[신간의 별자리x] 우리/미술, 그리고 ‘슬픔의 박물관’

Deposits of clastic carbonate-dominated (calciclastic) sedimentary slope systems in the rock record have been identified mostly as linearly-consistent carbonate apron deposits, even though most ancient clastic carbonate slope deposits fit the submarine fan systems better. Calciclastic submarine fans are consequently rarely described and are poorly understood. Subsequently, very little is known especially in mud-dominated calciclastic submarine fan systems. Presented in this study are a sedimentological core and petrographic characterisation of samples from eleven boreholes from the Lower Carboniferous of Bowland Basin (Northwest England) that reveals a >250 m thick calciturbidite complex deposited in a calciclastic submarine fan setting. Seven facies are recognised from core and thin section characterisation and are grouped into three carbonate turbidite sequences. They include: 1) Calciturbidites, comprising mostly of highto low-density, wavy-laminated bioclast-rich facies; 2) low-density densite mudstones which are characterised by planar laminated and unlaminated muddominated facies; and 3) Calcidebrites which are muddy or hyper-concentrated debrisflow deposits occurring as poorly-sorted, chaotic, mud-supported floatstones. These

Phd by thesis

The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

Progress, since the ITER Physics Basis publication (ITER Physics Basis Editors et al 1999 Nucl. Fusion 39 2137–2664), in understanding the processes that will determine the properties of the plasma edge and its interaction with material elements in ITER is described. Experimental areas where significant progress has taken place are energy transport in the scrape-off layer (SOL) in particular of the anomalous transport scaling, particle transport in the SOL that plays a major role in the interaction of diverted plasmas with the main-chamber material elements, edge localized mode (ELM) energy deposition on material elements and the transport mechanism for the ELM energy from the main plasma to the plasma facing components, the physics of plasma detachment and neutral dynamics including the edge density profile structure and the control of plasma particle content and He removal, the erosion of low- and high-Z materials in fusion devices, their transport to the core plasma and their migration at the plasma edge including the formation of mixed materials, the processes determining the size and location of the retention of tritium in fusion devices and methods to remove it and the processes determining the efficiency of the various fuelling methods as well as their development towards the ITER requirements. This experimental progress has been accompanied by the development of modelling tools for the physical processes at the edge plasma and plasma–materials interaction and the further validation of these models by comparing their predictions with the new experimental results. Progress in the modelling development and validation has been mostly concentrated in the following areas: refinement in the predictions for ITER with plasma edge modelling codes by inclusion of detailed geometrical features of the divertor and the introduction of physical effects, which can play a major role in determining the divertor parameters at the divertor for ITER conditions such as hydrogen radiation transport and neutral–neutral collisions, modelling of the ion orbits at the plasma edge, which can play a role in determining power deposition at the divertor target, models for plasma–materials and plasma dynamics interaction during ELMs and disruptions, models for the transport of impurities at the plasma edge to describe the core contamination by impurities and the migration of eroded materials at the edge plasma and its associated tritium retention and models for the turbulent processes that determine the anomalous transport of energy and particles across the SOL. The implications for the expected performance of the reference regimes in ITER, the operation of the ITER device and the lifetime of the plasma facing materials are discussed.

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Chapter 4: Power and particle control

https://pure.mpg.de/pubman/item/item_2141173_1/component/file_2141172/RothPSI08_Paper_v9.pdf

Recent analysis of key plasma wall interactions issues for ITER

D retention in C- and O-contaminated tungsten during D + irradiation

Abstract Carbon-based codeposits formed in carbon-containing fusion devices have the potential to dominate tritium retention in the torus. One of the tritium removal techniques currently being studied is thermo-oxidation, which is unique in its ability to remove tritium from codeposits without mechanical intervention in the torus and in its ability to remove tritium from codeposits in tile gaps and shaded areas. In preparation for an oxidation experiment planned to be performed in DIII-D, we have investigated the potential collateral effects of thermo-oxidation on DIII-D in-vessel components. Laboratory oxidation experiments were performed at 2 Torr ([approximately]270 Pa) and 15 Torr ([approximately]2 kPa) O2 pressure and temperatures in the range 100 to 350°C (373 to 623 K) for 2 to 8 h. After oxidation, components were examined for visual or mechanical change, and when appropriate, mass changes were also obtained. In some cases, optical diagnostics were also performed. The specimens were mostly spare/surplus components and spanned a wide variety of materials and functions, e.g., cryopump components; structural, mechanical, and diagnostic components; and fast-wave antennas. The effect of oxidation was found to be negligible for nearly all DIII-D components and materials tested.

J.W. Davis

Papers

D retention in C- and O-contaminated tungsten during D + irradiation

Assessment of Collateral Effects of Thermo-Oxidation on DIII-D In-Vessel Components in Preparation for In Situ Oxidation in DIII-D

Removal of deuterium from carbon-based codeposits by hydrogen isotope exchange

Bombardment of a Si pitch grating by C+ ions at an inclined incident angle parallel to the structure

Carbon erosion due to simultaneous H+ and e− impact