Pellet injection advances to next stage in the US

Researchers at the Oak Ridge National Laboratory (ORNL) have developed a continuous extruder for fusion fuel and are advancing state-of-the-art fuelling and plasma control for ITER. Reliable, high-speed continuous fuelling is essential for ITER to meet its goal of operating at 500 MW for several minutes at a time.

The latest pellet injection experiments using US ITER prototype designs were performed during the week of 22 July at the DIII-D Tokamak operated by General Atomics in San Diego, California. The conceptual design review for the ITER pellet injection system was completed earlier this year, and preparations are now underway for full-scale prototype testing.

The task of the pellet injection system is to provide plasma fuelling, while also lessening the impact of plasma instabilities due to large transient heat loads. The ITER pellet injectors must operate continuously, which is very different from most existing tokamak pellet injectors. The ITER machine also requires a higher rate of pellet fuelling throughput.

According to Dave Rasmussen, team leader for the US ITER pellet injection and disruption mitigation systems, „The ITER pellet injectors will require an increase in the deuterium-tritium mass flow and duration by a factor of 1,000 compared to present systems.”

To produce the pellets, researchers developed a twin-screw extruder which shapes a continuous ice stream of deuterium-tritium fuel into specific diameters and lengths.

„There are existing extruders used on tokamaks today, but they cannot meet the requirements of ITER. On most current installations, extruders have only needed to supply a few seconds of fuel pellets at a time, but the ITER Tokamak will require almost an hour of a continuous ice stream for pellet injection. The ORNL twin-screw extruder is designed to meet the requirements of ITER,” notes Mark Lyttle, a project engineer for the US ITER pellet injection and disruption mitigation systems.

Multiple pellet injectors will be installed on the ITER Tokamak, with up to two injectors at each of three locations on the machine. Some locations will be used more for fuelling while others will be deployed for lessening the impact of plasma instabilities known as edge localized modes (ELMs) by a technique called pellet ELM pacing. The pellet injector can also insert impurity pellets made of argon, neon, or nitrogen into the Tokamak for plasma impurity studies. The pellet injectors must also be able to handle tritium, a radioactive isotope of hydrogen with a half-life of about 12 years, safely.

„There is a 30-year technology development history at ORNL behind the ITER pellet injection design,” says Lyttle.

Under testing now is a 1:5 scale pellet twin-screw extruder. „We do plan to build a full-scale prototype and test it at the Spallation Neutron Source cryogenic facility at ORNL, where we have access to a supply of supercritical helium. Supercritical helium at only 5 degrees above absolute zero is used as the coolant to form the pellet ice and we are lucky to have one of the few facilities in the world that can supply our needs here at ORNL,” says Lyttle.

Other key upcoming activities for researchers and engineers are tests of a propellant gas recirculation loop for the pellet injection system using a tritium-compatible vacuum pump. The recirculation loop supplies the pressurized propellant gas and assures that the gas used to accelerate the pellets is not injected into the vacuum chamber of the ITER Tokamak during the fuelling process.

„Initial tests on the pumping speed look promising,” observes Lyttle. This pump has been tested with helium gas and soon will be tested with hydrogen gas. Ultimately, the pump and loop will undergo a multi-year „lifetime” test to assure its readiness for the ITER pellet injection system, where 99.9% availability is required.

For the original article and more news from US ITER, click here.

Melting tungsten for a good cause

Over the past two years ITER physicists and engineers, along with many scientific colleagues within the fusion research community, have been working to establish the design and physics basis for a modified divertor—the component located at the bottom of the huge ITER vacuum vessel responsible for exhausting most of the heat and all of the particles which will continuously flow out of ITER’s fusion plasmas. 

Our current Baseline begins plasma operations with divertor targets armoured with carbon fibre composite (CFC) material in the regions that will be subject to the highest heat flux densities. After the initial years of ITER exploitation, in which only hydrogen or helium will be used as plasma fuel producing no nuclear activation, this divertor is to be replaced. The replacement—a variant of the first component but fully armoured with tungsten—would be the heat and particle flux exhaust workhorse once the nuclear phase, using deuterium and then deuterium/tritium fuel, begins.

In 2011 the ITER Organization proposed to eliminate the first divertor and instead go for the full-tungsten („full-W”) version right from the start. This makes more operational sense and has the potential for substantial cost savings. By June 2013, the design was at a sufficiently advanced stage and we were confident that the necessary tungsten high heat flux handling technology was mature enough to invite external experts to examine our progress during the full-W divertor Final Design Review

But making a choice to begin operations with tungsten in the most severely loaded regions of the divertor is not just a question of having a design ready to build. 

Tungsten, a refractory metal with high melting temperature (3400 Celsius), is a much more difficult material than carbon when it comes to handling very high heat loads and running the plasmas which ITER will require to reach good fusion performance. Why? For two principal reasons: as a metal, tungsten will melt if the heat flux placed on it is high enough; also, as an element with high atomic number it can only be tolerated in minute concentrations in the burning plasma core.

Carbon, on the other hand, does not melt but sublimes (passing directly from solid to vapour) and is low atomic number, so can be tolerated in much higher quantities in the core plasma. Unfortunately, carbon is a difficult option for ITER nuclear phase operations as a result of its great capacity for swallowing up precious tritium fuel and efficiently trapping it inside the vacuum vessel. Tungsten retains fusion fuel only at comparatively low levels.

Why is melting such a problem? Because a melted metal surface is no longer the flat, pristine surface which is installed when the component is new. One of the ways the ITER divertor is able to handle the enormous power flux densities which will be carried along the magnetic field lines connecting to the target surfaces is to make the target intersect the field lines at very glancing angles, so that the power is spread over a wider surface. But a small angle means that any non-flat feature on the surface will receive a higher-than-average heat flux and can be further melted, producing a cascade effect.

The ITER full-W divertor design goes to great lengths to make sure that there is no possibility—on any of the many thousands of high heat flux handling elements—of an edge sticking up (for example, as a result of mechanical misalignment) that could overheat and begin to melt under the relentless bombardment these components receive during high power operation. However ITER’s size means that it will have the capacity to reach a value of stored energy in the plasma more than a factor of 10 higher than the largest currently operating tokamak, JET (EU). When some of this energy is released in a rapid burst (for example due to very transient magnetohydrodynamic events such as ELMs), some melting is possible—even if all edges have been hidden by clever design.

We intend to stop this happening as much as possible by applying ELM control techniques, but occasional larger events cannot always be excluded. So one of the big physics questions we have tried to answer over the past two years is: what exactly happens when a burst of energy, sufficient to melt tungsten, strikes our divertor targets?

Until recently we had only rather complex computer simulations with which to establish the physics design specifications. One of the main worries was not just that energy bursts could roughen up and damage divertor component surfaces, but that the very rapid melting induced by the burst could lead to the expulsion, or spraying, of micro-droplets of tungsten back into the plasma leading to intolerable contamination and a decrease in performance.

The computer simulations say this shouldn’t happen, but the process of melt ejection is so complex that experiment is the only sure test. But how to test the behaviour under conditions which only ITER can create? Well, as far as tokamaks are concerned, the only place where this was even conceivable was at JET, in which natural ELM energy bursts can be generated at levels similar to those expected for controlled ELMs in ITER. The problem is that these comparatively benign transients will not melt a tungsten surface!

In an experiment proposed and planned jointly between JET and the ITER Organization over the past two years, a small region of one of the full-W modules in the JET divertor was carefully modified to create a situation which every divertor designer would do anything to avoid—a deliberately misaligned edge.

The JET divertor modules are made up of about 9,000 small tungsten plates („W lamellas”), bound together by a complex spring loading system. The lamellas are only 5 mm wide and about 60 mm long with 1 mm gaps between neighbouring elements. For the experiment, a few lamellas were machined to make a single element stand up out of the crowd, presenting an edge of about 1.5 mm on average to the plasma in one of the hottest zones of the divertor.
The result: reassuringly unsurprising! Although there was some evidence suggesting the occasional ejection of very small droplets from the melted area, there was very little impact on the confined plasma. As the ELM plasma bursts repetitively melted the edge of the misaligned lamella, the molten material continuously migrated away from the heat deposition zone, accumulating harmlessly into a small mass of re-solidified tungsten (see video at left, courtesy of EFDA-JET). The JET plasmas with 3 MA of plasma current were able to produce ELM plasma pulses very similar to the lowest amplitude events we need to guarantee for 15 MA operation in ITER—a fact which makes the experiments very relevant from a plasma physics point of view.

Much more analysis is required to see how the results can be matched quantitatively by simulation, but the observations are clearly in qualitative agreement with theory. That’s the most reassuring part: that physics codes used to assist in component design for ITER tomorrow can be validated on experiments performed today. We will have to wait another year now for the damaged lamella to be retrieved from JET before the full picture of these important experiments can be completed, but this is already extremely valuable physics input for the important decisions coming up later this year with regard to our divertor strategy.

Green light for ITER’s blanket design

After three days and 29 presentations, a comprehensive design review with probably the largest participation in the history of the ITER project was completed last week. More than 80 experts from the ITER Organization, Domestic Agencies and industry attended the Final Design Review of the ITER blanket system.

„The development and validation of the final design of the blanket system is a major achievement on our way to deuterium-tritium operation—the main goal of the ITER project,” Blanket Integrated Product Team Leader (BIPT) and Section Leader Rene Raffray concluded at the end of the meeting, obviously relieved at the success of this tremendous endeavour. „We are looking at a first-of-a-kind fusion blanket which will operate in a first-of-a-kind fusion experimental reactor.”

The ITER blanket system provides the physical boundary for the plasma and contributes to the thermal and nuclear shielding of the vacuum vessel and the external machine components such as the superconducting magnets operating in the range of 4 Kelvin (-269°C). Directly facing the ultra-hot plasma and having to cope with large electromagnetic forces, while interacting with major systems and other components, the blanket is arguably the most critical and technically challenging component in ITER.

The blanket consists of 440 individual modules covering a surface of 600 m2, with more than 180 design variants depending on the segments’ position inside the vacuum vessel and their functionality. Each module consists of a shield block and first wall, together measuring 1 x 1.5 metres and weighing up to 4.5 tons—dimensions  that not only demand sophisticated remote handling in view of maintenance requirements during deuterium-tritium operation, but also an approach to attaching the modules which is far from trivial when considering the enormous electromagnetic forces. 

The first wall is made out of shaped „fingers.” These fingers are individually attached to a poloidal beam, the structural backbone of each first wall panel through which the cooling water will be distributed. Depending on their position inside the vacuum vessel, these panels are subject to different heat fluxes. Two different kinds of panels have been developed: a normal heat flux panel designed for heat fluxes of up to 2 MW/m2 and an enhanced heat flux panel designed for heat fluxes of up to 4.7 MW/m2.

The enhanced heat flux panels are located in areas of the vacuum vessel with greater plasma-wall interaction and they make use of the hyper-vapotron technology which is similar to that used for the divertor dome elements. All panels are designed for up to 15,000 full power cycles and are planned to be replaced at least once during ITER’s lifetime. A sophisticated R&D program is currently under way in Japan for the development of remote handling tools to dismantle and precisely re-position the panels.  

Due to the high heat deposition expected during plasma operation—the blanket is designed to take a maximum thermal load of 736 MW—ITER will be the first fusion device with an actively cooled blanket. The cooling water is fed to and from the shield blocks through manifolds and branch pipes. Furthermore, the modules have to provide passage for the multiple plasma diagnostic technologies, for the viewing systems, and for the plasma heating systems.

Because of its low plasma-contamination properties, beryllium has been chosen as the element to cover the first wall. Other materials used for the blanket system are CuCrZr for the heat sink, ITER-grade steel 316L(N)-IG for the  steel structure, Inconel 718 for the bolts and cartridges, an aluminium-bronze alloy for the pads that will buffer the electromechanical loads acting on the segments, and alumina for the insulating layer. 

The procurement of the 440 shield blocks is equally shared between China and Korea. The first wall panels will be manufactured by Europe (50%), Russia (40%) and China (10%). Russia will, in addition, provide the flexible supports, the key pads and the electrical straps. The assembly of the blanket is scheduled for the second assembly phase of the ITER machine starting in May 2021 and lasting until August 2022. The work will be performed with the help of two in-vessel transporters working in parallel.

In assessing the work presented at the Final Design Review, Andre Grosman, deputy head of Magnetic Fusion Research Institute at CEA and chair of the review panel, enthusiastically commended the BIPT for its achievements since the Preliminary Design Review in December 2011 which were „beyond the expectation of the panel.” He added: „We have singled out the continuity and benefit of the work done by the ITER Organization and the Domestic Agencies within the BIPT framework with a sharing of risk and information among all stakeholders.”

The panel nevertheless pointed out some remaining issues, including a few challenging issues that need to be addressed at the project level. But thanks to the excellent quality of work performed by the BIPT, the ITER blanket design can today be called „approved.” The BIPT can now turn its focus to addressing the feedback received at the Final Design Review, applying the final touches to the design, and preparing for the Procurement Arrangements, where fabrication is handed over to the Domestic Agencies, starting at the end of 2013.

EUR 83 million contract signed for Liquid Helium Plant

The ITER Tokamak will rely on the largest cryogenic plant (cryoplant) infrastructure ever built. Three liquid helium plants, working in parallel, will provide a total average cooling capacity of 75 kW at 4.5 K and a maximum cumulated liquefaction rate of 12,300 litres/hour.

On Tuesday, 11 December, ITER Director-General Osamu Motojima and the Managing Director of Air Liquide Advanced Technologies, Xavier Vigor, signed the contract for ITER’s three identical liquid helium (LHe) plants. The contract comprises the design, manufacturing, installation and commissioning of the LHe plants, which are adapted to the long-term, uninterrupted operation of the ITER Tokamak. The contract is worth EUR 83 million.

The cryoplant and cryo-distribution system will supply cooling for the ITER superconducting magnets to confine and stabilize the plasma. They will also provide the refrigeration for the cryosorption panels that are necessary to evacuate the helium ashes stemming from the fusion reaction and to assure the required vacuum for the cryostat and the vacuum vessel. All these users require helium cryogen at different temperature levels ranging from 4.5 K, to 50 K and up to 80 K.

The key design requirement is to cope with ITER’s large dynamic heat loads ranging from 40 to 110 kW at 4.5 K mainly deposited in the magnets due to magnetic field variation and neutron production from deuterium-tritium fusion reactions. At the same time, the system must be able to cope with the regular regeneration of the cryopumps.

Manufacturing of the LHe plant main components will start after design finalization in 2014. The first compressor station will be delivered at the end of 2015 and the LHe plants will be ready for the cool-down of sub-systems in 2018.

„This is a major milestone not only for the cryogenic system but for the whole project,” said the Head of the ITER Plant Engineering Division, Luigi Serio. „The contract covers the principal component that will drive the cool-down of the machine, seting the pace toward First Plasma.”

„We are very happy and excited to participate in the great ITER adventure,” Xavier Vigor said. „Be assured that we, the team from Air Liquide, are fully committed to making ITER a success.”

Air Liquide is the world leader in gases for industry, health and the environment, and is present in 80 countries with 46,200 employees. Oxygen, nitrogen, hydrogen and rare gases have been at the core of Air Liquide’s activities since its creation in 1902. In 2011, the Group’s revenues amounted to EUR 14.5 billion, of which more than 80% were generated outside France.

Deuterium from a quantum sieve

A metal-organic framework separates hydrogen isotopes more efficiently than previous methods

Deuterium is the heavy twin brother of hydrogen; however, it is more than 20 times rarer than identical twins. It accounts for only 0.015 percent of natural hydrogen and is twice as heavy as the light isotope.

There is no chemical difference between the two isotopes: both deuterium and ordinary hydrogen react with oxygen to form water. Its double mass allows researchers to lay a trail to elucidate chemical reactions or metabolic processes, however. They dispatch a compound containing deuterium into the processes and analyze in which conversion product it turns up. And this is only one of the tasks that deuterium fulfils in research. It may even become an inexhaustible and climate-neutral fuel in future.

This would be the case if nuclear fusion becomes so technically mature that energy is generated on Earth using the same process that also occurs in the Sun. This produces much less radioactive waste than nuclear fission.

In a cooperation established within the DFG German Research Foundation’s priority program „Porous Metal-Organic Frameworks” (SPP 1362), a team of scientists from the Max Planck Institute for Intelligent Systems in Stuttgart, Jacobs University Bremen and the University of Augsburg have now been able to enrich deuterium contained in hydrogen more efficiently than with conventional methods.

The findings are reported in the journal Advanced Materials. The researchers discovered that a certain metal-organic framework, abbreviated MOF, absorbs deuterium more easily than common hydrogen at temperatures below minus 200 degrees Celsius.

Read more here. 

Tore Supra ready to go WEST

On the other side of the CEA fence, in Cadarache, sits a large tokamak which played an important role in the definition of ITER. Tore Supra, a CEA-Euratom device which began operating in 1988, was the first tokamak to successfully implement superconducting magnets and actively-cooled plasma-facing components.

Over the past twenty-four years, Tore Supra has explored the physics of long-duration plasma pulses, reaching a record of 6.5 minutes in December 2003.

In 2000-2002, Tore Supra was equipped with a new carbon-carbon fibre (CFC) „limiter” — the equivalent of the divertor in ITER — capable of withstanding an ITER-relevant heat load of 10 MW per square metre.

This project, named CIEL for Composants Internes Et Limiteurs, demonstrated that, while CFC performs very well in terms of power handling and compatibility with the plasma, its use results in substantial erosion caused by the physico-chemical reactions between the carbon of the limiter and the hydrogen (deuterium) in the plasma. Further experiments in JET have confirmed these observations.

Now, there are not many options when it comes to choosing the material of a divertor. Fifty years of experience in tokamak technology have narrowed them to two: it’s either CFC or tungsten, their respective advantages or disadvantages depending on the plasma regimes they are exposed to. (More here).

In ITER, it was originally planned to begin operations with a CFC divertor and replace it with a tungsten one before the start of nuclear operation (deuterium + tritium) in 2026. After years of discussions, panels and reviews, a new plan was established and ITER is now considering doing without the first-phase CFC divertor.

Indeed, substantial cost reductions would be achieved by installing a tungsten divertor right from the start and operate it well into the nuclear phase. This solution would also provide for an early training, during the non-nuclear phase of ITER operation, on how to operate with a tungsten divertor.

The ITER Members, however, have not yet reached a unanimous position on this issue.

Whatever ITER decides eventually, the tungsten option must be explored and this is what Tore Supra’s WEST project (W Environment in Steady-state Tokamak, where „W” is the chemical symbol of tungsten) is about.

„ITER success is CEA’s top priority,” says Alain Bécoulet, the Head of CEA-IRFM (Institut de Recherche sur la Fusion Magnétique) which operates Tore Supra. „By installing an ITER-like full tungsten divertor in Tore Supra, we can turn our platform into a test-bench on ITER critical path. We can thus contribute to reducing the risk and to saving time and money for ITER. WEST is not something we would add to Tore Supra like we did with CIEL. It’s more like Tore Supra becomes WEST to serve ITER.”

The CIEL project provided IRFM with a strong experience in cooperating with the industry. Adapting Tore Supra to accommodate a full tungsten divertor — 500 components with a total of 15,000 tungsten tiles — is a challenge the Institute is ready to take on. (All carbon will have to be taken out of the device; in-vacuum vessel magnetic coils will need to be installed in order to modify the plasma shape from circular to „D-shaped” and heating systems will have to be adapted to the new configuration.)

The formal decision to go WEST is due to be taken by CEA at the end of 2012; Bécoulet is optimistic: partners are showing interest and „customers” other than ITER appear eager to utilize the future test bench as well. „All fusion machines, present and projected,” he says „are expected to go tungsten.”

Bringing a timely answer to ITER interrogations means that Tore Supra, which Bécoulet calls „a technological jewel”, should prepare to go WEST early in 2013 and be ready for the first experiments in 2015.

Click here to view an animation of the WEST project.

One more step towards the final green light

On 29 July, a new milestone was reached in the licensing process of ITER. A little more than one month after being notified that our proposals on the Tokamak’s operational conditions and design fulfilled the French safety requirements, we have now received from the Autorité de Sûreté Nucléaire (ASN)  the draft of the Décret d’Autorisation de Création — the final green light from the French Authorities to create our installation.
We are currently analyzing this draft and we will soon send back our comments to ASN. Then, a discussion will be organized with a college of ASN experts and at long last the final decree will be published — hopefully before the end of the year.

This is a lengthy, complex, demanding — sometimes frustrating… — process. But I must say it is also a very good process. ITER is the first fusion installation that will receive a full nuclear licence. And this is very important, not only for us here at ITER but for the whole worldwide fusion community.

We have always claimed that fusion is safe and in the past two years, we went through an exceptionally strict and challenging process to demonstrate that it is indeed. Now an independent body of experts, with a deserved reputation for being among the „toughest” in the world, is in the process of validating our claim. And again, this is a first: no fusion installation, not even JET or TFTR which, at one point implemented deuterium + tritium fusion, went through this process.

Twenty-seven years have passed since President Reagan and Secretary Gorbatchev met in Geneva and laid the ground for the project of an international experimental fusion reactor „for the benefit of all mankind”.

We all feel a deep satisfaction in seeing these 27 years of hard work and dedication now converging into a decision that, in many ways, is historical.

Common controls in ITER and IFMIF

On the 20th and 21st of August several meetings took place at Rokkasho (Japan) between the CODAC  teams in charge of the machine protection and interlocks of ITER and the International Fusion Materials Irradiation Facility (IFMIF) team.

IFMIF is one of the projects of the Broader Approach Agreement between Japan and Europe, which was signed to support ITER and achieve an early realization of Fusion Energy for peaceful purpose. In particular, IFMIF must present results in parallel with ITER operation since these will allow the design of DEMO by qualifying the materials capable to withstand the neutron flux that a commercial nuclear nusion reactor will undergo.

The aim of the meetings was to establish a first contact between the controls groups of both „brother” organisations focusing on the development of the machine protection systems. The sessions started with a seminar by Antonio Vergara (ITER) summarising his experience on the design, implementation and commissioning of machine protection systems for high energy physics accelerators like the Large Hadron Collider at CERN and how the lessons learnt can be applied to ITER and IFMIF interlocks. The presentation was followed by a  series of meetings organised by the IFMIF/ Engineering Validation and Engineering Design Activities (EVEDA) Project Leader,  Juan Knaster.

IFMIF plant will bombard suitable materials reaching more than 20 displacements per atom (dpa)/year (this value means that in average an atom has been displaced from its lattice 20 times per year). This would allow to obtain within a few years of operation the expected 150 dpa at the end of life of a commercial reactor; and with neutrons at an energy spectrum around the 14 MeV (typical of a Deuterium-Tritium nuclear fusion).

The neutron flux will be obtained by accelerating at 40 MeV two parallel beams of 125 mA Deuteron current and make them collide onto a Liquid Lithium screen. The Accelerator validation will be achieved by the installation, commissioning and operation of the Linear IFMIF Accelerator Prototype ( LIPAc) which will accelerate a current of 125 mA Deuterons at 9 MeV. The current status of the LIPAc control, safety and machine protection systems were presented and discussed.

The LIPAc, like ITER, is also based on in-kind procurements. The collaborating organizations in Japan and Europe are in charge of building and installing the different plant systems of the accelerator’s prototype including their local controls, safety and machine protection systems. The international team in Rokkasho is in charge of the development of the central control systems and the entire integration and commissioning.

Not surprisingly, they are facing many of the issues and challenges related to the integration of the I&C systems that the CODAC team at ITER has been solving during the last years by the development of tools such as the Plant Control Design Handbook (a new version will be released  at the beginning of 2013) and the CODAC Core System software.

One of the main meeting conclusions was that the similarities between the two project control systems – the fact that both are based on EPICS and the equivalent procurement strategy – makes a more detailed analysis of the potential collaboration between ITER and IFMIF very desirable.  The potential to share the developed tools, and procedures and apply  knowledge and lessons learnt from the ITER controls and interlocks teams to the design and implementation of the LIPAc control systems could  result in an efficient and cost-effective collaborative approach.