Low Tritium Inventory Research Articles

Repair of heat load damaged plasma–facing material using the wire-based laser metal deposition process

Due to its unique properties tungsten is a promising candidate as plasma-facing-material (PFM) in future nuclear fusion reactors. Tungsten features an exceptionally high melting point, high thermal conductivity, low tritium inventory and comparatively low erosion rate under plasma loading [1]. But given the extreme loads on the PFM during operation of a fusion reactor, the lifetime of plasma-facing components (PFC)s is limited. Currently, it is planned to replace damaged PFCs when they reach the end of their service life. However, the lifetime of PFCs could be increased by in situ repair using additive manufacturing technology (AM) in the form of direct-energy-deposition (DED). The wire–based laser metal deposition process (LMD-w) meets several necessary conditions for operation in the vessel and could be used for performing such in situ repairs.It was investigated if the LMD-w process is able to heal thermal induced surface cracks and roughening by remelting the substrate during deposition of tungsten. For this purpose, tungsten samples of 12 × 12 × 5mm3, which later served as substrate plates for the LMD-w experiments, were treated with combined steady-state and transient thermal loads in the electron beam facility JUDITH 2. These samples were brazed to a copper cooling structure and exposed to 105 thermal shocks of 0.5 ms duration and an intensity of Labs = 0.55 GW m−2 (FHF = 12 MW s0,5 m−2) at a base temperature of Tbase = 700 °C. This way, edge localized mode (ELM) like thermal load damage was induced on the tungsten samples. On these samples, different LMD-w and laser remelting process strategies were performed. Subsequently, these samples were analyzed, and it was examined that the healing of the pre-damaged substrate material was successful. In parallel, the laser remelting process was modeled in a thermal transient finite element method(FEM) simulation in order to gain an insight into the temperatures prevailing in the material during the process.

Fusing together an outline design for sustained fuelling and tritium self-sufficiency.

Ensuring tritium fuel self-sufficiency while maintaining continuous and high-specification fuel flow to the tokamak via a low tritium inventory and controllable fuel cycle is a significant challenge to the STEP plant design. Effective and high-quality fuelling and exhaust design is required to sustain and control a stable plasma, whereas fuel sufficiency is required to prevent depletion of available tritium supply. Concerns regarding the lack of tritium availability preventing continuous tritium import are countered by breeding, where highly energetic neutrons from the core fusion reactions interact with lithium atoms suspended in the surrounding breeder blanket to produce tritium. The compact nature of STEP prohibits the integration of inboard breeder blankets posing a significant challenge for the design team looking to ensure more tritium is bred and made available than consumed within the core plasma. This paper outlines how purposeful technology selection and system architecting has converged on the outline of a conceivable and tritium-capable fuel cycle and breeder blanket design. Before introducing the STEP fuel cycle design outline and summarizing the approach undertaken to address the challenges facing plasma fuelling, key aspects of fuel self-sufficiency are discussed. This includes discussing a proposed helium-cooled liquid lithium breeder blanket and possible technology options for tritium extraction from lithium. Lastly, there is a brief process modelling overview, which emphasizes the central contribution of various employed modelling methods. Reflections on the presented fuel cycle design outline conclude that substantial development work is still required to realize a continuous tritium fuel cycle design and overcome the major challenges posed by tritium and lithium handling. Reflections on the presented breeder blanket design proposal conclude that while many substantial risks and blockers remain to achieve fuel self-sufficiency, high breeding ratios are expected to be achievable with a compact spherical tokamak configuration. Nonetheless, it is recognized that further consideration is required to ensure that the selection of liquid lithium as a breeder medium provides the overall simplest route to a self-sufficient and realizable design.This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

614 タングステンと銅との接合材料の硬さに及ぼす高温照射効果の影響(OS4-3 オーガナイズドセッション《材料・組織と加工》)

Since plasma facing components for fusion reactor devices are exposed to very high heat load and plasma particle attacks, they are required to have high thermal resistance, high thermal conductivity, thermal shock resistance, radiation damage resistance, etc. Tungsten and copper joint materials are expected to use as the plasma facing components with high performances for the next fusion devices, so they have high joint strength, low erosion and low tritium inventory. However, it is feared that the mechanical properties and the microstructures change and degrade by neutron radiation damage and heat radiation. In this study, it aims to clarify the effect of irradiation damage on hardness of tungsten and copper joint materials at high temperature.

Mechanical properties and microstructural characterizations of potassium doped tungsten

Abstract Tungsten is a very promising candidate material for plasma facing components in fusion reactor due to its high melting temperature, high thermal conductivity, low tritium inventory and low erosion rate under plasma loading ( Mitteau et al., 2007 ). The main drawback is the embrittlement at low temperature. The potassium doped tungsten grade WVWM produced by Plansee AG could be a potential plasma facing material for future nuclear fusion facilities and reactors such as ITER and especially DEMO. For a better understanding of both recrystallization and ductile to brittle transition temperature, tensile tests are performed on potassium doped tungsten (WVWM), up to 2000 °C at different loading rates (0.2 and 42 mm/min). The mechanical properties are highly dependent on the microstructure. The fracture surfaces after tensile testing are microstructurally assessed by scanning electron microscopy (SEM) and transmission electron microscopy (TEM) is used to investigate the original specimen as the beginning point.

The ITER tritium systems

ITER is the first fusion machine fully designed for operation with equimolar deuterium–tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium–tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium. The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium–tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept. Not only the interfaces between the primary fuel cycle systems – being procured through different Participant Teams – but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping – again procured through different Participant Teams – and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task.

05/01520 Implications of convective scrape-off layer transport for fusion reactors with solid and liquid walls

The blanket design for a force-free helical reactor (FFHR) is presented, which is a demo-relevant heliotron-type D-T fusion reactor based on the first all-superconducting-coils device, LHD (large helical device) under construction in NIFS at present. For the goal of a self-ignited reactor of 3 GW thermal output, the design parameters at the first stage for concept definition of FFHR have been investigated. The main feature of FFHR is a force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use a high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt FLiBe as a self-cooling tritium breeder for mainly safety reasons owing to the low tritium inventory, low reactivity with air and water, low pressure operation, and low MHD resistance compatible with a high magnetic field. In particular, as common issues in fusion reactors, the FLiBe blanket system in FFHR is expressed in detail by showing engineering possibilities to overcome key issues on tritium permeation, material corrosion, heat transfer, operation pressure, etc. The basic design for maintenance and repair of the blanket is also discussed.

Fire, A Test Bed for ARIES-RS/AT Advanced Physics and Plasma Technology

The overall vision for FIRE is to develop and test the fusion plasma physics and plasma technologies needed to realize capabilities of the ARIES-RS/AT power plant designs. The mission of FIRE is to attain, explore, understand and optimize a fusion dominated plasma which would be satisfied by producing DT fusion plasmas with nominal fusion gains ~10, self-driven currents of ≈ 80%, fusion power ~ 150 - 300 MW and pulse lengths up to 40 s. Achieving these goals will require the deployment of several key fusion technologies under conditions approaching those of ARIES-RS/AT. The FIRE plasma configuration with strong plasma shaping, a double null pumped divertor and all metal plasma facing components is a 40% scale model of the ARIES-RS/AT plasma configuration. “Steady-state” advanced tokamak modes in FIRE with high β, high bootstrap fraction and 100% non-inductive current drive are suitable for testing the physics of the ARIES-RS/AT operating modes. The development of techniques to handle power plant relevant exhaust power while maintaining low tritium inventory is a major objective for a burning plasma experiment. The FIRE H-modes and AT-modes result in fusion power densities from 3 - 10 MWm-3 and neutron wall loading from 2 - 4 MW m-2 which are at the levels expected from the ARIES-RS/AT design studies.

Effect of TiO 2 on the reduction of lithium titanate induced by H 2 in the sweep gas

Lithium titanate (Li 2TiO 3) has attracted the attention of many researchers because of good tritium recovery at low temperature, chemical stability, low tritium inventory, etc. In this work, the effects of TiO 2 doping of Li 2TiO 3 on the resistance to the environmental attack in air and hydrogen reduction under Ar+0.1%H 2 flow were examined. Based on the results, the TiO 2 doping (from 3.7 to 30 wt%) was found to improve the resistance to the environmental attack for very long exposure times in air at room temperature. The hydrogen reaction rates of Li 2TiO 3 and Li 4Ti 5O 12 were almost similar up to 800 °C and that of Li 4Ti 5O 12 was higher in the temperature range from 800 to 950 °C.

Tritium inventory and recovery in next-step fusion devices

Future fusion devices will use tritium and deuterium fuel. Because tritium is both radioactive and expensive, it is absolutely necessary that there be an understanding of the tritium retention characteristics of the materials used in these devices as well as how to recover the tritium. There are three materials that are strong candidates for plasma-facing-material use in next-step fusion devices. These are beryllium, tungsten, and carbon. While beryllium has the disadvantage of high sputtering and low melting point (which limits its power handling capabilities in divertor areas), it has the advantages of being a low-Z material with a good thermal conductivity and the ability to get oxygen from the plasma. Due to beryllium's very low solubility for hydrogen, implantation of beryllium with deuterium and tritium results in a saturated layer in the very near-surface with limited inventory (J. Nucl. Mater. 273 (1999) 1). Unfortunately, there are nuclear reactions generated by neutrons that will breed tritium and helium in the material bulk (J. Nucl. Mater. 179 (1991) 329). This process will lead to a substantial tritium inventory in the bulk of the beryllium after long-term neutron exposure (i.e. well beyond the operation life time of a next-step reactor like ITER). Tungsten is a high-Z material that will be used in the divertor region of next-step devices (e.g. ITER) and possibly as a first wall material in later devices. The divertor is the preferred location for tungsten use because net erosion is very low there due to low sputtering and high redeposition. While experiments are still continuing on tritium retention in tungsten, present data suggest that relatively low tritium inventories will result with this material (J. Nucl. Mater. 290–293 (2001) 505). For tritium inventories, carbon is the problem material. Neutron damage to the graphite can result in substantial bulk tritium retention (J. Nucl. Mater. 191–194 (1992) 368), and codeposition of the sputtered carbon with the tritium from the plasma will produce a layer of carbonaceous material potentially containing kilograms of tritium in the cooler areas of the tokamak (J. Vac. Sci. Technol. A5 (1987) 2286). This paper reviews the tritium retention mechanisms for the three materials discussed above. Tritium removal techniques, including those used in situ to minimize in-vessel inventories as well as those used to reduce contamination prior to waste disposal, are discussed.

The tritium fuel cycle of ITER-FEAT

The Tritium Plant of ITER-FEAT is essential for the operation of the machine after the initial hydrogen phase, as tritium will be produced from DD fusion reactions. Within the fuel cycle of the Tokamak deuterium and later also tritium will be provided to the Fuelling Systems, and the unburned DT fraction recovered from the exhaust gases. The design of the tritium fuel cycle has to be based upon well proven technology to assure the safe handling of tritium along with credible accountancy, low tritium inventory, low generation of wastes and a high reliability of all components throughout the lifetime of ITER-FEAT.

In-situ Tritium Recovery Experiments of Blanket In-pile Mockup with Li2TiO3 Pebble Bed in Japan

Lithium titanate (Li2TiO3) is one of the most attractive tritium breeders for breeding blanket in fusion reactor from view points of low tritium inventory, high chemical stability and so on. The data on the performance of a blanket mockup with pebble bed under neutron irradiation is needed for the design of breeding blanket. To obtain such data, two kinds of the blanket in-pile mockups with Li2TiO3 pebble bed were developed and the in-situ tritium recovery experiments were carried out in the Japan Materials Testing Reactor (JMTR). In these studies, effects of various parameters, i.e., irradiation temperature, sweep gas flow rate, etc., on the tritium recovery behavior from Li2TiO3 pebble bed were evaluated. It was found that the tritium recovery (R) to tritium generation (G) ratio (R/G) increased with increasing the temperature of Li2TiO3 pebble bed and was saturated when the temperature of Li2TiO3 pebble bed at the outside edge exceeded 300°C. Additionally, the sweep gas flow rate in the range of 100 to 900 cm3/min affected very little the tritium recovery from Li2TiO3 pebble bed. A good prospect for the design of breeding blankets using Li2TiO3 pebble bed was obtained from these results of in-situ experiments.

In-situ Tritium Recovery Experiments of Blanket In-pile Mockup with Li2TiO3 Pebble Bed in Japan.

Lithium titanate (Li2TiO3) is one of the most attractive tritium breeders for breeding blanket in fusion reactor from view points of low tritium inventory, high chemical stability and so on. The data on the performance of a blanket mockup with pebble bed under neutron irradiation is needed for the design of breeding blanket. To obtain such data, two kinds of the blanket in-pile mockups with Li2TiO3 pebble bed were developed and the in-situ tritium recovery experiments were carried out in the Japan Materials Testing Reactor (JMTR). In these studies, effects of various parameters, i.e., irradiation temperature, sweep gas flow rate, etc., on the tritium recovery behavior from Li2TiO3 pebble bed were evaluated. It was found that the tritium recovery (R) to tritium generation (G) ratio (R/G) increased with increasing the temperature of Li2TiO3 pebble bed and was saturated when the temperature of Li2TiO3 pebble bed at the outside edge exceeded 300°C. Additionally, the sweep gas flow rate in the range of 100 to 900 cm3/min affected very little the tritium recovery from Li2TiO3 pebble bed. A good prospect for the design of breeding blankets using Li2TiO3 pebble bed was obtained from these results of in-situ experiments.

Study of safety concept for a helical-type fusion reactor, FFHR

Safety concepts for a force free helical reactor (FFHR), of which conceptual designs have been studied in parallel with the construction of the large helical device (LHD), are presented. The main safety features expected for the FFHR are, steady-state plasma operation, no dangerous current disruptions, and the selection of a molten-salt Flibe blanket for a tritium breeder which has low tritium inventory, self cooling effect and low chemical activity. The blanket reaches such high temperatures that the tritium permeation rate increases and this tritium must be continuously recovered. Then the first wall is expected to have a low tritium inventory. Based on the classification of the plant systems and components, it is shown that the high mobility and large radioactivity release accident accompanied with the energy release is avoidable. Fundamental safety would be secured by multiple protection systems. To prevent escalation to a severe accident caused by the fracture of any of the large radioactivity confinement components, passive and reliable active protection systems are proposed. Thus, FFHR will be designed to be as safe as possible.

Compact toroid fuelling for ITER

Deep fuelling of ITER may be achievable by compact toroid (CT) injection. The advantages of deep fuelling include avoiding edge density limits by fuelling beyond the transport barriers, profile peaking to reach ignition, profile control to improve current drive, low tritium inventory, etc. In this paper, a conceptual design of a CT fueller for ITER is presented. In the proposed design, a CT injector fires 2.2-mg toroids of DT fuel into ITER at a rate of up to 20 Hz and a speed of 300 km s −1. Each CT adds ≈0.3% to the Tokamak plasma inventory. The fuelling rate is 20 Pa m 3 s −1, which is based on achieving 10% burnup by deep fuelling (beyond r/ a∼0.5), and with a parallel gas puffing system. Each CT is ≈ 7% of the minor radius in size. The CT Fueller system would consume 8 MWe, with ≈2 MW deposited in the Tokamak.

Blanket design using FLiBe in helical-type fusion reactor FFHR

The blanket design for a force-free helical reactor (FFHR) is presented, which is a demo-relevant heliotron-type D-T fusion reactor based on the first all-superconducting-coils device, LHD (large helical device) under construction in NIFS at present. For the goal of a self-ignited reactor of 3 GW thermal output, the design parameters at the first stage for concept definition of FFHR have been investigated. The main feature of FFHR is a force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use a high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt FLiBe as a self-cooling tritium breeder for mainly safety reasons owing to the low tritium inventory, low reactivity with air and water, low pressure operation, and low MHD resistance compatible with a high magnetic field. In particular, as common issues in fusion reactors, the FLiBe blanket system in FFHR is expressed in detail by showing engineering possibilities to overcome key issues on tritium permeation, material corrosion, heat transfer, operation pressure, etc. The basic design for maintenance and repair of the blanket is also discussed.

Recent results on vacuum pumping and fuel clean-up

Considerable R&D effort has gone into the development of a Torus Exhaust Vacuum Pumping System over the past years. The selected primary pump is of cycling type with discontinuous regeneration, solid cryosorbent for helium pumping and no fuel separation. The pump includes a throttle/regeneration valve to maintain the divertor pressure, minimize the thermal load on the pump and permit regeneration. Modelling of various torus cryopump configurations has been carried out and a research program aimed at developing a primary torus cryopump with rapid regeneration cycles is in progress. Panel tests with tritium will provide additional data required for ITER design. The design, fabrication and demonstration of a model cryopump with integral inlet valve is in progress. Experimental work and modelling to develop a high speed pumping system for the neutral beam injector is under way. Practical experience under ITER relevant conditions and in the presence of tritium has accumulated on the performance of oil-free pumps in various tritium laboratories. Recent world-wide developments in the field of fuel clean-up have been impressive. Several concepts based on palladium/silver diffusors combined with catalytic as well as with electrochemical process steps have the potential to achieve the high overall tritium decontamination factors and low tritium inventory levels required by ITER. Basic research activities aimed at characterizing and optimizing the envisioned essential components for the various ITER fuel clean-up concepts are in progress. Tests with an industrially manufactured facility of technical scale employing realistic concentrations of tritium and using an advanced operation and safety concept have been performed. Optimization of process components with respect to performance, operability and safety criteria is proceeding in parallel with integrated loop tests. Endurance tests to demonstrate complete processes with tritium are in preparation in several laboratories. Basic research aimed at developing a process for the recovery of tritium from the helium glow discharge cleaning exhaust stream is under way. The routine operation with tritium of large civilian research laboratories under conditions specific to the respective sites will contribute significantly to a realistic design of the ITER Tritium Plant. For the operation of the Tritium Plant and tritium accountancy much effort has gone into the improvement of known analytical techniques and the further development of calorimetry.

D-T operation on TFTR

The Tokamak Fusion Test Reactor (TFTR) has operated safely and routinely with tritium fuel since November 1993. Experiments conducted during this time have demonstrated improved plasma confinement properties, the first measurements of alpha heating and transport and the ICRF heating and current drive of D-T plasmas. Reactor level fusion power densities up to 2.8 MW m −3 have been attained as well as more than 10 MW of fusion power and 6.5 MJ of fusion energy per pulse. In support of these accomplishments, the TFTR equipment has met or exceeded the original design values. The toroidal magnetic field has been upgraded from 5.2 to 6.0 Tesla (on-axis), the neutral beams have injected up to 40 MW with a combination of deuterium and tritium fueling and the ICRF systems have been operated at various frequencies, from 30 to 64 MHz, to support an assortment of fast wave experiments. From November 1993 through July 1996, there have been 19 725 plasma shots. These include 841 deuterium-tritium pulses which have produced a total of 4.8×10 20 D-T neutrons and 1.4 GJ of fusion energy. More than 864 kCi (almost 90 g) of tritium have been processed through the TFTR facility while constrained to a site limit (`in-process') of 50 kCi. A low tritium inventory cryogenic distillation system has been recently installed and tested to provide on-site repurification of the plasma exhaust. Maintenance and repair activities have become routine in this facility. Hundreds of calibrations, repairs and installations requiring work on tritium contaminated and activated systems have been performed. Utilizing As Low As Reasonably Achievable (ALARA) principles, the radiation exposure to workers has been comparable to that experienced during the previous deuterium-only operating phase. The total annual site boundary dose, due to all pathways (prompt neutron radiation, activated air and controlled tritium release), has been less than 3 μSv, or 3% of the design limit of 100 μSv and 0.3% of the regulatory limit of 1 mSv.

Compact Toroid Fueling for ITER

Experimental and theoretical work indicates that deep fueling of ITER may be possible by Compact Toroid (CT) injection. CT velocities sufficient for center fueling of a reactor have been demonstrated in the RACE device. CT injections into the TdeV tokamak have achieved central penetration at 1.4 T, and have increased the particle inventory by more than 30% without disruption. Tests on the MARAUDER device have achieved CT mass-densities suitable for injection into 5 T tokamaks. Techniques for producing multiple-shot CT's with passive electric switching are being tested on CTIX. The advantages of deep fueling by CT injection include profile peaking to reach ignition, profile control, low tritium inventory and others. In this paper, the CT experimental results are summarized, a conceptual design of CT fueler for ITER is presented, and the implications on ITER operation and fuel cycle are discussed.

ITER blanket designs

The ITER first wall, blanket, and shield system is being designed to handle 1.5 ± 0.3 GW of fusion power and 3 MWa m −2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design.

A Reactor-Permeator for Reduction of Tritium Oxide on Iron

Tritiated water (Q{sub 2}O) is produced during fusion fuel purification or air detritiation. Before recovering the tritium by isotope separation, the Q{sub 2}O needs to be reduced to form Q{sub 2} gas. The reduction of tritiated water on iron is an alternative to electrolysis and gas-shift reactors. It allows a simple, compact, configuration with low tritium inventory. The reactor design incorporates a palladium alloy permeator which extracts the Q{sub 2}. Tests on a commercial iron-based catalyst showed a high reactivity and no degradation with repeated cycling. The optimum temperature for water reduction was 375-395{degree}C, and for iron regeneration using hydrogen, 470-495{degree}C. The first prototype reactor-permeator decomposed 9.5 g water in 8 hrs using 210 g iron. The time needed for iron regeneration was reduced to 16 hrs by recirculating the hydrogen. A pilot-scale reactor permeator is now under development: it should be capable of reducing 35 kg of water per year, operating at 1 bar. Attention to the choice of structural materials will minimize tritium carryover into the water produced during regeneration. 7 refs., 2 figs.