will escape from them quickly under low temperatures, which is conducive to keeping the mechanical properties of RAFMs.

Deuterium retention removal in China reduced activation ferritic-martensitic steels through thermal desorption and hydrogen isotope exchange

Effective removal of deuterium and tritium trapping in fusion reactor materials is of great importance both to recycle fuel and prevent the performance degradation. Experiments on deuterium removal in China Low Activation Martensitic (CLAM) steel and China Low-activation Ferritic (CLF-1) steel through thermal desorption and hydrogen isotope exchange have been performed. The results indicate that CLAM steel has only one desorption peak at 578K with an activation energy of 21kJ/mol, while CLF-1 steel has two desorption peaks, whose corresponding activation energy is 31kJ/mol at 563K and 94kJ/mol at 797K respectively. After 2h thermal desorption, deuterium removal efficiency for CLAM is more than 90% above 373K. Yet for CLF-1, deuterium removal efficiency is less than 90% below 573K. Hydrogen isotope exchange can slightly enhance deuterium release. Overall, Deuterium retention in both materials can be removed effectively at relatively low temperature (<673K) through thermal desorption due to the high hydrogen diffusivity, especially for CLAM steels.

Implementation of ferritic steel as in vessel wall: Lessons learnt and follow up

ASDEX Upgrade (AUG) is the only tokamak in Europe to have low activation ferritic steel as the inner vessel wall facing component. Together with the massive tungsten tiles in the lower divertor, AUG is the tokamak with the closest DEMO wall. The project is a first step towards the extensive use of ferritic steel in future fusion reactors. For example, the test blanket module of ITER is planned to have a ferritic steel wall and thick tungsten tiles as a plasma facing component.The ‘ad hoc’ ferritic steel built with low activation capability is known as Eurofer. As the low activation property is not a requirement for AUG, the material selected for the project is the martensitic steel P92 which is the most similar material to Eurofer from a magnetic point of view. The purpose of the project is to improve understanding of the magnetic perturbation of the ferritic steel both on the plasma and magnetic probes, evaluating and controlling these effects. Additionally, the effect of the additional forces on the supporting structure has been addressed.Bearing this in mind, in 2013 a step wise program has been started and part of the W coated graphite tiles in the region of the inner column were replaced by steel tiles [1]. The first campaign did not suffer any particular issue related to the new material. According to the calculations, the plasma was almost unperturbed, thanks also to the toroidal symmetry of the tiles inside the vessel, and the magnetic probe measurements were properly corrected [2].Inspection of the machine pointed out some hardware problems. The graphite tiles adjacent to the steel tiles were damaged. The graphite tiles had broken edges in 5 from 64 positions and notches in many others. The coating of the graphite and steel tiles, made of tungsten and TiO respectively, was damaged. At first glance it was clear that the steel tiles were moving but it was definitely unexpected. In understanding the process, the location of the damage was the crucial hint. In fact all failures were located at the boundary between 2 vacuum vessel octants. To justify this failure mode inside the vessel, a hypothesis (about current flowing in the heat shield supporting structure) was made and FEM analyses were carried out in this direction. With extreme caution, in 2015 just one additional row of steel tiles was added together with diagnostics that confirmed the hypothesis. Now that a clear understanding of the problem has been reached, the project to add further rows of steel tiles can be continued. For the next campaign it is planned to replace all the tiles in the middle region of the heat shield together with stiffening and modification of the supporting structure.In this paper the learning process from the damage of the tiles and its causes, from the FEM analysis results to the data diagnostics will be reported. The future plans for steel tiles in AUG will be discussed.

Diffusive transport parameters of deuterium through China reduced activation ferritic-martensitic steels

Reduced Activation Ferritic/Martensitic (RAFM) steels have been considered as the most promising candidate structure materials for a fusion reactor. In the recent decades, two new types of RAFM steels, called China Low Activation Martensitic (CLAM) steel and China Low-activation Ferritic (CLF-1) steel, have been developed. The gas evolution permeation technique has been used to investigate diffusive transport parameters of deuterium through CLAM and CLF-1 over the temperature range 623 ∼ 873 K at deuterium pressure of 105 Pa. The resultant transport parameters are: Φ (mol. m−1 s−1 Pa−1/2) = 5.40 × 10−8 exp (−46.8 (kJ. mol−1)/RT), D(m2 s−1) = 3.81 × 10−7 exp(−24.0(kJ. mol−1)/RT) and S (mol. m−3 Pa−1/2) = 1.42 × 10−1 exp(−22.8(kJ. mol−1)/RT) for CLAM; while Φ(mol m−1 s−1 Pa−1/2) = 1.76 × 10−8 exp(−43.9(kJ. mol−1)/RT), D(m2. s−1) = 1.02 × 10−7 exp(−16.9(kJ. mol−1)/RT) and S(mol. m−1 Pa−1/2) = 1.73 × 10−1 exp(−27.0(kJ. mol−1) /RT) for CLF-1. The results show that CLAM is more permeable than CLF-1, thus it is easier for hydrogen isotopes to transport and be removed.

Control of substrate oxidation in MOD ceramic coating on low-activation ferritic steel with reduced-pressure atmosphere

An Er2O3 ceramic coating fabricated using the metal–organic decomposition (MOD) method on a Cr2O3-covered low-activation ferritic steel JLF-1 substrate was examined to improve hydrogen permeation barrier performance of the coating. The Cr2O3 layer was obtained before coating by heat treating the substrate at 700°C under reduced pressures of <5×10−3Pa and 5Pa. The Cr2O3 layer was significantly stable even with heat treatment at 700°C in air. This layer prevented further production of Fe2O3, which has been considered to degrade coating performance. An MOD Er2O3 coating with a smooth surface was successfully obtained on a Cr2O3-covered JLF-1 substrate by dip coating followed by drying and baking. Preprocessing to obtain a Cr2O3 layer would provide flexibility in the coating process for blanket components and ducts. Moreover, the Cr2O3 layer suppressed hydrogen permeation through the JLF-1 substrate. While further optimization of the coating fabrication process is required, it would be possible to suppress hydrogen permeation significantly by multilayers of Cr2O3 and MOD oxide ceramic.

Study of safety features and accident scenarios in a fusion DEMO reactor

After the Fukushima Dai-ichi nuclear accident, a need for assuring safety of fusion energy has grown in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of Broader Approach DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO concept. This concept employs in-vessel components that are cooled by pressurized water and built of a low activation ferritic steel (F82H), contains solid pebble beds made of lithium-titanate (Li2TiO3) and beryllium–titanium (Be12Ti) for tritium breeding and neutron multiplication, respectively. It is shown that unlike the energies expected in ITER, the enthalpy in the first wall/blanket cooling loops is large compared to the other energies expected in the reference DEMO concept. Reference accident event sequences in the reference DEMO in this study have been analyzed based on the Master Logic Diagram and Functional Failure Mode and Effect Analysis techniques. Accident events of particular concern in the DEMO have been selected based on the event sequence analysis and the hazard assessment.

Oxide coating fabrication by metal organic decomposition method for liquid blanket systems

Performance tests and process improvement for ceramic coatings fabricated by the metal organic decomposition (MOD) method have been conducted to obtain tritium permeation barrier and MHD insulator in liquid cooled blanket systems. In the present study, fabrication of MOD Er2O3 coating on a low activation ferritic steel substrate has been tested under reduced pressure condition. Measured hydrogen permeation reduction factors of >100 indicate that MOD coating fabrication under reduced pressure would be effective for suppression oxidation of a substrate and achieving superior coating performances. While crystallinities of MOD coating layers are significantly lower compared with sintered bulk materials, results of electrical conductivity and breakdown voltage measurements indicate that MOD Er2O3 coatings would have sufficient performances as an MHD insulation coating in a Li/V-alloy blanket system. Cathodoluminescence measurement using scanning electron microscope (SEM) is successfully applied to microscopic characterization of MOD coating layers.

Manufacture of Vacuum Plasma Spraying Tungsten with Homogenous Texture on Reduced Activation Ferritic Steel at about 873 K

The key to improving the heat load of vacuum plasma sprayed tungsten coatings on low activation ferritic steel maintained at low temperatures is elimination of stratified low-density layers with many large pores, in which thermal cracks propagate preferentially. The low-density layers are formed owing to the deposition of large solidified tungsten particles, which remain mainly at the periphery of the spray stream. In this study, by shading this periphery, partially homogeneous tungsten coatings without large pores were successfully obtained. The coatings are expected to show good heat load, which is feasible for nuclear fusion applications.

Analysis of Carbon-Bearing Materials for Use as First Wall Armor in the High Average Power Laser (HAPL) Chamber

The characterization of lifetime-component capabilities of various chamber armors is a critical path to the development of the High Average Power Laser (HAPL) reactor design. Previous studies have examined tungsten as an armor material to protect the low-activation ferritic steel first wall from x-ray and ion damage.Carbon-bearing materials are of interest as candidate armor materials due to their desirable thermal and mechanical properties. This analysis examines and compares several carbon-bearing materials: silicon carbide, graphite, engineered graphitic materials and carbon nanotube composites.The transient thermal response of these materials was simulated with the BUCKY 1-D radiation hydrodynamics code utilizing the standardized HAPL x-ray and ion threat spectra. Evacuated and buffer gas filled bare-walled configurations were simulated.

Conceptual design study of fusion DEMO plant at SWIP

The basic definition and development strategy of the DEMO plant based on the Chinese fusion power plant (FPP) program are presented briefly. A conceptual design study of fusion HCSB-DEMO reactor with a fusion power of 2550 MW and a neutron wall loading of 2.3 MW/m 2 is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li 4SiO 4), the structure material of low-activation ferritic steel (LAF/M) and Be neutron multiplier based on Chinese ITER HCSB-TBM design foundation are considered. The design parameters, preliminary analyses and the basic structure as well as development strategy of HCSB-DEMO reactor are introduced.

Assessment of the activation, decay heat, and waste disposal of the US helium-cooled ceramic breeder test blanket module in ITER

The radioactivity inventory and decay heat in the US helium-cooled ceramic breeder (HCCB) test blanket module (TBM) have been assessed at shutdown and for several times thereafter. The sub-module will have its own FW and structural container box that houses the breeder and beryllium pebble bed units, arranged in an edge-on-configuration. Low activation ferritic steel (F82H) is used as the structure and helium is used as a coolant. The breeder beds are made of Li 2TiO 3 pebbles in which lithium has been enriched up to 75% in Li-6. Pulsed operation mode is assumed. During operation in the D-T phase, the total heating rate in the TBM is ∼263 kW. The total amount of tritium generated in the breeder and the beryllium multiplier is ∼9 g and 0.07 g, respectively, after reaching the 0.3 MWa/m 2 fluence limit. At shutdown, the total radioactivity and decay heat levels are ∼0.89 MCi and ∼0.002 MW, respectively. These values drop sharply after 1 min to ∼0.098 MCi and ∼0.0006 MW. The contribution from the F82H structure is the dominant one up to ∼10 years following shutdown. After ∼10 years, the contribution to the total activation and decay heat from the breeder material is the dominant one due to the generated tritium. The WDR of various components are far below unity and thus are well within ITER regulatory guidelines.

Safety Assessment of the ARIES Compact Stellarator Design

ARIES-CS is a 1000 MW(electric) compact stellarator conceptual fusion power plant design. This power plant design contains many innovative features to improve the physics, engineering, and safety performance of the stellarator concept. ARIES-CS utilizes a dual-cooled lead lithium blanket that employs low-activation ferritic steel as a structural material, with the first wall cooled by helium and the breeding zone self-cooled by flowing lead lithium. In this paper we examine the safety and environmental performance of ARIES-CS by reporting radiological inventories, decay heat, and radioactive waste management options and by examining the response of ARIES-CS to accident conditions. These accidents include conventional loss of coolant and loss of flow events, an ex-vessel loss of coolant event, and an in-vessel loss of coolant with bypass event that mobilizes in-vessel radioactive inventories (e.g., tritium and erosion dust from plasma-facing components). Our analyses demonstrate that the decay heat can be safely removed from ARIES-CS and the facility can meet the no-evacuation requirement.

Corrosion characteristics of low activation ferritic steel, JLF-1, in liquid lithium in static and thermal convection conditions

The compatibility of JLF-1(Fe–9Cr–2W–0.1C), a reduced activation ferritec/martensitic (RAFM) steel with static and flowing lithium (Li) was investigated. The corrosion characteristics were studied by means of measurement of weight losses and scanning electron microscope (SEM)/energy dispersive X-ray spectrometer (EDS) analysis on the surfaces and cross-sections of the specimens. The weight losses of JLF-1 specimens showed saturation at the temperature of 500 °C and 600 °C. This is possibly due to the saturation of Fe, Cr in bulk Li or formation of saturated layer of dissolved elements in liquid Li near the specimen surface. In the corrosion test in a thermal convection loop, the corrosion rate at 500 °C for 250 h was significantly larger than that obtained in the static test in an identical condition. After Li exposure, the phase transformation from martensite to ferrite was found on the specimens. The chemical analysis results and the loss of carbides suggested that the phase change was caused by the depletion of carbon. At the same time, selective depletion of alloy elements, such as Cr and W was detected by EDS on the surface. The result of Vickers hardness test showed that obvious softening occurred on the surface of the specimens after Li exposure and the depth of the softened region was consistent with that of the phase transformation. The flowing Li enhanced the weight loss, phase change and hardness reduction due to the mass transfer.

Failure Strength Measurements of VPS Tungsten Coatings for HAPL First Wall Armor

The High Average Power Laser (HAPL) project is pursuing development of an IFE power reactor using a solid first wall chamber. Tungsten has been chosen as the primary candidate armor material protecting the low activation ferritic steel chamber wall structure. The tungsten armor is less than 1-mm thick and is applied by vacuum plasma spraying (VPS). The failure strength of the tungsten-armor is critical, which is measured using a state-of-the-art spallation technology developed at UCLA. A nano-second laser is used to propagate a compression/tension stress wave through the composite layered structure. The tensile strength in the coating is then related to the displacement velocity of the free surface of the tungsten coating. VPS tungsten coated steel samples were tested using the laser spallation technique and coating strengths were evaluated and are reported.

Three-Dimensional Nuclear Assessment for the Chamber of Z-Pinch Power Plant

Detailed three-dimensional nuclear analyses have been carried out for the chamber of a power plant concept that utilizes the Z-Pinch driven inertial confinement technology with a target yield of 3 GJ and repetition rate of 0.1 Hz per chamber. The elliptical chamber concept was modeled with the double-layered Recyclable Transmission Lines (RTL). Thick liquid jets are utilized to breed tritium, absorb energy, and shield the chamber wall. Two liquid breeder options were considered; the molten salt Flibe and the LiPb eutectic (Li17Pb83). The chamber wall is made of the low activation ferritic steel alloy F82H. While both breeders have the potential for achieving tritium self-sufficiency, the thermal power is ∼6.5% higher with LiPb. However, a 55% thicker jet zone is required with LiPb to provide adequate chamber wall shielding. A thicker chamber wall is required with LiPb to reduce the nuclear energy leakage below 1%. The chamber wall does not need replacement except for the top part around the jet nozzles. Helium production in the chamber wall protected by LiPb is much lower than that with Flibe. Rewelding is possible only in the lower part of chamber wall below the pool.