Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE). (C) 2016 Elsevier B.V. All rights reserved.

Manufacturing design has been developed by Hyundai Heavy Industries Co., Ltd. (HHI) and Korea Domestic Agency (KODA) to manufacture real product. As the first step of development of the manufacturing design, fabrication feasibility study was carried out in accordance with the RCC-MR 2007. Fabrication sequence was drafted and engineering analyses have been performed to guarantee structural integrity of manufacturing design and to minimize welding distortion. Several design modification proposals were derived during fabrication feasibility study. Fabrication sequence and manufacturing techniques have been established based on the results of a fabrication feasibility study. 3D multi-part model was developed based on the reference multi-body model and manufacturing drawings have been produced. A full scale mock-up of the lower port stub extension (PSE) has been fabricated in order to verify weldability, applicability of non-destructive examination (NDE) and welding distortion. Several qualifications were also conducted to confirm that the manufacturing design satisfies the requirements for bending, solution heat treatment, welding and NDE. Manufacturing and inspection plans (MIP) have been prepared to start real fabrication. In this paper, manufacturing design procedure and major technical results of manufacturing design development are presented for ITER W lower port. (C) 2015 Elsevier B.V. All rights reserved.

The ITER vacuum vessel (W) is under manufacturing by four domestic agencies after completion of engineering designs that have been approved by the Agreed Notified Body (ANB). Manufacturing designs of the VV have been being completed, component by component, by accommodating requirements of the RCC-MR 2007 edition. Manufacturing of the W first sector has been started in February 2012 in Korea and in-wall shielding in May 2013 in India. EU will start manufacturing of its first sector from September 2013 and Russia the upper port by the end of 2013. All DAs have manufactured several mock-ups including real-size ones to justify/qualify and establish manufacturing techniques and procedures. (C) 2014 Elsevier B.V. All rights reserved.

A revision of the ITER Project-Level Load Specification (to be used for all systems of the ITER machine) was implemented in April 2012. This revision supports ITER's licensing by accommodating requests from the French regulator to maintain consistency with the plasma physics database and our present understanding of plasma transients and electro-magnetic (EM) loads, to investigate the possibility of removing unnecessary conservatism in the load requirements and to review the list and definition of incidental cases. The purpose of this paper is to present the impact of this 2012 revision of the ITER Project-Level Load Specification (IS) on the ITER Vacuum Vessel (W) loads and the main structural margins required by the applicable French code, RCC-MR. (C) 2014 Elsevier B.V. All rights reserved.

Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011-2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The W support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering, the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs' manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R&D was launched for ISI. Conceptual design of the W instrumentation has been developed. The W baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure. (C) 2013 ITER Organization. Published by Elsevier B.V. All rights reserved.

After implementing a few design modifications (referred to as the “Modified Reference Design”) in 2009, the Vacuum Vessel (VV) design had been stabilized. The VV design is being finalized, including interface components such as support rails and feedthroughs for the in-vessel coils. It is necessary to make adjustments to the locations of the blanket supports and manifolds to accommodate design modifications to the in-vessel coils. The VV support design is also being finalized considering a structural simplification. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. The detailed layout of ferritic steel plates and borated steel plates was optimized based on the toroidal field ripple analysis. A dynamic test on the inter-modular key to support the blanket modules was performed to measure the dynamic amplification factor (DAF). An R&D program has started to select and qualify the welding and cutting processes for the port flange lip seal. The ITER VV material 316 L(N) IG was already qualified and the Modified Reference Design was approved by the Agreed Notified Body (ANB) in accordance with the Nuclear Pressure Equipment Order procedure.

SS316L(N)-IG (ITER grade) has been selected as the main structural material for the ITER vacuum vessel (VV), considering its high mechanical strength at operating temperatures, water chemistry properties, excellent fabrication characteristics, and low cost relative to other candidates. The ITER VV is a class-2 box structure as defined in RCC-MR, 2007 edition, which was selected as the code for the design and construction. This paper describes materials, applied code and regulatory requirements, baseline fabrication procedures, and assembly on the site.

According to recent design review results, the original reference vacuum vessel (VV) design was selected with a number of modifications including 3D shaping of the outboard inner shell. The VV load conditions were updated based on reviews of the plasma disruption and vertical displacement event (VDE) database. The lower port gussets have been reinforced based on structural analysis results, including non-linear buckling. Design of in-vessel coils for the mitigation of edge localized modes (ELM) and plasma vertical stabilization (VS) has progressed. Design of the in-wall-shielding (IWS) has progressed in details. The detailed layout of ferritic steel plates and borated steel plates is optimized based on the toroidal field ripple analysis. The procurement arrangements (PAs) for the VV including ports and IWS have been prepared or signed. Final design reviews were carried out to check readiness for the PA signature. The procedure for licensing the ITER VV according to the French Order on Nuclear Pressure Equipment (ESPN) has started and conformity assessment is being performed by an Agreed Notified Body (ANB). A VV design description document, VV load specification document, hazard and stress analysis reports and particular material appraisal were submitted according to the guideline and RCC-MR requirements.

A novel concept for incorporating an iron core transformer within a axisymmetric toroidal plasma containment device with a high neutron flux is described. This design enables conceptual design of low aspect ratio devices which employ standard transformer-driven plasma startup by using all-metal high resistance inserts between the toroidal field windings. This design avoids the inherent problems of a multi-turn air core transformer which will inevitably suffer from strong neutron bombardment and hence lose the integrity of its insulation, both through long term material degradation and short term neutron induced conductivity. A full 3-dimensional model of the concept has been developed within the MAXWELL program and the resultant loop voltage calculated. The utility of the result is found to be dependent on the resistivity of the high resistance inserts. Useful loop voltage time histories have been obtained using expected resistivities.

Procurement arrangements for ITER key components including the vacuum vessel (VV) have been signed and the ITER activities are now fully devoted towards construction. Final design reviews have been carried out for the main vessel and ports. One of the design review topics is the selection of materials, material procurement, and assessment of material performance during operation. The width of the inner shell splice plates was increased from 120mm to 160mm to minimize risk during the assembly of the Thermal shields and the VV. Instead of facet shaping, 3D shaping was introduced for the outboard inner shell. The material qualification procedures have been started for VV structural materials such as 316L(N) IG for licensing as a nuclear pressure equipment component. In accordance with the regulatory requirements and quality requirements for operation, common material specifications have been prepared in collaboration with the domestic agencies.

The vacuum vessel (VV) design is being finalized including interface components, such as the support rails and feedthroughs of coils for mitigation of edge localized modes (ELM) and vertical stabilization (VS) of the plasma (ELM/VS coils). It was necessary to make adjustments in the locations of the blanket supports and manifolds to accommodate the design modifications in the ELM/VS coils. The lower port gussets were reinforced to keep a sufficient margin under the increased VV load conditions. The VV support design is being finalized as well, with an emphasis on structure simplification. The design of the in-wall shielding (IWS) has progressed, considering assembly and required tolerances. The layout of ferritic steel plates and borated steel plates will be optimized based on on-going toroidal field ripple analysis. The VV instrumentation was defined in detail. Strain gauges, thermocouples, displacement meters and accelerometers shall be installed to monitor the status of the VV in normal and off-normal conditions to confirm all safety functions are performed correctly. The ITER VV design was preliminarily approved, and the VV materials including 316L(N) IG were already qualified by the Agreed Notified Body (ANB) according to the procedure of Nuclear Pressure Equipment Order.

RATIONALE: Esophageal remodeling in Eosinophilic Esophagitis (EoE), a food-allergic disease, includes epithelial hyperplasia and subepithelial fibrosis. Epithelial Mesenchymal Transition (EMT) is a process whereby non-motile epithelial cells de-differentiate into motile mesenchymal-like cells during chronic inflammation, and can further differentiate into myofibroblasts. Since the mechanism of esophageal fibrosis in EoE is poorly understood, our objective was to determine whether EMT occurs in EoE and contributes to its pathogenesis. METHODS: Sections from 60 formalin-fixed, paraffin-embedded biopsies from patients with EoE (17), GERD (7), indeterminate esophagitis (15), and normal esophagus (21) were evaluated for EMT by two-color immunofluorescence staining for cells expressing cytokeratin (epithelial) and vimentin (mesenchymal) markers. Sections were analyzed in a blinded fashion by confocal microscopy and graded for the presence/degree of EMT using a 6-point scale. Mean EMT indices/hpf were analyzed for relationships to diagnosis, eosinophils/hpf, and indices for eosinophil peroxidase (EPX) and TGF-beta immunostaining, and esophageal fibrosis. RESULTS: The highest EMT index was associated with a patient diagnosis of EoE (3.25±0.28), followed by indeterminate esophagitis(2.69±0.29)>GERD(1.62±0.39)>normal(1.06±0.15). EMT was positively correlated (all p<0.01) with eosinophils/hpf (r=0.691), EPX (r=0.738) and TGF-beta (r=0.520) staining indices, and esophageal fibrosis scores (r=0.644). CONCLUSIONS: This is the first identification of EMT in EoE. EMT was observed principally in EoE, to a lesser degree in indeterminate esophagitis with fibrosis, whereas it was observed only once in GERD, and not in normal esophagus. The significant correlations of EMT to esophageal eosinophils and their activation, and measures of fibrosis, suggest that EMT may contribute to the subepithelial fibrosis characteristic of EoE pathogenesis.

The ITER vacuum vessel (VV) is one of the most critical components in the ITER project. It is on the critical path in the construction schedule and it is also a safety important class component (SIC), providing the first confinement barrier.As a result of reviews and the latest physics analyses, design requirements have been updated (e.g. ELM/VS coils) and a few design changes have to be implemented. This paper covers the updates of the VV vertical and horizontal EM load conditions during asymmetric VDEs, the design analysis of the ELM/VS coils and their interfaces to the VV, the blanket manifold design and the preparation of the technical specification in preparation for the procurement arrangement to be signed. (C) 2009 K. Ioki. Published by Elsevier B.V. All rights reserved.

The National Spherical Torus Experiment (NSTX) has been operating successfully since February of 1999. A unique element of NSTX is the center solenoid or OH coil that from the start has been a design challenged by the low aspect ratio/geometry of the device. To achieve this low aspect ratio the OH coil's outer diameter is constrained to a narrow profile creating the need for creative design solutions concerning cooling connections, lead orientation, and insulation schemes. The original design has succeeded overall, but NSTX run time has been lost due to coil reliability issues. It was decided in the last year that it would be prudent to fabricate a new OH coil and have it available as an upgrade to the experiment. The experience of operating and maintaining the OH coil has provided the basis for an improved OH coil design. A collaboration was arranged with ASIPP in China to fabricate a spare coil for NSTX. The new OH coil will incorporate both design improvements intended to increase reliability as well as upgrades that will provide flexibility during future operation by allowing for an expanded operational profile. This paper summarizes and reviews these design and reliability improvements.

An overview of the research capabilities and the future plans on the MA-class National Spherical Torus Experiment (NSTX) at Princeton is presented. NSTX research is exploring the scientific benefits of modifying the field line structure from that in more conventional aspect ratio devices, such as the tokamak. The relevant scientific issues pursued on NSTX include energy confinement, MHD stability at high β, non-inductive sustainment, solenoid-free start-up, and power and particle handling. In support of the NSTX research goal, research tools are being developed by the NSTX team. In the context of the fusion energy development path being formulated in the US, an ST-based Component Test Facility (CTF) and, ultimately a high β Demo device based on the ST, are being considered. For these, it is essential to develop high performance (high β and high confinement), steady-state (non-inductively driven) ST operational scenarios and an efficient solenoid-free start-up concept. We will also briefly describe the Next-Step-ST (NSST) device being designed to address these issues in fusion-relevant plasma conditions.

We describe the preparation of novel, nanostructured model catalysts for electrocatalytic studies, consisting of Pt nanoparticles with well-defined particle size and separation supported on glassy carbon. The model catalysts are fabricated by colloidal lithography (CL) and characterized with respect to their morphology and their electrochemical and electrocatalytic properties by scanning electron microscopy, atomic force microscopy, X-ray photoelectron spectroscopy, cyclic voltammetry and CO electrooxidation. The resulting data – with comparatively large particles – closely resemble those obtained on polycrystalline Pt electrodes, demonstrating that CL is a viable method for the preparation of well-defined model systems with successively smaller structures for electrocatalytic studies.

A spherical torus (ST) fusion energy development path which is complementary to proposed tokamak burning plasma experiments such as ITER is described. The ST strategy focuses on a compact Component Test Facility (CTF) and higher performance advanced regimes leading to more attractive DEMO and Power Plant scale reactors. To provide the physics basis for the CTF an intermediate step needs to be taken which we refer to as the ''Next Step Spherical Torus'' (NSST) device and examine in some detail herein. NSST is a ''performance extension'' (PE) stage ST with the plasma current of 5-10 MA, R = 1.5 m, and Beta(sub)T less than or equal to 2.7 T with flexible physics capability. The mission of NSST is to: (1) provide a sufficient physics basis for the design of CTF, (2) explore advanced operating scenarios with high bootstrap current fraction/high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, and (3) contribute to the general plasma/fusion science of high beta toroidal plasmas. The NSST facility is designed to utilize the Tokamak Fusion Test Reactor (or similar) site to minimize the cost and time required for the design and construction.

The spherical torus (ST) fusion energy development path is complementary to the tokamak burning plasma experiment such as ITER as it focuses toward the compact Component Test Facility (CTF) and higher toroidal beta regimes to improve the design of DEMO and a Power Plant. To support the ST development path, one option of a Next Step Spherical Torus (NSST) device is examined. NSST is a “performance extension” (PE) stage ST with a plasma current of 5 10 MA, R = 1.5, BT ≤ 2.7 T with flexible physics capability to 1) Provide a sufficient physics basis for the design of the CTF, 2) Explore advanced operating scenarios with high bootstrap current fraction / high performance regimes, which can then be utilized by CTF, DEMO, and Power Plants, 3) Contribute to the general plasma / fusion science of high β toroidal plasmas. The NSST facility is designed to utilize the TFTR site to minimize the cost and time required for the construction. 1. Spherical Torus Fusion Energy Development Path The potential of the Spherical Torus (ST) configuration to enable attractive fusion energy was discussed in a number of recent papers.[1-3] The engineering feasibility of a single-turn center leg for the toroidal field coils was identified as a key element for attractive ST reactors [1], which in turn pointed to the importance of solenoid-free startup and sustainment of the ST plasmas. The use of single-turn center leg also provides the possibility of compact driven steady state burning ST plasma. These plasmas would have R ~ 1+ m [2,3] assuming standard ST physics performance (such as βT ~ 25%) to produce substantial neutron wall loading (WL ~ 1 MW/m or higher) at a modest total fusion power (~ 50 MW or higher). Such an approach can be a strong candidate for the Component Test Facility (CTF) (also called the volume neutron source or VNS) needed in the test and development of reliable fusion nuclear components [4] for a DEMO power plant. The importance of CTF in the accelerated development of fusion energy was recently recognized more broadly in the U.S. fusion research community [5]. The higher end of the CTF performance can in principle include the potential for producing net electrical power under high performance plasma conditions, if high performance fusion nuclear components can also be developed (such as for WL ~ 5 MW/m 2 or higher). One can envision that the CTF facility would begin the operation at the level of WL ~ 12 MW/m , and progressively upgrade the test components to handle higher WL while improving ST plasma performance. The CTF device is therefore expected to satisfy stringent operational requirements [3], such as complete modularity of all fusion core components (including the single-turn center leg) to permit rapid change out for replacement under full remote conditions. The CTF device should also achieve the high neutron fluence (~6 MW-yr/m or higher) required in the testing program. The expected high tritium consumption required for component testing puts a premium on highly compact devices that maximize WL while minimizing the total fusion power. Due to the limited supply of tritium anticipated for the next few decades, tritium breeding for tritium selfsufficiency becomes a critical requirement. For this purpose, the fusion neutrons lost to the

Vertical stability and control are examined for a tokamak configuration intended to be a generic representation of next step devices. Vertical stability calculations show that a critical resistive wall location can be determined for realistic structures and that the introduction of small amounts of low resistivity material to an all steel structure can significantly reduce the vertical instability growth rate. Vertical control simulations show that internal control coils require significantly less feedback power than external coils, and that low resistivity materials can allow either very low feedback powers or the coils to be located externally with reasonable feedback powers.

Strategies for the improvement of quasiaxisymmetric stellarator configurations are explored. Calculations of equilibrium flux surfaces for candidate configurations are also presented. One optimization strategy is found to generate configurations with improved neoclassical confinement, simpler coils with lower current density, and improved flux surface quality relative to previous designs. The flux surface calculations find significant differences in the extent of islands and stochastic regions between candidate configurations. (These calculations do not incorporate the predicted beneficial effects of perturbed bootstrap currents.) A method is demonstrated for removing low order islands from candidate configurations by relatively small modifications of the configuration. One configuration is identified as having particularly desirable properties for a proposed experiment.