The operational results of the KSTAR PF cryo-circuit

The Koreas Superconducting Tokamak Advanced Research (KSTAR) PF cryo-circuit is designed for cooling the fourteen superconducting magnets (Nb3Sn, NbTi) and structures. Those are cooled down by the supercritical helium (4.5 K and 5.5 bar) of a forced flow (pressure gradient: 2 bar) in order to maintain the supercritical state of the helium. To supply a large amount of supercritical helium (>370 g/s), a circulator was inserted into the PF cryo-circuit. The compressed supercritical helium is distributed to five helium manifolds with cryogenic valves and supplied to each PF magnet. While the PF magnets had been operating, the mass flow rate reduced and the pressure head of the circulator was fluctuated depending on the PF magnet operation scenario. These phenomena could damage the circulator and could stop it during operation. Therefore, by-pass valve, which is parallel with in-line valve and is connected with inlet and outlet of the magnet, was opened in order to reduce of the circulator's pressure head. In this paper, we focused on the helium behavior of the superconducting magnet when the by-pass valve was opened in order to release the pressure head of the circulator and the results will be presented.     

Commissioning and initial operation of KSTAR superconducting tokamak

The commissioning and the initial operation for the first plasma in the KSTAR device have been accomplished successfully without any severe failure preventing the device operation and plasma experiments. The commissioning is classified into four steps: vacuum commissioning, cryogenic cool-down commissioning, magnet system commissioning, and plasma discharge.Vacuum commissioning commenced after completion of the tokamak and basic ancillary systems construction. Base pressure of the vacuum vessel was about 3 × 10^?6 Pa and that of the cryostat about 2.7 × 10^?4 Pa, and both levels meet the KSTAR requirements to start the cool-down operation. All the SC magnets were cooled down by a 9 kW rated cryogenic helium facility and reached the base temperature of 4.5 K in a month. The performance test of the superconducting magnet showed that the joint resistances were below 3 nΩ and the resistance to ground after cool-down was over 1 GΩ. An ac loss test of each PF coil made by applying a dc biased sinusoidal current showed that the coupling loss was within the KSTAR requirement with the coupling loss time constant less than 35 ms for both Nb₃Sn and NbTi magnets. All the superconducting magnets operated in stable without quench for long-time dc operation and with synchronized pulse operation by the plasma control system (PCS). By using an 84 GHz ECH system, second harmonic ECH assisted plasma discharges were produced successfully with loop voltage of less than 3 V. By the real-time feedback control, operation of 100 kA plasma current with pulse length up to 865 ms was achieved, which also meet the first plasma target of 100 kA and 100 ms. The KSTAR device will be operated to meet the missions of steady-state and high-beta achievement by system upgrades and collaborative researches.     

Thermal analysis of toroidal field coil in EAST at 3.7 K

The thermal performance of toroidal field (TF) coil is studied at 3.7 K in Experimental Advanced Superconducting Tokamak device (EAST) to obtain the higher stability for the higher plasma parameters operation. It is a good way to lower the operating temperature of TF coil to acquire the higher stability margin. This paper describes the structure and cooling process design of TF coil and case firstly. Based on the thermal load in the case, the thermal performance of the TF coil is performed at the plasma disruption state. The helium temperature in the cable-in-conduit conductor (CICC) and case is evaluated during the 1.5 MA plasma disruptions. Then, the experimental results of TF coil which has been cooled at 3.7 K and discharged in 10 kA are shown including the thermal loss evaluation. Finally, the thermal stability performance of TF coil is analyzed according to the 3.7 K experimental results and the stability prediction is performed at 1.5 MA plasma current operations.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

Progress of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO [1] has been developed at Karlsruhe Institute of Technology (KIT) since a couple of years with the goal of reaching a heat flux of 10 MW/m² anticipated for DEMO. The reference concept HEMJ (He-cooled modular divertor with multiple-jet cooling) is based on the use of small cooling fingers – each composed of a tungsten tile brazed to a tungsten alloy thimble – as well as on impingement jet cooling with helium at 10 MPa, 600 °C. The cooling fingers are connected to the main structure of ODS Eurofer steel by brazing in combination with a mechanical interlock. This paper reports progress to date of the design accompanying R&Ds, i.e. primarily the fabrication technology and HHF experiments. For the latter a combined helium loop and electron beam facility (200 kW, 40 keV) at Efremov Institute, St. Petersburg, Russia, has been used. This facility enables mock-up testing at a nominal helium inlet temperature of 600 °C, a pressure of 10 MPa, and a maximal pressure head of 0.5 MPa. HHF test results till now confirm well the divertor design performance. In the recent test series in early 2010 the first breakthrough was achieved when a mock-up has survived over 1000 cycles at 10 MW/m² unscathed.     

Start-up phase of the HELOKA-LP low pressure helium test facility for IFMIF irradiation modules

As part of the Engineering Validation and Engineering Design Activities (EVEDA) for the International Fusion Materials Irradiation Facility (IFMIF) [1], it is foreseen to design and test a 1:1 scale prototype of the IFMIF High Flux Test Module (HFTM) [2]. The module has been designed to be cooled by a low pressure helium gas flowing through minichannels to remove the nuclear heat.The Helium Loop Karlsruhe-Low Pressure (HELOKA-LP) has been designed to provide coolant at 1:1 HFTM operational conditions: massflow 12–120 g/s, inlet pressure 0.3–0.6 MPa, inlet temperature RT – 250 °C. A secondary objective is to use the experience gained with HELOKA-LP for the planning of the IFMIF helium cooling system.The facility has been put into operation in 2009, and has since then been in a test and optimization phase. It was proven, that the above mentioned requirements for the facility are achieved. The paper describes the process layout and components of the facility. The performance is characterized by the results of several steady state and transient benchmark tests. Typical start-up and transition times relevant for the operation mode in the IFMIF irradiation campaigns are obtained. Additionally first results on the impurity ingress and the cooling gas chemistry are described.     

Deuterium and helium retentions of V–4Cr–4Ti alloy used as first wall of breeding blanket in a fusion reactor

The deuterium and helium retention properties of V–4Cr–4Ti alloy were investigated by thermal desorption spectroscopy (TDS). Ion energies of deuterium and helium were taken at 1.7 and 5 keV, respectively. The retained amount of deuterium in the sample irradiated at 380 K increased with the ion fluence and was not saturated to fluence of up to 1 × 10²³ D/m². For the irradiation at 773 K, 0.1% of implanted deuterium was retained at the highest fluence. For the helium ion irradiation at room temperature, three groups of desorption peaks appeared at around 500, 850, and 1200 K in the TDS spectrum. In the lower fluence region (<1 × 10²¹ He/m²), the retained helium desorbed mainly at around 1200 K. With increasing fluence, the amount desorbed at 500 K increased. Total amount of retained helium in the samples saturated at fluence up to 5 × 10²¹ He/m² and saturation level was 2.7 × 10²¹ He/m².     

Construction and initial operation of MHD PbLi facility at UCLA

A magnetohydrodynamic flow facility MaPLE (Magnetohydrodynamic PbLi Experiment) that utilizes molten eutectic alloy lead–lithium (PbLi) as working fluid has been constructed and tested at University of California, Los Angeles. The loop operation parameters are: maximum magnetic field 1.8 T, PbLi temperature up to 350 °C, maximum PbLi flow rate with/without a magnetic field 15/50 l/min, maximum pressure head 0.15 MPa. The paper describes the loop itself and its major components, basic operation procedures, experience of handling PbLi, initial loop testing, flow diagnostics and current and near-future experiments. The obtained test results of the loop and its components have demonstrated that the new facility is fully functioning and ready for experimental studies of magnetohydrodynamic, heat and mass transfer phenomena in PbLi flows and also can be used in mock up testing in conditions relevant to fusion applications.     

Experimental study and analysis of the purge gas pressure drop across the pebble beds for the fusion HCPB blanket

The lithium ceramic and beryllium pebble beds of the breeder units (BU), in the fusion breeding blanket, are purged by helium to extract the bred tritium. Therefore, the objective of this study is to support the design of the BU purge gas system by studying the effect of pebbles diameter, packing factor, pebble bed length, and flow inlet pressure on the purge gas pressure drop. The pebble bed was formed by packing glass pebbles in a rectangular container (56 mm × 206 mm × 396 mm) and was integrated into a gas loop to be purged by helium at BU-relevant pressures (1.1–3.8 bar). To determine the pressure drop across the pebble bed, the static pressure was measured at four locations along the pebble bed as well as at the inlet and outlet locations. The results show: (i) the pressure drop significantly increases with decreasing the pebbles diameter and slightly increases with increasing the packing factor, (ii) for a constant inlet flow velocity, the pressure drop is directly proportional to the pebble bed length and inlet pressure, and (iii) predictions of Ergun's equation agree well with the experimental values of the pressure drop.     

Conceptual design of a helium heater for high temperature applications

The Karlsruhe Advanced Technologies Helium Loop (KATHELO) has been designed for testing divertor modules as well as qualifying materials for high heat flux, high temperature (up to 800 °C) and high pressure (10 MPa) applications. The test section inlet temperature level is controlled using a process electrical heater. To cope with the extreme operating conditions, a special design of this unit has been proposed. In this paper the conceptual design of the unit will be presented and the impact of the coupling between the cold and hot helium gas on the overall efficiency of the loop will be investigated. The detailed thermal-hydraulic analysis of the feed through of the hot helium into the low temperature pressure vessel using ANSYS CFX will be presented. The impact of the design choices on the overall energy budget of the loop will be analyzed using RELAP5-3D.     

Vacancy-type defects and hardness of helium implanted CLAM steel studied by positron-annihilation spectroscopy and nano-indentation technique

China Low Activation Martensitic (CLAM) steel was implanted with helium up to 1e + 16/cm² at 300–873 K using 140 keV helium ions. Vacancy-type defects induced by implantation were investigated with positron beam Doppler broadening technique, and then nano-hardness measurements were performed to investigate helium-induced hardening effect. He implantation produced a large number of vacancy-type defects in CLAM steel, and the concentration of vacancy-type defects decreased with increasing temperature. Vacancy–helium complexes were main defects at different temperatures. Irradiation induced hardening was observed at all irradiation temperatures, and the peak value of hardness was at 473 K. The result suggested that both vacancy–helium complexes and helium bubbles had contribution to irradiation induced hardening. The decomposition and annihilation of irradiation-induced defects became more and more significant with increasing temperature, which induced the increment of hardness became more and more small.     

Experimental investigation on cryogenic hydrogen adsorption of molecular sieves

Tritium extraction system (TES) is one of the most important components in the helium cooled solid breeder test blanket modules (TBMs) of ITER. TES will extract various isotopic species of hydrogen by the liquid nitrogen cooled molecular sieve adsorber beds (MSB). The cryogenic hydrogen adsorption properties of several kinds of molecular sieves have been investigated at the pressure of hydrogen of 100 Pa, 200 Pa, and 0.2 MPa in order to offer the suitable molecular sieve for the MSB in TES. The saturated hydrogen adsorption capacities of the MS5A-2 and MS13X-2 have been measured at 100 Pa hydrogen pressure. To demonstrate the hydrogen extraction from continuous He–H₂ purge gases, the MS5A-2 has been tested in circulating 99.79% He–0.21% H₂ mixture with a flow rate of 16.8 L/min. The results show that the globular MS5A-2 with a diameter of 3–5 mm can adsorb/desorb hydrogen quickly. The saturated hydrogen adsorption capacity of MS5A-2 is 7.55 ml g^?1 (NTP) and MS5A-2 could effectively extract trace hydrogen from mixture gases. As a result, this type of molecular sieve can be the candidate of the one in the MSB in ITER TBM.     

Preparation and investigation of aluminized coating and subsequent heat treatment on 9Cr–1Mo Grade 91 steel

Iron aluminide inner coating with alumina top layer is being considered as a potential solution for tritium permeation barrier and mitigating MHD pressure drop for liquid metal blanket concepts in the fusion reactor systems. Hot-dip aluminizing with subsequent heat treatment seems to offer a good possibility to produce aluminized coating with alumina top layer. 9Cr–1Mo Grade 91 steel samples were hot dipped in Al melt containing 2.25 wt% of Si at 750 °C for 3 min. Heat treatment was performed at 650, 750 and 950 °C for 5 h; samples were either air cooled or furnace cooled. Coatings have been evaluated by SEM, EDX, X-ray diffraction, microhardness, scratch adhesion and Raman spectroscopy. The thickness of the layers and phases formed were influenced by the heat treatment adopted. Fe₂Al₅ was the major phase present in the samples heat treated at 650/750 °C, whereas FeAl and α-Fe(Al) primarily made up the outer and inner layers respectively in the samples heat treated at 950 °C. Cooling method deployed affected the hardness. Air cooled samples had comparatively higher hardness than furnace cooled samples. The scratch test showed the adhesion for the samples heat treated at 950 °C was much better as compared to the samples heat treated at 650/750 °C. Raman spectroscopy analysis showed the presence of both α-Al₂O₃ and γ-Al₂O₃ on the surface of the samples heat treated at 950 °C, while Fe₃O₄ was present in the furnace cooled sample only.     

Updating the design of the feeder components for the ITER magnet system

The ITER superconducting magnet system generates an average heat load of 23 kW at 4 K to the cryoplant, from nuclear and thermal radiation, conduction and electromagnetic heating, and requires current supplies 10–68 kA to 48 individual coils. The helium flow to remove this heat, consisting of supercritical helium at pressures up to 1.0 MPa and temperature between 4.3 and 4.7 K, is distributed to the coils and structures through 30 separate feeder lines. The feeders also contain the electrical supplies to the coil, helium supply pipes and the instrumentation lines, and are integrated with the current lead transitions to room temperature. The components consist of the in-cryostat feeders, the cryostat feedthroughs and the coil terminal boxes (CTBs). This paper discusses the functional requirements on the feeder system and presents the latest design concept and parameters of the feeder components.     

Design and test program of a simplified divertor dummy coil structure for the WEST project

In order to fully validate actively cooled tungsten plasma facing components (industrial fabrication, operation with long plasma duration), the implementation of a tungsten axisymmetric divertor structure in the tokamak Tore-Supra is studied. With this major upgrade, so-called WEST (Tungsten Environment in Steady state Tokamak), Tore-Supra will be able to address the problematic of long plasma discharges with a metallic divertor target.To do so, it is planned to install two symmetric divertor coils inside the vacuum vessel. This assembly, called divertor structure, is made up of two stainless steel casings containing a copper winding pack cooled by a pressurized hot water circuit (up to 180 °C, 4 MPa) and is designed to perform steady state plasma operation (up to 1000 s).The divertor structure will be a complex assembly ring of 4 m diameter representing a total weight of around 20 tons. The technical challenge of this component will be the implementation of angular sectors inside the vacuum vessel environment (TIG welding of the coil casing, induction brazing and electrical insulation of the copper winding). Moreover, this complex assembly must sustain harsh environmental conditions in terms of ultra high vacuum conditions, electromagnetical loads and electrical isolation (13 kV ground voltage) under high temperature.In order to fully validate the assembly and the performance of this complex component, the production of a scale one dummy coil is in progress.The paper will illustrate, the technical developments performed in order to finalize the design for the call for tender for fabrication. The progress and the first results of the simplified dummy coils will be also addressed.     

The procurement and testing of the stainless steel in-vessel panels of the Wendelstein 7-X Stellarator

320 In-vessel water cooled stainless steel panels, poloidal closure plates and pumping gap panels, covering an area of approximately 100 m², are used in Wendelstein7-X to protect the plasma vessel. The panels are manufactured at Deggendorf, Germany by MAN Diesel & Turbo SE. The panels consist of a laser welded sandwich of stainless steel plates together with a labyrinth of cooling channels and have a complicated geometry to fit the plasma vessel of Wendelstein 7-X. The hydraulic and mechanical stability requirements whilst maintaining the tight tolerances for the shape of the components are very demanding. The panels are designed to operate at up to an average heat load of 100 kW/m² and a maximum heat load of 200 kW/m² with a water velocity of approximately 2 m s^?1. High heat flux testing of an un-cooled panel at a time averaged load of 200 kW/m² for 10 s were successfully performed to support the start up phase of Wendelstein 7-X operation. Extensive testing both during manufacture and after delivery to IPP-Garching demonstrates the suitability of the delivered panels for their purpose.     

Helium-implanted CLAM steel and evolutionary behavior of defects investigated by positron-annihilation spectroscopy

China Low Activation Martensitic (CLAM) steel has been chosen as the primary candidate structural material for the first wall/blanket for fusion reactor. The excessive helium irradiation induced damage of CLAM steel at high temperatures and the evolution of defects were investigated in this paper. The samples were homogeneously implanted with 1e + 17 ions/cm² and 100 keV of helium at room temperature, 473, 673, and 873 K. Irradiation induced damage of CLAM steel and the annealing behavior of defects were probed by slow positron beam Doppler broadening technique. Helium implantation produced a large number of vacancy-type defects which bound with helium and formed helium–vacancy complexes. Target atoms’ displacement capacity was strengthened with rising irradiation temperatures, so the S parameter increased with increasing irradiation temperatures, and helium–vacancy complexes were main defects after helium implantation at damage layers. Helium bubbles would be unstable and the desorption of helium bubbles would promote the density of defects above 673 K. By analyzing the curves of S–W and annealing tests of irradiated specimen, it suggested that there werenot only one type of defect in damage layers. Though helium–vacancy complexes were primary defects after helium implanted, introducing excessive helium might also generated other point defects or dislocation loops in the material.     

Evaluation of soft error rates using nuclear probes in bulk and SOI SRAMs with a technology node of 90 nm

The difference of soft error rates (SERs) in conventional bulk Si and silicon-on-insulator (SOI) static random access memories (SRAMs) with a technology node of 90 nm has been investigated by helium ion probes with energies ranging from 0.8 to 6.0 MeV and a dose of 75 ions/μm². The SERs in the SOI SRAM were also investigated by oxygen ion probes with energies ranging from 9.0 to 18.0 MeV and doses of 0.14–0.76 ions/μm². The soft error in the bulk and SOI SRAMs occurred by helium ion irradiation with energies at and above 1.95 and 2.10 MeV, respectively. The SER in the bulk SRAM saturated with ion energies at and above 2.5 MeV. The SER in the SOI SRAM became the highest by helium ion irradiation at 2.5 MeV and drastically decreased with increasing the ion energies above 2.5 MeV, in which helium ions at this energy range generated the maximum amount of excess charge carriers in a SOI body. The soft errors occurred by helium ions were induced by a floating body effect due to generated excess charge carriers in the channel regions. The soft error occurred by oxygen ion irradiation with energies at and above 10.5 MeV in the SOI SRAM. The SER in the SOI SRAM gradually increased with energies from 10.5 to 13.5 MeV and saturated at 18 MeV, in which the amount of charge carriers induced by oxygen ions in this energy range gradually increased. The computer calculation indicated that the oxygen ions with energies above 13.0 MeV generated more excess charge carriers than the critical charge of the 90 nm node SOI SRAM with the designed over-layer thickness. The soft errors, occurred by oxygen ions with energies at and below 12.5 MeV, were induced by a floating body effect due to the generated excess charge carriers in the channel regions and those with energies at and above 13.0 MeV were induced by both the floating body effect and generated excess carriers. The difference of the threshold energy of the oxygen ions between the experiment and the computer calculation might be due to the difference between the designed and real structures.     

Towards assembly completion and preparation of experimental campaigns of Wendelstein 7-X in the perspective of a path to a stellarator fusion power plant

The superconducting stellarator device Wendelstein 7-X, currently under construction, is the key device for the proof of stellarator optimization principles. To establish the optimized stellarator as a serious candidate for a fusion reactor, reactor-relevant dimensionless plasma parameters must be achieved in fully integrated steady-state scenarios. After more than 10 years of construction time, the completion of the device is now approaching rapidly (mid-2014). We discuss the most important lessons learned during the device assembly and first experiences with coming major work packages. Those are (a) assembly of about 2500 large, water-cooled, 3d-shaped in-vessel component elements; (b) assembly of in total 14 superconducting current leads, one pair for each coil type; and (c) assembly of the device periphery including diagnostics and heating systems. In the second part we report on the present status of planning for the first operation phase (5–10 s discharge duration at 8 MW heating power), the completion and hardening of the device for full power steady-state operation, and the second operation phase (up to 30 min discharge duration at 10 MW heating power). It is the ultimate goal of operation phase one to develop credible and robust discharge scenarios for the high-power steady-state operation phase two. Beyond the improved equilibrium, confinement, and stability properties owing to stellarator optimization, this requires density control, impurity control, edge iota control as well as high density microwave heating. Of paramount importance is the operation of the island divertor, which is realized in the first operation phase as an inertially cooled conventional graphite target divertor. It will be replaced later on by the steady-state capable island divertor with its water-cooled carbon fiber reinforced carbon target elements.     

Tritium management and anti-permeation strategies for three different breeding blanket options foreseen for the European Power Plant Physics and Technology Demonstration reactor study

In DT fusion reactors like DEMO, the commonly accepted tritium (T) losses through the steam generator (SG) shall not exceed about 2 mg/d that are more than 5 orders of magnitude lower than the T production rate of about 360 g/d in the breeding blanket (BB). A very effective mitigation strategy is required balancing the size and efficiency of the processes in the breeding and cooling loops, and the availability and efficiency of anti-permeation barriers. A numerical study is presented using the T permeation code FUS-TPC that computes all T flows and inventories considering the design and operation of the BB, the SG, and the T systems. Many scenarios are numerically analyzed for three breeding blankets concepts – helium cooled pebbles bed (HCPB), helium cooled lithium lead (HCLL), and water cooled lithium lead (WCLL) – varying the T processes throughput and efficiency, and the permeation regimes through the BB and SG to be either surface-limited or diffusion-limited with possible permeation reduction factor. For each BB concept, we discuss workable operation scenarios and suggest specific anti-permeation strategies.