Design, fabrication, and testing of a helium-cooled module for the ITER divertor

The International Thermonuclear Reactor (ITER) will have a single-null divertor with total power flow of 200 MW and a peak heat flux of about 5 MW/m{sup 2}. The reference coolant for the divertor is water. However, helium is a viable alternative and offers advantages from safety considerations, such as excellent radiation stability and chemical inertness. In order to prove the feasibility of helium cooling at ITER relevant heat flux conditions, General Atomics designed, fabricated, and tested a helium-cooled divertor module. The module was made from dispersion strengthened copper, with a heat flux surface 25 mm wide and 80 mm long, designed for twice the ITER divertor heat flux. Different techniques were examined to enhance the heat transfer, which in turn reduced the flow and pumping power required to cool the module. It was concluded that an extended surface was the most practical solution. An optimization study was performed to find the best extended surface parameters. The optimum extended surface geometry consisted of fins: 10 mm high, 0.4 mm thick with a 1 mm pitch. It was estimated to require a pumping power of 150 W to remove 20 kW of power. This is more than an order of magnitude reduction in pumping power requirement, compared to smooth surface. The module was fabricated by electric discharge machining (EDM) process. The testing was carried out at SNLA during August 1993. The testing confirmed the design calculations. The peak heat flux during the test was 10 MW/m{sup 2} applied over a surface area of 20 cm{sup 2}. The pumping power calculated from flow rate and pressure drop measurement was about 160 W, which was less than 1% of the power removed. It is planned to test the module to higher temperature limits and higher heat fluxes during coming months. As a result of this effort we conclude that helium cooling of the ITER divertor is feasible without requiring a very large helium pressure or a large pumping power

Engineering design of a radiative divertor for DIII-D

A new divertor configuration is being developed for the DIII-D tokamak. This divertor will operate in the radiative mode. Experiments and modeling form the basis for the new design. The Radiative Divertor reduces the heat flux on the divertor plates by dispersing the power with radiation in the divertor region. In addition, the Radiative Divertor structure will allow density control in plasma shapes required for advanced tokamak operation. The divertor structure allows for operation in either double-null or single-null plasma configurations. Four independently controlled divertor cryopumps will enable pumping at either the inboard (upper and lower) or the outboard (upper and lower) divertor plates. An upgrade to the DIII-D cryogenic system is part of this project. The increased capabilities of the cryogenic system will allow delivery of liquid helium and nitrogen to the three new cryopumps. The Radiative Divertor design is very flexible, and will allow physics studies of the effects of slot width and length. Radiative Divertor diagnostics are being designed in parallel to provide comprehensive measurements for diagnosing the divertor. The Radiative divertor installation is scheduled for late 1996. Engineering experience gained in the DIII-D Advanced Divertor program form a foundation for the design work on the Radiative Divertor

The DIII-D Radiative Divertor Project: Status and plans

New divertor hardware is being designed and fabricated for the Radiative Divertor modification of the DIII-D tokamak. The installation of the hardware has been separated into two phases, the first phase starting in October of 1996 and the second and final phase, in 1998. The phased approach enables the continuation of the divertor characterization research in the lower divertor while providing pumping for density control in high triangularity, single- or double-null advanced tokamak discharges. When completed, the Radiative Divertor Project hardware will provide pumping at all four strike points of a double-null, high triangularity discharge and provide baffling of the neutral particles from transport back to the core plasma. By puffing neutral gas into the divertor region, a reduction in the heat flux on the target plates will be be demonstrated without a large rise in core density. This reduction in heat flux is accomplished by dispersing the power with radiation in the divertor region. Experiments and modeling have formed the basis for the new design. The capability of the DIII-D cryogenic system is being upgraded as part of this project. The increased capability of the cryogenic system will allow delivery of liquid helium and nitrogen to three new cryopumps. Physics studies on the effects of slot width and length can be accomplished easily with the design of the Radiative Divertor. The slot width can be varied by installing graphite tiles of different geometry. The change in slot length, the distance from the X-point to the target plate, requires relocating the structure vertically and can be completed in about 6-8 weeks. Radiative Divertor diagnostics are being designed to provide comprehensive measurements for diagnosing the divertor. Required diagnostic modifications will be minimal for Phase 1, but extensive for Phase 2 installation. These Phase 2 diagnostics will be required to fully diagnose the high triangularity discharges in the divertor slots

Manufacturing development of low activation vanadium alloys

General Atomics is developing manufacturing methods for vanadium alloys as part of a program to encourage the development of low activation alloys for fusion use. The culmination of the program is the fabrication and installation of a vanadium alloy structure in the DIII-D tokamak as part of the Radiative Divertor modification. Water-cooled vanadium alloy components will comprise a portion of the new upper divertor structure. The first step, procuring the material for this program has been completed. The largest heat of vanadium alloy made to date, 1200 kg of V-4Cr-4Ti, has been produced and is being converted into various product forms. Results of many tests on the material during the manufacturing process are reported. Research into potential fabrication methods has been and continues to be performed along with the assessment of manufacturing processes particularly in the area of joining. Joining of vanadium alloys has been identified as the most critical fabrication issue for their use in the Radiative Divertor Program. Joining processes under evaluation include resistance seam, electrodischarge (stud), friction and electron beam welding. Results of welding tests are reported. Metallography and mechanical tests are used to evaluate the weld samples. The need for a protective atmosphere during different welding processes is also being determined. General Atomics has also designed, manufactured, and will be testing a helium-cooled, high heat flux component to assess the use of helium cooled vanadium alloy components for advanced tokamak systems. The component is made from vanadium alloy tubing, machined to enhance the heat transfer characteristics, and joined to end flanges to allow connection to the helium supply. Results are reported

HTR2008-58015 ROTOR SCALE MODEL TESTS FOR POWER CONVERSION UNIT OF GT-MHR

ABSTRACT A power-generating unit with the high-temperature helium reactor (GT-MHR) has a turbomachine (TM) that is intended for both conversion of coolant thermal energy into electric power in the direct gas-turbine cycle, and provision of helium circulation in the primary circuit. The vertically oriented TM is placed in the central area of the power conversion unit (PCU). TM consists of a turbocompressor (TC) and a generator. Their rotors are joined with a diaphragm coupling and supported by electro-magnetic bearings (EMB). The complexity and novelty of the task of the full electromagnetic suspension system development requires thorough stepwise experimental work, from small-scale physical models to full-scale specimen. On this purpose, the following is planned within the framework of the GT-MHR Project: investigations of the "flexible" rotor small-scale mockup with electro-magnetic bearings ("Minimockup" test facility); tests of the radial EMB; tests of the position sensors; tests of the TM rotor scale model; tests of the TM catcher bearings (CB) friction pairs; tests of the CB mockups; tests of EMB and CB pilot samples and investigation of the full-scale electromagnetic suspension system as a part of full-scale turbocompressor tests. The rotor scale model (RSM) tests aim at investigation of dynamics of rotor supported by electromagnetic bearings to validate GT-MHR turbomachine serviceability. Like the full-scale turbomachine rotor, the RSM consist of two parts: the generator rotor model and the turbocompressor rotor model that are joined with a coupling. Both flexible and rigid coupling options are tested. Each rotor is supported by one axial and two radial EMBs. The rotor is arranged vertically. The RSM rotor length is 10.54 m, and mass is 1171 kg. The designs of physical model elements, namely of the turbine, compressors, generator and exciter, are simplified and performed with account of rigid characteristics, which are identical to those of the full-scale turbomachine elements. INTRODUCTION A power-generating unit with the high-temperature helium reacto

DESIGN, FABRICATION, INSTALLATION, TESTING AND INITIAL RESULTS OF IN-VESSEL CONTROL COILS FOR DIII-D

OAK-B135 Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. In 2000, these coils also demonstrated benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modeling has shown that substantial performance improvements could be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and were power tested successfully after several bakes to 350 C. A full set of twelve internal coils and related magnetic sensors are now operational in the DIII-D tokamak. The design requirements for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils wa sin joining of both the copper conductor and the stainless tube without overheating the polyamide insulator. Elastic-plastic analysis was used to demonstrate acceptable thermal stresses during baking conditions. Analysis determined the optimum water cooling channel diameter. The coils were tested in high toroidal field to the limit of the power supply of 4.5 kA DC with inductance-limited current for frequencies between 300 Hz and 1000 Hz. Recent results are presented

DESIGN, FABRICATION, INSTALLATION AND TESTING OF IN-VESSEL CONTROL COILS FOR DIII-D

OAK A271 DESIGN, FABRICATION, INSTALLATION AND TESTING OF IN-VESSEL CONTROL COILS FOR DIII-D. Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. Recently, these coils have also demonstrated significant benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modeling has shown that substantial performance improvements can be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and have been tested successfully. installation of a set of twelve internal coils and magnetic sensors in the DIII-D tokamak is to be completed in December 2002. The design requirement for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator

Design and analysis of the DII-D radiative divertor water-cooled structures

The Radiative Divertor is a major modification to the divertor of DIII-D and is being designed and fabricated for installation in late 1996. The Radiative Divertor Program (RDP) will enhance the dissipative processes in the edge and divertor plasmas to reduce the heat flux and plasma erosion at the divertor target. This approach will have major implications for the heat removal methods used in future devices. The divertor is of slot-type configuration designed to minimize the flow of sputtered and injected impurities back to the core plasma. The new divertor will be composed of toroidally continuous, Inconel 625 water-cooled rings of sandwich construction with an internal water channel, incorporating seam welding to provide the water-to-vacuum seal as well as structural integrity. The divertor structure is designed to withstand electromagnetic loads as a result of halo currents and induced toroidal currents. It also accommodates the thermal differences experienced during the 400 {degrees}C bake used on DIII-D. A low Z plasma-facing surface is provided by mechanically attached graphite tiles. Water flow through the rings will inertially cool these tiles which will be subjected to 38 MW, 10 second pulses. Current schedules call for detailed design in 1996 with installation completed in March 1997. A full size prototype, one-quarter of one ring, is being built to validate manufacturing techniques, machining, roll-forming, and seam welding. The experience and knowledge gained through the fabrication of the prototype is discussed. The design of the electrically isolated (5 kV) vacuum-to-air water feedthroughs supplying the water-cooled rings is also discussed

Interaction between SNAI2 and MYOD enhances oncogenesis and suppresses differentiation in Fusion Negative Rhabdomyosarcoma

Rhabdomyosarcoma (RMS) is an aggressive pediatric malignancy of the muscle, that includes Fusion Positive (FP)-RMS harboring PAX3/7-FOXO1 and Fusion Negative (FN)-RMS commonly with RAS pathway mutations. RMS express myogenic master transcription factors MYOD and MYOG yet are unable to terminally differentiate. Here, we report that SNAI2 is highly expressed in FN-RMS, is oncogenic, blocks myogenic differentiation, and promotes growth. MYOD activates SNAI2 transcription via super enhancers with striped 3D contact architecture. Genome wide chromatin binding analysis demonstrates that SNAI2 preferentially binds enhancer elements and competes with MYOD at a subset of myogenic enhancers required for terminal differentiation. SNAI2 also suppresses expression of a muscle differentiation program modulated by MYOG, MEF2, and CDKN1A. Further, RAS/MEK-signaling modulates SNAI2 levels and binding to chromatin, suggesting that the differentiation blockade by oncogenic RAS is mediated in part by SNAI2. Thus, an interplay between SNAI2, MYOD, and RAS prevents myogenic differentiation and promotes tumorigenesis