Vacuum Vessel Material Research Articles

Integral analysis of the effect of material dimension and composition on tokamak neutronics *

The neutronics performance of a tokamak has been identified as an important factor in designing a fusion power plant. The design of the tokamak should not only meet operational parameters such as sufficient tritium breeding, but also safety parameters such as low structural material activation. This paper investigates the impacts of the neutronics metrics for the ARC-class tokamak, a compact tokamak with an immersion blanket, by perturbing the first five layers of structural material—first wall, inner vacuum vessel, coolant salt channel, neutron multiplier, and outer vacuum vessel. The goal of this work is to provide insight into shaping and scaling the flux on each layer to obtain optimized operational and safety metrics through quantification of the responses from each perturbation. Results show that increased first wall thickness can increase the tritium breeding ratio (TBR) in specific configurations with high 6Li enrichments and that vacuum vessels decrease TBR for low-6Li enrichment configurations. It was also found that the neutron multiplier can either increase or decrease TBR depending on the configuration. The response of metrics to the change in layer thickness and enrichment also varies depending on the vacuum vessel material. The integral impacts of 6Li enrichment, layer thicknesses, and vacuum vessel material choice are investigated and presented in this paper.

Effect of the various tritium breeding materials on the tritium breeding ratio in ARC fusion reactor

This study deals with TBR (Tritium Breeding Ratio) performance of the four different types of molten salt (FLiBe, FLiNaK, FLiNaBe, FLiNaRb) which has four different Li-6 enrichment ratios (0%, 30%, 60%, 90%) Moreover, two different vacuum vessel materials (SiC composite and V4Cr4Ti alloy) have been used in ARC reactor geometry. In models with TBR ≥1.05, it has been desired to examine the sustainability of the tritium production of the reactor by removing the neutron multiplication layer from the design without using the toxic Beryllium element. The neutronic performance studies have been examined in OpenMC, which is an open-source Monte Carlo neutron and photon transport code. The TBR values of the different models obtained and the change of these values according to Li-6 enrichment were examined. Obtaining result showed that the FLiBe molten salt and V4Cr4Ti alloy vacuum vessel wall material could be the best choice for the ARC reactor.

Development of a thermo-mechanical behaviour model adapted to the ITER vacuum vessel material

The ITER machine has been classified as a Basic Nuclear Installation French nuclear regulator (INB n°174), which implies that it will be the first fusion reactor to go through complete French nuclear licencing. The combination of mechanic and electromagnetic phenomena with the heat loads caused by neutron streaming requires a multi-physics approach to the damage assessment, which has not yet been implemented in the common nuclear codes and standards.The general damage prevention methodology consists in guaranteeing the structural integrity of a component. The development of design rules has mainly two origins: prevention of damage from monotonic mechanical loads and prevention of damage from repeated application of loads. In most cases, structural integrity is justified within a linear elastic behaviour but when this route is not enough to respect the design criteria, several non-linear approaches to the material's mechanical behaviour can be considered, requiring more elaborated demonstration of the design compliance. Nevertheless, the models proposed in the nuclear model database are sometimes not sufficient to properly describe the experimentally observed cyclic plasticity behaviour and, in particular, the ratcheting and shakedown phenomena.According to ITER community experts in materials and analyses, a thermo-mechanical behaviour model fitting the ITER Tokamak materials data will guarantee the best prediction of the damage considering a nuclear and a multi-physic loading condition.This paper describes the assessment of the non-linear behaviour of Vacuum Vessel (VV) material with a strong thermo-mechanical coupling and a damage parameter to prevent crack initiation. More precisely, Chaboche's models available in the literature (elasto-(visco)-plasticity models, with various types of hardening) have been enriched in order to explicitly take into account the influence of the temperature on the mechanical behaviour and, reciprocally, the influence of the mechanical behaviour on the temperature. Mechanical cycling tests have been performed on the VV constitutive material, emphasizing on the progressive deformation state up to failure mode, i.e., ratcheting. The proposed models have been tested on a homogeneous problem and the results compared with uniaxial test results; this type of simulation is commonly called “0D”analysis. The last part of this document describes the finite element implementation of the constitutive material model and its application to the ITER VV welded support.

Design Challenges and Activation Concerns for ARIES Vacuum Vessel

Research has been conducted to find the optimal steel to use in the vacuum vessel (VV) of ARIES power plants. The VV should meet several design criteria, including activation and fabrication requirements. Seven different types of steel were examined in order to determine which steel would be the best candidate for the ARIES VV. The main concerns are related to activation, properties under irradiation, and fabrication of a sizable VV. Steels generating high-level waste (such as 316-SS) were excluded from possible material choices. As a VV material, there is the necessity for a carefully controlled the post-weld-heat-treatment at ~750°C after assembly, welding, and rewelding. For this particular reason, the F82H FS is not suitable for the ARIES VV. The newly developed 3Cr-3WV bainitic FS meets the activation requirements and has the potential to satisfy the fabrication requirements for the ARIES VV. It is recommended for further consideration because of several advantages over other candidate steels.

Fabrication Preparation of ITER Vacuum Vessel—Material Considerations, Regulatory Requirements, and Fabrication Plans

SS 316 L(N)-IG (ITER grade) has been selected as the main structural material for the ITER vacuum vessel (VV), considering its high mechanical strength at operating temperatures, water chemistry properties, excellent fabrication characteristics, and low cost relative to other candidates. The ITER VV is a class-2 box structure as defined in RCC-MR, 2007 edition, which was selected as the code for the design and construction. This paper describes materials, applied code and regulatory requirements, baseline fabrication procedures, and assembly on the site.

Non-destructive tritium measurements of Mk IIA divertor tile by BIXS

β-Ray-induced X-ray spectrometry was applied to determine the amounts and depth profiles of tritium captured by graphite tiles used in the PTE1 campaign and by carbon fiber composite tiles in the DTE1 campaign of JET. Although the X-ray spectra observed for both carbon tiles consisted of characteristic and bremsstrahlung X-rays, the intensities and shapes of the X-ray spectra were quite different between these tiles. The analyses of the X-ray spectra showed that the amount of tritium retained in the latter tile was on average 15 500 times greater than that in the former tile, and the tritium depth profile extended to 60 μm for the latter tile compared to 5 μm for the former tile. In addition, characteristic X-rays corresponding to Cl, Cr, Fe, Ni and Mo appeared in the latter tile, where Ni showed the highest concentration, indicating that considerable erosion and deposition of vacuum vessel materials took place in the DTE1 campaign.

Characterization of non-magnetic Mn–Cr steel as a low induced activation material for vacuum vessels

The JT-60SU (Super Upgrade) program is under discussion at Japan Atomic Energy Research Institute (JAERI). Its design optimization activity requires the vacuum vessel material to be non-magnetic, very strong and with low induced activation. However, there is no suitable material available to fulfill all the requirements. JAERI started to develop a new material for the vacuum vessel together with the Japan Steel Works (JSW). Chemical composition and metallurgical processes were optimized and a new steel named VC9, which has the composition of Cr – 16 wt%, Mn – 15.5 wt%, C – 0.2 wt%, and N – 0.2 wt% with non-magnetic single γ phase, was selected as a candidate material. Here, mechanical properties and weldability of VC9 were examined and the results were compared with those of type 316 or 316L stainless steel. It was shown that VC9 has good mechanical properties and weldability.

Metal-hydride Characterization and Mechanical Properties of Ti-6A1-4V Alloy

Hydrogen absorption/dessociation properties and mechanical properties were examined in a Ti-6A1–4V alloy exposed to highly pressurized hydrogen gas of 0.1–5 MPa at elevated temperatures of 373 K to 773 K. Pressure-Concentration-Temperature (P-C-T) test results of the materials exhibit a large amount of hydrogen absorption up to 3 wt% H at 473 K with plateaus as observed in typical hydrogen absorption materials and the specimens tested are broken into pieces with the same level of hydrogen contents. Hydrogen dissociation from the material is not observed below 773 K. The results of bending tests and fractography show no degradation of mechanical properties of the materials with hydrogen contents of less than 0.1 wt% H. It is concluded from these test results that there is no mechanical degradation of Ti-6A1-4V vacuum vessel material for the tokamak device, JT-60SU due to metal-hydride reaction between Ti-6A1-4V and fuel hydrogen induced by hydrogen absorption mechanism into Ti-6A1-4V under the hydrogen pressure of 0.01–1 Pa H between room temperature and 773 K, in which JT-60SU is being designed, mainly because oxide layer formed on Ti-6A1-4V restrains metal-hydrogen reaction into the material.

Metal-hydride Characterization and Mechanical Properties of Ti-6Al-4V Alloy.

Hydrogen absorption/dessociation properties and mechanical properties were examined in a Ti-6A1–4V alloy exposed to highly pressurized hydrogen gas of 0.1–5 MPa at elevated temperatures of 373 K to 773 K. Pressure-Concentration-Temperature (P-C-T) test results of the materials exhibit a large amount of hydrogen absorption up to 3 wt% H at 473 K with plateaus as observed in typical hydrogen absorption materials and the specimens tested are broken into pieces with the same level of hydrogen contents. Hydrogen dissociation from the material is not observed below 773 K. The results of bending tests and fractography show no degradation of mechanical properties of the materials with hydrogen contents of less than 0.1 wt% H. It is concluded from these test results that there is no mechanical degradation of Ti-6A1-4V vacuum vessel material for the tokamak device, JT-60SU due to metal-hydride reaction between Ti-6A1-4V and fuel hydrogen induced by hydrogen absorption mechanism into Ti-6A1-4V under the hydrogen pressure of 0.01–1 Pa H between room temperature and 773 K, in which JT-60SU is being designed, mainly because oxide layer formed on Ti-6A1-4V restrains metal-hydrogen reaction into the material.

Gas desorption properties of low-activation ferritic steel as a blanket or a vacuum vessel material

The gas desorption properties of low-activation ferritic steel (F82H) were evaluated by thermal desorption spectroscopy (TDS), and compared with those of austenitic steels (316SS and 316LSS). The major desorbed gas species of F82H were CO, H 2O, CO 2 and H 2. The fraction of oxygen-containing species such as CO, CO 2 and H 2O to the total outgassing amount was 98%. The gas desorption behavior was very similar to the case of 316SS. The gas desorption spectra for F82H show the required prebaking temperature becomes higher than 700 K. For the F82H sample exposed to atmosphere, the gas desorption properties were examined. As the time of exposure increase, desorption amounts of H 2O and CO largely increased. This increase may be due to the corrosion in atmosphere.

Hydrogen permeation properties and surface structure of BN-coated stainless steel membrane

We developed boron nitride-coated stainless steel for vacuum vessel materials, with low adsorption rate of oxygen and carbonates. In extremely high vacuum, of a range 10 −10 Pa, the main component of the residual gas is hydrogen. We found a low adsorption rate of hydrogen on the surface of precipitated BN. A precipitated BN layer can also lower the hydrogen permeation rate of a stainless steel membrane. BN-coating has an effect on the reduction of outgassing due to, not only adsorbed hydrogen on the surface of vessel materials, but also contained hydrogen in vessel materials. We successfully coated BN on a stainless steel surface by using surface segregation from BN and the stainless steel co-deposited film. The RHEED pattern of BN showed the diffraction rings and spots of hexagonal-BN which has its C-axis orientated to the surface normal. The image using transmission electron microscopy showed a lattice of h-BN basal plane, and that the diameter of the BN crystals was 50–80 nm. This crystal structure inhibited hydrogen permeation.

Design of the ITER vacuum vessel

The ITER vacuum vessel is a major safety barrier and must support electromagnetic loads during plasma disruptions and vertical displacement events (VDE) and withstand plausible accidents without losing confinement. The vacuum vessel has a double wall structure to provide structural and electrical continuity in the toroidal direction. The inner and outer shells and poloidal stiffening ribs between them are joined by welding, which gives the vessel the required mechanical strength. The space between the shells will be filled with steel balls and plate inserts to provide additional nuclear shielding. Water flowing in this space is required to remove nuclear heat deposition, which is 0.2—2.5% of the total fusion power. The minor and major radii of the tokamak are 3.9 m and 13 m respectively, and the overall height is 15 m. The total thickness of the vessel wall structure is 0.4−0.7 m. The inboard and outboard blanket segments are supported from the vacuum vessel. The support structure is required to withstand a large total vertical force of 200–300 MN due to VDE and to allow for differential thermal expansion. The first candidate for the vacuum vessel material is Inconel 625, due to its higher electric resistivity and higher yield strength, even at high temperatures. Type 316 stainless steel is also considered a vacuum vessel material candidate, owing to its large database and because it is supported by more conventional fabrication technology.

Study of the parametric variation for argon glow discharge conditioning of SS304

SS304 is the most commonly used vacuum vessel material for fusion devices and accelerators. Carbon in SS is the main impurity of concern for these applications. Glow Discharge Conditioning (GDC) is used to remove this impurity in the form of volatile gas production which can be monitored using a Quadrupole Mass Spectrometer (QMS). In this paper variations in the important experimental parameters like discharge pressure, current density, wall temperature and surface pretreatment and their effects on the efficiency of Ar GDC has been discussed. This has been studied with the help of two parameters, desorption yield and total throughput of carbon removed. The efficiency increased with decrease in discharge pressure, increase in current density and increase in wall temperature. The extent of impurity removal was minimal for a polished and baked surface.

Compact Ignition Tokamak Vacuum Vessel Material Selectiona

The Compact Ignition Tokamak (CIT) vacuum vessel is envisioned to be a thick wall torus with a Dee shaped cross section. It must withstand violent impact forces from plasma disruptions. Operating conditions include a base pressure of 10−8 torr and a 350°C base temperature. A suitable vacuum vessel material of construction must combine adequate elevated temperature strength and desired physical properties compatible with the operating service conditions. CIT performance and design requirements also dictate that fabricability and availability critically influence selection. Similarly, a requirement that the vessel shall have a low level waste activation further affects the appropriate choice. Based upon these selection criteria two materials were identified, that used in a unique hybrid design, satisfy both performance and decommissioning requirements.

VACUUM ULTRAVIOLET SPECTRA OF CT-6B TOKAMAK PLASMA

A spectrographic survey of the impurity lines was made on the CT-6B Tokamak with I metre grazing incidence spectrograph. The observed wavelength ranges were from 1580 to 550 ? and from 460 to 43 ?, using 300 lines/mm and 1200 lines/mm gratings respecrively. The survey revealed that the main impurities in CT-6B Tokamak plasma are Mo for high Z impurities and C, N, O for low Z impurities. There are a large number of OV and OVI lines with intense emission. We also observed MoVIIl resonance lines 234.314 ?, 235.510 ? and 237.215 ?. The shortest observed wavelengths are the second order of OVII lines 21.6020 ? and 21804 ?. The first order of these two lines situate beyond the film and can not be recorded. According tc the features of the spectra and the ionization potentials of OV, OVI as well as MoVII, the average electron temperature for CT-6B Tokamak is estimated to be about 140 eV. Oxygen is the main impurity in CT-6B Tokamak, and its lines emission is the main portion of energy loss in CT-6B Tokamak. Lines of the main components of the vacuum-vessel material such as Fe, Cr, Ni and Ti have not been found.

Ｎ，Ｂ，Ｃｅを添加したステンレス鋼の真空容器用材料としての評価

Ｎ，Ｂ，Ｃｅを添加したステンレス鋼の真空容器用材料としての評価

The materials aspects of the JT-60 device

JT-60 is the JAERI Tokamak, of which the objective is to achieve the break-even plasma condition making use of a hydrogen plasma. The construction of this device will be completed in the spring of 1985. The following seven items of JT-60 are discussed from the point of view of materials. 1. (1) Vacuum vessel material: Inconel 625, 2. (2) Liner plates as the first wall: Molybdenum and Inconel 625, 3. (3) Fixed limiters, magnetic limiter plates and armor plates for neutral beam injection: Molybdenum, 4. (4) Coating of the first wall: TiC, 5. (5) In-situ coating of TiC within the torus, 6. (6) Non-magnetic structural materials, and 7. (7) Conductor materials for magnet coils.