ELFT HÍRADÓ

[Sandor Zoletnik <zoletnik AT rmki.kfki.hu>]

Kedves Kollegak!

Az EURATOM szabalyozott termonuklearis fuzios kutatasi programja
ujabb palyazatokat irt ki a programhoz kapcsolodo technologia problemak
megoldasara. Ezek nagy resze igen specialis technologiat es
felkeszultseget igenyel, de kivalogattam azokat a projecteket amikre
esetleg magyar intezmenyek, cegek is esellyel palyazhatnanak. Ezek cimet
es rovid leirasat mellekelem. Akik ugy erzik, hogy valamelyik temaban
esellyel tudnanak palyazani, azok kerem jelentkezzenek nalam, mivel
marcius 9-ig kell szandeknyilatkozatot tenni.

A felsoroltakon kivul nagy vonalakban a kovetkezo temakban vannak meg
palyazatok: ipari szupravezeto technikak, specialis hegesztesi modszerek,
nagyteljesitmenyu mikrohullamu es semleges nyalab futesi eljarasok,
tricium kezelesi modszerek, in-situ anyagvizsgalati modszerek, kornyezeti
hatas es nukearis balaseti kovetkezmeny vizsgalatok.

Szivelyes udvozlettel           Zoletnik Sandor
                        A magyar EURATOM fuzios szovetseg koordinatora

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Kivonat a palyazhato temak listajabol:

Title: Radiation effects on fusion magnet insulations  
Ref: TW1-TMS/RADEFF 
     Magnet insulation materials as currently used in industry for the
     fabrication of large coils (also for the TF and CS Model coils) is
     actually being tested. The properties are just at the limit to resist
     the nuclear radiation expected in ITER. Therefore it is desirable to
     find in this respect improved materials. The characterization of such
     materials after irradiation will be carried out. The mechanical, fatigue
     and electrical properties of such materials at 77K and RT will be
     measured.

Title: Irradiation Effects in Ceramics for Heating and Current Drive, and 
Ref: TW1-TPD/IRRCER
       Diagnostics Systems
     To provide the required testing and screening of small prototype
     components, and the necessary insulator research and data base for
     heating and current drive, and diagnostic systems for a "Next-Step"
     fusion device. Work will be carried out on transmission components
     including widows, and fibre optics, and on sensors including bolometers
     and pressure gauges. Specific ceramic issues to be addressed will
     include mechanical properties and, electrical degradation including
     in-reactor. For heating and current drive systems insulator
     characterization for NBI and LH will be carried out.

Title: Development of Tritium Analytical Devices
Ref: TW1-TTF/TR12  
     Assessment of performance of Micro Gas Chromatography, the principal
     analytical tool foreseen for the ITER Tritium Plant, in combination with
     the key processing (experimental and infrastructural) systems of TLK.
     This investigation should aim to establish the range of gas compositions
     which can be analysed, and the associated sensitivities, sample
     processing times and other process parameters, any incompatibilities
     with expected gas species and limitations in ability to discriminate
     between species, as well as reliability and maintenance aspects. The
     capabilities of commercially available equipment, and any modifications
     needed for the specific applications in tritium plants should be
     documented.

Title: Determination (ex-situ) of Tritium Content in Plasma Facing Components
Ref: TW1-TTF/TR14  
     This task focuses on the determination of the tritium content of PFC
     specimens (tiles, flakes and dust) from JET and other machines (as
     available), prior to, and following, detritiation. The determinations
     are to be made by methods which include high-sensitivity calorimetry and
     complete combustion for bulk determinations, and surface techniques.
     (The development of detritiation methods is carried out in other tasks
     within the ongoing ITER and JET technology programmes.) An assessment as
     to the achievable residual tritium content, and therefore the ultimate
     disposal options, should be made.
     

Title: In BEAM Mechanical testing of CuCrZr
Ref: TW1-TVV/BEAM  
     The copper components in ITER are subject to cyclic thermal and
     mechanical loads at the same time as being bombarded with a neutron
     flux. At the present time it is not known what the dynamic effect of the
     cyclic applied stress will have on the damage accumulation and thus on
     the mechanical performance of the copper alloy. It has already been
     shown that Copper Alloys are sensitive to the dynamic interaction
     between fatigue and creep. This task is intended to investigate the
     effect of cyclic stress on the evolution of the microstructure during
     particle irradiation. At present there is no way to study systematically
     the interaction of continuous bombardment on the fatigue lifetime of
     these copper alloys. This study should help to define the extent of any
     effect on the fatigue life of simultaneous irradiation and cyclic
     mechanical loading.

     The temperature to be investigated at 10?C. At this temperature the
     effect of irradiation and then fatigue testing is weel known. The
     materials to be investigated shall be commercially available alloys, as
     would be used for the manufacture of the blanket modules.
     
     The intended start date for this work is 01/06/2001.

Title: INCONEL 625+ AND MARTENSITIC STEEL IRRADIATION INDUCED STRESS 
RELAXATION
Ref: TW1-TVV/BOLMAT  
  
     While there are some data on the effects of irradiation on the
     mechanical properties of Inconel 625+ and PH 13-8 Mo, little information
     is available about its dimensional stability due to irradiation creep
     effects. Therefore, it is the objective of this Task to create a set of
     data for loading conditions typical of the actual operation conditions
     of the bolts in ITER.
     
     Two types of experiments can be performed. One is based on conventional
     pressurised tube technique and is aimed at accurate measurement of
     deformation due to irradiation creep and its comparison with the data
     base derived from the same type of tests. The other one uses prestressed
     bolts, in configurations as close as possible to the ITER loading.
     Experiments are planned for irradiations to several dose levels (0.5, 1
     and 2 dpa), at temperatures in the range of 300?C to 350?C, and under
     pre stresses loads typical of the ITER conditions. The experiments shall
     include some tensile testing for control purposes.
     
     
     TW1-TVV/COP Irradiation testing of as-fabricated CuCrZr base alloy
     including Creep-fatigue
     
     The use of CuAl25 alloy has shown increasing problems with low fracture
     toughness and more recently a reduction in fatigue lifetime when exposed
     to combined creep loads. At the same time fabrication route have been
     developed for the manufacture of the blanket modules which allow CuCrZr
     to be used in a near optimum state, the mechanical and physical
     properties of the alloy CuCrZr being sensitive to the thermal history of
     this alloy. This task has two objectives.
     
     The first objective is to study creep-fatigue effects in CuCrZr
     especially in the irradiated state at temperatures typical of the
     blanket operation. The second objective is to determine the detailed
     mechanical and physical properties of this alloy after the manufacturing
     heat treatments. In particular, tensile strength, fracture toughness and
     fatigue strength and electrical conductivity are required in the
     temperature range typical of the blanket module during operation. The
     material to be investigated shall be a commercially available alloy, as
     would be used for the manufacture of the blanket modules. Crucial to
     this study would be the effect of irradiation levels up to 0.3 dpa.
     
     The intended start date for this work is 01/04/2001.

Title: Titanium alloy IRRADIATION testing
Ref: TW1-TVV/TITAN

The ITER first wall/shield modules are attached to the backplate by a
set of four radial flexible supports. Ti-6Al-4V (alpha + beta alloy) is
selected for fabricating the flexible cartridges due to its excellent
mechanical properties and low Young's modulus.

The objective of this task is to investigate the effects of irradiation
on tensile, fatigue, and fracture toughness properties of this alloy,
and includes microstructural examination. Neutron irradiation is
performed to a dose of 0.3 dpa at 150?C. Testing is performed at the
same temperatures. Previous work has been performed at 50?C and 350?C
showing the alpha alloy has a higher resistance to irradiation that the
alpha + beta alloys but this needs to be confirmed at the actual
operational temperature of the flexible cartridges.

Title: Development of in-situ measurement technique for tritium inventory in 
plasma facing materials
Ref: TW1-TVP/TINV   

     In a next step D/T fusion device, the accurate predictions of tritium
     retention in Plasma Facing Component is important both from the point of
     view of the safety and the physics performance.
     
     However, for reliable prediction of T-retention, a detailed knowledge
     and understanding on the mechanisms, which are responsible for
     T-retention are essential.
     
     After a decade experimental investigation and theoretical study on D/T
     plasma wall interaction, D/T transport in materials as well as neutron
     effects, the progress on understanding of the complex of T-retention is
     significant, with the consequence that the prediction of T-retention can
     be performed more accurately, including the following mechanisms
     
     1. Tritium retention via implantation, 
     2. Bulk retention via bulk diffusion,  
     3. Tritium retention via neutron transmutation, 
     4. Tritium retention via  co- deposition, 
     5. Tritium retention via neutron damage induced trapping sites and  
     6. Tritium retention in dust and flakes
     
     In real machine operation, in-situ monitoring of the tritium inventory
     is required to observe the strict safety regulations. Initial R&D on
     such measuring methods had been performed in the past with encouraging
     results.
     
     These efforts shall be continued within this Task with:
     
     a) Systematic experimental investigation of present existing tritium
     detection and analysis techniques in plasma facing materials (C, Be, W,
     Mo), and selection of the most promising technique by laboratory testing.
     
     b) Verification of selected technique in plasma facility or in tokamak.

Title: Erosion & re-deposition of W: Impact of Oxygen impurity in Hydrogen 
Ref: TW1-TVP/TU2

     Apart from the low impurity limit tolerable in a fusion plasma, high-Z
     materials possess high surface binding energies and consequently low
     physical sputtering yields by light ions. In particular, the erosion of
     high-Z materials, e.g. W, Mo by light ions may become very marginal, if
     the plasma temperature is kept, Te<25 eV. However, some investigations
     showed that the oxygen impurity may reduce the surface binding energy of
     W. Consequently, the threshold potential of D sputtering is reduced to
     <27 eV, instead of 178 Ev. It implies, the erosion of W will be
     enhanced, if plasma contains oxygen impurity. Therefore, the knowledge
     of oxygen impact on the behavior of high-Z materials is essential for
     use as amour materials in fusion devices. Blister formation at divertor
     conditions is also serious concern.
     
     The objective of this Task is to complement the data base for tungsten
     as a reference plasma facing material for ITER, with respect to the
     following properties:
     
     a.) Erosion by simultaneous H+ and O+ impurity impact; as a function of
     O+/H+ ratios and target temperature.
     
     b) D retention in W as a function of blister formation, flux, fluence,
     O+/D+ ratios and target temperature

Title: Neutron Effects on physical & Mechanical Properties of Tungsten
Ref: TW1-TVP/TU3

     The assessed neutron flux for next step fusion device, e.g. ITER will be
     around 3.5-9.0 x 1014cm-2s-1 for the first wall, whilst the neutron flux
     for the divertor is around 1-3 x 1014 cm-2s-1, for which leads to damage
     of around 10-20 dpa for the first wall and 3-6 dpa for the divertor for
     1 full power year of operation. In the framework of European Fusion R &
     D programs, an extensive effort on neutron effects on plasma facing
     components is being undertaken to investigate the effects of neutron
     doses and irradiation temperature on the thermal conductivity,
     mechanical properties, dimensional stability and tritium inventory of
     various carbon based materials. However, the knowledge about the neutron
     effects on the physical & mechanical properties is still rare.
     
     The objective of this Task is to investigate the neutron effects on the
     properties of tungsten as a function of neutron dose and irradiation
     temperature

Title: Erosion & redeposition of W and Mo: Impact of Oxygen Impurity 
Ref: TW1-TVP/TUMO1

     Apart from the low impurity limit tolerable in fusion plasma, high-Z
     materials posses high surface binding energies and consequently low
     physical sputtering yields by light ions. In particular, the erosion of
     high-Z materials, e.g. W, Mo by light ions may become very marginal, if
     the plasma temperature is kept, Te<25 eV. However, some investigations
     showed that the oxygen impurity may reduce the surface binding energy of
     W, consequently, the threshold potential of D sputtering is reduce to
     <27 eV, instead of 178 Ev. It implies, the erosion of W will be
     enhanced, if plasma contains oxygen impurity. Therefore, the knowledge
     of oxygen impact on the behavior of high-Z materials is essential for
     use as amour materials in fusion devices. Blister formation at divertor
     conditions is also serious concern.
     
     The objective of this Task is to complement the data base on tungsten as
     an ITER reference armour material, and to later extend it to molybdenum.
     In particular, the following measurements shall be performed in a plasma
     facility:
     
     a) Erosion by simultaneous O+ impurity impact in plasma; as a function
     of Te , flux density, O+/H+/D+ ratios and target temperature.
     
     b) Blister formation as function of Te, flux, fluence, O+/D+ ratios and
     target temperature

Title: Microwave inspection and imaging
Ref: TW1-TVA/MIC   

     The object of this task is to investigate the possibility of using
     electromagnetic waves for the non-optical in-vessel inspection of a next
     step tokamak, especially in emergency situations when a quick survey is
     required. This feasibility study seeks ways for the practical
     exploitation of the advantages theoretically inherent in the application
     of microwave techniques. The practical aim of such a system would be to
     reconstruct an image of the inside of the vessel. The study, assuming a
     positive outcome, will include proposals for a laboratory demonstration
     (to be carried out in 2002).