Butt Joint Tool Commissioning

ITER Central Solenoid uses butt joints for connecting the pancakes in the CS module. The principles of the butt joining of the CICC were developed by the JAPT during CSMC project. The difference between the CSMC butt joint and the CS butt joint is that the CS butt joint is an in-line joint, while the CSMC is a double joint through a hairpin jumper. The CS butt joint has to carry the hoop load. The straight length of the joint is only 320 mm, and the vacuum chamber around the joint has to have a split in the clamp shell. These requirements are challenging. Fig.1 presents a CSMC joint, and Fig.2 shows a CS butt joint. The butt joint procedure was verified and demonstrated. The tool is capable of achieving all specified parameters. The vacuum in the end was a little higher than the target, which is not critical and readily correctable. We consider, tentatively that the procedure is established. Unexpectedly, we discover significant temperature nonuniformity in the joint cross section, which is not formally a violation of the specs, but is a point of concern. All testing parameters are recorded for QA purposes. We plan to modify the butt joining tool to improve its convenience of operation and provide all features necessary for production of butt joints by qualified personnel

Butt Joint Tool Status: ITER-US-LLNL-NMARTOVETSKY-01312007

Butt joint tool vacuum vessel has been built at C&H Enterprise, Inc. Leak checking and loading tests were taken place at the factory. The conductor could not be pumped down better than to 500 mtorr and therefore we could not check the sealing mechanism of the seal around conductor. But the rest of the vessel, including the flat gasket, one of the difficult seals worked well, no indication of leak at sensitivity 1e-7 l*torr/sec. The load test showed fully functional system of the load mechanism. The conductors were loaded up to 2200 kgf (21560 N) and the pressure between the butts was uniform with 100% of the contact proved by pressure sensitive film. The status of the butt joint tool development is reported

Testing Large CICC in Short Sample Configuration and Predicting Their Performance in Large Magnets

It is well known that large Nb3Sn Cable-in-Conduit Conductors (CICC) do not always completely utilize current carrying capacity of the strands they are made of. The modern state of theory is not accurate enough to eliminate CICC full scale testing. Measuring properties of large CICC is not a simple task due to variety of parameters that need to be controlled, like temperature, exposure of all the strands to the peak magnetic field, mass flow and particular nonuniform current distribution. The paper presents some measurement issues of CICC testing in a short sample test facility, particularly, conditions for uniform current distribution and effect of twist pitches on the critical current

SULTAN measurement and qualification: ITER-US-LLNL-NMARTOVETSKY- 092008

Measuring the characteristics of full scale ITER CICC at SULTAN is the critical qualification test. If volt-ampere characteristic (VAC) or volt-temperature characteristic (VTC) are distorted, the criterion of 10 uV/m may not be a valid criterion to judge the conductor performance. Only measurements with a clear absence or low signals from the current distribution should be considered as quantitatively representative, although in some obvious circumstances one can judge if a conductor will meet or fail ITER requirements. SULTAN full scale ITER CICC testing should be done with all measures taken to ensure uniform current redistribution. A full removal of Cr plating in the joint area and complete solder filling of the joints (with provision of the central channel for helium flow) should be mandatory for DC qualification samples for ITER. Also, T and I should be increased slowly that an equilibrium could be established for accurate measurement of Tcs, Ic and N. It is also desirable to go up in down in current and/or temperature (within stable range) to make sure that the equilibrium is reached

Effect of Twist Pitch in the Strands on the Saturation and Losses in the Nb3Sn Strands for the ITER TF CICC

ITER TF coils will see a significant longitudinal magnetic field in the event of the plasma disruption. This abrupt change of magnetic fields results in the appearance of an additional electrical field in the strands. The mechanism of this electrical field is the induced currents that expel the flux from the strands. This effect was known since the late 1970's [1-3] and most of the details necessary for the analyses given in this report are presented in [4]. Let's assume for simplicity a zero transport current in the strand. When a longitudinal pulsed field is applied, the outer filaments will carry an induced current repelling the change of flux. The current density of this current is 'critical' in the simplification of Bean's critical state model, where superconducting transition is represented as j=j{sub c} at any non-zero electrical field and zero where the electrical field has not penetrated. In reality, since the current density is roughly logarithmic with the electrical field, E=E{sub c}*exp[(j-j{sub c})/jo], Bean's model is just a simplification, and current density is slightly nonuniform in the outer filament and more so for the interior strands. The inner portion of the filaments will carry a current of the opposite sign. Even in the Bean's model it is not uniform, but the assumption that it is uniform and less than critical simplifies mathematics significantly and does not deviate far from the real current density distribution. In certain circumstances, the average electrical field in the strands will be high enough to exceed the take-off electrical field averaged across the cross section. In this case, the multifilamentary strand will become unstable and will experience transition to the normal state. With zero transport current, it will eventually recover, of course. This phenomenon is analogous to the flux jump. If the strand carries a transport current, the situation becomes more complicated. If it goes unstable and the transport current is higher than the cryostability limit (by Stekly), or if there are enough losses to bring the temperature above the current sharing temperature taking into account limited heat capacity of the CICC, the strand will not recover, and the CICC will go normal. Conservatively, we will consider that if we find an instantaneous unstable situation, it is not acceptable. In presence of a transport current, the situation is sensitive to the direction of the strand twist, direction of the pulsed field and direction of the transport current. Recently, ITER decided to increase the twist pitch of the TF strands from 15 mm to 30 mm to improve the stability of the strands against the longitudinal field. In this report we will quantify the effects of this proposed change and perform a trade off study. The issue is that by increasing the twist pitch of the strands we not only increase the coupling losses in the transverse magnetic field, as expected in classical multifilamentary composite superconductors, but also increase the hysteresis losses in the strands with internal tin. In classical multifilamentary superconductors, twist pitch change should not cause an increase of the hysteresis losses in the transverse field. However the high Nb3Sn content internal tin strands develop transverse links, which couple the filaments into clusters. These links turn out to contribute a significant fraction to hysteresis losses [5]. If we project the results of [5] onto the ITER proposal to increase the twist pitch from 15 to 30 mm, we should expect the hysteresis losses to increase by a factor of two, which will likely disqualify strands with 30 mm twist pitch. This very strand twisted to 15 mm twist pitch would likely pass the ITER criteria. So, increasing the twist pitch has a very negative consequence and we need to make sure that it is absolutely necessary. Recently, A. Vostner (private communication) reported preliminary results on the losses in candidate TF strands. In agreement with what was reported in [5]; he found that TF strands with 15 mm twist pitch have hysteresis losses about half of what the strands with 30 mm twist pitch have

Summary Report-fy2006 Iter Work Accomplished

Six parties (EU, Japan, Russia, US, Korea, China) will build ITER. The US proposed to deliver at least 4 out of 7 modules of the Central Solenoid. Phillip Michael (MIT) and I were tasked by DoE to assist ITER in development of the ITER CS and other magnet systems. We work to help Magnets and Structure division headed by Neil Mitchell. During this visit I worked on the selected items of the CS design and carried out other small tasks, like PF temperature margin assessment

Development of the Butt Joint for the ITER Central Solenoid

The ITER Central Solenoid (CS) requires compact and reliable joints for its Cable-in-Conduit Conductor (CICC). The baseline design is a diffusion bonded butt joint. In such a joint the mating cables are compacted to a very low void fraction in a copper sleeve and then heat treated. After the heat treatment the ends are cut, polished and aligned against each other and then diffusion bonded under high compression in a vacuum chamber at 750 C. The jacket is then welded on the conductor to complete the joint, which remarkably does not require more room than a regular conductor. This joint design is based on a proven concept developed for the ITER CS Model Coil that was successfully tested in the previous R&D phase

Testing Short Samples of ITER Conductors and Projection of Their Performance in ITER Magnets

Qualification of the ITER conductor is absolutely necessary. Testing large scale conductors is expensive and time consuming. To test straight 3-4m long samples in a bore of a split solenoid is a relatively economical way in comparison with fabrication of a coil to be tested in a bore of a background field solenoid. However, testing short sample may give ambiguous results due to different constraints in current redistribution in the cable or other end effects which are not present in the large magnet. This paper discusses processes taking place in the ITER conductor, conditions when conductor performance could be distorted and possible signal processing to deduce behavior of ITER conductors in ITER magnets from the test data

Central Solenoid Insert Technical Specification

The US ITER Project Office (USIPO) is responsible for the ITER central solenoid (CS) contribution to the ITER project. The Central Solenoid Insert (CSI) project will allow ITER validation the appropriate lengths of the conductors to be used in the full-scale CS coils under relevant conditions. The ITER Program plans to build and test a CSI to verify the performance of the CS conductor. The CSI is a one-layer solenoid with an inner diameter of 1.48 m and a height of 4.45 m between electric terminal ends. The coil weight with the terminals is approximately 820 kg without insulation. The major goal of the CSI is to measure the temperature margin of the CS under the ITER direct current (DC) operating conditions, including determining sensitivity to load cycles. Performance of the joints, ramp rate sensitivity, and stability against thermal or electromagnetic disturbances, electrical insulation, losses, and instrumentation are addressed separately and therefore are not major goals in this project. However, losses and joint performance will be tested during the CSI testing campaign. The USIPO will build the CSI that will be tested at the Central Solenoid Model Coil (CSMC) Test Facility at the Japan Atomic Energy Agency (JAEA), Naka, Japan. The industrial vendors (the Suppliers) will report to the USIPO (the Company). All approvals to proceed will be issued by the Company, which in some cases, as specified in this document, will also require the approval of the ITER Organization. Responsibilities and obligations will be covered by respective contracts between the USIPO, called Company interchangeably, and the industrial Prime Contractors, called Suppliers. Different stages of work may be performed by more than one Prime Contractor, as described in this specification. Technical requirements of the contract between the Company and the Prime Contractor will be covered by the Fabrication Specifications developed by the Prime Contractor based on this document and approved by the Company and ITER. The Fabrication Specifications may reflect some national requirements and regulations that are not fully provided here. This document presents the ITER CSI specifications

Optimization of a 3x3 focusing array for heavy ion drivers

A heavy ion driver for inertial fusion will accelerate an array of beams through common induction cores and then direct the beams onto the DT target. An array of quadrupole focusing magnets is used to prevent beam expansion from space charge forces. In the array, the magnet fields from the coils embracing the beams are coupled, which reduces the cost of superconductor and increases the focusing power. The challenges in designing such an array are meeting the strict requirements for the quadrupole field inside the beam pipes and preventing stray fields outside. We report our optimization effort on designing such an array and show that 3 x 3 or larger arrays are feasible and practical to build with flat racetrack coils