| 1. |
- Carninci, P, et al.
(författare)
-
The transcriptional landscape of the mammalian genome
- 2005
-
Ingår i: Science (New York, N.Y.). - : American Association for the Advancement of Science (AAAS). - 1095-9203 .- 0036-8075. ; 309:5740, s. 1559-1563
-
Tidskriftsartikel (refereegranskat)abstract
- This study describes comprehensive polling of transcription start and termination sites and analysis of previously unidentified full-length complementary DNAs derived from the mouse genome. We identify the 5′ and 3′ boundaries of 181,047 transcripts with extensive variation in transcripts arising from alternative promoter usage, splicing, and polyadenylation. There are 16,247 new mouse protein-coding transcripts, including 5154 encoding previously unidentified proteins. Genomic mapping of the transcriptome reveals transcriptional forests, with overlapping transcription on both strands, separated by deserts in which few transcripts are observed. The data provide a comprehensive platform for the comparative analysis of mammalian transcriptional regulation in differentiation and development.
|
|
| 2. |
- Lamalle, P. U., et al.
(författare)
-
Expanding the operating space of ICRF on JET with a view to ITER
- 2006
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 46:2, s. 391-400
-
Tidskriftsartikel (refereegranskat)abstract
- This paper reports on ITER-relevant ion cyclotron resonance frequency (ICRF) physics investigated on JET in 2003 and early 2004. Minority heating of helium three in hydrogen plasmas-(He-3)H-was systematically explored by varying the 3 He concentration and the toroidal phasing of the antenna arrays. The best heating performance (a maximum electron temperature of 6.2 keV with 5 MW of ICRF power) was obtained with a preferential wave launch in the direction of the plasma current. A clear experimental demonstration was made of the sharp and reproducible transition to the mode conversion heating regime when the 3 He concentration increased above similar to 2%. In the latter regime the best heating performance (a maximum electron temperature of 8 keV with 5 MW of ICRF power) was achieved with dipole array phasing, i.e. a symmetric antenna power spectrum. Minority heating of deuterium in hydrogen plasmas-(D)H-was also investigated but was found inaccessible because this scenario is too sensitive to impurity ions with Z/A = 1/2 such as C6+, small amounts of which directly lead into the mode conversion regime. Minority heating of up to 3% of tritium in deuterium plasmas was systematically investigated during the JET trace tritium experimental campaign (TTE). This required operating JET at its highest possible magnetic field (3.9 to 4 T) and the ICRF system at its lowest frequency (23 MHz). The interest of this scenario for ICRF heating at these low concentrations and its efficiency at boosting the suprathermal neutron yield were confirmed, and the measured neutron and gammay ray spectra permit interesting comparisons with advanced ICRF code simulations. Investigations of finite Larmor radius effects on the RF-induced high-energy tails during second harmonic (omega = 2 omega(c)) heating of a hydrogen minority in D plasmas clearly demonstrated a strong decrease in the RF diffusion coefficient at proton energies similar to 1 MeV in agreement with theoretical expectations. Fast wave heating and current drive experiments in deuterium plasmas showed effective direct electron heating with dipole phasing of the antennas, but only small changes of the central plasma current density were observed with the directive phasings, in particular at low single pass damping. New investigations of the heating efficiency of ICRF antennas confirmed its strong dependence on the parallel wavenumber spectrum. Advances in topics of a more technological nature are also summarized: ELM studies using fast RF measurements, the successful experimental demonstration of a new ELM-tolerant antenna matching scheme and technical enhancements planned on the JET ICRF system for 2006, they being equally strongly driven by the preparation for ITER.
|
|
| 3. |
- Lamalle, P.U, et al.
(författare)
-
Expanding the operating space of ICRF on JET with a view to ITER
- 2006
-
Ingår i: Nucl. Fusion. ; 46, s. 391-
-
Tidskriftsartikel (refereegranskat)abstract
- Mercury (Hg) has been used for millennia in many applications, primarily in artisanal mining and as an electrode in the chlor–alkali industry. It is anthropogenically emitted as a pollutant from coal fired power plants and naturally emitted, primarily from volcanoes. Its unique chemical characteristics enable global atmospheric transport and it is deposited after various processes, ultimately ending up in one of its final sinks, such as incorporated into deep sediment or bioaccumulated, primarily in the marine environment. All forms of Hg have been established as toxic, and there have been no noted biological benefits from the metal.Throughout time, there have been notable incidents of Hg intoxication documented, and the negative health effects have been documented to those chronically or acutely exposed. Today, exposure to Hg is largely diet or occupationally dependent, however, many are exposed to Hg from their amalgam fillings. This paper puts a tentative monetary value on Hg polluted food sources in the Arctic, where local, significant pollution sources are limited, and relates this to costs for strategies avoiding Hg pollution and to remediation costs of contaminated sites in Sweden and Japan. The case studies are compiled to help policy makers and the public to evaluate whether the benefits to the global environment from banning Hg and limiting its initial emission outweigh the benefits from its continued use or lack of control of Hg emissions. The cases we studied are relevant for point pollution sources globally and their remediation costs ranged between 2500 and 1.1 million US$ kg−1 Hg isolated from the biosphere. Therefore, regulations discontinuing mercury uses combined with extensive flue gas cleaning for all power plants and waste incinerators is cost effective.
|
|
| 4. |
- Hirai, T., et al.
(författare)
-
R&D on full tungsten divertor and beryllium wall for JET ITER-like wall project
- 2007
-
Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 82:15-24, s. 1839-1845
-
Tidskriftsartikel (refereegranskat)abstract
- The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. In order to perform a fully integrated material test JET has launched the ITER-like Wall Project with the aim of installing a full metal wall during the next major shutdown. As a result of R&D projects in 2005-2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies.
|
|
| 5. |
- Maier, H., et al.
(författare)
-
Tungsten and beryllium armour development for the JET ITER-like wall project
- 2007
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 47:3, s. 222-227
-
Tidskriftsartikel (refereegranskat)abstract
- For the ITER-like wall project at JET the present main chamber CFC tiles will be exchanged with Be tiles and in parallel a fully tungsten-clad divertor will be prepared. Therefore three R&D programmes were initiated: Be coatings on Inconel as well as Be erosion markers were developed for the first wall of the main chamber. High heat flux screening and cyclic loading tests carried out on the Be coatings on Inconel showed excellent performance, above the required power and energy density. For the divertor a conceptual design for a bulk W horizontal target plate was investigated, with the emphasis on minimizing electromagnetic forces. The design consisted of stacks of W lamellae of 6 mm width that were insulated in the toroidal direction. High heat flux tests of a test module were performed with an electron beam at an absorbed power density Up to 9 MW m(-2) for more than 150 pulses and finally with increasing power loads leading to surface temperatures in excess of 3000 degrees C. No macroscopic failure occurred during the test while SEM showed the development of micro-cracks on the loaded surface. For all other divertor parts R&D was performed to provide the technology to coat the 2-directional CFC material used at JET with thin tungsten coatings. The W-coated CFC tiles were subjected to heat loads with power densities ranging up to 23.5 MW m(-2) and exposed to cyclic heat loading for 200 pulses at 10.5 MW m(-2). All coatings developed cracks perpendicular to the CFC fibres due to the stronger contraction of the coating upon cool-down after the heat pulses.
|
|
| 6. |
- Rosanvallon, S., et al.
(författare)
-
Tritium related studies within the JET Fusion Technology work programme
- 2005
-
Ingår i: Fusion science and technology. - 1536-1055 .- 1943-7641. ; 48:1, s. 268-273
-
Tidskriftsartikel (refereegranskat)abstract
- The JET Fusion Technology (FT) work programme was launched in 2000, in the frame of the European Fusion Development Agreement, to address issues related to JET and ITER. In particular, there are four topics related to tritium being investigated Based on the experience gained on the existing tokamaks, first calculations indicate that in-vessel tritium retention could represent a burden for ITER operation. Therefore erosion/deposition studies are being performed in order to better understand the layer co-deposition and tritium retention processes in tokamaks. Moreover, testing of in-situ detritiation processes, in particular laser and flash lamp treatments, should assess detritiation techniques for in-vessel components in the ITER-relevant JET configuration. To reduce the constraints on waste disposal, dedicated procedures are being developed for detritiation Of metals, graphite, carbon-fibre composites, process and housekeeping waste. During the operational and decommissioning phases of a fusion reactor, many processes will produce tritiated water. Key components for an ITER relevant water detritiation facility are being studied experimentally with the aim of producing a complete design that could be implemented and tested at JET. This paper describes these topics of the FT-programme, the strategy developed and the results obtained so far.
|
|
| 7. |
- Grisolia, C., et al.
(författare)
-
JET contributions to ITER technology issues
- 2006
-
Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 81:07-jan, s. 149-154
-
Tidskriftsartikel (refereegranskat)abstract
- The Joint European Torus (JET) fusion machine is the only device capable of operation with tritium and of handling Be and therefore is best suited to the study of tritium and fusion-related issues. A large variety of activities are performed within the JET fusion technology task force (FT-TF). In this paper, some topics such as erosion/deposition and material transport, characterisation of flakes and detritiation techniques are highlighted. Recent examples of results obtained on waste management studies are also given. Data on some ITER-relevant components that have been tested at JET, such as a pumping cryopanel and hardened optics fibers, are presented. In all fields, the work to be addressed in future JET work programmes is discussed.
|
|
| 8. |
- Hirai, T., et al.
(författare)
-
Characterization and heat flux testing of beryllium coatings on Inconel for JET ITER-like wall project
- 2007
-
Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 166-170
-
Tidskriftsartikel (refereegranskat)abstract
- In order to perform a fully integrated material test, JET has launched the ITER-like wall project with the aim of installing a full metal wall during the next major shutdown. The material foreseen for the main chamber wall is bulk Be at the limiters and Be coatings on inconel tiles elsewhere. R&D process comprises global characterization ( structure, purity etc) of the evaporated films and testing of their performance under heat loads. The major results are (i) the layers have survived energy loads of 20 MJ m(-2) which is significantly above the required level of 5 - 10 MJ m(-2), (ii) melting limit of beryllium coating would be at the energy level of 30 MJ m(-2), (iii) cyclic thermal load of 10 MJ m(-2) for up to 50 cycles have not induced any noticeable damage such as flaking or detachment.
|
|
| 9. |
- Matthews, G. F., et al.
(författare)
-
Overview of the ITER-like wall project
- 2007
-
Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 137-143
-
Tidskriftsartikel (refereegranskat)abstract
- Work is in progress to completely replace, in 2008/9, the existing JET CFC tiles with a configuration of plasma facing materials consistent with the ITER design. The ITER-like wall (ILW) will be created with a combination of beryllium ( Be), tungsten ( W), W-coated CFC and Be-coated inconel tiles, with the material depending on the local anticipated heat flux and geometry. It is part of an integrated package of JET enhancements whose aim is to develop an understanding of the ITER materials issues and develop the techniques required to operate with inductive and advanced scenarios as close as possible to ITER parameters. Over 4000 tiles will be replaced and the ILW will accommodate additional heating up to at least 50 MW for 10 s. This paper describes the scientific background to the project, the technical objectives, the material configuration selected, the R&D behind the practical realization of the objectives and the generic problems associated with the Be tiles ( power handling capacity and disruption induced eddy currents). One of the objectives is to maintain or improve the existing CFC tile power handling performance which has been achieved in most cases by hiding bolt holes, optimizing tile size and profile and introducing castellations on plasma facing surfaces.
|
|
| 10. |
- Piazza, G., et al.
(författare)
-
R&D on tungsten plasma facing components for the JET ITER-like wall project
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 367, s. 1438-1443
-
Tidskriftsartikel (refereegranskat)abstract
- Currently, the primary ITER materials choice is a full beryllium main wall with carbon fibre composite at the divertor strike points and tungsten on the upper vertical targets and dome. The full tungsten divertor option is a possibility for the subsequent D-T phase. Neither of the ITER material combinations of first wall and divertor materials has ever been tested in a tokamak. To collect operational experience at JET with ITER relevant material combination (Be, C and W) would reduce uncertainties and focus the preparation for ITER operations. Therefore, the ITER-like wall project has been launched to install in JET a tungsten divertor and a beryllium main wall. This paper describes the R&D activities carried out for the project to develop an inertially cooled bulk tungsten divertor tile, to fully characterise tungsten coating technologies for CFC divertor tiles and to develop erosion markers for use as diagnostics on beryllium tiles.
|
|
