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Sample records for iter candidate beryllium

  1. Neutron irradiation behavior of ITER candidate beryllium grades

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B.; Gorokhov, V.A.; Nikolaev, G.N. [A.A.Bochvar All-Russia Scientific Research Inst. of Inorganic Materials (VNIINM), Moscow (Russian Federation); Melder, R.R.; Ostrovsky, Z.E.

    1998-01-01

    Beryllium is one of the main candidate materials both for the neutron multiplier in a solid breeding blanket and for the plasma facing components. That is why its behaviour under the typical for fusion reactor loading, in particular, under the neutron irradiation is of a great importance. This paper presents mechanical properties, swelling and microstructure of six beryllium grades (DshG-200, TR-30, TshG-56, TRR, TE-30, TIP-30) fabricated by VNIINM, Russia and also one - (S-65) fabricated by Brush Wellman, USA. The average grain size of the beryllium grades varied from 8 to 25 {mu}m, beryllium oxide content was 0.8-3.2 wt. %, initial tensile strength was 250-680 MPa. All the samples were irradiated in active zone of SM-3 reactor up to the fast neutron fluence (5.5-6.2) {center_dot} 10{sup 21} cm{sup -2} (2.7-3.0 dpa, helium content up to 1150 appm), E > 0.1 MeV at two temperature ranges: T{sub 1} = 130-180degC and T{sub 2} = 650-700degC. After irradiation at 130-180degC no changes in samples dimensions were revealed. After irradiation at 650-700degC swelling of the materials was found to be in the range 0.1-2.1 %. Beryllium grades TR-30 and TRR, having the smallest grain size and highest beryllium oxide content, demonstrated minimal swelling, which was no more than 0.1 % at 650-700degC and fluence 5.5 {center_dot} 10{sup 21} cm{sup -2}. Tensile and compression test results and microstructure parameters measured before and after irradiation are also presented. (author)

  2. Tritium behavior in ITER beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1990-10-01

    The beryllium neutron multiplier in the ITER breeding blanket will generate tritium through transmutations. That tritium constitutes a safety hazard. Experiments evaluating tritium storage and release mechanisms have shown that most of the tritium comes out in a burst during thermal ramping. A small fraction of retained tritium is released by thermally activated processes. Analysis of recent experimental data shows that most of the tritium resides in helium bubbles. That tritium is released when the bubbles undergo swelling sufficient to develop porosity that connects with the surface. That appears to occur when swelling reaches about 10--15%. Other tritium appears to be stored chemically at oxide inclusions, probably as Be(OT) 2 . That component is released by thermal activation. There is considerable variation in published values for tritium diffusion through the beryllium and solubility in it. Data from experiments using highly irradiated beryllium from the Idaho National Engineering Laboratory showed diffusivity generally in line with the most commonly accepted values for fully dense material. Lower density material, planned for use in the ITER blanket may have very short diffusion times because of the open structure. The beryllium multiplier of the ITER breeding blanket was analyzed for tritium release characteristics using temperature and helium production figures at the midplane generated in support of the ITER Summer Workshop, 1990 in Garching. Ordinary operation, either in Physics or Technology phases, should not result in the release of tritium trapped in the helium bubbles. Temperature excursions above 600 degree C result in large-scale release of that tritium. 29 refs., 10 figs., 3 tabs

  3. Beryllium application in ITER plasma facing components

    International Nuclear Information System (INIS)

    Raffray, A.R.; Federici, G.; Barabash, V.; Cardella, A.; Jakeman, R.; Ioki, K.; Janeschitz, G.; Parker, R.; Tivey, R.; Pacher, H.D.; Wu, C.H.; Bartels, H.W.

    1997-01-01

    Beryllium is a candidate armour material for the in-vessel components of the International Thermonuclear Experimental Reactor (ITER), namely the primary first wall, the limiter, the baffle and the divertor. However, a number of issues arising from the performance requirements of the ITER plasma facing components (PFCs) must be addressed to better assess the attractiveness of Be as armour for these different components. These issues include heat loading limits arising from temperature and stress constraints under steady state conditions, armour lifetime including the effects of sputtering erosion as well as vaporisation and loss of melt during disruption events, tritium retention and permeation, and chemical hazards, in particular with respect to potential Be/steam reaction. Other issues such as fabrication and the possibility of in-situ repair are not performance-dependent but have an important impact on the overall assessment of Be as PFC armour. This paper describes the present view on Be application for ITER PFCs. The key issues are discussed including an assessment of the current level of understanding based on analysis and experimental data; and on-going activities as part of the ITER EDA R and D program are highlighted. (orig.)

  4. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    Energy Technology Data Exchange (ETDEWEB)

    Ulrickson, M.A. [ed.] [Sandia National Labs., Albuquerque, NM (United States); Manly, W.D. [Oak Ridge National Lab., TN (United States); Dombrowski, D.E. [Brush Wellman, Inc., Cleveland, OH (United States)] [and others

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers.

  5. Report of a technical evaluation panel on the use of beryllium for ITER plasma facing material and blanket breeder material

    International Nuclear Information System (INIS)

    Ulrickson, M.A.; Manly, W.D.; Dombrowski, D.E.

    1995-08-01

    Beryllium because of its low atomic number and high thermal conductivity, is a candidate for both ITER first wall and divertor surfaces. This study addresses the following: why beryllium; design requirements for the ITER divertor; beryllium supply and unirradiated physical/mechanical property database; effects of irradiation on beryllium properties; tritium issues; beryllium health and safety; beryllium-coolant interactions and safety; thermal and mechanical tests; plasma erosion of beryllium; recommended beryllium grades for ITER plasma facing components; proposed manufacturing methods to produce beryllium parts for ITER; emerging beryllium materials; proposed inspection and maintenance techniques for beryllium components and coatings; time table and costs; and the importance of integrating materials and manufacturing personnel with designers

  6. Characterization of plasma sprayed beryllium ITER first wall mockups

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R.G.; Vaidya, R.U.; Hollis, K.J. [Los Alamos National Lab., NM (United States). Material Science and Technology Div.

    1998-01-01

    ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/m{sup 2} without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface. (author)

  7. Characterization of Plasma Sprayed Beryllium ITER First Wall Mockups

    International Nuclear Information System (INIS)

    Castro, Richard G.; Vaidya, Rajendra U.; Hollis, Kendall J.

    1997-10-01

    ITER first wall beryllium mockups, which were fabricated by vacuum plasma spraying the beryllium armor, have survived 3000 thermal fatigue cycles at 1 MW/sq m without damage during high heat flux testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. The thermal and mechanical properties of the plasma sprayed beryllium armor have been characterized. Results are reported on the chemical composition of the beryllium armor in the as-deposited condition, the through thickness and normal to the through thickness thermal conductivity and thermal expansion, the four-point bend flexure strength and edge-notch fracture toughness of the beryllium armor, the bond strength between the beryllium armor and the underlying heat sink material, and ultrasonic C-scans of the Be/heat sink interface

  8. Erosion of beryllium under ITER - Relevant transient plasma loads

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Kurbatova, L. A.; Porezanov, N. P.; Podkovyrov, V. L.; Muzichenko, A. D.; Zhitlukhin, A. M.; Gervash, A. A.; Safronov, V. M.

    2015-08-01

    Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2-0.5 MJ/m2 at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology.

  9. Beryllium-copper reactivity in an ITER joining environment

    International Nuclear Information System (INIS)

    Odegard, B.C.; Cadden, C.H.; Yang, N.Y.C.

    1998-01-01

    Beryllium-copper reactivity was studied using test parameters being considered for use in the ITER reactor. In this application, beryllium-copper tiles are produced using a low-temperature copper-copper diffusion bonding technique. Beryllium is joined to copper by first plating the beryllium with copper followed by diffusion bonding the electrodeposited (ED) copper to a wrought copper alloy (CuNiBe) at 450 C, 1-3 h using a hot isostatic press (HIP). In this bonded assembly, beryllium is the armor material and the CuNiBe alloy is the heat sink material. Interface temperatures in service are not expected to exceed 350 C. For this study, an ED copper-beryllium interface was subjected to diffusion bonding temperatures and times to study the reaction products. Beryllium-copper assemblies were subjected to 350, 450 and 550 C for times up to 200 h. Both BeCu and Be 2 Cu intermetallic phases were detected using scanning electron microscopy and quantitative microprobe analysis. Growth rates were determined experimentally for each phase and activation energies for formation were calculated. The activation energies were 66 mol and 62 kJ mol -1 for the BeCu and Be 2 Cu, respectively. Tensile bars were produced from assemblies consisting of coated beryllium (both sides) sandwiched between two blocks of Hycon-3. Tensile tests were conducted to evaluate the influence of these intermetallics on the bond strength. Failure occurred at the beryllium-copper interface at fracture strengths greater than 300 MPa for the room-temperature tests. (orig.)

  10. Erosion of beryllium under ITER – Relevant transient plasma loads

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Nikolaev, G.N.; Kurbatova, L.A.; Porezanov, N.P.; Podkovyrov, V.L.; Muzichenko, A.D.; Zhitlukhin, A.M.; Gervash, A.A.; Safronov, V.M.

    2015-01-01

    Highlights: • We study the erosion, mass loss/gain and surface structure evolution of Be/CuCrZr mock-ups, armored with beryllium of TGP-56FW grade after irradiation by deuterium plasma heat load of 0.5 MJ/m 2 at 250 °C and 500 °C. • Beryllium mass loss/erosion under plasma heat load at 250 °C is rather small (no more than 0.2 g/m 2 shot and 0.11 μm/shot, correspondingly, after 40 shots) and tends to decrease with increasing number of shots. • Beryllium mass loss/erosion under plasma heat load at 500 °C is much higher (∼2.3 g/m 2 shot and 1.2 μm/shot, correspondingly, after 10 shot) and tends to decrease with increasing the number of shots (∼0.26 g/m 2 pulse and 0.14 μm/shot, correspondingly, after 100 shot). • Beryllium erosion value derived from the measurements of profile of irradiated surface is much higher than erosion value derived from mass loss data. - Abstract: Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2–0.5 MJ/m 2 at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology

  11. Erosion of beryllium under ITER – Relevant transient plasma loads

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.B., E-mail: igkupr@gmail.com [A.A. Bochvar High Technology Research Institute of Inorganic Materials, Rogova St. 5a, 123060 Moscow (Russian Federation); Nikolaev, G.N.; Kurbatova, L.A.; Porezanov, N.P. [A.A. Bochvar High Technology Research Institute of Inorganic Materials, Rogova St. 5a, 123060 Moscow (Russian Federation); Podkovyrov, V.L.; Muzichenko, A.D.; Zhitlukhin, A.M. [TRINITI, Troitsk, Moscow reg. (Russian Federation); Gervash, A.A. [Efremov Research Institute, S-Peterburg (Russian Federation); Safronov, V.M. [Project Center of ITER, Moscow (Russian Federation)

    2015-08-15

    Highlights: • We study the erosion, mass loss/gain and surface structure evolution of Be/CuCrZr mock-ups, armored with beryllium of TGP-56FW grade after irradiation by deuterium plasma heat load of 0.5 MJ/m{sup 2} at 250 °C and 500 °C. • Beryllium mass loss/erosion under plasma heat load at 250 °C is rather small (no more than 0.2 g/m{sup 2} shot and 0.11 μm/shot, correspondingly, after 40 shots) and tends to decrease with increasing number of shots. • Beryllium mass loss/erosion under plasma heat load at 500 °C is much higher (∼2.3 g/m{sup 2} shot and 1.2 μm/shot, correspondingly, after 10 shot) and tends to decrease with increasing the number of shots (∼0.26 g/m{sup 2} pulse and 0.14 μm/shot, correspondingly, after 100 shot). • Beryllium erosion value derived from the measurements of profile of irradiated surface is much higher than erosion value derived from mass loss data. - Abstract: Beryllium will be used as a armor material for the ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of the ITER first wall. This paper presents the results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility. The Be/CuCrZr mock-ups were exposed to up to 100 shots by deuterium plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat loads range of 0.2–0.5 MJ/m{sup 2} at different temperature of beryllium tiles. The temperature of Be tiles has been maintained about 250 and 500 °C during the experiments. After 10, 40 and 100 shots, the beryllium mass loss/gain under erosion process were investigated as well as evolution of surface microstructure and cracks morphology.

  12. The development of beryllium plasma spray technology for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Castro, R.G.; Elliott, K.E.; Hollis, K.J.; Watson, R.D.

    1999-01-01

    Over the past five years, four international parties, which include the European Communities, Japan, the Russian Federation and the United States, have been collaborating on the design and development of the International Thermonuclear Experimental Reactor (ITER), the next generation magnetic fusion energy device. During the ITER Engineering Design Activity (EDA), beryllium plasma spray technology was investigated by Los Alamos National Laboratory as a method for fabricating and repairing and the beryllium first wall surface of the ITER tokamak. Significant progress has been made in developing beryllium plasma spraying technology for this application. Information will be presented on the research performed to improve the thermal properties of plasma sprayed beryllium coatings and a method that was developed for cleaning and preparing the surface of beryllium prior to depositing plasma sprayed beryllium coatings. Results of high heat flux testing of the beryllium coatings using electron beam simulated ITER conditions will also be presented

  13. Fluid simulation of beryllium transport in the ITER gaseous divertor

    International Nuclear Information System (INIS)

    Knoll, D.A.; Campbell, R.B.; McHugh, P.R.

    1994-01-01

    The transport of either intrinsic or injected impurities will play a crucial role in the energy loss mechanisms in the ITER gaseous/cold plasma target divertor. Both 1-D and 2-D multi-charge state fluid codes are used to model the transport of beryllium in the ITER SOL. Our major conclusion is that in order to model the containment of impurities, the background flow field must be known in detail. Comparing 1-D and 2-D solutions, hydrogen flow reversal plays an important role in the entrainment process. Further, the flow of particles from the core plasma also has a strong impact on the resultant entrainment of the impurities in both 1-D and 2-D. It is imperative that those components of poloidal velocity due to E x B and diamagnetic drifts be included in the models. (orig.)

  14. Reducing the cost of S-65C grade beryllium for ITER first wall applications

    International Nuclear Information System (INIS)

    Kaczynski, D.; Sato, K.; Savchuk, V.V.; Shestakov, V.P.

    2004-01-01

    Beryllium is the current material of choice for plasma-facing components in ITER. The present design is for 10 mm thick beryllium tiles bonded to an actively cooled copper substrate. Brush Wellman grade S65C beryllium is preferred grade off beryllium for these tiles. S65C has the best resistance to low-cycle thermal fatigue than any other beryllium grad in the world. S65C grade beryllium has been successfully deployed in fusion reactors for more than two decades, most recently in the JET reactor. This paper will detail a supply chain to produce the most cost-effective S65C plasma facing components for ITER. This paper will also propose some future work too demonstrate the best technology for bonding beryllium to copper. (author)

  15. Modelling of radiation impact on ITER Beryllium wall

    Science.gov (United States)

    Landman, I. S.; Janeschitz, G.

    2009-04-01

    In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.

  16. Modelling of radiation impact on ITER Beryllium wall

    Energy Technology Data Exchange (ETDEWEB)

    Landman, I.S. [Forschungszentrum Karlsruhe, IHM, FUSION, P.O. Box 3640, 76021 Karlsruhe (Germany)], E-mail: igor.landman@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, IHM, FUSION, P.O. Box 3640, 76021 Karlsruhe (Germany)

    2009-04-30

    In the ITER H-Mode confinement regime, edge localized instabilities (ELMs) will perturb the discharge. Plasma lost after each ELM moves along magnetic field lines and impacts on divertor armour, causing plasma contamination by back propagating eroded carbon or tungsten. These impurities produce enhanced radiation flux distributed mainly over the beryllium main chamber wall. The simulation of the complicated processes involved are subject of the integrated tokamak code TOKES that is currently under development. This work describes the new TOKES model for radiation transport through confined plasma. Equations for level populations of the multi-fluid plasma species and the propagation of different kinds of radiation (resonance, recombination and bremsstrahlung photons) are implemented. First simulation results without account of resonance lines are presented.

  17. Movement of liquid beryllium during melt events in JET with ITER-like wall

    International Nuclear Information System (INIS)

    Sergienko, G; Huber, A; Brezinsek, S; Coenen, J W; Mertens, Ph; Philipps, V; Samm, U; Arnoux, G; Matthews, G F; Nunes, I; Riccardo, V; Sirinelli, A; Devaux, S

    2014-01-01

    The ITER-like wall recently installed in JET comprises solid beryllium limiters and a combination of bulk tungsten and tungsten-coated carbon fibre composite divertor tiles without active cooling. During a beryllium power handling qualification experiment performed in limiter configuration with 5 MW neutral beam injection input power, accidental beryllium melt events, melt layer motion and splashing were observed locally on a few beryllium limiters in the plasma contact areas. The Lorentz force is responsible for the observed melt layer movement. To move liquid beryllium against the gravity force, the current flowing from the plasma perpendicularly to the limiter surface must be higher than 6 kA m −2 . The thermo-emission current at the melting point of beryllium is much lower. The upward motion of the liquid beryllium against gravity can be due to a combination of the Lorentz force from the secondary electron emission and plasma pressure force. (paper)

  18. Testing of beryllium marker coatings in PISCES-B for the JET ITER-like wall

    International Nuclear Information System (INIS)

    Widdowson, A.; Baldwin, M.J.; Coad, J.P.; Doerner, R.P.; Hanna, J.; Hole, D.E.; Matthews, G.F.; Rubel, M.; Seraydarian, R.; Xu, H.

    2009-01-01

    Beryllium has been chosen as the first wall material for ITER. In order to understand the issues of material migration and tritium retention associated with the use of beryllium, a largely beryllium first wall will be installed in JET. As part of the JET ITER-like wall, beryllium tiles with marker coatings are proposed as a diagnostic tool for studying the erosion and deposition of beryllium around the vessel. The nominal structure for these coatings is a ∼10 μm beryllium surface layer separated from the beryllium tile by a 2-3 μm metallic inter-layer. Two types of coatings are tested here; one with a nickel inter-layer and one with a copper/beryllium mixed inter-layer. The coating samples were deposited by DC magnetron sputtering at General Atomics and were exposed to deuterium plasma in PISCES-B. The results of this testing show that the beryllium/nickel marker coating would be suitable for installation in JET.

  19. Technical issues and solutions on ITER first wall beryllium application. Industrial viewpoint

    International Nuclear Information System (INIS)

    Iwadachi, T.; Uda, M.; Ito, M.; Miyakawa, M.; Ibuki, M.

    2004-01-01

    Beryllium is selected as reference armor material of ITER primary first wall and is joined to the copper alloy heat sink such as CuCrZr or Dispersion Strengthened Copper (DSCu) Various joining technologies have been successfully developed and the manufacturing possibilities of large size first wall panels with beryllium armor has been demonstrated. Based on such results, further technical improvement is needed to reduce manufacturing cost and ensure the reliability of joining in actual size first wall. The technical issues to optimize the fabrication process of beryllium attachment were shown in this paper from an industrial point of view. Determination of the optimum size and the surface qualities of beryllium tiles are important issues in term of the material specification to ensure joining reliability and to reduce cost. The consolidation method and the finish machining methods of beryllium tiles are also critical in terms of material cost. These items should be determined by paying concern to the accommodation of the joining methods. The selections of slitting methods for attached beryllium have a great influence on fabrication cost. In the actual fabrication of beryllium attachment, safety provisions for exposure to beryllium in working environment and the recycling of the waste from the fabrication processes will be concerned sufficiently. (author)

  20. Experimental Simulation of Beryllium Armour Damage Under ITER-like Intense Transient Plasma Loads

    Energy Technology Data Exchange (ETDEWEB)

    Kupriyanov, I.; Basaleev, E.; Nikolaev, G.; Kurbatova, L., E-mail: igkupr@gmail.com [A.A. Bochvar High Technology Research Institute of Inorganic Material, Moscow (Russian Federation); Podkovyrov, V.; Zhitlukhin, A. [SRC RF TRINITI, Troitsk (Russian Federation); Khimchenko, L. L. [Project Centre of ITER, Moscow (Russian Federation)

    2012-09-15

    Full text: Beryllium will be used as a plasma facing material in the next generation of tokamaks such as ITER. During plasma operation in ITER, the plasma facing materials and components will be suffered by different kinds of loading which may affect their surface or their joint to the heat sink. In addition to quasi-stationary loadings which are caused by the normal cycling operation, the plasma facing components and materials may also be exposed to the intense short transient loads like disruptions, ELMs. All these events may lead to beryllium surface melting, cracking, evaporation and erosion. It is expected that the erosion of beryllium under transient plasma loads such as ELMs and disruptions will mainly determine a lifetime of ITER first wall. To obtain the experimental data for the evaluation of the beryllium armor lifetime and dust production under ITER-relevant transient loads, the advanced plasma gun QSPA-Be facility has been constructed in Bochvar Institute. This paper presents recent results of the experiments with Russian beryllium of TGP-56FW ITER grade. The mock-ups of a special design armored with two beryllium targets (80 x 80 x 10 mm{sup 3}) were tested by hydrogen plasma streams (5 cm in diameter) with pulse duration of 0.5 ms and heat load of 0.5 and 1.0 MJ/m{sup 2}. Experiments were performed at RT temperature. The evolution of surface microstructure and profile, cracks morphology and mass loss/gain under erosion process on the beryllium surface exposed to up to 250 shots will be presented and discussed. (author)

  1. Development and experimental study of beryllium window for ITER radial X-ray camera

    Energy Technology Data Exchange (ETDEWEB)

    Chen, Zhaoxi [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Jin, Guangxu [Materion Brush (United States); Chen, Kaiyun; Chen, Yebin; Song, Yuntao [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Hu, Liqun, E-mail: lqhu@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Niu, Luying; Sheng, Xiuli; Cheng, Yong; Lu, Kun [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2013-12-15

    Highlights: • The thickness of the beryllium foil is chosen as 80 μm to guarantee its safety under high pressure differential in accident events. • Using low purity of beryllium as the transition material, the effect of thermal stress caused by diffusion bonding process can be reduced. • Sealing ring and honeycomb-like supports are designed and used in the mechanical clamped beryllium window to enhance its sealing and safety performance. • The beryllium windows have good performance under severe working conditions like high temperature baking, vibration or impact load. -- Abstract: Radial X-ray camera (RXC) is a diagnostic device planned to be installed in the ITER Equatorial Port no. 12. Beryllium window will be installed between the inner and outer camera of RXC, which severs as the transmission photocathode substrate and also the vacuum isolation component. In this paper the design and manufacture process of two types of beryllium windows were introduced. Although 50 μm thickness of beryllium foil is the best choice, the 80 μm one with X-ray threshold of 1.34 keV was selected for safety consideration. Using the intermediate layer (low purity of beryllium) between the beryllium foil and the stainless steel base flange is an effective strategy to limit the welding thermal deformation and thermal stress of the thin foil caused by bonding between different materials. By using ANSYS software, the feasibility of the aperture design was analyzed and validated. Metal sealing ring was applied in the mechanical clamped beryllium window for its good stability under high temperature and neutron radiation. Although both of the hollow metal sealing ring with 0.03 mm silver coating and the pure silver sealing ring can satisfy the sealing requirement, the later one was chosen to produce the final product. Two hours 240 °C high temperature baking test, two hours 3.3 Hz vibration test and fatigue test were performed on the two types of beryllium windows. Based on the

  2. Fabrication of ITER first wall mock-ups with beryllium armour

    International Nuclear Information System (INIS)

    Mohri, K.; Nomoto, Y.; Uda, M.; Enoeda, M.; Akiba, M.

    2004-01-01

    This paper presents the fabric ability development for the ITER first wall through the fabrication of a real size first wall panel mock-up without beryllium armor and a partial mock-up of the first wall panel with beryllium armor. Microscopic observation and mechanical test of the hot isostatic pressed Be/Cu-alloy joints were also performed of which results showed good bond ability of the joints. Finally the fabrication procedure of the ITER first wall panel has been established. (author)

  3. Beryllium assessment and recommendation for application in ITER plasma facing components

    Energy Technology Data Exchange (ETDEWEB)

    Barabash, V.; Tanaka, S.; Matera, R. [ITER Joint Central Team, Muenchen (Germany)

    1998-01-01

    The design status of the ITER Plasma Facing Components (PFC) is presented. The operational conditions of the armour material for the different components are summarized. Beryllium is the reference armour material for the Primary Wall, Baffle and Limiter and the back-up material for the Divertor Dome. The activities on the selection of the Be grades and the joining technologies are reviewed. (author)

  4. Status of RF beryllium characterization for ITER Fist Wall

    Science.gov (United States)

    Kupriyanov, I. B.; Nikolaev, G. N.; Roedig, M.; Gervash, A. А.; Linke, I. J.; Kurbatova, L. A.; Perevalov, S. I.; Giniyatulin, R. N.

    2011-10-01

    The status of RF R&D activities in production and characterization of TGP-56FW beryllium grade is presented. The results of metallographic studies of microstructure and cracks morphology are reported for full-scale Be tiles (56 × 56 × 10 mm) subjected to VDE simulation tests in TSEFEY-M testing facility (VDE-10 MJ/m 2 during 0.1 s, 1 shot ) and following low cycle thermal fatigue tests (500 thermal cycles at 1.5 MW/m 2). First results of plasma disruption tests ( E = 1.2-5 MJ/m 2, 5 ms), which were obtained during the realization of Thermal Shock/VDE Qualification program of RF beryllium in JUDITH-1 facility, are also discussed.

  5. Status of RF beryllium characterization for ITER Fist Wall

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Nikolaev, G.N.; Roedig, M.; Gervash, A.A.; Linke, I.J.; Kurbatova, L.A.; Perevalov, S.I.; Giniyatulin, R.N.

    2011-01-01

    The status of RF R and D activities in production and characterization of TGP-56FW beryllium grade is presented. The results of metallographic studies of microstructure and cracks morphology are reported for full-scale Be tiles (56 x 56 x 10 mm) subjected to VDE simulation tests in TSEFEY-M testing facility (VDE-10 MJ/m 2 during 0.1 s, 1 shot) and following low cycle thermal fatigue tests (500 thermal cycles at 1.5 MW/m 2 ). First results of plasma disruption tests (E = 1.2-5 MJ/m 2 , 5 ms), which were obtained during the realization of Thermal Shock/VDE Qualification program of RF beryllium in JUDITH-1 facility, are also discussed.

  6. Beryllium

    Science.gov (United States)

    Foley, Nora K.; Jaskula, Brian W.; Piatak, Nadine M.; Schulte, Ruth F.; Schulz, Klaus J.; DeYoung,, John H.; Seal, Robert R.; Bradley, Dwight C.

    2017-12-19

    Beryllium is a mineral commodity that is used in a variety of industries to make products that are essential for the smooth functioning of a modern society. Two minerals, bertrandite (which is supplied domestically) and beryl (which is currently supplied solely by imports), are necessary to ensure a stable supply of high-purity beryllium metal, alloys, and metal-matrix composites and beryllium oxide ceramics. Although bertrandite is the source mineral for more than 90 percent of the beryllium produced globally, industrial beryl is critical for the production of the very high purity beryllium metal needed for some strategic applications. The current sole domestic source of beryllium is bertrandite ore from the Spor Mountain deposit in Utah; beryl is imported mainly from Brazil, China, Madagascar, Mozambique, and Portugal. High-purity beryllium metal is classified as a strategic and critical material by the Strategic Materials Protection Board of the U.S. Department of Defense because it is used in products that are vital to national security. Beryllium is maintained in the U.S. stockpile of strategic materials in the form of hot-pressed beryllium metal powder.Because of its unique chemical properties, beryllium is indispensable for many important industrial products used in the aerospace, computer, defense, medical, nuclear, and telecommunications industries. For example, high-performance alloys of beryllium are used in many specialized, high-technology electronics applications, as they are energy efficient and can be used to fabricate miniaturized components. Beryllium-copper alloys are used as contacts and connectors, switches, relays, and shielding for everything from cell phones to thermostats, and beryllium-nickel alloys excel in producing wear-resistant and shape-retaining high-temperature springs. Beryllium metal composites, which combine the fabrication ability of aluminum with the thermal conductivity and highly elastic modulus of beryllium, are ideal for

  7. Beryllium

    International Nuclear Information System (INIS)

    1988-01-01

    In this data sheet the occurrence, ore processing, chemical and physical properties and the uses of beryllium and its alloys is presented. The hazards involved in the use of beryllium and its compounds in the laboratory are discussed with particular reference to its toxicity, carcinogenicity, handling, storage, disposal, fire prevention and the principal health hazards. Further reading is suggested. (UK)

  8. Beryllium mock-ups development and ultrasonic testing for ITER divertor conditions

    International Nuclear Information System (INIS)

    Barabash, V.R.; Bykov, V.A.; Giniyatulin, R.N.; Gervash, A.A.; Gurieva, T.M.; Egorov, K.E.; Komarov, V.L.; Korolkov, M.D.; Mazul, I.V.; Gitarsky, L.S.; Strulia, I.L.; Sizenev, V.S.; Pronyakin, V.T.

    1995-01-01

    At the present time beryllium is considered as the most suitable armour material for the ITER divertor application. Different types of Be-divertor mock-up construction are compared in the report. Two different technologies of beryllium tiles joining to a heat sink body are analysed: high temperature brazing and thermodiffusion bonding. The comparative analysis of different constructions has been performed on the basis of 2-D finite element calculation for temperatures and stresses. The main parameters and diagnostic capabilities of electron beam facility for HHF testing of beryllium mock-ups are described. The first results of HHF tests of ''beryllium-copper saddle-MAGT tube'' and ''beryllium-copper plate-SS body'' mock-ups are presented. The reasons of the damages during the HHF are analysed. The technique of ultrasonic testing of the thermodifussion bonding and brazing quality for beryllium-copper joints is presented. The recorded results are prepared in the form of ultrasound grams. The testing results are compared with the metallographic analysis. (orig.)

  9. Summary of beryllium qualification activity for ITER first-wall applications

    International Nuclear Information System (INIS)

    Barabash, V; Eaton, R; Hirai, T; Kupriyanov, I; Nikolaev, G; Wang Zhanhong; Liu Xiang; Roedig, M; Linke, J

    2011-01-01

    Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

  10. Summary of beryllium qualification activity for ITER first-wall applications

    Science.gov (United States)

    Barabash, V.; Eaton, R.; Hirai, T.; Kupriyanov, I.; Nikolaev, G.; Wang, Zhanhong; Liu, Xiang; Roedig, M.; Linke, J.

    2011-12-01

    Beryllium is considered as an armor material for the ITER first wall. The ITER Final Design Report 2001 identified the reference grades S-65C vacuum hot pressed (VHP) from Brush Wellman and DShG-200 from the Russian Federation. These grades have been selected based on excellent thermal fatigue/shock behavior and the available comprehensive database. Later, Chinese and Russian ITER Parties proposed their new grades: CN-G01 (from China) and TGP-56FW (from Russia). To assess the performance of these new grades, the ITER Organization, Chinese and Russian Parties established a program for the characterization of these materials. A summary of the published data and new results are presented in the paper. It was concluded that the proposed Chinese (CN-G01) and Russian (TGP-56FW) beryllium grades can be accepted. Three grades of beryllium are now available for the armor application for the ITER first wall: S-65, CN-G01 and TGP-56FW.

  11. Beryllium

    International Nuclear Information System (INIS)

    Hansen, N.B.

    1980-01-01

    A method for determination of beryllium in minerals and rocks is described. The method comprises microanalysis and trace analysis. Because of the toxidity of beryllium the method is designed for determination of a hitherto unknown small amount, 1-10 nanogram Be. With the optimal amount for determination, 3 ng Be, the relative error of the method is 10%. The description includes an inventory of chemicals and apparatus, also an example of application of the method on the mineral epididymite. In brief, the sample is melted with sodium carbonate and sodium tetra borate; when required the sample in advance is fumed with hydrogen fluoride and sulphuric acid to evaporate silica. The residuum is dissolved in water and hydrogen chloride, upon which the solution is made to volume. In the Ring oven interfering compounds are masked with EDTA. Beryllium is settled with chrome azurol and ammonia. Beryllium is identified and evaluated in comparison with previously produced standards. (author)

  12. Beryllium

    International Nuclear Information System (INIS)

    Hansen, N.B.

    1979-01-01

    A method for determination of beryllium in minerals and rocks is described. The method comprises microanalysis and trace analysis. Because of the toxidity of beryllium the method is designed for determination of a hitherto unknown small amount, 1-10 nanogram Be. With the optimal amount for determination, 3 ng Be, the relative error of the method is 10%. The description includes an inventory of chemicals and apparatus, also an example of application of the method on the mineral epididymite. In brief, the sample is melted with sodium carbonate and sodium tetra borate; when required the sample in advance is fumed with hydrogen fluoride and sulphuric acid to evaporate silica. The residuum is dissolved in water and hydrogen chloride, upon which the solution is made to volume. In the Ring oven interfering compounds are masked with EDTA. Beryllium is settled with chrome azurol and ammonia. Beryllium is identified and evaluated in comparison with previously produced standards. (author)

  13. Application of beryllium as first wall armour for ITER primary, baffle and limiter modules

    International Nuclear Information System (INIS)

    Cardella, A.; Barabash, V.; Ioki, K.; Yamada, M.; Mazul, I.; Merola, M.; Strebkov, Y.

    2000-01-01

    During the engineering design activities of the ITER project, beryllium has been selected as the armour material for the first wall of the primary, baffle and limiter blanket modules. These components have different requirements according to their function, so the armour design and its joining technology has been developed in order to withstand different operating and loading conditions. Extensive R and D has been performed to develop, select and characterise the beryllium material and the joining techniques. In parallel, beryllium plasma spray coating has been developed, mainly as a possible in situ repair method for locally damaged areas. For the reduced technical objectives / reduced cost (RTO/RC) ITER project, it is proposed to maintain Be as the reference armour material and to optimise the manufacturing technologies in order to minimise costs. The paper presents the rationale of the design choices for the application of beryllium to the blanket first wall and gives an overview of the R and D performed and the results achieved. (orig.)

  14. Recent results on high thermal energy load testing of beryllium for ITER first wall application

    Science.gov (United States)

    Kupriyanov, I. B.; Roedig, M.; Nikolaev, G. N.; Kurbatova, L. A.; Linke, J.; Gervash, A. A.; Giniyatulin, R. N.; Podkovyrov, V. L.; Muzichenko, A. D.; Khimchenko, L.

    2011-12-01

    In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m-2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m-2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m-2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm3) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m-2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

  15. Recent results on high thermal energy load testing of beryllium for ITER first wall application

    International Nuclear Information System (INIS)

    Kupriyanov, I B; Nikolaev, G N; Kurbatova, L A; Roedig, M; Linke, J; Gervash, A A; Giniyatulin, R N; Podkovyrov, V L; Muzichenko, A D; Khimchenko, L

    2011-01-01

    In this paper, progress in the high heat flux (HHF) qualification testing of TGP-56FW beryllium grade for ITER first wall applications is presented. Two actively cooled Be/CuCrZr brazing mock-ups were tested under complex thermal loading conditions in the electron beam facility JUDITH-1 (step 1: vertical displacement event test at 40 MJ m - 2, 0.3 s, 1 shot; step 2: disruption tests at 3 MJ m - 2, 1 shot, Δt=5 ms; step 3: repetitive fatigue test at 80 MW m - 2, 1000 shots, Δt=25 ms). After testing, metallographic investigations on the microstructure and crack morphology were carried out. The results of these studies of Be tiles are reported and discussed. The overall results of TGP-56FW grade qualification testing have demonstrated the reliable performance capability of TGP-56FW for application as the armor of the ITER first wall. In addition, the results of first experiments with TGP-56FW and S-65C beryllium grades in the QSPA-Be plasma gun facility are also reported. In these experiments, beryllium tiles (80×80×10 mm 3 ) were tested in a hydrogen plasma stream (5 cm in diameter) with pulse duration 0.5 ms and heat loads of 0.5-2 MJ m - 2. Experiments were performed at room temperature. The evolution of the surface microstructure and mass loss of beryllium exposed to up to 100 shots is presented.

  16. Progress in production and characterization of RF beryllium for ITER fist wall

    International Nuclear Information System (INIS)

    Kupriyanov, I.B.; Nikolaev, G.N.; Khomutov, A.M.; Gervash, A.; Romanov, P.V.

    2007-01-01

    Full text of publication follows: Beryllium was chosen as the armor material for the Fist Wall and Port Limiter of ITER. As results of ITER EDA activities the reference beryllium grade S-65C (USA, Brush Wellman) and the DshG-200 grade (RF, Bochvar Institute) were selected as materials for armor of ITER FW. The main criterion of a choice of given grades was their highest resistance against High Heat Flux loading shown by results of thermal fatigue/shock test experiments carried out earlier in SNL (USA) and FZ Juelich (Germany). Recently in RF the new Be grade TGP-56FW has been proposed, which is a modified analogue of DshG-200 grade with improved properties, and also a low-waste manufacturing process for production of Be tiles for armor of ITER FW has been developed. In comparison with DshG-200 grade the main advantage of TGP-56FW grade consists in higher level of strength properties and plasticity as well as higher uniformity of physical-mechanical properties in transverse and longitudinal directions relatively pressing axis. In this paper the results of RF R and D activities in production and characterization of TGP- 56FW beryllium grade for FW armor are presented. The results of HHF tests (thermal fatigue/disruption simulation) on TSEFEY-M testing facility for full-scale beryllium tiles produced from TGP-56FW grade are also reported. A new in-pile integrated test experiment for Be mock-ups that combines thermal cycling (2x10 4 cycles at 0.5 MW/m 2 ) and neutron irradiation (about 0.6 dpa) is discussed.(authors)

  17. Vacuum hot-pressed beryllium and TiC dispersion strengthened tungsten alloy developments for ITER and future fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Xiang, E-mail: xliu@swip.ac.cn [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Chen, Jiming; Lian, Youyun; Wu, Jihong; Xu, Zengyu; Zhang, Nianman; Wang, Quanming; Duan, Xuro [Southwestern Institute of Physics, P.O. Box 432, Chengdu 610041, Sichuan (China); Wang, Zhanhong; Zhong, Jinming [Northwest Rare Metal Material Research Institute, CNMC, Ningxia Orient Group Co. Ltd.,No.119 Yejin Road, Shizuishan City, Ningxia,753000 (China)

    2013-11-15

    Beryllium and tungsten have been selected as the plasma facing materials of the ITER first wall (FW) and divertor chamber, respectively. China, as a participant in ITER, will share the manufacturing tasks of ITER first-wall mockups with the European Union and Russia. Therefore ITER-grade beryllium has been developed in China and a kind of vacuum hot-pressed (VHP) beryllium, CN-G01, was characterized for both physical, and thermo-mechanical properties and high heat flux performance, which indicated an equivalent performance to U.S. grade S-65C beryllium, a reference grade beryllium of ITER. Consequently CN-G01 beryllium has been accepted as the armor material of ITER-FW blankets. In addition, a modification of tungsten by TiC dispersion strengthening was investigated and a W–TiC alloy with TiC content of 0.1 wt.% has been developed. Both surface hardness and recrystallization measurements indicate its re-crystallization temperature approximately at 1773 K. Deuterium retention and thermal desorption behaviors of pure tungsten and the TiC alloy were also measured by deuterium ion irradiation of 1.7 keV energy to the fluence of 0.5–5 × 10{sup 18} D/cm{sup 2}; a main desorption peak at around 573 K was found and no significant difference was observed between pure tungsten and the tungsten alloy. Further characterization of the tungsten alloy is in progress.

  18. Beryllium and graphite performance in ITER during a disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Ehst, D.A.; Gahl, J.

    1993-09-01

    Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed

  19. Beryllium and graphite performance in ITER during a disruption

    International Nuclear Information System (INIS)

    Hassanein, A.; Ehst, D.A.; Gahl, J.

    1994-01-01

    Plasma disruptions are considered one of the most limiting factors for successful operation of magnetic fusion reactors. During a disruption, a sharp, rapid release of energy strikes components such as the divertor or limiter plates. Severe surface erosion and melting of these components may then occur. The amount of material eroded from both ablation and melting is important to the reactor design and component lifetime. The anticipated performance of both beryllium and graphite as plasma-facing materials during such abnormal events is analyzed and compared. Recent experimental data obtained with both plasma guns and electron beams are carefully evaluated and compared to results of analytical modeling, including vapor shielding effect. Initial results from plasma gun experiments indicate that the Be erosion rate is about five times larger than that for a graphite material under the same disruption conditions. Key differences between simulation experiments and reactor disruption on the net erosion rate, and consequently on the lifetime of the divertor plate, are discussed in detail. The advantages and disadvantages of Be over graphite as a divertor plasma-facing material are discussed. ((orig.))

  20. Status of the beryllium tile bonding qualification activities for the manufacturing of the ITER first wall

    International Nuclear Information System (INIS)

    Mitteau, Raphaël; Eaton, R.; Perez, G.; Zacchia, F.; Banetta, S.; Bellin, B.; Gervash, A.; Glazunov, D.; Chen, J.

    2015-01-01

    The preparation of the manufacturing of the ITER first wall involves a qualification stage. The qualification aims at demonstrating that manufacturers can deliver the needed reliability and quality for the beryllium to copper bond, before the manufacturing can commence. The qualification is done on semi-prototype, containing relevant features relative to the beryllium armour (about 1/6 of the panel size). The qualification is done by the participating parties, firstly by a manufacturing semi-prototype and then by testing it under heat flux. One semi-prototype is manufactured and is being tested, and further from other manufacturers are still to come. The qualification programme is accompanied by bond defect investigations, which aim at defining defect acceptance criteria. Qualification and defect acceptance programme are supported by thermal and stress analyses, with good agreement regarding the thermal results, and some insights about the governing factors to bond damage.

  1. Development of a copper alloy to beryllium HIP bonding technology for the ITER first wall

    International Nuclear Information System (INIS)

    Sherlock, P.; Peacock, A.T.; Mc Callum, A.D.

    2005-01-01

    The primary first wall (PFW) panels of the ITER blanket concept comprise a bi-metallic copper alloy/stainless steel water-cooled heatsink faced with a plasma facing material. Precipitation strengthened CuCrZr is one option for the copper alloy of the heatsink; beryllium, in the form of tiles is an option for the plasma facing material. Over recent years, the technology needed to HIP bond the beryllium tiles to CuCrZr alloy has been developed. This paper describes small samples and larger mock-ups produced during the development of this HIP bonding technology and outlines how structural analyses were used to gain an understanding of the bonding process and refine the design

  2. Behaviour of molten beryllium with ITER reference CFC SEPCARB NB31 under moisture

    Energy Technology Data Exchange (ETDEWEB)

    Lipa, M. [CEA/Cadarache-DSM/DRFC, Euratom-Association, Saint Paul Lez Durance F-13108 (France)], E-mail: manfred.lipa@cea.fr; Linke, J. [Forschungszentrum Juelich, Euratom-Association, Juelich D-52425 (Germany); Martin, G. [CEA/Cadarache-DSM/DRFC, Euratom-Association, Saint Paul Lez Durance F-13108 (France); Wessel, E. [Forschungszentrum Juelich, Euratom-Association, Juelich D-52425 (Germany)

    2007-10-15

    A dramatic exothermic reaction has been observed in Tore Supra with molten aluminium enclosed in a thermally loaded 2D carbon fibre reinforced carbon composite (CFC) tile, exposed after component removal to atmospheric moisture. This led to an important local destruction of the graphite matrix and fibre texture. Beryllium, as aluminium, a carbide forming metal, is foreseen for the ITER first wall armour. Since it can be expected that sputtered, molten and/or evaporated beryllium material will be transported to the hot CFC based divertor surface during normal and off-normal plasma operation, a preliminary reaction experiment has been performed in order to investigate to which extent the Be contaminated graphite matrix and fibre texture of 3D-CFC SEPCARB NB31 could be locally damaged ('erosion') following a subsequent long lasting air venting or a supposed in-vessel ingress of coolant (steam) event (ICE)

  3. Status of the beryllium tile bonding qualification activities for the manufacturing of the ITER first wall

    Energy Technology Data Exchange (ETDEWEB)

    Mitteau, Raphaël, E-mail: Raphael.mitteau@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Eaton, R.; Perez, G. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France); Zacchia, F.; Banetta, S.; Bellin, B. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Gervash, A.; Glazunov, D. [Efremov Research Institute, 189631 St. Petersburg (Russian Federation); Chen, J. [Southwestern Institute of Physics, Huangjing Road, Chengdu 610225 (China)

    2015-10-15

    The preparation of the manufacturing of the ITER first wall involves a qualification stage. The qualification aims at demonstrating that manufacturers can deliver the needed reliability and quality for the beryllium to copper bond, before the manufacturing can commence. The qualification is done on semi-prototype, containing relevant features relative to the beryllium armour (about 1/6 of the panel size). The qualification is done by the participating parties, firstly by a manufacturing semi-prototype and then by testing it under heat flux. One semi-prototype is manufactured and is being tested, and further from other manufacturers are still to come. The qualification programme is accompanied by bond defect investigations, which aim at defining defect acceptance criteria. Qualification and defect acceptance programme are supported by thermal and stress analyses, with good agreement regarding the thermal results, and some insights about the governing factors to bond damage.

  4. Failure analysis of beryllium tile assembles following high heat flux testing for the ITER program

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C. H.; Yang, N. Y. C.

    2000-01-01

    The following document describes the processing, testing and post-test analysis of two Be-Cu assemblies that have successfully met the heat load requirements for the first wall and dome sections for the ITER (International Thermonuclear Experimental Reactor) fusion reactor. Several different joint assemblies were evaluated in support of a manufacturing technology investigation aimed at diffusion bonding or brazing a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Judicious selection of materials and coatings for these assemblies was essential to eliminate or minimize interactions with the highly reactive beryllium armor material. A thin titanium layer was used as a diffusion barrier to isolate the copper heat sink from the beryllium armor. To reduce residual stresses produced by differences in the expansion coefficients between the beryllium and copper, a compliant layer of aluminum or aluminum-beryllium (AlBeMet-150) was used. Aluminum was chosen because it does not chemically react with, and exhibits limited volubility in, beryllium. Two bonding processes were used to produce the assemblies. The primary process was a diffusion bonding technique. In this case, undesirable metallurgical reactions were minimized by keeping the materials in a solid state throughout the fabrication cycle. The other process employed an aluminum-silicon layer as a brazing filler material. In both cases, a hot isostatic press (HIP) furnace was used in conjunction with vacuum-canned assemblies in order to minimize oxidation and provide sufficient pressure on the assemblies for full metal-to-metal contact and subsequent bonding. The two final assemblies were subjected to a suite of tests including: tensile tests and electron and optical metallography. Finally, high heat flux testing was conducted at the electron beam testing system (EBTS) at Sandia National Laboratories, New Mexico. Here, test mockups were fabricated and subjected to normal heat loads to

  5. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    Science.gov (United States)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2009-04-01

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  6. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany)], E-mail: bazylev@ihm.fzk.de; Janeschitz, G. [Forschungszentrum Karlsruhe, Fusion, P.O. Box 3640, 76021 Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [Forschungszentrum Karlsruhe, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

    2009-04-30

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  7. Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B., E-mail: boris.bazylev@kit.edu [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Igitkhanov, Yu.; Landman, I.; Pestchanyi, S. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, D-76021 Karlsruhe (Germany); Loarte, A. [ITER Organisation, Cadarache, 13108 Saint Paul Lez Durance Cedex (France)

    2011-10-01

    Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

  8. Erosion simulation of first wall beryllium armour after ITER transient heat loads and runaway electrons action

    International Nuclear Information System (INIS)

    Bazylev, B.; Igitkhanov, Yu.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2011-01-01

    Beryllium is foreseen as plasma facing armour for the first wall (FW) in ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) and runaway electrons impact are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting, evaporation, and melt motion, which determine the life-time of the plasma facing components. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the J x B forces are analyzed for bulk Be and different sizes of Be-brushes. The damage of the FW due to heat loads caused by runaway electrons is numerically simulated.

  9. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    International Nuclear Information System (INIS)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2009-01-01

    The beryllium is foreseen as plasma facing armour for the first wall in the ITER in form of Be-clad blanket modules in macrobrush design with brush size about 8-10 cm. In ITER significant heat loads during transient events (TE) are expected at the main chamber wall that may leads to the essential damage of the Be armour. The main mechanisms of metallic target damage remain surface melting and melt motion erosion, which determines the lifetime of the plasma facing components. Melting thresholds and melt layer depth of the Be armour under transient loads are estimated for different temperatures of the bulk Be and different shapes of transient loads. The melt motion damages of Be macrobrush armour caused by the tangential friction force and the Lorentz force are analyzed for bulk Be and different sizes of Be-brushes. The damage of FW under radiative loads arising during mitigated disruptions is numerically simulated.

  10. Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall

    Energy Technology Data Exchange (ETDEWEB)

    Riccardo, V., E-mail: valeria.riccardo@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Lomas, P.; Matthews, G.F. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Nunes, I. [Associação EURATOM-IST, IPFN – Laboratório Associado, IST, Lisbon (Portugal); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Thompson, V. [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); Villedieu, E. [CEA, IRFM, F-13108 Saint-Paul-lez-Durance (France); JET-EFDA, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom)

    2013-10-15

    Highlights: ► 40 m{sup 2} of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets.

  11. Design, manufacture and initial operation of the beryllium components of the JET ITER-like wall

    International Nuclear Information System (INIS)

    Riccardo, V.; Lomas, P.; Matthews, G.F.; Nunes, I.; Thompson, V.; Villedieu, E.

    2013-01-01

    Highlights: ► 40 m 2 of plasma facing surface covered with bulk Be re-using existing supports, designed for C-based tiles (hence for much lower disruption loads). ► Optimization of power handling to allow compatibility with higher (×1.5) and longer (×2) neutral beam power. ► Beryllium re-cycling. ► Machining and cleaning to ultra high vacuum standards of <350 μm thin castellations in Be. ► Quality control to minimize installation problems (proto-types, full scale jigs, inspections). -- Abstract: The aim of the JET ITER-like wall project was to provide JET with the plasma facing material combination now selected for the DT phase of ITER (bulk beryllium main chamber limiters and a full tungsten divertor) and, in conjunction with the upgraded neutral beam heating system, to achieve ITER relevant conditions. The design of the bulk Be plasma facing components had to be compatible with increased heating power and pulse length, as well as to reuse the existing tile supports originally designed to cope with disruption loads from carbon based tiles and be installed by remote handling. Risk reduction measures (prototypes, jigs, etc.) were implemented to maximize efficiency during the shutdown. However, a large number of clashes with existing components not fully captured by the configuration model occurred. Restarting the plasma on the ITER-like Wall proved much easier than for the carbon wall and no deconditioning by disruptions was observed. Disruptions have been more threatening than expected due to the reduced radiative losses compared to carbon, leaving most of the plasma magnetic energy to be conducted to the wall and requiring routine disruption mitigation. The main chamber power handling has achieved and possibly exceeded the design targets

  12. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    Energy Technology Data Exchange (ETDEWEB)

    Bazylev, B.; Janeschitz, G. [Forschungszentrum Karlsruhe GmbH, FZK, Karlsruhe (Germany); Landman, I.; Pestchanyi, S. [FZK-Forschungszentrum Karlsruhe, Association Euratom-FZK, Technik und Umwelt, Karlsruhe (Germany); Loarte, A. [EFDA Close Support Unit Garching, Garching bei Munchen(Germany)

    2007-07-01

    Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10{sup 4} ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m{sup 2} in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m{sup 2} and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m{sup 2} in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be

  13. Erosion simulation of first wall beryllium armour under ITER transient heat loads

    International Nuclear Information System (INIS)

    Bazylev, B.; Janeschitz, G.; Landman, I.; Pestchanyi, S.; Loarte, A.

    2007-01-01

    Full text of publication follows: Operation of ITER at high fusion gain is assumed to be the H-mode. A characteristic feature of this regime is the transient release of energy from the confined plasma onto divertor and the first wall by multiple ELMs (about 10 4 ELMs per ITER discharge), which can play a determining role in the erosion rate and lifetime of these components. It is expected that about 50-70 % of the ELM energy releases onto divertor armour and the rest is dumped onto the First Wall (FW) armour. The expected energy heat loads on the ITER divertor and FW during Type I ELM are in range 0.5 - 4 MJ/m 2 in timescales of 0.3-0.6 ms. In case of the ITER disruptions the material evaporated from the divertor expands into the SOL and generates significant radiation heating of the FW armour up to several GW/m2 during a few milliseconds that can also lead to the its melting and noticeable damage. Beryllium macro-brush armour (Be-brushes) is foreseen as plasma FW facing component (PFC) in ITER. During the intense transient events in ITER the surface melting, melt motion, melt splashing and evaporation are seen as the main mechanisms of Be-erosion. The expected erosion of the ITER plasma facing components under transient energy loads can be properly estimated by numerical simulations using the codes MEMOS and PHEMOBRID validated against experimental data obtained at the plasma gun facilities QSPA-T, MK-200UG and QSPA-Kh50 that provide a way to simulate the energy loads expected in ITER in laboratory experiments. The numerical simulations were carried out for the expected ITER ELMs for the heat loads in the range 0.5 - 3.0 MJ/m 2 and the timescale up 0.6 ms and ITER disruptions for the heat loads in the range 2 - 13 MJ/m 2 in timescales of 1-5 ms. Radiation heat loads at the FW armour from the vapour expanded into the SOL were calculated using the codes FOREV-2 and TOKES for both ITER ELM and ITER disruption scenarios. Melt layer damage of the Be FW macro

  14. Theoretical prediction of thermodynamic properties of tritiated beryllium molecules and application to ITER source term

    Energy Technology Data Exchange (ETDEWEB)

    Virot, F., E-mail: francois.virot@irsn.fr; Barrachin, M.; Souvi, S.; Cantrel, L.

    2014-10-15

    Highlights: • Standard enthalpies of formation of BeH, BeH{sub 2}, BeOH, Be(OH){sub 2} have been calculated. • The impact of hydrogen isotopy on thermodynamic properties has been shown. • Speciation in the vacuum vessel shows that the main tritiated species is tritiated steam. • Beryllium hydroxide and hydride could exist during an accidental event. - Abstract: By quantum chemistry calculations, we have evaluated the standard enthalpies of formation of some gaseous species of the Be-O-H chemical system: BeH, BeH{sub 2}, BeOH, Be(OH){sub 2} for which the values in the referenced thermodynamic databases (NIST-JANAF [1] or COACH [2]) were, due to the lack of experimental data, estimated or reported with a large uncertainty. Comparison between post-HF, DFT approaches and available experimental data allows validation of the ability of an accurate exchange-correlation functional, VSXC, to predict the thermo-chemical properties of the beryllium species of interest. Deviation of enthalpy of formation induced by changes in hydrogen isotopy has been also calculated. From these new theoretically determinated data, we have calculated the chemical speciation in conditions simulating an accident of water ingress in the vacuum vessel of ITER.

  15. Behaviour of Molten Beryllium with ITER Reference CFC NB31 (SNECMA) Under Moisture

    International Nuclear Information System (INIS)

    Lipa, M.; Martin, G.; Linke, J.

    2006-01-01

    A dramatic exothermic reaction with aluminium, a carbide forming metal, has been observed in Tore Supra. A small rod of 30 mm 3 , acting as a temperature proof, was enclosed in a blind hole of a thermally loaded low density PAN fiber CFC 1001Z block (SGL), which reached a temperature of about 1300 o C during plasma operation. The molten aluminium had penetrated the carbon matrix through to the block's front surface. After component removal and roughly 2 months of exposure to air in the laboratory, the CFC in front of the blind hole was found to have been locally destroyed over a crater-shaped structure of 2 cm diameter. This was due to an enhanced decomposition of aluminium carbide to aluminium hydroxide. Beryllium (Be), also a carbide forming metal, is used on the ITER first wall. Carbon reinforced carbon (CFC) of type NB31 (Snecma) covers the vertical divertor targets. It is expected that beryllium material will be transported during normal and/or off normal plasma operation to the carbon based divertor targets to form beryllium carbide. During air venting or a supposed accidental in-vessel water leak event, it will react exothermically under moisture to beryllium oxide. In order to investigate to which extent the CFC structure could be modified or eventually destroyed, this reaction process has been simulated with a CFC block NB31 of size 16 x 32 x 20 mm 3 , where about 40 mm 3 of Be S65 C (Brush Wellmann) has been placed in a previously drilled blind hole of 4 mm diameter oriented parallel to the high conductivity pitch fibers. When melted, by heating the CFC block, the Be penetrated in the carbon matrix through to the block's front surface. The front surface of the CFC was then exposed to humidity (tap water) for about 2 weeks and then stored for a further 2 months in a closed vinyl bag under atmospheric pressure after which the sample was analysed. After the exposure of the CFC to humidity, reaction products have been detected at the surface of the carbon

  16. Beryllium layer response to ITER-like ELM plasma pulses in QSPA-Be

    Directory of Open Access Journals (Sweden)

    N.S. Klimov

    2017-08-01

    Full Text Available Material migration in ITER is expected to move beryllium (Be eroded from the first wall primarily to the tungsten (W divertor region and to magnetically shadowed areas of the wall itself. This paper is concerned with experimental study of Be layer response to ELM-like plasma pulses using the new QSPA-Be plasma gun (SRC RF TRINITI. The Be layers (1→50µm thick are deposited on special castellated Be and W targets supplied by the ITER Organization using the Thermionic Vacuum Arc technique. Transient deuterium plasma pulses with duration ∼0.5ms were selected to provide absorbed energy densities on the plasma stream axis for a 30° target inclination of 0.2 and 0.5MJm−2, the first well below and the second near the Be melting point. This latter value is close to the prescribed maximum energy density for controlled ELMs on ITER. At 0.2MJm−2 on W, all Be layer thicknesses tested retain their integrity up to the maximum pulse number, except at local defects (flakes, holes and cracks and on tile edges. At 0.5MJm−2 on W, Be layer melting and melt layer agglomeration are the main damage processes, they happen immediately in the first plasma impact. Melt layer movement was observed only near plasma facing edges. No significant melt splashing is observed in spite of high plasma pressure (higher than expected in ITER. Be layer of 10µm thick on Be target has higher resistance to plasma irradiation than 1 and 55µm, and retain their integrity up to the maximum pulse number at 0.2MJm−2. For 1µm and 55µm thick on Be target significant Be layer losses were observed at 0.2MJm−2.

  17. European Fusion Programme. ITER task T23: Beryllium characterisation. Progress report. Tensile tests on neutron irradiated and reference beryllium

    International Nuclear Information System (INIS)

    Moons, F.

    1996-02-01

    As part of the European Technology Fusion Programme, the irradiation embrittlement characteristics of the more ductile and isotopic grades of beryllium manufactured by Brush Wellman has been investigated using modern powder production and consolidation techniques . This study was initiated in support of the development and evaluation of beryllium as a neutron multiplier for the solid breeder blanket design concepts proposed for a DEMO fusion power reactor. Four different species of beryllium: S-200 F (vacuum hot pressed, 1.2 wt% BeO), S-200FH (hot isostatic pressed, 0.9 wt% BeO), S-65 (vacuum hot pressed, 0.6 wt% BeO), S-65H (hot isostatic pressed, 0.5 wt% BeO) have been compared. Three batches of the beryllium have been investigated, a neutron batch, a thermal control batch and a reference batch. Neutron irradiation has been performed at temperatures between 175 and 605 degrees Celsius up to a neutron fluence of 2.1 10 25 n.m -2 (E> 1 MeV) or 750 appm He. The results of the tensile tests are summarized

  18. Analysis and design of the beryllium tiles for the JET ITER-like wall project

    International Nuclear Information System (INIS)

    Thompson, V.; Krivchenkov, Y.; Riccardo, V.; Vizvary, Z.

    2007-01-01

    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. Over 4000 tiles will be replaced and the ILW will accommodate additional heating up to at least 50 MW for 10 s. 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, optimising tile size and profile and introducing castellations on plasma facing surfaces. This paper describes the generic problems associated with the Be tiles (power handling capacity and disruption induced eddy currents) and illustrates the solution selected for the inner wall guard limiter (IWGL) where the present CFC tiles will be replaced with Be

  19. Thermal fatigue testing of a diffusion-bonded beryllium divertor mock-up under ITER relevant conditions

    International Nuclear Information System (INIS)

    Youchison, D.L.; Guiniiatouline, R.; Watson, R.D.

    1994-01-01

    Thermal response and thermal fatigue tests of four 5 mm thick beryllium tiles on a Russian divertor mock-up were completed on the Electron Beam Test System at Sandia National Laboratories. The beryllium tiles were diffusion bonded onto an OFHC copper saddleblock and a DSCu (MAGT) tube containing a porous coating. Thermal response tests were performed on the tiles to an absorbed heat flux of 5 MW/m 2 and surface temperatures near 300 degrees C using 1.4 MPa water at 5.0 m/s flow velocity and an inlet temperature of 8-15 degrees C. One tile was exposed to incrementally increasing heat fluxes up to 9.5 MW/m 2 and surface temperatures up to 690 degrees C before debonding at 10 MW/m 2 . A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m 2 , while another debonded after 6800 cycles. In all cases, fatigue failure occurred in the intermetallic layers between the beryllium and copper. No fatigue cracking of the bulk beryllium was observed. During thermal cycling, a gradual loss of porous coating produced increasing sample temperatures. These experiments indicate that diffusion-bonded beryllium tiles can survive several thousand thermal cycles under ITER relevant conditions without failure. However, the reliability of the diffusion bonded Joint remains a serious issue

  20. Thermal load testing of erosion-monitoring beryllium marker tile for the ITER-Like Wall Project at JET

    International Nuclear Information System (INIS)

    Hirai, T.; Linke, J.; Rubel, M.; Coad, J.P.; Likonen, J.; Lungu, C.P.; Matthews, G.F.; Philipps, V.; Wessel, E.

    2008-01-01

    ITER-Like Wall Project has been launched at JET in order to perform a fully integrated test of plasma-facing materials. During the next major shutdown a full metal wall will be installed: tungsten in the divertor and beryllium in the main chamber. Beryllium erosion is one of key issues to be addressed. Special marker tiles have been designed for this purpose. Test coupons of such markers have been manufactured and examined. The performance test under high power deposition was carried in the electron beam facility JUDITH. The results of material characterization before and after high heat flux loads are presented. The samples survived, without macroscopic damage, power loads of up to 4.5 MW/m 2 for 10 s (surface temperature ∼650 deg. C) and 50 cyclic loads at 3.5 MW/m 2 lasting 10 s each (surface temperature ∼600 deg. C)

  1. Thermal load resistance of erosion-monitoring beryllium maker tile for JET ITER like wall project

    International Nuclear Information System (INIS)

    Hirai, T.; Linke, J.; Sundelin, P.; Rubel, M.; Coad, J.P.; Matthews, G.F.; Lungu, C.P.

    2007-01-01

    The ITER reference materials, beryllium (Be), carbon fibre composite (CFC) and tungsten (W), have been tested separately in tokamaks. An integrated test demonstrating both compatibility of metal plasma facing components with high-power operation and acceptable tritium retention has not yet been carried out. At JET, the size, magnetic field strength and high plasma current allow to conducting tests with the combination of the materials. Thus, the ITER-like Wall (ILW) project has been launched. In the project, Be will be the plasmafacing material on the main chamber wall of JET. To assess the erosion of the Be tiles, a Be marker tile was proposed and designed. The test samples which simulate the JET Be marker tile have been produced in MEdC, Romania in order to study the thermal load resistance of the JET Be marker (20 x 20 mm 2 size with 30 mm height). The marker tile sample consists of bulk Be, high-Z interlayer (2-3 μm Ni coating) and 8-9 μm Be coating. Thermionic Vacuum Arc (TVA) techniques based on the electron-induced evaporation have been selected for this purpose. In the present work, the global characterization of the maker tile samples and thermal load tests were performed. After the pre-characterization (microstructure observation by scanning electron microscope and elemental analysis by means of Wavelength Dispersive X-ray Spectroscopy and Energy Dispersive X-ray Spectroscopy), the thermal loading tests were performed in the electron beam facility JUDITH. The coating consisted of tiny platelets of ∝0.1 um in diameter and localized larger platelets of 1 um in diameter. The surface and bulk temperature were observed during the tests. In the screening thermal load test, the samples were loaded to 6 MW/m 2 for 10 s. The layers did not show any macroscopic damages at up to 4.5 MW/m 2 for 10 s (45 MJ/m 2 ). However, the coating delaminated and the maker was damaged when the thermal loading reached at 5 MW/m 2 (∝50 MJ/m 2 ). Cyclic heat load tests were

  2. Investigation of high purity beryllium for the International Thermonuclear Experimental Reactor (ITER), Task 002. Final report

    International Nuclear Information System (INIS)

    Vagin, S.P.

    1995-05-01

    The report includes a description of experimental abilities of Solid Structure Research Laboratory of IAE NNC RK, a results of microstructural characterization of A-4 grade polycrystal Beryllium produced at the Ulba metal plant and a technical project-for irradiation experiments. Technical project contains a detailed description of five proposed experiments, clearing behavior of Beryllium materials under the influence of irradiation, temperature, helium and hydrogen accumulation. Complex irradiation jobs, microstructural investigations and mechanical tests are planned in the framework of these experiments

  3. Solid state bonding of beryllium-copper for an ITER first wall application

    Energy Technology Data Exchange (ETDEWEB)

    Odegard, B.C. Jr.; Cadden, C.H. [Sandia National Labs., Livermore, CA (United States)

    1998-01-01

    Several different joint assemblies were evaluated in support of a manufacturing technology for diffusion bonding a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Because beryllium reacts with all but a few elements to form intermetallic compounds, this study considered several different surface treatments as a means of both inhibiting these reactions and promoting a good diffusion bond between the two substrates. All diffusion bonded assemblies used aluminum or an aluminum-beryllium composite (AlBeMet-150) as the interfacial material in contact with beryllium. In most cases, explosive bonding was utilized as a technique for joining the copper alloy heat sink to an aluminum or AlBeMet-150 substrate, which was subsequently diffusion bonded to an aluminum coated beryllium tile. In this approach, a 250 {mu}m thick titanium foil was used as a diffusion barrier between the copper and aluminum to prevent the formation of Cu-Al intermetallic phases. In all cases, a hot isostatic pressing (HIP) furnace was used in conjunction with canned assemblies in order to minimize oxidation and apply sufficient pressure on the assembly for excellent metal-to-metal contact and subsequent bonding. Several different processing schedules were evaluated during the course of this study; bonded assemblies were produced that failed outside the bond area indicating a 100% joint efficiency. (author)

  4. Solid state bonding of beryllium-copper for an ITER first wall application

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.

    1998-02-01

    Several different joint assemblies were evaluated in support of a manufacturing technology for diffusion bonding a beryllium armor tile to a copper alloy heat sink for fusion reactor applications. Because beryllium reacts with all but a few elements to form intermetallic compounds, this study considered several different surface treatments as a means of both inhibiting these reactions and promoting a good diffusion bond between the two substrates. A diffusion bonded assemblies used aluminum or an aluminum-beryllium composite (AlBeMet-150) as the interfacial material in contact with beryllium. In most cases, explosive bonding was utilized as a technique for joining the copper alloy heat sink to an aluminum or AlBeMet-150 substrate, which was subsequently diffusion bonded to an aluminum coated beryllium tile. In this approach, a 250 microm thick titanium foil was used as a diffusion barrier between the copper and aluminum to prevent the formation of Cu-Al intermetallic phases. In all cases, a hot isostatic pressing (HIP) furnace was used in conjunction with canned assemblies in order to minimize oxidation and apply sufficient pressure on the assembly for excellent metal-to-metal contact and subsequent bonding. Several different processing schedules were evaluated during the course of this study; bonded assemblies were produced that failed outside the bond area indicating a 100% joint efficiency

  5. Evaluation of the cool-down behaviour of ITER FW beryllium tiles for an early failure detection

    Directory of Open Access Journals (Sweden)

    Thomas Weber

    2016-12-01

    Full Text Available The design of the first wall in ITER foresees several hundred thousand beryllium tiles, which are bonded to the water-cooled CuCrZr supporting structure. Due to the nature of a Tokamak reactor this bonding is faced to thermal fatigue. Since the failure of a single tile might already have a major impact on the operability of ITER, comprehensive high heat flux tests are performed on prototypes prior to the acceptance of manufacturing procedures. For a deeper understanding of the temperature curves, which were and will be measured by IR devices of these first wall prototypes, thermo-mechanical FEM simulations shall demonstrate the possibilities of an early bonding failure detection. Hereby, the maximum temperatures for each cycle as well as the cool-down behaviour are the input data.

  6. The structure and thermal properties of plasma-sprayed beryllium for the International Thermonuclear Experimental Reactor (ITER)

    International Nuclear Information System (INIS)

    Castro, R.G.; Bartlett, A.; Elliott, K.E.; Hollis, K.J.

    1996-01-01

    Plasma spraying is being studied for in situ repair of damaged Be and W plasma facing surfaces for ITER, the next generation magnetic fusion energy device, and is also being considered for fabricating Be and W plasma-facing components for the first wall of ITER. Investigators at LANL's Beryllium Atomization and Thermal Spray Facility have concentrated on investigating the structure-property relation between as-deposited microstructures of plasma sprayed Be coatings and resulting thermal properties. In this study, the effect of initial substrate temperature on resulting thermal diffusivity of Be coatings and the thermal diffusivity at the coating/Be substrate interface (interface thermal resistance) was investigated. Results show that initial Be substrate temperatures above 600 C can improve the thermal diffusivity of the Be coatings and minimize any thermal resistance at the interface between the Be coating and Be substrate

  7. High heat flux testing of ITER ICH&CD antenna beryllium faraday screen bars mock-ups

    International Nuclear Information System (INIS)

    Courtois, X.; Meunier, L.; Kuznetsov, V.; Beaumont, B.; Lamalle, P.; Conchon, D.; Languille, P.

    2016-01-01

    Highlights: • ITER ICH&CD antenna beryllium faraday screen bars mock-ups were manufactured. • The mock-ups are submitted to high heat loads to test their heat exhaust capabilities. • The mock-ups withstand without damage the design limit load. • Lifetime is gradually reduced when the heat load is augmented beyond the design limit. • Thermal and mechanical behavior are reproducible, and coherent with the calculation. - Abstract: The Faraday Screen (FS) is the plasma facing component of ITER ion cyclotron heating antennas shielding. The requirement for the high heat exhaust, and the limitation of the temperatures to minimize strain and thus offer sufficient resistance to fatigue, imply the need for high conductivity materials and a high cooling flow rate. The FS bars are constructed by a hipping process involving beryllium tiles, a pure copper layer, a copper chrome zirconium alloy for the cooling channel and a stainless steel backing strip. Two FS bars small scale mock-ups were manufactured and tested under high heat flux. They endured 15,000 heating cycles without degradation under nominal heat flux, and revealed growing flaws when the heat flux was progressively augmented beyond. In this case, the ultrasonic test confirms a strong delamination of the Be tiles.

  8. Fast brazing development for the joining of the beryllium armor layer for the ITER First Wall panels

    International Nuclear Information System (INIS)

    Buodot, C.; Boireau, B.; Lorenzetto, P.; Macel, D.

    2006-01-01

    In order to reduce cost and manufacturing time induction brazing is being developed as an alternative to Hot Isostatic Pressing for the joining of the beryllium armor onto the copper alloy heat sink material for the manufacture of First Wall panels for the ITER Blanket. The copper alloy that is currently adopted by ITER is a Copper Chromium Zirconium alloy. Its good mechanical properties are obtained by precipitation hardening by means of an ageing heat treatment at a temperature of about 480 o C. In order to avoid over-ageing and keep acceptable mechanical properties, brazing at higher temperatures must therefore be done as fast as possible. The flat geometry of a panel is not familiar for induction process; nevertheless, a development work was done validating the feasibility of joining beryllium tiles onto a copper chromium zirconium flat surface of a panel by induction brazing process. The development was done in 2 stages: validation of the capability of the induction process to realise a heat cycle on a dummy panel and in parallel, validation of the brazing parameters giving acceptable mechanical results on the beryllium CuCrZr joint. A flat pancake inductor was manufactured and tested on a dummy panel in an induction brazing vessel manufactured for this purpose. Several heating cycles were done with the aim of defining a cycle that gives uniform temperature at the interface of all the beryllium tiles on the entire panel surface. These cycles gave us a temperature range in which the brazing can be performed. A special device for brazing small mock up was also manufactured. This was for the metallurgical characterisation program. Many brazing samples where done and mechanically characterised. Unfortunately, this first metallurgical stage led to unacceptably low shear test values. A complete analysis of this non conformance put in evidence that the bad results were due to the braze material that was not adapted to this process. By changing the braze material

  9. Low cycle thermal fatigue testing of beryllium grades for ITER plasma facing components

    International Nuclear Information System (INIS)

    Watson, R.D.; Youchison, D.L.; Dombrowski, D.E.; Guiniatouline, R.N.; Kupriynov, I.B.

    1996-01-01

    A novel technique has been used to test the relative low cycle thermal fatigue resistance of different grades of US and Russian beryllium, which is proposed as plasma facing armor for fusion reactor first wall, limiter, and divertor components. The 30 kW electron beam test system at Sandia National Laboratories was used to sweep the beam spot along one direction at 1 Hz. This produces a localized temperature ''spike'' of 750 degree C for each pass of the beam. Large thermal stresses in excess of the yield strength are generated due to very high spot heat flux, 250 MW/m 2 . Cyclic plastic strains on the order of 0.6% produced visible cracking on the heated surface in less than 3000 cycles. An in-vacuo fiber optic borescope was used to visually inspect the beryllium surfaces for crack initiation. Grades of US beryllium tested included: S-65C, S- 65H, S-200F, S-200F-H, SR-200, I-400, extruded high purity, HIP'd spherical powder, porous beryllium (94% and 98% dense), Be/30% BeO, Be/60% BeO, and TiBe 12 . Russian grades included: TGP-56, TShGT, DShG-200, and TShG-56. Both the number of cycles to crack initiation, and the depth of crack propagation, were measured. The most fatigue resistant grades were S-65C, DShG-200, TShGT, and TShG-56. Rolled sheet Be (SR-200) showed excellent crack propagation resistance in the plane of rolling, despite early formation of delamination cracks. Only one sample showed no evidence of surface melting, Extruded (T). Metallographic and chemical analyses are provided. Good agreement was found between the measured depth of cracks and a 2-D elastic-plastic finite element stress analysis

  10. Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

    Energy Technology Data Exchange (ETDEWEB)

    Pérez, Germán, E-mail: german.perez.pichel@gmail.com; Mitteau, Raphaël; Eaton, Russell; Raffray, René

    2015-12-15

    Highlights: • Bonding defects at the ITER first wall beryllium armour are studied. • Experimental and analytical methods are combined. • Models supporting test results interpretation are proposed. • Guidelines for new experimental protocols are suggested. • Contribution to the definition of defects acceptance criteria. - Abstract: The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme. This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

  11. Investigation on bonding defects in ITER first wall beryllium armour components by combining analytical and experimental methods

    International Nuclear Information System (INIS)

    Pérez, Germán; Mitteau, Raphaël; Eaton, Russell; Raffray, René

    2015-01-01

    Highlights: • Bonding defects at the ITER first wall beryllium armour are studied. • Experimental and analytical methods are combined. • Models supporting test results interpretation are proposed. • Guidelines for new experimental protocols are suggested. • Contribution to the definition of defects acceptance criteria. - Abstract: The reliability of the plasma facing components (PFCs) is essential for the efficient plasma operation in a fusion machine. This concerns especially the bond between the armour tiles facing the plasma and the heat sink material (copper alloy). The different thermal expansions of the bonded materials cause a stress distribution in the bond, which peaks at the bond edge. Under cyclic heat flux and accounting for the possible presence of bonding defects, this stress could reach a level where the component might be jeopardised. Because of the complexity of describing realistically by analyses and models the stress evolution in the bond, “design by experiments” is the main procedure for defining and qualifying the armour joint. Most of the existing plasma operation know-how on actively cooled PFCs has been obtained with carbon composite armour tiles. In ITER, the tiles of the first wall are made out of beryllium, which means that the know-how is progressively adapted to this specific bimetallic pair. Nonetheless, analyses are still performed for supporting the R&D experimental programme. This paper: explores methods for combining experimental results with finite element and statistical analyses; benchmarks test results; proposes hypothesis and rationales consistent with test results interpretations; suggests guidelines for defining possible further experimental protocols; and contributes to the definition of defects acceptance criteria.

  12. Beryllium technology workshop, Clearwater Beach, Florida, November 20, 1991

    International Nuclear Information System (INIS)

    Longhurst, G.R.

    1991-12-01

    This report discusses the following topics: beryllium in the ITER blanket; mechanical testing of irradiated beryllium; tritium release measurements on irradiated beryllium; beryllium needs for plasma-facing components; thermal conductivity of plasma sprayed beryllium; beryllium research at the INEL; Japanese beryllium research activities for in-pile mockup tests on ITER; a study of beryllium bonding of copper alloy; new production technologies; thermophysical properties of a new ingot metallurgy beryllium product line; implications of beryllium:steam interactions in fusion reactors; and a test program for irradiation embrittlement of beryllium at JET

  13. Beryllium technology workshop, Clearwater Beach, Florida, November 20, 1991

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R.

    1991-12-01

    This report discusses the following topics: beryllium in the ITER blanket; mechanical testing of irradiated beryllium; tritium release measurements on irradiated beryllium; beryllium needs for plasma-facing components; thermal conductivity of plasma sprayed beryllium; beryllium research at the INEL; Japanese beryllium research activities for in-pile mockup tests on ITER; a study of beryllium bonding of copper alloy; new production technologies; thermophysical properties of a new ingot metallurgy beryllium product line; implications of beryllium:steam interactions in fusion reactors; and a test program for irradiation embrittlement of beryllium at JET.

  14. Reaction-diffusion modeling of hydrogen in beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Wensing, Mirko; Matveev, Dmitry; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik (Germany)

    2016-07-01

    Beryllium will be used as first-wall material for the future fusion reactor ITER as well as in the breeding blanket of DEMO. In both cases it is important to understand the mechanisms of hydrogen retention in beryllium. In earlier experiments with beryllium low-energy binding states of hydrogen were observed by thermal desorption spectroscopy (TDS) which are not yet well understood. Two candidates for these states are considered: beryllium-hydride phases within the bulk and surface effects. The retention of deuterium in beryllium is studied by a reaction rate approach using a coupled reaction diffusion system (CRDS)-model relying on ab initio data from density functional theory calculations (DFT). In this contribution we try to assess the influence of surface recombination.

  15. RF Sheath-Enhanced Plasma Surface Interaction Studies using Beryllium Optical Emission Spectroscopy in JET ITER-Like Wall

    Energy Technology Data Exchange (ETDEWEB)

    Agarici, G. [Fusion for Energy (F4E), Barcelona, Spain; Klepper, C Christopher [ORNL; Colas, L. [French Atomic Energy Commission (CEA); Krivska, Alena [Ecole Royale Militaire, Brussels Belgium; Bobkov, V. [Max-Planck-Institut fur Plasmaphysik, EURATOM Association, Garching, Germany; Jacquet, P. [Culham Centre for Fusion Energy (CCFE), Abingdon, UK; Delabie, Ephrem G. [ORNL; Giroud, C. [EURATOM / UKAEA, UK; Kirov, K K. [Association EURATOM-CCFE, Abingdon, UK; Lasa Esquisabel, Ane [ORNL; Lerche, E. [ERM-KMS, Association EURATOM-Belgian State, Brussels, Belgium; Dumortier, P. [ERM-KMS, Association EURATOM-Belgian State, Brussels, Belgium; Durodie, Frederic [Ecole Royale Militaire, Brussels Belgium

    2017-10-01

    A dedicated study on JET-ILW, deploying two types of ICRH antennas and spectroscopic observation spots at two outboard, beryllium limiters, has provided insight on long-range (up to 6m) RFenhanced plasma-surface interactions (RF-PSI) due to near-antenna electric fields. To aid in the interpretation of optical emission measurements of these effects, the antenna near-fields are computed using the TOPICA code, specifically run for the ITER-like antenna (ILA); similar modelling already existed for the standard JET antennas (A2). In the experiment, both antennas were operated in current drive mode, as RF-PSI tends to be higher in this phasing and at similar power (∼0.5 MW). When sweeping the edge magnetic field pitch angle, peaked RF-PSI effects, in the form of 2-4 fold increase in the local Be source,are consistently measured with the observation spots magnetically connect to regions of TOPICAL-calculated high near-fields, particularly at the near-antenna limiters. It is also found that similar RF-PSI effects are produced by the two types of antenna on similarly distant limiters. Although this mapping of calculated near-fields to enhanced RF-PSI gives only qualitative interpretion of the data, the present dataset is expected to provide a sound experimental basis for emerging RF sheath simulation model validation.

  16. ITER Task T227 (1995): Solubility, diffusion, and desorption of hydrogen isotopes in beryllium and tungsten

    International Nuclear Information System (INIS)

    Thompson, D.A.; Macauley-Newcombe, R.G.

    1996-04-01

    This report contains results of thermal absorption and ion-beam analysis experiments performed in 1995 for ITER task T227. The materials studies were Be, BeO and W. The Be samples used are single crystals provided by Dr. H. Kawamura of J.A.E.R.I.; the BeO samples are polycrystalline semiconductor substrates provided by the Brush Wellman Company (USA); and the W samples are single crystals and are-melted polycrystals from the Johnson-Matthey Company (USA). All samples are very high purity and 100% density. (author) 16 refs., 1 tab., 10 figs

  17. R(and)D on full tungsten divertor and beryllium wall for JET ITER-like Wall Project

    International Nuclear Information System (INIS)

    Hirai, T.; Maier, H.; Rubel, M.

    2006-01-01

    The ITER-like Wall Project was initiated at JET, with the goal of testing the reference material combination chosen for ITER: beryllium (Be) in the main chamber (wall and limiters) and tungsten (W) in the divertor. The major aims are to study the tritium retention, material mixing, melt layer behavior and to optimize plasma operation scenarios with a full metal wall. The project requires major design and engineering efforts in R(and)D: (i) bulk W tile, (ii) W coatings on carbon fibre composites (CFC) (iii) Be coatings on Inconel, (iv) Be marker tiles. For the W divertor, two R(and)D tasks were initiated: (1) development of a conceptual design for a bulk W tile as the main outer divertor target plate, and (2) W coating selection from 14 different samples produced by various techniques for the other divertor plates and neutral beam shine. The bulk W tile must withstand power loads of 7 MW/m 2 for 10 s. JET divertor plates are not actively cooled, therefore, heat capacity of the tiles is an important design parameter. In addition to power handling, mechanical structural stability under electromagnetic forces and compatibility with remote handling are the key requirements in the design. The design has been completed. The test-tile survived 100 pulses at 7 MW/m 2 for 10 s in the electron beam facility, JUDITH. The W coatings with different thickness, thin ( 2 and 200 pulses at 10 MW/m 2 for 5 s. In all tested samples cracks developed perpendicularly to the fiber bundles in CFC because of contraction of the coating in the cooling phase. Coatings were also exposed to 1000 ELM-like loading pulses. The thin coatings showed fatigue leading to delamination, whereas for thick coatings better resistance in ELM-like loading. As a result of R(and)D a full W divertor was decided: bulk metal at the outer divertor and W coating at other areas. Be-related R(and)D activities are in two areas. Production of 8-9 μm layers on inner wall cladding Inconel tiles ensures the full coating of

  18. Measurements of beryllium sputtering yields at JET

    Science.gov (United States)

    Jet-Efda Contributors Stamp, M. F.; Krieger, K.; Brezinsek, S.

    2011-08-01

    The lifetime of the beryllium first wall in ITER will depend on erosion and redeposition processes. The physical sputtering yields for beryllium (both deuterium on beryllium (Be) and Be on Be) are of crucial importance since they drive the erosion process. Literature values of experimental sputtering yields show an order of magnitude variation so predictive modelling of ITER wall lifetimes has large uncertainty. We have reviewed the old beryllium yield experiments on JET and used current beryllium atomic data to produce revised beryllium sputtering yields. These experimental measurements have been compared with a simple physical sputtering model based on TRIM.SP beryllium yield data. Fair agreement is seen for beryllium yields from a clean beryllium limiter. However the yield on a beryllium divertor tile (with C/Be co-deposits) shows poor agreement at low electron temperatures indicating that the effect of the higher sputtering threshold for beryllium carbide is important.

  19. The feasibility of beryllium as structural material for the ITER plasma-facing components (PFC)

    International Nuclear Information System (INIS)

    Vieider, G.; Cardella, A.; Gorenflo, H.

    1993-01-01

    Be as plasma-facing armour has attractive features including excellent plasma compatibility, no T-retention via co-deposition and the potential for in-situ repair via plasma spraying. In order to avoid the bonding of the Be-armour to a heatsink structure in e.g., Cu-alloys, the ITER Joint Central Team (JCT) proposed for the divertor tubular elements with monolithic Be, both as plasma-facing and structural material. The analysis of these Be-tubes with 5 mm wall thickness at a heat load of 5 MW/m 2 showed that even for the most favourable assumptions thermal stresses exceed by far the allowed values according to design codes. Damage by neutrons and disruptions would worsen further the case for Be as monolithic plasma-facing and structural material. For PFC at heat flux significantly above 1 MW/m 2 it appears evident that Be should be used merely as armour bonded to a suitable structural material as heatsink. (orig.)

  20. Tungsten and beryllium armour development for the JET ITER-like wall project

    International Nuclear Information System (INIS)

    Maier, H.; Hirai, T.; Rubel, M.; Neu, R.; Mertens, Ph.; Greuner, H.; Hopf, Ch.; Matthews, G.F.; Neubauer, O.; Piazza, G.; Gauthier, E.; Likonen, J.; Mitteau, R.; Maddaluno, G.; Riccardi, B.; Philipps, V.; Ruset, C.; Lungu, C.P.; Uytdenhouwen, I.

    2007-01-01

    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 and 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 0 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 and 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

  1. Tungsten and beryllium armour development for the JET ITER-like wall project

    Science.gov (United States)

    Maier, H.; Hirai, T.; Rubel, M.; Neu, R.; Mertens, Ph.; Greuner, H.; Hopf, Ch.; Matthews, G. F.; Neubauer, O.; Piazza, G.; Gauthier, E.; Likonen, J.; Mitteau, R.; Maddaluno, G.; Riccardi, B.; Philipps, V.; Ruset, C.; Lungu, C. P.; Uytdenhouwen, I.; EFDA contributors, JET

    2007-03-01

    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 °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.

  2. Study of the chemical interaction between the beryllium powders of different particles size and the air in the temperature range 500-1000degC form the viewpoint of ITER safety

    Energy Technology Data Exchange (ETDEWEB)

    Davydov, D.A. [State Scientific Center of Russian Federation, Moscow (Russian Federation); Konovalov, Y.V.; Gorokhov, V.A.; Levin, V.B.; Chekhlatov, G.M.; Khomutov, A.M.

    1998-01-01

    Under an effect of some factors characteristic for the ITER- operating condition a dense beryllium facing plasma can transit into various forms, changing its structural states. As a result of the bombardment of beryllium plasma facing components by ion fluxes, the production of a dust including the particles from a few micrometers to a few millimeters in size is possible. The specific features in the behaviour of various beryllium forms under emergency conditions are of an essential interest from the viewpoint of ITER safety. Some grades of powders of different average particles size (14-31 micron) have been produced in a given study, and their chemical interaction at high temperatures with air (500-1100degC), test duration effects simulating the emergency situation at ITER in the first approximation have been studied. The temperature dependence of beryllium powders (different particles size after disc abrased) interaction with air in the temperature range 500-1000degC at the exposure of 5 hours long for each temperature and kinetic dependence of interaction of these powders with air at 800degC for the exposure from half an hour to 7 hours long were studied. An analysis of granulometric weight fraction in the metallic and oxidized beryllium powders with different particles size has been done by the photosedimentational technique with the instrument `Analysette-20`. Construction of a mathematical model for the chemical interaction of beryllium powders with air at high temperatures have been carried out. (author)

  3. Compatibility of ITER candidate structural materials with static gallium

    International Nuclear Information System (INIS)

    Luebbers, P.R.; Michaud, W.F.; Chopra, O.K.

    1993-12-01

    Tests were conducted on the compatibility of gallium with candidate structural materials for the International Thermonuclear Experimental Reactor, e.g., Type 316 SS, Inconel 625, and Nb-5 Mo-1 Zr alloy, as well as Armco iron, Nickel 270, and pure chromium. Type 316 stainless steel is least resistant to corrosion in static gallium and Nb-5 Mo-1 Zr alloy is most resistant. At 400 degrees C, corrosion rates are ∼4.0, 0.5, and 0.03 mm/yr for type 316 SS, Inconel 625, and Nb-5 Mo- 1 Zr alloy, respectively. The pure metals react rapidly with gallium. In contrast to findings in earlier studies, pure iron shows greater corrosion than nickel. The corrosion rates at 400 degrees C are ≥88 and 18 mm/yr, respectively, for Armco iron and Nickel 270. The results indicate that at temperatures up to 400 degrees C, corrosion occurs primarily by dissolution and is accompanied by formation of metal/gallium intermetallic compounds. The solubility data for pure metals and oxygen in gallium are reviewed. The physical, chemical, and radioactive properties of gallium are also presented. The supply and availability of gallium, as well as price predictions through the year 2020, are summarized

  4. Beryllium R and D for blanket application

    Energy Technology Data Exchange (ETDEWEB)

    Dalle Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Longhurst, G.R. [Idaho National Engineering Lab., Idaho Falls (United States); Kawamura, H. [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-10-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.) 29 refs.

  5. Beryllium R and D for blanket application

    International Nuclear Information System (INIS)

    Dalle Donne, M.; Scaffidi-Argentina, F.; Kawamura, H.

    1998-01-01

    The paper describes the main problems and the R and D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point. (orig.)

  6. Beryllium R&D for blanket application

    Science.gov (United States)

    Donne, M. Dalle; Longhurst, G. R.; Kawamura, H.; Scaffidi-Argentina, F.

    1998-10-01

    The paper describes the main problems and the R&D for the beryllium to be used as neutron multiplier in blankets. As the four ITER partners propose to use beryllium in the form of pebbles for their DEMO relevant blankets (only the Russians consider the porous beryllium option as an alternative) and the ITER breeding blanket will use beryllium pebbles as well, the paper is mainly based on beryllium pebbles. Also the work on the chemical reactivity of fully dense and porous beryllium in contact with water steam is described, due to the safety importance of this point.

  7. Irradiation tests of ITER candidate Hall sensors using two types of neutron spectra

    International Nuclear Information System (INIS)

    Duran, I.; Bolshakova, I.; Holyaka, R.; Viererbl, L.; Lahodova, Z.; Sentkerestiova, J.; Bem, P.

    2010-01-01

    We report on irradiation tests of InSb based Hall sensors at two irradiation facilities with two distinct types of neutron spectra. One was a fission reactor neutron spectrum with a significant presence of thermal neutrons, while another one was purely fast neutron field. Total neutron fluence of the order of 10 16 cm -2 was accumulated in both cases, leading to significant drop of Hall sensor sensitivity in case of fission reactor spectrum, while stable performance was observed at purely fast neutron spectrum. This finding suggests that performance of this particular type of Hall sensors is governed dominantly by transmutation. Additionally, it further stresses the need to test ITER candidate Hall sensors under neutron flux with ITER relevant spectrum.

  8. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures: Preliminary results

    International Nuclear Information System (INIS)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.

    1993-01-01

    Candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at temperatures of either 60 or 250 degrees C. Preliminary results have been obtained for several of these materials irradiated at 60 degrees C. The results show that irradiation at this temperature reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The unloading compliance technique developed for the subsize disk compact specimens works quite well, particularly for materials with lower toughness. Specimens of materials with very high toughness deform excessively, and this results in experimental difficulties

  9. Beryllium production using beryllium fluoride

    International Nuclear Information System (INIS)

    Hubler, Carlos Henrique

    1993-01-01

    This work presents the beryllium production by thermal decomposition of the ammonium beryllium fluoride, followed by magnesium reduction, obtained in the small pilot plant of the Brazilian National Nuclear Energy Commission - Nuclear Engineering Institute

  10. An investigation of pulsed phase thermography for detection of disbonds in HIP-bonded beryllium tiles in ITER normal heat flux first wall (NHF FW) components

    Energy Technology Data Exchange (ETDEWEB)

    Bushell, J., E-mail: joe.bushell@amec.com [AMEC Foster Wheeler, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (United Kingdom); Sherlock, P. [AMEC Foster Wheeler, Booths Hall, Chelford Road, Knutsford, Cheshire WA16 8QZ, England (United Kingdom); Mummery, P. [School of Mechanical, Aerospace and Civil Engineering, University of Manchester, England (United Kingdom); Bellin, B.; Zacchia, F. [Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, Barcelona (Spain)

    2015-10-15

    Highlights: • Pulsed phase thermography was trialled on Be-tiled plasma facing components. • Two components, one with known disbonds, one intact, were inspected and compared. • Finite element analysis was used to verify experimental observations. • PPT successfully detected disbonds in the failed component. • Good agreement found with ultrasonic test, though defect geometry was uncertain. - Abstract: Pulsed phase thermography (PPT) is a non destructive examination (NDE) technique, traditionally used in the Aerospace Industry for inspection of composite structures, which combines characteristics and benefits of flash thermography and lock-in thermography into a single, rapid inspection technique. The aim of this work was to evaluate the effectiveness of PPT as a means of inspection for the bond between the beryllium (Be) tiles and the copper alloy (CuCrZr) heatsink of the ITER NHF FW components. This is a critical area dictating the functional integrity of these components, as single tile detachment in service could result in cascade failure. PPT has advantages over existing thermography techniques using heated water which stress the component, and the non-invasive, non-contact nature presents advantages over existing ultrasonic methods. The rapid and non-contact nature of PPT also gives potential for in-service inspections as well as a quality measure for as-manufactured components. The technique has been appraised via experimental trials using ITER first wall mockups with pre-existing disbonds confirmed via ultrasonic tests, partnered with finite element simulations to verify experimental observations. This paper will present the results of the investigation.

  11. An investigation of pulsed phase thermography for detection of disbonds in HIP-bonded beryllium tiles in ITER normal heat flux first wall (NHF FW) components

    International Nuclear Information System (INIS)

    Bushell, J.; Sherlock, P.; Mummery, P.; Bellin, B.; Zacchia, F.

    2015-01-01

    Highlights: • Pulsed phase thermography was trialled on Be-tiled plasma facing components. • Two components, one with known disbonds, one intact, were inspected and compared. • Finite element analysis was used to verify experimental observations. • PPT successfully detected disbonds in the failed component. • Good agreement found with ultrasonic test, though defect geometry was uncertain. - Abstract: Pulsed phase thermography (PPT) is a non destructive examination (NDE) technique, traditionally used in the Aerospace Industry for inspection of composite structures, which combines characteristics and benefits of flash thermography and lock-in thermography into a single, rapid inspection technique. The aim of this work was to evaluate the effectiveness of PPT as a means of inspection for the bond between the beryllium (Be) tiles and the copper alloy (CuCrZr) heatsink of the ITER NHF FW components. This is a critical area dictating the functional integrity of these components, as single tile detachment in service could result in cascade failure. PPT has advantages over existing thermography techniques using heated water which stress the component, and the non-invasive, non-contact nature presents advantages over existing ultrasonic methods. The rapid and non-contact nature of PPT also gives potential for in-service inspections as well as a quality measure for as-manufactured components. The technique has been appraised via experimental trials using ITER first wall mockups with pre-existing disbonds confirmed via ultrasonic tests, partnered with finite element simulations to verify experimental observations. This paper will present the results of the investigation.

  12. Fracture toughness of irradiated candidate materials for ITER first wall/blanket structures

    International Nuclear Information System (INIS)

    Alexander, D.J.; Pawel, J.E.; Grossbeck, M.L.; Rowcliffe, A.F.; Shiba, Kiyoyuki

    1994-01-01

    Disk compact specimens of candidate materials for first wall/blanket structures in ITER have been irradiated to damage levels of about 3 dpa at nominal irradiation temperatures of either 90 or 250 degrees C. These specimens have been tested over a temperature range from 20 to 250 degrees C to determine J-integral values and tearing moduli. The results show that irradiation at these temperatures reduces the fracture toughness of austenitic stainless steels, but the toughness remains quite high. The toughness decreases as the test temperature increases. Irradiation at 250 degrees C is more damaging than at 90 degrees C, causing larger decreases in the fracture toughness. Ferritic-martensitic steels are embrittled by the irradiation, and show the lowest toughness at room temperature

  13. Reactivity test between beryllium and copper

    International Nuclear Information System (INIS)

    Kawamura, H.; Kato, M.

    1995-01-01

    Beryllium has been expected for using as plasma facing material on ITER. And, copper alloy has been proposed as heat sink material behind plasma facing components. Therefore, both materials must be joined. However, the elementary process of reaction between beryllium and copper alloy does not clear in detail. For example, other authors reported that beryllium reacted with copper at high temperature, but it was not obvious about the generation of reaction products and increasing of the reaction layer. In the present work, from this point, for clarifying the elementary process of reaction between beryllium and copper, the out-of-pile compatibility tests were conducted with diffusion couples of beryllium and copper which were inserted in the capsule filled with high purity helium gas (6N). Annealing temperatures were 300, 400, 500, 600 and 700 degrees C, and annealing periods were 100, 300 and 1000h. Beryllium specimens were hot pressed beryllium, and copper specimens were OFC (Oxygen Free Copper)

  14. Beryllium allergy

    International Nuclear Information System (INIS)

    Schoenherr, S.; Pevny, I.

    1989-12-01

    Beryllium is not only a high potent allergen, but also a fotoallergen and can provoke contact allergic reactions, fotoallergic reactions, granulomatous skin reactions, pulmonary granulomatous diseases and sometimes even systemic diseases. The authors present 9 own cases of a patch test positive beryllium allergy, 7 patients with relevant allergy and 5 patients with an allergic contact stomatitis. (author)

  15. The High Aspect Ratio Design (HARD): A candidate ITER concept with improved technology phase performance

    International Nuclear Information System (INIS)

    Nevins, W.M.; Perkins, L.J.; Wesley, J.C.

    1992-10-01

    The High Aspect Ratio Design (HARD) International Thermonuclear Experimental Reactor (ITER) concept developed by the US ITER team is an alternate to the low-aspect-ratio ITER design developed by the ITER participants during the Conceptual Design Activity (CDA). The CDA design, referred to hereafter as ITER CDA, has an aspect ratio, A, of 2.79, a toroidal magnetic field, B T , of 4.85 T, and a plasma current, I p , of 22 MA for operation with an ignited plasma. In contrast, HARD employs higher aspect ratio, A = 4.0, higher toroidal field, B T = 7.11 T, and lower plasma current, I p = 14.8 MA for ignition operation. The cross sections of the two designs are compared in. The parameters and performance of HARD and ITER CDA for inductively driven ignition operation are compared in. The HARD parameters provide the same ignition performance (ignition margin evaluated against ITER-89P confinement scaling) as ITER CDA in a device with comparable size and cost. However, the reason for advancing HARD rather than ITER CDA as the ITER design concept is not inductively driven ignition performance but HARD's significantly enhanced potential to achieve the technology testing and steady-state operation goals of the ITER objectives with non-inductive current drive

  16. Iter

    Science.gov (United States)

    Iotti, Robert

    2015-04-01

    ITER is an international experimental facility being built by seven Parties to demonstrate the long term potential of fusion energy. The ITER Joint Implementation Agreement (JIA) defines the structure and governance model of such cooperation. There are a number of necessary conditions for such international projects to be successful: a complete design, strong systems engineering working with an agreed set of requirements, an experienced organization with systems and plans in place to manage the project, a cost estimate backed by industry, and someone in charge. Unfortunately for ITER many of these conditions were not present. The paper discusses the priorities in the JIA which led to setting up the project with a Central Integrating Organization (IO) in Cadarache, France as the ITER HQ, and seven Domestic Agencies (DAs) located in the countries of the Parties, responsible for delivering 90%+ of the project hardware as Contributions-in-Kind and also financial contributions to the IO, as ``Contributions-in-Cash.'' Theoretically the Director General (DG) is responsible for everything. In practice the DG does not have the power to control the work of the DAs, and there is not an effective management structure enabling the IO and the DAs to arbitrate disputes, so the project is not really managed, but is a loose collaboration of competing interests. Any DA can effectively block a decision reached by the DG. Inefficiencies in completing design while setting up a competent organization from scratch contributed to the delays and cost increases during the initial few years. So did the fact that the original estimate was not developed from industry input. Unforeseen inflation and market demand on certain commodities/materials further exacerbated the cost increases. Since then, improvements are debatable. Does this mean that the governance model of ITER is a wrong model for international scientific cooperation? I do not believe so. Had the necessary conditions for success

  17. Research of beryllium safety issues

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Dolan, T.J.; Hankins, M.R.; Pawelko, R.J.

    1993-01-01

    Beryllium has been identified as a leading contender for the plasma-facing material in ITER. Its use has some obvious advantages, but there are also a number of safety concerns associated with it. The Idaho National Engineering Laboratory (INEL) has undertaken a number of studies to help resolve some of these issues. One issue is the response of beryllium to neutron irradiation. We have tested samples irradiated in the Advanced Test Reactor (ATR) and are currently preparing to make measurements of the change in mechanical properties of beryllium samples irradiated at elevated temperatures in the Fast Flux Test Facility (FFTF) and the Experimental Breeder Reactor II (EBR-II) at the INEL. Mechanical tests will be conducted at the irradiation temperatures of 375-550 C. Other experiments address permeation and retention of implanted tritium in plasma-sprayed beryllium. In one test the porosity of the material allowed 0.12% of implanted ions and 0.17% of atoms from background gas pressure to pass through the foil with essentially no delay. For comparison, similar tests on fully dense hot-rolled, vacuum melted or sintered powder foils of high purity beryllium showed only 0.001% of implanting ions to pass through the foil, and then only after a delay of several hours. None of the molecular gas appeared to permeate these latter targets. An implication is that plasma-sprayed beryllium may substantially enhance recycling of tritium to the plasma provided it is affixed to a relatively impermeable substrate. (orig.)

  18. Beryllium poisonings

    International Nuclear Information System (INIS)

    Alibert, S.

    1959-03-01

    This note reports a bibliographical study of beryllium toxicity. Thus, this bibliographical review addresses and outlines aspects and issues like aetiology, cases of acute poisoning (cutaneous manifestations, pulmonary manifestations), chronic poisoning (cutaneous, pulmonary and bone manifestations), excretion and localisation, and prognosis

  19. Preliminary results for explosion bonding of beryllium to copper

    International Nuclear Information System (INIS)

    Butler, D.J.; Dombrowski, D.E.

    1995-01-01

    This program was undertaken to determine if explosive bonding is a viable technique for joining beryllium to copper substrates. The effort was a cursory attempt at trying to solve some of the problems associated with explosive bonding beryllium and should not be considered a comprehensive research effort. There are two issues that this program addressed. Can beryllium be explosive bonded to copper substrates and can the bonding take place without shattering the beryllium? Thirteen different explosive bonding iterations were completed using various thicknesses of beryllium that were manufactured with three different techniques

  20. Hydrogen transport behavior of beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Hankins, M.R.; Longhurst, G.R.; Pawelko, R.J. (Idaho National Engineering Lab., EG and G Idaho, Inc., Idaho Falls, ID (United States)); Macaulay-Newcombe, R.G. (Dept. of Engineering Physics, Univ. Hamilton, ON (Canada))

    1992-12-01

    Beryllium is being evaluated for use as a plasma-facing material in the International Thermonuclear Experimental Reactor (ITER). One concern in the evaluation is the retention and permeation of tritium implanted into the plasma-facing surface. We performed laboratory-scale studies to investigate mechanisms that influence hydrogen transport and retention in beryllium foil specimens of rolled powder metallurgy product and rolled ingot cast beryllium. Specimen characterization was accomplished using scanning electron microscopy. Auger electron spectroscopy, and Rutherford backscattering spectrometry (RBS) techniques. Hydrogen transport was investigated using ion-beam permeation experiments and nuclear reaction analysis (NRA). Results indicate that trapping plays a significant role in permeation, re-emission, and retention, and that surface processes at both upstream and downstream surfaces are also important. (orig.).

  1. The status of beryllium technology for fusion

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Longhurst, G.R. E-mail: gx1@inel.gov; Shestakov, V.; Kawamura, H

    2000-12-01

    Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented.

  2. The status of beryllium technology for fusion

    International Nuclear Information System (INIS)

    Scaffidi-Argentina, F.; Longhurst, G.R.; Shestakov, V.; Kawamura, H.

    2000-01-01

    Beryllium was used for a number of years in the Joint European Torus (JET), and it is planned to be used extensively on the lower heat-flux surfaces of the reduced technical objective/reduced cost international thermonuclear experimental reactor (RTO/RC ITER). It has been included in various forms in a number of tritium breeding blanket designs. There are technical advantages but also a number of safety issues associated with the use of beryllium. Research in a variety of technical areas in recent years has revealed interesting issues concerning the use of beryllium in fusion. Progress in this research has been presented at a series of International Workshops on Beryllium Technology for Fusion. The most recent workshop was held in Karlsruhe, Germany on 15-17 September 1999. In this paper, a summary of findings presented there and their implications for the use of beryllium in the development of fusion reactors are presented

  3. Beryllium for fusion application - recent results

    International Nuclear Information System (INIS)

    Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

    2002-01-01

    The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described

  4. Beryllium for fusion application - recent results

    Science.gov (United States)

    Khomutov, A.; Barabash, V.; Chakin, V.; Chernov, V.; Davydov, D.; Gorokhov, V.; Kawamura, H.; Kolbasov, B.; Kupriyanov, I.; Longhurst, G.; Scaffidi-Argentina, F.; Shestakov, V.

    2002-12-01

    The main issues for the application of beryllium in fusion reactors are analyzed taking into account the latest results since the ICFRM-9 (Colorado, USA, October 1999) and presented at 5th IEA Be Workshop (10-12 October 2001, Moscow Russia). Considerable progress has been made recently in understanding the problems connected with the selection of the beryllium grades for different applications, characterization of the beryllium at relevant operational conditions (irradiation effects, thermal fatigue, etc.), and development of required manufacturing technologies. The key remaining problems related to the application of beryllium as an armour in near-term fusion reactors (e.g. ITER) are discussed. The features of the application of beryllium and beryllides as a neutron multiplier in the breeder blanket for power reactors (e.g. DEMO) in pebble-bed form are described.

  5. ITER oriented issues-2 (etc.)

    International Nuclear Information System (INIS)

    Goryayev, G.V.; Savchuk, V.V.; Shakhvorostov, Yu. V.

    2004-01-01

    The study analyzes the possibilities of utilization beryllium ingots produced at UMZ (Ulba Metallurgical Plant) for the purpose of ITER program. The results of comparative analysis of specification requirement to S-65 grade chemical compound and statistics data of UMZ beryllium ingots impurities content are presented. It has been demonstrated that beryllium industrial ingots produced at UMZ can be used for a production of powders and billets conforming the requirements of ITER specification. Beryllium ingots production flow chart, description of basic process equipment, the layout of metallurgical production upgrade, the results of such upgrade implementation are complimentary data to this study. The study illustrated with explanatory drawings. (author)

  6. Some aspects of beryllium disposal in Kazakhstan

    International Nuclear Information System (INIS)

    Shestakov, V.; Chikhray, Y.; Shakhvorostov, Yr.

    2004-01-01

    Historically in Kazakhstan all disposals of used beryllium and beryllium wasted materials were stored and recycled at JSC ''Ulba Metallurgical Plant''. Since Ulba Metallurgical Plant (beside beryllium and tantalum production) is one of the world largest complex producers of fuel for nuclear power plants as well it has possibilities, technologies and experience in processing toxic and radioactive wastes related with those productions. At present time only one operating Kazakhstan research reactors (EWG1M in Kurchatov) contains beryllium made core. The results of current examination of that core allow using it without replacement long time yet (at least for next five-ten years). Nevertheless the problem how to utilize such irradiated beryllium becomes actual issue for Kazakhstan even today. Since Kazakhstan is the member of ITER/DEMO Reactors Projects and is permanently considered as possible provider of huge amount of beryllium for those reactors so that is the reason for starting studies of possibilities of large scale processing/recycling of such disposed irradiated beryllium. It is clear that the Ulba Metallurgical Plant is considered as the best site for it in Kazakhstan. The draft plan how to organize experimental studies of irradiated beryllium disposals in Kazakhstan involving National Nuclear Center, National University (Almaty), JSC ''Ulba Metallurgical Plant'' (Ust-Kamenogorsk) would be presented in this paper as well as proposals to arrange international collaboration in that field through ISTC (International Science Technology Center, Moscow). (author)

  7. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    Energy Technology Data Exchange (ETDEWEB)

    Tanchuk, Victor, E-mail: Victor.Tanchuk@sintez.niiefa.spb.su [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Alexandrov, Evgeny [Institution “Project Center ITER”, 1, Akademika Kurchatova sq., 123182 Moscow (Russian Federation); Batyunin, Alexander; Kashchuk, Yuri [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Korban, Svetlana; Lyublin, Boris [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation); Obudovsky, Sergey [State Research Center of Russian Federation Troitsk Institute for Innovation and Fusion Research, ul. Pushkovykh, vladenie 12, 142190 Troitsk, Moscow Region (Russian Federation); Senik, Konstantin [Scientific Technical Center SINTEZ, D.V. Efremov Institute, 196641 St. Petersburg (Russian Federation)

    2013-10-15

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity.

  8. Thermal–hydraulic analysis of a candidate design for ITER divertor neutron flux monitor (DNFM)

    International Nuclear Information System (INIS)

    Tanchuk, Victor; Alexandrov, Evgeny; Batyunin, Alexander; Kashchuk, Yuri; Korban, Svetlana; Lyublin, Boris; Obudovsky, Sergey; Senik, Konstantin

    2013-01-01

    The key role in direct measurement of the ITER fusion power is assigned to the neutron diagnostic system for measurement of total neutron flux of the D–D and D–T fusion reaction with the help of a neutron flux monitor located under the divertor dome. High plasma heat loads in this position implies stringent requirements for the detector design and its cooling system to ensure the required temperature operation regime of the neutron detector. The paper describes the neutron flux monitor design developed in close collaboration with IO ITER diagnostic division. Two numerical models (hydraulic and thermal) built up to simulate the water flow in the cooling system and the temperature state of detector components are also presented and discussed. The numerical investigations carried out on the developed models have shown that only good thermal contact between the shell of the detector blocks and water-cooled casing of the monitor (fit, brazing) will provide the required temperature operation regimes of the most temperature-sensitive IFC electrodes. The obtained high temperature of the detector supports makes necessary an auxiliary direct cooling of the supports or their redesign so as to provide their higher thermal conductivity

  9. Method for welding beryllium

    Science.gov (United States)

    Dixon, R.D.; Smith, F.M.; O`Leary, R.F.

    1997-04-01

    A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs.

  10. Method for welding beryllium

    International Nuclear Information System (INIS)

    Dixon, R.D.; Smith, F.M.; O'Leary, R.F.

    1997-01-01

    A method is provided for joining beryllium pieces which comprises: depositing aluminum alloy on at least one beryllium surface; contacting that beryllium surface with at least one other beryllium surface; and welding the aluminum alloy coated beryllium surfaces together. The aluminum alloy may be deposited on the beryllium using gas metal arc welding. The aluminum alloy coated beryllium surfaces may be subjected to elevated temperatures and pressures to reduce porosity before welding the pieces together. The aluminum alloy coated beryllium surfaces may be machined into a desired welding joint configuration before welding. The beryllium may be an alloy of beryllium or a beryllium compound. The aluminum alloy may comprise aluminum and silicon. 9 figs

  11. Recommended design correlations for S-65 beryllium

    International Nuclear Information System (INIS)

    Billone, M.C.

    1995-01-01

    The properties of tritium and helium behavior in irradiated beryllium are reviewed, along with the thermal-mechanical properties needed for ITER design analysis. Correlations are developed to describe the performance of beryllium in a fusion reactor environment. While this paper focuses on the use of beryllium as a plasma-facing component (PFC) material, the correlations presented here can also be used to describe the performance of beryllium as a neutron multiplier for a tritium breeding blanket. The performance properties for beryllium are subdivided into two categories: properties which do not change with irradiation damage to the bulk of the material; and properties which are degraded by neutron irradiation. The irradiation-independent properties described within are: thermal conductivity, specific heat capacity, thermal expansion, and elastic constants. Irradiation-dependent properties include: yield strength, ultimate tensile strength, plastic tangent modulus, uniform and total tensile elongation, thermal and irradiation-induced creep strength, He-induced swelling and tritium retention/release. The approach taken in developing properties correlations is to describe the behavior of dense, pressed S-65 beryllium -- the material chosen for ITER PFC application -- as a function of temperature. As there are essentially no data on the performance of porous and/or irradiated S-65 beryllium, the degradation of properties with as-fabricated porosity and irradiation are determined from the broad data base on S-200F, as well as other types and grades, and applied to S-65 beryllium by scaling factors. The resulting correlations can be used for Be produced by vacuum hot pressing (VHP) and cold-pressing (CP)/sintering(S)/hot-isostatic-pressing (HIP). The performance of plasma-sprayed beryllium is discussed but not quantified

  12. Effect of transient heating loads on beryllium

    International Nuclear Information System (INIS)

    Kupriyanov, Igor B.; Porezanov, Nicolay P.; Nikolaev, Georgyi N.; Kurbatova, Liudmila A.; Podkovyrov, Vyacheslav L.; Muzichenko, Anatoliy D.; Zhitlukhin, Anatoliy M.; Khimchenko, Leonid N.; Gervash, Alexander A.

    2014-01-01

    Highlights: • We study the effect of transient plasma loads on beryllium erosion and surface microstructure. • Beryllium targets were irradiated by plasma streams with energy of 0.5–1 MJ/m 2 at ∼250 °C. • Under plasma loads 0.5–1 MJ/m 2 cracking of beryllium surface is rather slight. • Under 0.5 MJ/m 2 the mass loss of Be is no more than 0.2 g/m 2 shot and decreasing with shots number. • Under 1 MJ/m 2 maximum mass loss of beryllium was 3.7 g/m 2 shot and decreasing with shots number. - Abstract: Beryllium will be used as a plasma facing material for ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of ITER first wall. The results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility are presented. The Be/CuCrZr mock-ups were exposed to upto 100 shots by deuterium plasma streams with pulse duration of 0.5 ms at ∼250 °C and average heat loads of 0.5 and 1 MJ/m 2 . Experiments were performed at 250 °C. The evolution of surface microstructure and cracks morphology as well as beryllium mass loss are investigated under erosion process

  13. The beryllium production at Ulba metallurgical plant (Ust-Kamenogrsk, Kazakhstan)

    Energy Technology Data Exchange (ETDEWEB)

    Dvinskykh, E.M.; Savchuk, V.V.; Tuzov, Y.V. [Ulba Metallurgical Plant (Zavod), Ust-Kamenogorsk, Abay prospect 102 (Kazakhstan)

    1998-01-01

    The Report includes data on beryllium production of Ulba metallurgical plant, located in Ust-Kamenogorsk (Kazakhstan). Beryllium production is showed to have extended technological opportunities in manufacturing semi-products (beryllium ingots, master alloys, metallic beryllium powders, beryllium oxide) and in production of structural beryllium and its parts. Ulba metallurgical plant owns a unique technology of beryllium vacuum distillation, which allows to produce reactor grades of beryllium with a low content of metallic impurities. At present Ulba plant does not depend on raw materials suppliers. The quantity of stored raw materials and semi-products will allow to provide a 25-years work of beryllium production at a full capacity. The plant has a satisfactory experience in solving ecological problems, which could be useful in ITER program. (author)

  14. Study on neutron irradiation behavior of beryllium as neutron multiplier

    Energy Technology Data Exchange (ETDEWEB)

    Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    1998-03-01

    More than 300 tons beryllium is expected to be used as a neutron multiplier in ITER, and study on the neutron irradiation behavior of beryllium as the neutron multiplier with Japan Materials Testing Reactor (JMTR) were performed to get the engineering data for fusion blanket design. This study started as the study on the tritium behavior in beryllium neutron reflector in order to make clear the generation mechanism on tritium of JMTR primary coolant since 1985. These experiences were handed over to beryllium studies for fusion study, and overall studies such as production technology of beryllium pebbles, irradiation behavior evaluation and reprocessing technology have been started since 1990. In this presentation, study on the neutron irradiation behavior of beryllium as the neutron multiplier with JMTR was reviewed from the point of tritium release, thermal properties, mechanical properties and reprocessing technology. (author)

  15. Mechanical properties of irradiated beryllium

    International Nuclear Information System (INIS)

    Beeston, J.M.; Longhurst, G.R.; Wallace, R.S.

    1992-01-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10 25 n/m 2 (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.)

  16. Mechanical properties of irradiated beryllium

    Science.gov (United States)

    Beeston, J. M.; Longhurst, G. R.; Wallace, R. S.; Abeln, S. P.

    1992-10-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 × 10 25 n/m 2 ( E > 1 MeV) at an irradiation temperature of 75°C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium.

  17. Safety handling of beryllium for fusion technology R and D

    International Nuclear Information System (INIS)

    Yoshida, Hiroshi; Okamoto, Makoto; Terai, Takayuki; Odawara, Osamu; Ashibe, Kusuo; Ohara, Atsushi.

    1992-07-01

    Feasibility of beryllium use as a blanket neutron multiplier, first wall and plasma facing material has been studied for the D-T burning experiment reactors such as ITER. Various experimental work of beryllium and its compounds will be performed under the conditions of high temperature and high energy particle exposure simulating fusion reactor conditions. Beryllium is known as a hazardous substance and its handling has been carefully controlled by various health and safe guidances and/or regulations in many countries. Japanese regulations for hazardous substance provide various guidelines on beryllium for the protection of industrial workers and environment. This report was prepared for the safe handling of beryllium in a laboratory scale experiments for fusion technology R and D such as blanket development. Major items in this report are; (1) Brief review of guidances and regulations in USA, UK and Japan. (2) Safe handling and administration manuals at beryllium facilities in INEL, LANL and JET. (3) Conceptual design study of beryllium handling facility for small to mid-scale blanket R and D. (4) Data on beryllium toxicity, example of clinical diagnosis of beryllium disease, and environmental occurence of beryllium. (5) Personnel protection tools of Japanese Industrial Standard for hazardous substance. (author) 61 refs

  18. Chronic Beryllium Disease

    Science.gov (United States)

    ... who are exposed to beryllium will not experience health effects. Studies have shown that on average, 1 – 6 percent of exposed workers develop beryllium sensitization, although the rates can be ...

  19. Laser fabrication of beryllium components

    International Nuclear Information System (INIS)

    Hanafee, J.E.; Ramos, T.J.

    1995-08-01

    Working with the beryllium industry on commercial applications and using prototype parts, the authors have found that the use of lasers provides a high-speed, low-cost method of cutting beryllium metal, beryllium alloys, and beryllium-beryllium oxide composites. In addition, they have developed laser welding processes for commercial structural grades of beryllium that do not need a filler metal; i.e., autogenous welds were made in commercial structural grades of beryllium by using lasers

  20. Beryllium chemistry and processing

    CERN Document Server

    Walsh, Kenneth A

    2009-01-01

    This book introduces beryllium; its history, its chemical, mechanical, and physical properties including nuclear properties. The 29 chapters include the mineralogy of beryllium and the preferred global sources of ore bodies. The identification and specifics of the industrial metallurgical processes used to form oxide from the ore and then metal from the oxide are thoroughly described. The special features of beryllium chemistry are introduced, including analytical chemical practices. Beryllium compounds of industrial interest are identified and discussed. Alloying, casting, powder processing, forming, metal removal, joining and other manufacturing processes are covered. The effect of composition and process on the mechanical and physical properties of beryllium alloys assists the reader in material selection. The physical metallurgy chapter brings conformity between chemical and physical metallurgical processing of beryllium, metal, alloys, and compounds. The environmental degradation of beryllium and its all...

  1. Beryllium processing technology review for applications in plasma-facing components

    International Nuclear Information System (INIS)

    Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

    1993-07-01

    Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included

  2. Beryllium processing technology review for applications in plasma-facing components

    Energy Technology Data Exchange (ETDEWEB)

    Castro, R.G.; Jacobson, L.A.; Stanek, P.W.

    1993-07-01

    Materials research and development activities for the International Thermonuclear Experimental Reactor (ITER), i.e., the next generation fusion reactor, are investigating beryllium as the first-wall containment material for the reactor. Important in the selection of beryllium is the ability to process, fabricate and repair beryllium first-wall components using existing technologies. Two issues that will need to be addressed during the engineering design activity will be the bonding of beryllium tiles in high-heat-flux areas of the reactor, and the in situ repair of damaged beryllium tiles. The following review summarizes the current technology associated with welding and joining of beryllium to itself and other materials, and the state-of-the-art in plasma-spray technology as an in situ repair technique for damaged beryllium tiles. In addition, a review of the current status of beryllium technology in the former Soviet Union is also included.

  3. The structure, properties and performance of plasma-sprayed beryllium for fusion applications

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Elliott, K.E.

    1995-01-01

    Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasma-spray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H 2 gas additions to improve the melting of the beryllium powder and transferred-arc cleaning to improve the bonding between the beryllium coatings and the underlying surface. Information will also be presented on thermal fatigue tests which were done on beryllium coated ISX-B beryllium limiter tiles using 10 sec cycle times with 60 sec cooldowns and an International Thermonuclear Experimental Reactor (ITER) relevant divertor heat flux slightly in excess of 5 MW/m 2

  4. Beryllium. Evaluation of beryllium hydroxide industrial processes. Pt. 3

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1991-01-01

    This work continues the 'Beryllium' series. It is a historical review of different industrial processes of beryllium hydroxide obtention from beryllium ores. Flowsheats and operative parameters of five plants are provided. These plants (Degussa, Brush Beryllium Co., Beryllium Corp., Murex Ltd., SAPPI) were selected as representative samples of diverse commercial processes in different countries. (Author) [es

  5. Beryllium R and D for fusion applications

    International Nuclear Information System (INIS)

    Scaffidi-Argentina, F.; Longhurst, G.R.; Shestakov, V.; Kawamura, H.

    2000-01-01

    Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal

  6. Beryllium R and D for fusion applications

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F. E-mail: francesco.scaffidi@iket.fzk.de; Longhurst, G.R.; Shestakov, V.; Kawamura, H

    2000-11-01

    Beryllium is one of the primary candidates as both plasma-facing material (PFM) and neutron multiplier in the next-step fusion reactors. Both sintered-product blocks and pebbles are considered in fusion reactor designs. Beryllium evaporated on carbon tiles has also been used in Joint European Torus (JET) and may be considered for other designs. Future efforts are directed toward the pebble form of beryllium. Research and evaluations of data are underway to determine the most attractive material processing approaches in terms of fabrication cost and quality; technical issues associated with heat transfer; thermal, mechanical and irradiation stability; safety and tritium release. Beryllium plasma-facing components will require periodic repair or replacement, therefore disposal or recycling of activated and tritiated beryllium will also be a concern. Beryllium as a component of the molten salt, Flibe is also being considered in novel approaches to the plasma-structure interface. This paper deals with the main issues related to the use of Be in a fusion reactor as both neutron multiplier and first wall material. These issues include potential reactions with steam during accidents and the health and environmental aspects of its use, reprocessing and reuse, or disposal.

  7. (Beryllium). Internal Report No. 137, Jan. 15, 1958; Le beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Mouret, P; Rigaud, A

    1959-07-01

    After a brief summary of the physical and chemical properties of beryllium, the various chemical treatments which can be applied to beryllium minerals either directly or after a physical enrichment are discussed. These various treatments give either the hydroxide or beryllium salts, from which either beryllium oxide or metallic beryllium can easily be obtained. The purification, analysis and uses of beryllium are also briefly discussed. (author)

  8. Beryllium coating on Inconel tiles

    International Nuclear Information System (INIS)

    Bailescu, V.; Burcea, G.; Lungu, C.P.; Mustata, I.; Lungu, A.M.; Rubel, M.; Coad, J.P.; Matthews, G.; Pedrick, L.; Handley, R.

    2007-01-01

    Full text of publication follows: The Joint European Torus (JET) is a large experimental nuclear fusion device. Its aim is to confine and study the behaviour of plasma in conditions and dimensions approaching those required for a fusion reactor. The plasma is created in the toroidal shaped vacuum vessel of the machine in which it is confined by magnetic fields. In preparation for ITER a new ITER-like Wall (ILW) will be installed on Joint European Torus (JET), a wall not having any carbon facing the plasma [1]. In places Inconel tiles are to be installed, these tiles shall be coated with Beryllium. MEdC represented by the National Institute for Laser, Plasma and Radiation Physics, Magurele, Bucharest and in direct cooperation with Nuclear Fuel Plant Pitesti started to coat Inconel tiles with 8 μm of Beryllium in accordance with the requirements of technical specification and fit for installation in the JET machine. This contribution provides an overview of the principles of manufacturing processes using thermal evaporation method in vacuum and the properties of the prepared coatings. The optimization of the manufacturing process (layer thickness, structure and purity) has been carried out on Inconel substrates (polished and sand blasted) The results of the optimization process and analysis (SEM, TEM, XRD, Auger, RBS, AFM) of the coatings will be presented. Reference [1] Takeshi Hirai, H. Maier, M. Rubel, Ph. Mertens, R. Neu, O. Neubauer, E. Gauthier, J. Likonen, C. Lungu, G. Maddaluno, G. F. Matthews, R. Mitteau, G. Piazza, V. Philipps, B. Riccardi, C. Ruset, I. Uytdenhouwen, R and D on full tungsten divertor and beryllium wall for JET TIER-like Wall Project, 24. Symposium on Fusion Technology - 11-15 September 2006 -Warsaw, Poland. (authors)

  9. Beryllium armour produced by evaporation-condensation technique

    International Nuclear Information System (INIS)

    Anisimov, A.; Frolov, V.; Moszherin, S.; Pepekin, G.; Pirogov, A.; Komarov, V.; Mazul, I.

    1997-01-01

    Beryllium, as armour material for ITER plasma facing components, has a limited erosion lifetime. In order to repair the surface of eroded tiles in-situ, Be-deposition technologies are under consideration. One of them uses the physical vapour deposition of beryllium on copper or beryllium substrate produced by a hot Be-target placed in the vicinity of this substrate. Three different options for using this technology for ITER Be-armour application are considered. The first option is the repair in-situ of eroded Be-tiles. The second option suggests the use of this technology to provide the joining of Be to Cu-substrate. The third option assumes the use of evaporated-condensed beryllium as a bulk tile material bonded to copper substrate by conventional joining (Brazing et al.) techniques. The first results and prospects of these approaches are presented below. (orig.)

  10. Extractive metallurgy of the beryllium

    International Nuclear Information System (INIS)

    Alonso, Neusa; Capocchi, Jose Deodoro Trani

    1995-01-01

    A bibliographic review is performed on the beryllium extractive metallurgy. The work describes the main type of ores and processes applied to the metallic beryllium production, beryllium oxide production using fluoride, sulfide and direct chlorination. The thermodynamic consideration are made on beryllium reduction processes, discussing the viability of the beryllium oxide and hallide reduction processes. Under the technological viewpoint, the Cu-Be alloys main production processes are discussed, and the main toxicity problems related with beryllium are mentioned

  11. The effect of processing parameters on plasma sprayed beryllium for fusion applications

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Jacobson, L.A.; Cowgill, D.F.; Snead, L.L.

    1993-01-01

    Plasma spraying is being investigated as a potential coating technique for applying thin (0.1--5mm) layers of beryllium on plasma facing surfaces of blanket modules in ITER and also as an in-situ repair technique for repairing eroded beryllium surfaces in high heat flux divertor regions. High density spray deposits (>98% of theoretical density) of beryllium will be required in order to maximize the thermal conductivity of the beryllium coatings. A preliminary investigation was done to determine the effect of various processing parameters (particle size, particle morphology, secondary gas additions and reduced chamber pressure) on the as-deposited density of beryllium. The deposits were made using spherical beryllium feedstock powder which was produced by centrifugal atomization at Los Alamos National Laboratory (LANL). Improvements in the as-deposited densities and deposit efficiencies of the beryllium spray deposits will be discussed along with the corresponding thermal conductivity and outgassing behavior of these deposits

  12. The effect of processing parameters on plasma sprayed beryllium for fusion applications

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Jacobson, L.W.; Cowgill, D.F.

    1993-01-01

    Plasma spraying is being investigated as a potential coating technique for applying thin (0.1-5mm) layers of beryllium on plasma facing surfaces of blanket modules in ITER and also as an in-situ repair technique for repairing eroded beryllium surfaces in high heat flux divertor regions. High density spray deposits (>98% of theoretical density) of beryllium will be required in order to maximize the thermal conductivity of the beryllium coatings. A preliminary investigation was done to determine the effect of various processing parameters (particle size, particle morphology, secondary gas additions and reduced chamber pressure) on the as-deposited density of beryllium. The deposits were made using spherical beryllium feedstock powder which was produced by centrifugal atomization at Los Alamos National Laboratory (LANL). Improvements in the as-deposited densities and deposit efficiencies of the beryllium spray deposits will be discussed along with the corresponding thermal conductivity and outgassing behavior of these deposits. (orig.)

  13. METHOD OF BRAZING BERYLLIUM

    Science.gov (United States)

    Hanks, G.S.; Keil, R.W.

    1963-05-21

    A process is described for brazing beryllium metal parts by coating the beryllium with silver (65- 75 wt%)-aluminum alloy using a lithium fluoride (50 wt%)-lithium chloride flux, and heating the coated joint to a temperature of about 700 un. Concent 85% C for about 10 minutes. (AEC)

  14. Preparation of beryllium hydride

    International Nuclear Information System (INIS)

    Roberts, C.B.

    1975-01-01

    A process is described for preparing beryllium hydride by the direct reaction of beryllium borohydride and aluminum hydride trimethylamine adduct. Volatile by-products and unreacted reactants are readily removed from the product mass by sublimation and/or evaporation. (U.S.)

  15. Plasma cleaning of ITER first mirrors

    Science.gov (United States)

    Moser, L.; Marot, L.; Steiner, R.; Reichle, R.; Leipold, F.; Vorpahl, C.; Le Guern, F.; Walach, U.; Alberti, S.; Furno, I.; Yan, R.; Peng, J.; Ben Yaala, M.; Meyer, E.

    2017-12-01

    Nuclear fusion is an extremely attractive option for future generations to compete with the strong increase in energy consumption. Proper control of the fusion plasma is mandatory to reach the ambitious objectives set while preserving the machine’s integrity, which requests a large number of plasma diagnostic systems. Due to the large neutron flux expected in the International Thermonuclear Experimental Reactor (ITER), regular windows or fibre optics are unusable and were replaced by so-called metallic first mirrors (FMs) embedded in the neutron shielding, forming an optical labyrinth. Materials eroded from the first wall reactor through physical or chemical sputtering will migrate and will be deposited onto mirrors. Mirrors subject to net deposition will suffer from reflectivity losses due to the deposition of impurities. Cleaning systems of metallic FMs are required in more than 20 optical diagnostic systems in ITER. Plasma cleaning using radio frequency (RF) generated plasmas is currently being considered the most promising in situ cleaning technique. An update of recent results obtained with this technique will be presented. These include the demonstration of cleaning of several deposit types (beryllium, tungsten and beryllium proxy, i.e. aluminium) at 13.56 or 60 MHz as well as large scale cleaning (mirror size: 200 × 300 mm2). Tests under a strong magnetic field up to 3.5 T in laboratory and first experiments of RF plasma cleaning in EAST tokamak will also be discussed. A specific focus will be given on repetitive cleaning experiments performed on several FM material candidates.

  16. Status of beryllium study for fusion in RF

    International Nuclear Information System (INIS)

    Khomutov, A.M.; Kupriyanov, I.B.; Markushkin, Yu.E.; Gervash, A.; Kolbasov, B.N.

    2004-01-01

    The main directions of research activities in the field of beryllium application science and technology carried out in Russia during 2001-2003 have been reviewed. The main results of these investigations have been highlighted. First wall and port-limier. The investigation on the actively cooled components with beryllium cladding is under progress objecting on the clarification of their ultimate thermo cycling capabilities. The study of behavior of bulk beryllium and the boundary region of the contact with the cooling structure under the intensive thermo cycling loading and neutron irradiation have been the object of consideration in particular. The works on the optimization and modification of industrial fabrication processes for commercial scaled production of beryllium tile were also under way. The influence of neutron irradiation. The new experimental data on the nuclear properties of several Russian beryllium grades has been obtained. The samples have been subjected to the high neutron dozes. The influence of low temperature (70-200degree C) neutron irradiation on the thermal conductivity has been examined in particular. The interrelations of the helium inventory and temperature of neutron irradiation with tritium release out of irradiated beryllium samples have been analyzed. The beryllium associated safety questions. The experiments on the modeling of normal working conditions and conditions imitating the plasma disruption events in ITER performance scenario have been continued. The new experimental information on the coefficient of pulverization of beryllium and the accumulation of deuterium in beryllium under the action of proton beam has been collected. The dependence of the reaction rate constant for the beryllium oxidation by the water vapor for different conditions has been analyzed. The compact, porous and powder beryllium samples have been tested at the wide range of temperature, pressure and duration of reaction with water vapor. The calculating

  17. Corrosion of beryllium

    International Nuclear Information System (INIS)

    Mueller, J.J.; Adolphson, D.R.

    1987-01-01

    The corrosion behavior of beryllium in aqueous and elevated-temperature oxidizing environments has been extensively studied for early-intended use of beryllium in nuclear reactors and in jet and rocket propulsion systems. Since that time, beryllium has been used as a structural material in les corrosive environments. Its primary applications include gyro systems, mirror and reentry vehicle structures, and aircraft brakes. Only a small amount of information has been published that is directly related to the evaluation of beryllium for service in the less severe or normal atmospheric environments associated with these applications. Despite the lack of published data on the corrosion of beryllium in atmospheric environments, much can be deduced about its corrosion behavior from studies of aqueous corrosion and the experiences of fabricators and users in applying, handling, processing, storing, and shipping beryllium components. The methods of corrosion protection implemented to resist water and high-temperature gaseous environments provide useful information on methods that can be applied to protect beryllium for service in future long-term structural applications

  18. Beryllium poisonings; Les intoxications par le beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Alibert, S.

    1959-03-15

    This note reports a bibliographical study of beryllium toxicity. Thus, this bibliographical review addresses and outlines aspects and issues like aetiology, cases of acute poisoning (cutaneous manifestations, pulmonary manifestations), chronic poisoning (cutaneous, pulmonary and bone manifestations), excretion and localisation, and prognosis.

  19. Mechanical properties of irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Beeston, J.M.; Longhurst, G.R.; Wallace, R.S. (EG and G Idaho, Inc., Idaho Falls, ID (United States). Idaho National Engineering Lab.); Abeln, S.P. (EG and G Rocky Flats, Inc., Golden, CO (United States))

    1992-10-01

    Beryllium is planned for use as a neutron multiplier in the tritium breeding blanket of the International Thermonuclear Experimental Reactor (ITER). After fabricating samples of beryllium at densities varying from 80 to 100% of the theoretical density, we conducted a series of experiments to measure the effect of neutron irradiation on mechanical properties, especially strength and ductility. Samples were irradiated in the Advanced Test Reactor (ATR) to a neutron fluence of 2.6 x 10[sup 25] n/m[sup 2] (E > MeV) at an irradiation temperature of 75deg C. These samples were subsequently compression-tested at room temperature, and the results were compared with similar tests on unirradiated specimens. We found that the irradiation increased the strength by approximately four times and reduced the ductility to approximately one fourth. Failure was generally ductile, but the 80% dense irradiated samples failed in brittle fracture with significant generation of fine particles and release of small quantities of tritium. (orig.).

  20. Study beryllium microplastic deformation

    International Nuclear Information System (INIS)

    Papirov, I.I.; Ivantsov, V.I.; Nikolaenko, A.A.; Shokurov, V.S.; Tuzov, Yu.V.

    2015-01-01

    Microplastic flow characteristics systematically studied for different varieties beryllium. In isostatically pressed beryllium it decreased with increasing particle size of the powder, increasing temperature and increasing the pressing metal purity. High initial values of the limit microelasticity and microflow in some cases are due a high level of internal stresses of thermal origin and over time it can relax slowly. During long-term storage of beryllium materials with high initial resistance values microplastic deformation microflow limit and microflow stress markedly reduced, due mainly to the relaxation of thermal microstrain

  1. Preparation of beryllium hydride

    International Nuclear Information System (INIS)

    Lowrance, B.R.

    1975-01-01

    A process is described for the preparation of beryllium hydride which comprises pyrolyzing, while in solution in a solvent inert under the reaction conditions, with respect to reactants and products and at a temperature in the range of about 100 0 to about 200 0 C, sufficient to result in the formation of beryllium hydride, a di-t-alkyl beryllium etherate wherein each tertiary alkyl radical contains from 4 to 20 carbon atoms. The pyrolysis is carried out under an atmosphere inert under the reaction conditions, with respect to reactants and products. (U.S.)

  2. Status of beryllium development for fusion applications

    International Nuclear Information System (INIS)

    Billone, M.C.; Donne, M.D.; Macaulay-Newcombe, R.G.

    1994-05-01

    Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma facing component of first wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold isostatic pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation

  3. New facility for post irradiation examination of neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi

    1995-01-01

    Beryllium is expected as a neutron multiplier and plasma facing materials in the fusion reactor, and the neutron irradiation data on properties of beryllium up to 800 degrees C need for the engineering design. The acquisition of data on the tritium behavior, swelling, thermal and mechanical properties are first priority in ITER design. Facility for the post irradiation examination of neutron irradiated beryllium was constructed in the hot laboratory of Japan Materials Testing Reactor to get the engineering design data mentioned above. This facility consist of the four glove boxes, dry air supplier, tritium monitoring and removal system, storage box of neutron irradiated samples. Beryllium handling are restricted by the amount of tritium;7.4 GBq/day and 60 Co;7.4 MBq/day

  4. Beryllium and copper-beryllium alloys; Beryllium und Kupfer-Beryllium-Legierungen

    Energy Technology Data Exchange (ETDEWEB)

    Nagel, Nikolaus [Materion Brush GmbH, Stuttgart (Germany). Operation and Quality/EH and S

    2017-02-15

    The light metal beryllium is a comparatively rare element, which today is primarily derived from bertrandite. It is mainly used as pure metal or in the form of copper-beryllium alloys, e.g., in automotive industry, aerospace, and electrical components. The wide range of applications is mainly attributed to the extremely high rigidity/density ratio. An overview of the history of the metal, its production, and recycling as well as the properties of CuBe alloys are given.

  5. The immunotoxicity of beryllium

    International Nuclear Information System (INIS)

    Reeves, A.L.

    1983-01-01

    In the disease berylliosis, granulomatous hypersensitivity is the specific immune response to tissue contact with a poorly soluble particle of beryllium compound, mediated through the accumulation and proliferation of reticuloendothelial cells. A review is given of the work accomplished since the 1950's and particularly since the 1970's to elucidate the nature and consequences of this response to beryllium and its compounds. (U.K.)

  6. Preparation of beryllium hydride

    International Nuclear Information System (INIS)

    Bergeron, C.R.; Baker, R.W.

    1975-01-01

    Beryllium hydride of high bulk density, suitable for use as a component of high-energy fuels, is prepared by the pyrolysis, in solution in an inert solvent, of a ditertiary-alkyl beryllium. An agitator introduces mechanical energy into the reaction system, during the pyrolysis, at the rate of 0.002 to 0.30 horsepower per gallon of reaction mixture. (U.S.)

  7. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    2001-01-01

    The divertor ''Large Project'' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  8. The ITER divertor cassette project

    International Nuclear Information System (INIS)

    Ulrickson, M.; Tivey, R.; Akiba, M.

    1999-01-01

    The divertor 'Large Project' was conceived with the aim of demonstrating the feasibility of meeting the lifetime requirements by employing the candidate armor materials of beryllium, tungsten (W) and carbon-fiber-composite (CFC). At the start, there existed only limited experience with constructing water-cooled high heat flux armored components for tokamaks. To this was added the complication posed by the need to use a silver-free joining technique that avoids the transmutation of n-irradiated silver to cadmium. The research project involving the four Home Teams (HTs) has focused on the design, development, manufacture and testing of full-scale Plasma Facing Components (PFCs) suitable for ITER. The task addressed all the issues facing ITER divertor design, such as providing adequate armor erosion lifetime, meeting the required armor-heat sink joint lifetime and heat sink fatigue life, sustaining thermal-hydraulic and electromechanical loads, and seeking to identify the most cost-effective manufacturing options. This paper will report the results of the divertor large project. (author)

  9. Beryllium coating produced by evaporation-condensation method and some their properties

    Energy Technology Data Exchange (ETDEWEB)

    Pepekin, G.I.; Anisimov, A.B.; Chernikov, A.S.; Mozherinn, S.I.; Pirogov, A.A. [SRI SIA Lutch., Podolsk (Russian Federation)

    1998-01-01

    The method of vacuum evaporation-condensation for deposition of beryllium coatings on metal substrates, considered in the paper, side by side with a plasma-spray method is attractive fon ITER application. In particular this technique may be useful for repair the surface of eroded tiles which is operated in a strong magnetic field. The possibility of deposition of beryllium coatings with the rate of layer growth 0.1-0.2 mm/h is shown. The compatibility of beryllium coating with copper or stainless steel substrate is provided due to intermediate barrier. The results of examination of microstructure, microhardness, porosity, thermal and physical properties and stability under thermal cycling of beryllium materials are presented. The value of thermal expansion coefficient and thermal conductivity of condensed beryllium are approximately the same as for industrial grade material produced by powder mettalurgy technique. However, the condensed beryllium has higher purity (up to 99.9-99.99 % wt.). (author)

  10. 5. IEA International workshop on beryllium technology for fusion. Book of abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

    The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be{sub 12}Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented.

  11. 5. IEA International workshop on beryllium technology for fusion. Book of abstracts

    International Nuclear Information System (INIS)

    2001-01-01

    The collection includes the abstracts of reports presented to the 5-th IEA international workshop on beryllium technology for fusion. The themes of reports are as follows: status of beryllium technology for fusion in Russia; manufacturing and testing of Be armoured first wall mock-up for ITER; development of the process of diffusion welding of metals stainless steel-copper-beryllium into a single composite; some features of beryllium-laser beam interaction; the effect of irradiation dose on tritium and helium release from neutron irradiated beryllium; thermal properties of neutron irradiated Be 12 Ti. The results of investigating the mechanical properties variation and swelling of beryllium under high temperature neutron irradiation are presented [ru

  12. Status of beryllium development for fusion applications

    International Nuclear Information System (INIS)

    Billone, M.C.; Macaulay-Newcombe, R.G.

    1995-01-01

    Beryllium is a leading candidate material for the neutron multiplier of tritium breeding blankets and the plasma-facing component of first-wall and divertor systems. Depending on the application, the fabrication methods proposed include hot-pressing, hot-isostatic-pressing, cold-isostatic-pressing/sintering, rotary electrode processing and plasma spraying. Product forms include blocks, tubes, pebbles, tiles and coatings. While, in general, beryllium is not a leading structural material candidate, its mechanical performance, as well as its performance with regard to sputtering, heat transport, tritium retention/release, helium-induced swelling and chemical compatibility, is an important consideration in first-wall/blanket design. Differential expansion within the beryllium causes internal stresses which may result in cracking, thereby affecting the heat transport and barrier performance of the material. Overall deformation can result in loading of neighboring structural material. Thus, in assessing the performance of beryllium for fusion applications, it is important to have a good database in all of these performance areas, as well as a set of properties correlations and models for the purpose of interpolation/extrapolation.In this current work, the range of anticipated fusion operating conditions is reviewed. The thermal, mechanical, chemical compatibility, tritium retention/release, and helium retention/swelling databases are then reviewed for fabrication methods and fusion operating conditions of interest. Properties correlations and uncertainty ranges are also discussed. In the case of the more complex phenomena of tritium retention/release and helium-induced swelling, fundamental mechanisms and models are reviewed in more detail. Areas in which additional data are needed are highlighted, along with some trends which suggest ways of optimizing the performance of beryllium for fusion applications. (orig.)

  13. ITER shielding blanket

    Energy Technology Data Exchange (ETDEWEB)

    Strebkov, Yu [ENTEK, Moscow (Russian Federation); Avsjannikov, A [ENTEK, Moscow (Russian Federation); Baryshev, M [NIAT, Moscow (Russian Federation); Blinov, Yu [ENTEK, Moscow (Russian Federation); Shatalov, G [KIAE, Moscow (Russian Federation); Vasiliev, N [KIAE, Moscow (Russian Federation); Vinnikov, A [ENTEK, Moscow (Russian Federation); Chernjagin, A [DYNAMICA, Moscow (Russian Federation)

    1995-03-01

    A reference non-breeding blanket is under development now for the ITER Basic Performance Phase for the purpose of high reliability during the first stage of ITER operation. More severe operation modes are expected in this stage with first wall (FW) local heat loads up to 100-300Wcm{sup -2}. Integration of a blanket design with protective and start limiters requires new solutions to achieve high reliability, and possible use of beryllium as a protective material leads to technologies. The rigid shielding blanket concept was developed in Russia to satisfy the above-mentioned requirements. The concept is based on a copper alloy FW, austenitic stainless steel blanket structure, water cooling. Beryllium protection is integrated in the FW design. Fabrication technology and assembly procedure are described in parallel with the equipment used. (orig.).

  14. Assessment of the feasibility and advantages of beryllium recycling

    International Nuclear Information System (INIS)

    Druyts, F.; Braet, J.; Ooms, L.

    2006-01-01

    This paper proposes a generic route for the recycling of beryllium from fusion reactors, based on critical issues associated with beryllium pebbles after their service life in the HCPB breeding blanket. These critical issues are the high tritium inventory, the presence of long-lived radionuclides (among which transuranics due to traces of uranium in the base metal), and the chemical toxicity of beryllium. On the basis of the chemical and radiochemical characteristics of the neutron irradiated beryllium pebbles, we describe a possible recycling route. The first step is the detritiation of the material. This can be achieved by heating the pebbles to 800 o C under an argon flow. The argon gas avoids oxidation of the beryllium, and at the proposed temperature the tritium inventory is readily released from the pebbles. In a second step, the released tritium can be oxidised on a copper oxide bed to produce tritiated water, which is consistent with the current international strategy to convert all kinds of tritiated waste into tritiated water, which can subsequently be treated in a water detritiation plant. Removal of radionuclides from the beryllium pebbles may be achieved by several types of chloride processes. The first step is to pass chlorine gas (in an argon flow) over the pebbles, thus yielding volatile BeCl 2 . This beryllium chloride can then be purified by fractional distillation. As a small fraction of the beryllium chloride contains the long-lived 10 Be isotope, 10BeCl 2 has to be separated from 9BeCl 2 , which could be achieved by centrifugal techniques. The product can then be reduced to obtain high-purity metallic beryllium. Two candidate reduction methods were identified: fused salt electrolysis and thermal decomposition. Both these methods require laboratory parametric studies to maximise the yield and achieve a high purity metal, before either process can be upgraded to a larger scale. The eventual product of the chloride reduction process must be a high

  15. Sintering of beryllium oxide

    International Nuclear Information System (INIS)

    Caillat, R.; Pointud, R.

    1955-01-01

    This study had for origin to find a process permitting to manufacture bricks of beryllium oxide of pure nuclear grade, with a density as elevated as possible and with standardized shape. The sintering under load was the technique kept for the manufacture of the bricks. Because of the important toxicity of the beryllium oxide, the general features for the preliminary study of the sintering, have been determined while using alumina. The obtained results will be able to act as general indication for ulterior studies with sintering under load. (M.B.) [fr

  16. Modeling tritium processes in plasma-facing beryllium

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Dolan, T.J.; Mulock, M.J.

    1995-01-01

    In this paper we present techniques and recommended parameters for modeling tritium implantation, trapping and release, and permeation, in beryllium-clad structures adjacent to the plasma. Among the features that should be considered are the effects of surface films, the mobility of beryllium through those films, damage caused by ion implantation, especially in regions where pitting may be expected, and bubble formation. Tritium transport parameters recommended are based on fits with experimental data and available theory. Estimates of inventories in ITER using these parameters are also given. 31 refs., 2 figs., 1 tab

  17. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report

  18. Permeation behavior of deuterium implanted into beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Nakamura, Hirofumi; Hayashi, Takumi; O' hira, Shigeru; Nishi, Masataka [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-09-01

    Study on Implantation Driven Permeation (IDP) behavior of deuterium through pure beryllium was investigated as a part of the research to predict the tritium permeation through the first wall components ITER (International Thermonuclear Experimental Reactor). The permeation experiments were carried out with two beryllium specimens, one was an unannealed specimen and the other was that annealed at 1173 K. The permeation flux was measured as a function of specimen temperature and incident ion flux. Surface analysis of specimen was also carried out after the permeation experiment. Permeation was observed only with the annealed specimen and no significant permeation was observed with unannealed specimen under the present experimental condition (maximum temperature: 685 K, detection limit: 1x10{sup 13} D atoms/m{sup 2}s). It could be attributed that the intrinsic lattice defects, which act as diffusion preventing site, decreased with the specimen annealing. Based on the result of steady and transient permeation behavior and surface analysis, it was estimated that the deuterium permeation implanted into annealed beryllium was controlled by surface recombination due to the oxide layer on the surface of the permeated side. (author)

  19. ITER blanket designs

    International Nuclear Information System (INIS)

    Gohar, Y.; Parker, R.; Rebut, P.H.

    1995-01-01

    The ITER first wall, blanket, and shield system is being designed to handle 1.5±0.3 GW of fusion power and 3 MWa m -2 average neutron fluence. In the basic performance phase of ITER operation, the shielding blanket uses austenitic steel structural material and water coolant. The first wall is made of bimetallic structure, austenitic steel and copper alloy, coated with beryllium and it is protected by beryllium bumper limiters. The choice of copper first wall is dictated by the surface heat flux values anticipated during ITER operation. The water coolant is used at low pressure and low temperature. A breeding blanket has been designed to satisfy the technical objectives of the Enhanced Performance Phase of ITER operation for the Test Program. The breeding blanket design is geometrically similar to the shielding blanket design except it is a self-cooled liquid lithium system with vanadium structural material. Self-healing electrical insulator (aluminum nitride) is used to reduce the MHD pressure drop in the system. Reactor relevancy, low tritium inventory, low activation material, low decay heat, and a tritium self-sufficiency goal are the main features of the breeding blanket design. (orig.)

  20. HEINBE; the calculation program for helium production in beryllium under neutron irradiation

    International Nuclear Information System (INIS)

    Shimakawa, Satoshi; Ishitsuka, Etsuo; Sato, Minoru

    1992-11-01

    HEINBE is a program on personal computer for calculating helium production in beryllium under neutron irradiation. The program can also calculate the tritium production in beryllium. Considering many nuclear reactions and their multi-step reactions, helium and tritium productions in beryllium materials irradiated at fusion reactor or fission reactor may be calculated with high accuracy. The calculation method, user's manual, calculated examples and comparison with experimental data were described. This report also describes a neutronics simulation method to generate additional data on swelling of beryllium, 3,000-15,000 appm helium range, for end-of-life of the proposed design for fusion blanket of the ITER. The calculation results indicate that helium production for beryllium sample doped lithium by 50 days irradiation in the fission reactor, such as the JMTR, could be achieved to 2,000-8,000 appm. (author)

  1. Advanced solutions for beryllium and tungsten plasma-facing components

    International Nuclear Information System (INIS)

    Ibbott, C.; Jakeman, R.; Ando, T.; Chiocchio, S.; Federici, G.; Heidl, H.; Tivey, R.; Falter, H.; Ciric, D.; Merola, M.; Vieider, G.; Ploechl, L.; Roedig, M.

    1998-01-01

    Beryllium and tungsten are candidate plasma-facing armour materials for the International Thermonuclear Experimental Reactor (ITER). These armours are proposed for areas with low heat flux (≤5 MW m -2 ); however, in the divertor, surface melting during abnormal events may occur. This paper reports the progress made in developing novel approaches to solving the difficulties posed in designing with these armours. A Be monoblock brazed to an OFHC 10 mm ID Cu tube using InCuSil 'ABA' braze alloy has survived 130 cycles of 10-11 MW m -2 for 6 s, with surface temperatures of 1250 C. No visible surface cracking occurred. The same monoblock was then exposed to several cycles of 20-22 MW m -2 for 8 s, creating a 2 mm deep molten layer. High cycle fatigue was then performed. The test results are detailed in this paper. Comparison between experimental and theoretical results are made. W and Cu have a large mismatch in their thermal expansion coefficients and two designs are proposed that minimise the interface stresses. These are: a 'brush'-like structure with rectangular fibres set in a Cu substrate using the 'active metal casting' (AMC) technique; and thin monoblocks (or lamellae) brazed or active metal cast onto a Cu tube. Analyses of the lamellae concept for steady-state heat loads of 5 MW m -2 are presented. Fatigue analyses show that both solutions are theoretically viable (∝10 4 cycles). A 'brush' mock-up has been manufactured and progress on its testing is reported. Results of all tests and their relevance to the ITER design are discussed. (orig.)

  2. Thermal effects on beryllium mirrors

    International Nuclear Information System (INIS)

    Weinswig, S.

    1989-01-01

    Beryllium is probably the most frequently used material for spaceborne system scan mirrors. Beryllium's properties include lightweightedness, high Young's modulus, high stiffness value, high resonance value. As an optical surface, beryllium is usually nickel plated in order to produce a higher quality surface. This process leads to the beryllium mirror acting like a bimetallic device. The mirror's deformation due to the bimetallic property can possibly degrade the performance of the associated optical system. As large space borne systems are designed and as temperature considerations become more crucial in the instruments, the concern about temporal deformation of the scan mirrors becomes a prime consideration. Therefore, two sets of tests have been conducted in order to ascertain the thermal effects on nickel plated beryllium mirrors. These tests are categorized. The purpose of this paper is to present the values of the bimetallic effect on typical nickel plated beryllium mirrors

  3. Beryllium. Its minerals. Pt. 1

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1990-01-01

    With this work a series of reports begins, under the generic name 'Beryllium', related to several aspects of beryllium technology. The target is to update, with critical sense, current bibliographic material in order to be used in further applications. Some of the most important beryllium ores, the Argentine emplacement of their deposits and world occurrence are described. Argentine and world production, resources and reserves are indicated here as well. (Author) [es

  4. Beryllium Metal Supply Options

    Science.gov (United States)

    1989-01-01

    Carcinogen Assessment Group (U.S. Environmental Protec- tion Agency, 1987). The National Institute of Occupational Safety and Health is currently examining...Epidemiology of beryllium intox- ication. Arch. Ind. Hyg. Occup . Med. 4:123-151. U. S. Environmental Protection Agency 1987. Health Assessment Document...1957 to 1961. He rejoined the Bureau of Mines in 1961 as the aluminum and bauxite commcdity specialist. In 1973 he became chief of the Division of

  5. Beryllium and zirconium

    International Nuclear Information System (INIS)

    Salesse, Marc

    1959-01-01

    Pure beryllium and zirconium, both isolated at about the same date but more than a century ago remained practically unused for eighty years. Fifteen years ago they were released from this state of inactivity by atomic energy, which made them into current metal a with an annual production which runs into tens of tons for the one and thousands for the other. The reasons for this promotion promise well for the future of the two metals, which moreover will probably find additional uses in other branches of industry. The attraction of beryllium and zirconium for atomic energy is easily explained. The curve of figure 1 gives the price per gram of uranium-235 as a function of enrichment: this price increases by about a factor of 3 on passing from natural uranium (0, 7 percent 235 U) to almost pure uranium-235. Because of their tow capture cross-section beryllium and zirconium make it possible, or at least easier, to use natural uranium and they thus enjoy an advantage the extent of which must be calculated for each reactor or fuel element project, but which is generally considerable. It will be seen later that this advantage should be based on figures which are even more favourable that would appear from the simple ratio 3 of the price of pure uranium- 235 contained in natural uranium. Reprint of a paper published in 'Industries Atomiques' - n. 1-2, 1959

  6. Characterization of shocked beryllium

    Directory of Open Access Journals (Sweden)

    Papin P.A.

    2012-08-01

    Full Text Available While numerous studies have investigated the low-strain-rate constitutive response of beryllium, the combined influence of high strain rate and temperature on the mechanical behavior and microstructure of beryllium has received limited attention over the last 40 years. In the current work, high strain rate tests were conducted using both explosive drive and a gas gun to accelerate the material. Prior studies have focused on tensile loading behavior, or limited conditions of dynamic strain rate and/or temperature. Two constitutive strength (plasticity models, the Preston-Tonks-Wallace (PTW and Mechanical Threshold Stress (MTS models, were calibrated using common quasi-static and Hopkinson bar data. However, simulations with the two models give noticeably different results when compared with the measured experimental wave profiles. The experimental results indicate that, even if fractured by the initial shock loading, the Be remains sufficiently intact to support a shear stress following partial release and subsequent shock re-loading. Additional “arrested” drive shots were designed and tested to minimize the reflected tensile pulse in the sample. These tests were done to both validate the model and to put large shock induced compressive loads into the beryllium sample.

  7. Deuterium/hydrogen isotope exchange on beryllium and beryllium nitride; Deuterium/Wasserstoff-Isotopenaustausch an Beryllium und Berylliumnitrid

    Energy Technology Data Exchange (ETDEWEB)

    Dollase, Petra; Eichler, Michael; Koeppen, Martin; Dittmar, Timo; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik (Germany)

    2016-07-01

    In the fusion experiments JET and ITER, the first wall is made up of beryllium. The use of nitrogen is discussed for radiative cooling in the divertor. This can react with the surface of the first wall to form beryllium nitride (Be{sub 3}N{sub 2}). The hydrogen isotopes deuterium and tritium, which react in the fusion reaction to helium and a neutron, are used as fuel. Since the magnetic confinement of the plasma is not perfect, deuterium and tritium ions are also found on the beryllium wall and can accumulate there. This should be avoided due to the radioactivity of tritium. Therefore the isotope exchange with deuterium is investigated to regenerate the first wall. We investigate the isotopic exchange of deuterium and protium in order to have not to work with radioactive tritium. The ion bombardment is simulated with an ion source. With voltages up to a maximum of 5 kV, deuterium and protic hydrogen ions are implanted in polycrystalline Be and Be{sub 3}N{sub 2}. The samples are then analyzed in situ using X-ray photoelectron spectroscopy (XPS) and thermal desorption spectroscopy (TDS). Subsequently, samples prepared under the same conditions are characterized ex-situ by means of nuclear reaction analysis (NRA). [German] In den Fusionsexperimenten JET und ITER besteht die erste Wand im Hauptraum aus Beryllium (Be). Zur Strahlungskuehlung im Divertor wird der Einsatz von Stickstoff diskutiert. Dieser kann mit der Oberflaeche der ersten Wand zu Berylliumnitrid (Be{sub 3}N{sub 2}) reagieren. Als Brennstoff werden die Wasserstoffisotope Deuterium und Tritium eingesetzt, die in der Fusionsreaktion zu Helium und einem Neutron reagieren. Da der magnetische Einschluss des Plasmas nicht perfekt ist, treffen auch Deuterium- und Tritiumionen auf die Berylliumwand auf und koennen sich dort anreichern. Das soll aufgrund der Radioaktivitaet von Tritium unbedingt vermieden werden. Daher wird zur Regenerierung der ersten Wand der Isotopenaustausch mit Deuterium untersucht. Wir

  8. Beryllium. Beryllium oxide, obtention and properties. Pt.4

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1991-01-01

    As a continuation of the 'Beryllium' series this work reviews several methods of high purity beryllia production. Diverse methods of obtention and purification from different beryllium compounds are described. Some chemical, mechanical and electrical properties related with beryllia obtention methods are summarized. (Author) [es

  9. Proceedings of the third IEA international workshop on beryllium technology for fusion

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi; Okamoto, Makoto

    1998-01-01

    This report is the Proceedings of the Third International Energy Agency International Workshop on Beryllium Technology for Fusion. The workshop was held on October 22-24, 1997, at the Sangyou Kaikan in Mito City with 68 participants who attended from the Europe, the Russian Federation, the Kazakstan, the United States and Japan. The topics for papers were arranged into 9 sessions; beryllium applications for ITER, production and characterization, chemical compatibility and corrosion, forming and joining, plasma/tritium interactions, beryllium coating, first wall applications, neutron irradiation effects, health and safety. To utilize beryllium in the pebble type blanket, a series of discussions were intensified in multiple view points such as the swelling, He/T release from beryllium pebble irradiated up to high He content, effective thermal conductivity, tritium permeation and coating, and fabrication cost, and so on. As the plasma facing material, life time of beryllium and coated beryllium, dust and particle production, joining, waste treatment, mechanical properties and deformation by swelling were discussed as important issues. Especially, it was recognized throughout the discussions that the comparative study by the different researchers should be carried out to establish the reliability of the data reported in the workshop and in others. To enhance the comparative study, the world wide collaboration for the relative evaluation of the beryllium was proposed by the International Organization Committee and the proposal was approved by all of the participants. The 45 of the presented papers are indexed individually. (J.P.N.)

  10. Plasma spraying of beryllium and beryllium-aluminum-silver alloys

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Elliott, K.E.; Jacobson, L.A.

    1994-01-01

    A preliminary investigation on plasma-spraying of beryllium and a beryllium-aluminum-4% silver alloy was done at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility (BATSF). Spherical Be and Be-Al-4%Ag powders, which were produced by centrifugal atomization, were used as feedstock material for plasma-spraying. The spherical morphology of the powders allowed for better feeding of fine (<38 μm) powders into the plasma-spray torch. The difference in the as-deposited densities and deposit efficiencies of the two plasma-sprayed powders will be discussed along with the effect of processing parameters on the as-deposited microstructure of the Be-Al-4%Ag. This investigation represents ongoing research to develop and characterize plasma-spraying of beryllium and beryllium-aluminum alloys for magnetic fusion and aerospace applications

  11. Plasma spraying of beryllium and beryllium-aluminum-silver alloys

    International Nuclear Information System (INIS)

    Castro, R.G.; Stanek, P.W.; Elliott, K.E.; Jacobson, L.A.

    1993-01-01

    A preliminary investigation on plasma-spraying of beryllium and a beryllium-aluminum 4% silver alloy was done at the Los Alamos National Laboratory's Beryllium Atomization and Thermal Spray Facility (BATSF). Spherical Be and Be-Al-4%Ag powders, which were produced by centrifugal atomization, were used as feedstock material for plasma-spraying. The spherical morphology of the powders allowed for better feeding of fine (<38 μm) powders into the plasma-spray torch. The difference in the as-deposited densities and deposit efficiencies of the two plasma-sprayed powders will be discussed along with the effect of processing parameters on the as-deposited microstructure of the Be-Al-4%Ag. This investigation represents ongoing research to develop and characterize plasma-spraying of beryllium and beryllium-aluminum alloys for magnetic fusion and aerospace applications

  12. Beryllium for first wall, limiter and divertor - a literature survey

    International Nuclear Information System (INIS)

    Schuster, A.; Smid, I.; Kny, E.

    1994-01-01

    A survey of the topical literature on beryllium as material for plasma interactive components in future fusion devices is given. The radiation damage which can be expected as a result of the neutron irradiation from ignited tokamak plasma is discussed. The response to high heat fluxes and simulation experiments in different test facilities are referred. Another focus will be on the material properties literature data, on joining techniques and on compatibility with other materials. The performance of a beryllium coated first wall at JET is reported. Some relevant literature on other candidate materials for plasma interactive components shall be considered

  13. Technical Basis for PNNL Beryllium Inventory

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, Michelle Lynn

    2014-07-09

    The Department of Energy (DOE) issued Title 10 of the Code of Federal Regulations Part 850, “Chronic Beryllium Disease Prevention Program” (the Beryllium Rule) in 1999 and required full compliance by no later than January 7, 2002. The Beryllium Rule requires the development of a baseline beryllium inventory of the locations of beryllium operations and other locations of potential beryllium contamination at DOE facilities. The baseline beryllium inventory is also required to identify workers exposed or potentially exposed to beryllium at those locations. Prior to DOE issuing 10 CFR 850, Pacific Northwest Nuclear Laboratory (PNNL) had documented the beryllium characterization and worker exposure potential for multiple facilities in compliance with DOE’s 1997 Notice 440.1, “Interim Chronic Beryllium Disease.” After DOE’s issuance of 10 CFR 850, PNNL developed an implementation plan to be compliant by 2002. In 2014, an internal self-assessment (ITS #E-00748) of PNNL’s Chronic Beryllium Disease Prevention Program (CBDPP) identified several deficiencies. One deficiency is that the technical basis for establishing the baseline beryllium inventory when the Beryllium Rule was implemented was either not documented or not retrievable. In addition, the beryllium inventory itself had not been adequately documented and maintained since PNNL established its own CBDPP, separate from Hanford Site’s program. This document reconstructs PNNL’s baseline beryllium inventory as it would have existed when it achieved compliance with the Beryllium Rule in 2001 and provides the technical basis for the baseline beryllium inventory.

  14. Solid state bonding of beryllium to copper and vanadium

    International Nuclear Information System (INIS)

    Floyd, D.R.; Liby, A.L.; Weaver, W.

    1993-01-01

    The intent of this effort was to demonstrate that ingot metallurgy (IM) beryllium (Be) can be bonded to dissimilar metals such as copper (Cu) or vanadium (V) at low temperatures by using silver (Ag) as a bonding aid. It is hoped that success at the coupon stage will stimulate more extensive studies of the mechanical and thermal integrity of such joints, leading ultimately to use of this technology to fabricate first wall structures for ITER. (orig.)

  15. Method for fabricating beryllium structures

    Science.gov (United States)

    Hovis, Jr., Victor M.; Northcutt, Jr., Walter G.

    1977-01-01

    Thin-walled beryllium structures are prepared by plasma spraying a mixture of beryllium powder and about 2500 to 4000 ppm silicon powder onto a suitable substrate, removing the plasma-sprayed body from the substrate and placing it in a sizing die having a coefficient of thermal expansion similar to that of the beryllium, exposing the plasma-sprayed body to a moist atmosphere, outgassing the plasma-sprayed body, and then sintering the plasma-sprayed body in an inert atmosphere to form a dense, low-porosity beryllium structure of the desired thin-wall configuration. The addition of the silicon and the exposure of the plasma-sprayed body to the moist atmosphere greatly facilitate the preparation of the beryllium structure while minimizing the heretofore deleterious problems due to grain growth and grain orientation.

  16. Doped beryllium lanthanate crystals

    International Nuclear Information System (INIS)

    1974-01-01

    Monocrystals of doped beryllium lanthanate, Be 2 Lasub(2-2x)Zsub(2x)O 5 --where Z may be any rare earth, but preferably neodymium, and x may have values between 0.001 and 0.2, but preferably between 0.007 and 0.015-- are recommended as laser hosts. They are softer and may be grown at a lower temperature than Y 3 A1 5 O 12 :Nd (YAG:Nd). Their chemical composition and preparation are described. An example of an optically pumped laser apparatus with this type of monocrystal as laser host is presented

  17. Investigations of the ternary system beryllium-carbon-tungsten and analyses of beryllium on carbon surfaces

    International Nuclear Information System (INIS)

    Kost, Florian

    2009-01-01

    Beryllium, carbon and tungsten are planned to be used as first wall materials in the future fusion reactor ITER. The aim of this work is a characterization of mixed material formation induced by thermal load. To this end, model systems (layers) were prepared and investigated, which give insight into the basic physical and chemical concepts. Before investigating ternary systems, the first step was to analyze the binary systems Be/C and Be/W (bottom-up approach), where the differences between the substrates PG (pyrolytic graphite) and HOPG (highly oriented pyrolytic graphite) were of special interest. Particularly X-ray photoelectron spectroscopy (XPS), low energy ion scattering (ISS) and Rutherford backscattering spectroscopy (RBS) were used as analysis methods. Beryllium evaporated on carbon shows an island growth mode, whereas a closed layer can be assumed for layer thicknesses above 0.7 nm. Annealing of the Be/C system induces Be 2 C island formation for T≥770 K. At high temperatures (T≥1170 K), beryllium carbide dissociates, resulting in (metallic) beryllium desorption. For HOPG, carbide formation starts at higher temperatures compared to PG. Activation energies for the diffusion processes were determined by analyzing the decreasing beryllium amount versus annealing time. Surface morphologies were characterized using angle-resolved XPS (ARXPS) and atomic force microscopy (AFM). Experiments were performed to study processes in the Be/W system in the temperature range from 570 to 1270 K. Be 2 W formation starts at 670 K, a complete loss of Be 2 W is observed at 1170 K due to dissociation (and subsequent beryllium desorption). Regarding ternary systems, particularly Be/C/W and C/Be/W were investigated, attaching importance to layer thickness (reservoir) variations. At room temperature, Be 2 C, W 2 C, WC and Be 2 W formation at the respective interfaces was observed. Further Be 2 C is forming with increasing annealing temperatures. Depending on the layer

  18. Thermal fatigue of beryllium

    International Nuclear Information System (INIS)

    Deksnis, E.; Ciric, D.; Falter, H.

    1995-01-01

    Thermal fatigue life of S65c beryllium castellated to a geometry 6 x 6 x (8-10)mm deep has been tested for steady heat fluxes of 3 MW/m 2 to 5 MW/m 2 and under pulsed heat fluxes (10-20 MW/m 2 ) for which the time averaged heat flux is 5 MW/m 2 . These tests were carried out in the JET neutral beam test facility A test sequence with peak surface temperatures ≤ 600 degrees C produced no visible fatigue cracks. In the second series of tests, with T max ≤ 750 degrees C evidence for fatigue appeared after a minimum of 1350 stress cycles. These fatigue data are discussed in view of the observed lack of thermal fatigue in JET plasma operations with beryllium PFC. JET experience with S65b and S65c is reviewed; recent operations with Φ = 25 MW/m 2 and sustained melting/resolidification are also presented. The need for a failure criterion for finite element analyses of Be PFC lifetimes is discussed

  19. Corrosion of beryllium oxide

    International Nuclear Information System (INIS)

    Elston, J.; Caillat, R.

    1958-01-01

    Data are reported on the volatilization rate of beryllium oxide in moist air depending on temperature and water vapour concentration. They are concerned with powder samples or sintered shapes of various densities. For sintered samples, the volatilization rate is very low under the following conditions: - temperature: 1300 deg. C, - water vapour concentration in moist air: 25 g/m 3 , - flow rate: 12 I/hour corresponding to a speed of 40 m/hour on the surface of the sample. For calcinated powders (1300 deg. C), grain growth has been observed under a stream of moist air at 1100 deg. C. For instance, grain size changes from 0,5 to at least 2 microns after 500 hours of exposure at this temperature. Furthermore, results data are reported on corrosion of sintered beryllium oxide in pressurized water. At 250 deg. C, under a pressure of 40 kg/cm 2 water is very slightly corrosive; however, internal strains are revealed. Finally, some features on the corrosion in liquid sodium are exposed. (author) [fr

  20. The INEL beryllium multiplication experiment

    International Nuclear Information System (INIS)

    Smith, J.R.; King, J.J.

    1991-03-01

    The experiment to measure the multiplication of 14-MeV neutrons in bulk beryllium has been completed. The experiment consists of determining the ratio of 56 Mn activities induced in a large manganese bath by a central 14-MeV neutron source, with and without a beryllium sample surrounding the source. In the manganese bath method a neutron source is placed at the center of a totally-absorbing aqueous solution of MnSo 4 . The capture of neutrons by Mn produces a 56 Mn activity proportional to the emission rate of the source. As applied to the measurement of the multiplication of 14- MeV neutrons in bulk beryllium, the neutron source is a tritium target placed at the end of the drift tube of a small deuteron accelerator. Surrounding the source is a sample chamber. When the sample chamber is empty, the neutrons go directly to the surrounding MnSO 4 solution, and produce a 56 Mn activity proportional to the neutron emission rate. When the chamber contains a beryllium sample, the neutrons first enter the beryllium and multiply through the (n,2n) process. Neutrons escaping from the beryllium enter the bath and produce a 56 Mn activity proportional to the neutron emission rate multiplied by the effective value of the multiplication in bulk beryllium. The ratio of the activities with and without the sample present is proportional to the multiplication value. Detailed calculations of the multiplication and all the systematic effects were made with the Monte Carlo program MCNP, utilizing both the Young and Stewart and the ENDF/B-VI evaluations for beryllium. Both data sets produce multiplication values that are in excellent agreement with the measurements for both raw and corrected values of the multiplication. We conclude that there is not real discrepancy between experimental and calculated values for the multiplication of neutrons in bulk beryllium. 12 figs., 11 tabs., 18 refs

  1. Offshoots from beryllium development programme

    International Nuclear Information System (INIS)

    Sharma, B.P.; Sinha, P.K.

    1995-01-01

    The paper briefly presents extraction and processing of beryllium metal as practiced in the beryllium facilities at Turbhe, New Bombay. These facilities have been set up to meet the indigenous requirements of the metal in space and nuclear science programmes. As offshoot of this beryllium development programme has been the development of a number of pyro and powder metallurgical equipment. Indigenous development of these pieces of equipment has been a professionally rewarding experience. Efforts are now on to promote these equipment for industrial use. (author). 6 refs., 6 figs., 2 tabs

  2. The Status of Beryllium Research for Fusion in the United States

    International Nuclear Information System (INIS)

    Glen R. Longhurst

    2003-01-01

    Use of beryllium in fusion reactors has been considered for neutron multiplication in breeding blankets and as an oxygen getter for plasma-facing surfaces. Previous beryllium research for fusion in the United States included issues of interest to fission (swelling and changes in mechanical and thermal properties) as well as interactions with plasmas and hydrogen isotopes and methods of fabrication. When the United States formally withdrew its participation in the International Thermonuclear Experimental Reactor (ITER) program, much of this effort was terminated. The focus in the U.S. has been mainly on toxic effects of beryllium and on industrial hygiene and health-related issues. Work continued at the INEEL and elsewhere on beryllium-containing molten salts. This activity is part of the JUPITER II Agreement. Plasma spray of ITER first wall samples at Los Alamos National Laboratory has been performed under the European Fusion Development Agreement. Effects of irradiation on beryllium structure are being studied at Oak Ridge National Laboratory. Numerical and phenomenological models are being developed and applied to better understand important processes and to assist with design. Presently, studies are underway at the University of California Los Angeles to investigate thermo-mechanical characteristics of beryllium pebble beds, similar to research being carried out at Forschungszentrum Karlsruhe and elsewhere. Additional work, not funded by the fusion program, has dealt with issues of disposal, and recycling

  3. The status of beryllium research for fusion in the United States

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Snead, L.L.; Abou-Sena, A.A.

    2004-01-01

    Use of beryllium in fusion reactor has been considered for neutron multiplication in breeding blankets an as an oxygen getter for plasma - facing surface. Previous beryllium research for fusion in the United States included issues of interest to fission (swelling an changes in mechanical and thermal properties) as well as interactions with plasmas and hydrogen isotopes and methods of fabrication. When the United States formally withdrew its participation in the International Experimental Reactor (ITER) program, much of this effort was terminated. The focus in the U.S. has been mainly on toxic effects of beryllium and on industrial hygiene and health-related issues. Work continued at the INEEL (Idaho National Engineering and Environmental Laboratory) and elsewhere on beryllium-containing molten salts. This activity is part of the JUPITER II Agreement. Plasma spray of ITER first wall samples at Los Alamos National Laboratory has been performed under the European Fusion Development Agreement. Effects of irradiation on beryllium structure are being studied at Oak Ridge National Laboratory. Numerical and phenomenological models are being developed and applied at the University of California Los Angels to investigate thermo-mechanical characteristics of beryllium pebble beds, similar to research being carried out at Forschungszentrum Karlsruhe and elsewhere. Additional work, not funded by the fusion program, has dealt with issues of disposal, and recycling. (author)

  4. Technical aspects of the joint JET-ISX-B beryllium limiter experiment

    International Nuclear Information System (INIS)

    Edmonds, P.H.; Dietz, K.J.; Mioduszewski, P.K.; Watson, R.D.; Emerson, L.C.; Gabbard, W.A.; Goodall, D.; Simpkins, J.E.; Yarber, J.L.

    1984-01-01

    An experiment has been performed on the Impurity Study Experiment (ISX-B) tokamak to test beryllium as a limiter material. Beryllium is an attractive candidate for a limiter and has been proposed for use in the Joint European Torus (JET) experiment. A temperature-controlled, segmented, beryllium top-rail limiter was located inside the plasma radius described by the existing titanium limiters. An extended set of diagnostics was added for measurement of scrapeoff and limiter parameters. These included visible and infrared monitoring systems, probes, and surface analysis experiments. Tokamak experiments included parameter surveys of both ohmically heated and neutral-beam-heated plasmas and an extended fluence test of the limiter

  5. Beryllium armored mockups for fusion high flux application

    International Nuclear Information System (INIS)

    Gervash, A.; Giniyatulin, R.; Mazul, I.; Watson, R.

    1998-01-01

    One of the main requirements to use Be as a candidate for plasma facing component in ITER is providing a reliable joint between Be and Cu-alloy heat sink structure. In this work authors present the results of recent activity on this way. To create Be/CuCrZr joints the unique fast e-beam brazing technology was developed in Russia. The numbers of Be/CuCrZr mock-ups were manufactured in Efremov Institute by fast e-beam brazing using Cu-Sn-In-Ni brazing alloy. These mock-ups were tested by Sandia Laboratory at the EBTS electron beam facility. The goals of the tests were to define the allowable dimensions of the armour tiles for the heat loads of more than 10 MW/m 2 , to find the limit of bond strength for the Be/CuCrZr joint and response to heat loads and to estimate the life time of the brazed tiles by thermo-cyclic testing. The screening and thermal fatigue results are presented. With the aim to check the applicability of developed fast brazing process to DS-Cu alloy (GlidCop) the actively cooled mock-up produced by fast brazing Be onto GlidCop structure was manufactured and tested in Efremov Institute, main results are discussed. Summarizing testing data authors show that the CuCrZr or DS-Cu heat sink armored with beryllium can survive high heat fluxes (≥10 MW/m 2 ) during thousand heating/cooling cycles without serious damaging in the armour material and its joint. (author)

  6. Fracture toughness of irradiated beryllium

    International Nuclear Information System (INIS)

    Beeston, J.M.

    1978-01-01

    The fracture toughness of nuclear grade hot-pressed beryllium upon irradiation to fluences of 3.5 to 5.0 x 10 21 n/cm 2 , E greater than 1 MeV, was determined. Procedures and data relating to a round-robin test contributing to a standard ASTM method for unirradiated beryllium are discussed in connection with the testing of irradiated specimens. A porous grade of beryllium was also irradiated and tested, thereby enabling some discrimination between the models for describing the fracture toughness behavior of porous beryllium. The fracture toughness of unirradiated 2 percent BeO nuclear grade beryllium was 12.0 MPa m/sup 1 / 2 /, which was reduced 60 percent upon irradiation at 339 K and testing at 295 K. The fracture toughness of a porous grade of beryllium was 13.1 MPa m/sup 1 / 2 /, which was reduced 68 percent upon irradiation and testing at the same conditions. Reasons for the reduction in fracture toughness upon irradiation are discussed

  7. ITER...ation

    International Nuclear Information System (INIS)

    Troyon, F.

    1997-01-01

    Recurrent attacks against ITER, the new generation of tokamak are a mix of political and scientific arguments. This short article draws a historical review of the European fusion program. This program has allowed to build and manage several installations in the aim of getting experimental results necessary to lead the program forwards. ITER will bring together a fusion reactor core with technologies such as materials, superconductive coils, heating devices and instrumentation in order to validate and delimit the operating range. ITER will be a logical and decisive step towards the use of controlled fusion. (A.C.)

  8. Safe waste management practices in beryllium facilities

    International Nuclear Information System (INIS)

    Bhat, P.N.; Soundararajan, S.; Sharma, D.N.

    2012-01-01

    Beryllium, an element with the atomic symbol Be, atomic number 4, has very high stiffness to weight ratio and low density. It has good electrical conductive properties with low coefficient of thermal expansion. These properties make the metal beryllium very useful in varied technological endeavours, However, beryllium is recognised as one of the most toxic metals. Revelation of toxic effects of beryllium resulted in institution of stringent health and safety practices in beryllium handling facilities. The waste generated in such facilities may contain traces of beryllium. Any such waste should be treated as toxic waste and suitable safe waste management practices should be adopted. By instituting appropriate waste management practice and through a meticulously incorporated safety measures and continuous surveillance exercised in such facilities, total safety can be ensured. This paper broadly discusses health hazards posed by beryllium and safe methods of management of beryllium bearing wastes. (author)

  9. The nature of beryllium disease

    International Nuclear Information System (INIS)

    Williams, W.J.

    1977-01-01

    The increasing use of beryllium in modern industry poses a continuing health hazard with a real risk of producing incapacitating disease and even death. Beryllium and its salts are very toxic, even in small doses and may produce lesions in any organ. The majority of cases follow inhalation and may cause either acute or chronic lung disease. Acute pulmonary disease is a form of chemical pneumonitis while the chronic disease is characterised by the production of granulomas and fibrosis. The skin may be affected with the finding of dermatitis, acute or chronic ulceration. Other organs commonly involved include the liver and kidneys. The pathology of beryllium disease is not specific and diagnosis depends on satisfying the following criteria - history of exposure, consistent clinical, radiographic and pathological finding, presence of beryllium in tissue/fluid and evidence of hypersensitivity. Recent development of 'in vitro' tests of hypersensitivity may prove of value in both diagnosis and prevention of disease. Beryllium disease responds to steroid therapy but the only sure treatment is avoidance of exposure. (author)

  10. Experimental and thermodynamic assessment of beryllium-replacement materials for CANDU brazed joints

    Energy Technology Data Exchange (ETDEWEB)

    Potter, K.N.; Ferrier, G.A.; Corcoran, E.C., E-mail: Kieran.Potter@rmc.ca [Royal Military College of Canada, Kingston ON, (Canada); Dimayuga, F.C. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

    2015-07-01

    Currently, appendages are joined to CANDU fuel elements via a brazing process, with beryllium as the filler material. A potential reduction in the occupational limit on airborne beryllium particulates has motivated research into alternative brazing materials. To this end, the Canadian nuclear industry has funded an initiative to identify and evaluate the suitability of several candidate brazing materials. This work describes contributions toward the assessment of alternative brazing materials from the Royal Military College of Canada (RMCC). An impact testing method was developed to evaluate the mechanical strength of candidate braze joints.Thermodynamic modelling was performed to predict the aqueous behaviour of each candidate material in CANDU coolant conditions characteristic of reactor shutdown, and corrosion experiments are underway to support modelling predictions.The results of these activities will assist in selecting a suitable replacement material for beryllium. (author)

  11. High heat flux tests on beryllium and beryllium-copper joints

    International Nuclear Information System (INIS)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.

    1997-01-01

    A large test program has been set up to evaluate the performance of beryllium as a plasma facing material for the divertor in thermonuclear fusion devices. Simulation of steady state heat loads of 5 MWm -2 and above on actively cooled divertor modules, and off-normal plasma conditions with energy densities in the range 1-7 MJm -2 , have been investigated. Thermal shock tests were carried out with the ITER reference grade S65-C and several Russian grades of beryllium. At incident energies up to 7 MJm -2 the best erosion behaviour is observed for S65-C and for TGP-56. Steady state heating tests with actively cooled Be/Cu mock-ups were performed at incident powers of up to 5.8 MWm -2 . All samples investigated in these tests did not show any indications of failure. A Be/Cu mock-ups with Incusil braze was loaded in thermal fatigue up to 500 cycles at an incident power of 4.8 MWm -2 . Up to the end of the experiment no temperature increase of the surface and no indication of failure was observed. (orig.)

  12. High-strength beryllium block

    International Nuclear Information System (INIS)

    Pinto, N.P.; Keith, G.H.

    1977-01-01

    Beryllium billets hot isopressed using fine powder of high purity have exceptionally attractive properties; average tensile ultimate, 0.2% offset yield strength and elongation are 590 MPa, 430 MPa and 4.0% respectively. Properties are attributed to the fine grain size (about 4.0 μm average diameter) and the relatively low levels of BeO present as fine, well-dispersed particles. Dynamic properties, e.g., fracture toughness, are similar to those of standard grade, high-purity beryllium. The modulus of beryllium is retained to very high stress levels, and the microyield stress or precision elastic limit is higher than for other grades, including instrument grades. Limited data for billets made from normal-purity fine powders show similar room temperature properties. (author)

  13. ITER convertible blanket evaluation

    International Nuclear Information System (INIS)

    Wong, C.P.C.; Cheng, E.

    1995-01-01

    Proposed International Thermonuclear Experimental Reactor (ITER) convertible blankets were reviewed. Key design difficulties were identified. A new particle filter concept is introduced and key performance parameters estimated. Results show that this particle filter concept can satisfy all of the convertible blanket design requirements except the generic issue of Be blanket lifetime. If the convertible blanket is an acceptable approach for ITER operation, this particle filter option should be a strong candidate

  14. Cause of pitting in beryllium

    International Nuclear Information System (INIS)

    Kershaw, R.P.

    1982-01-01

    Light microscopy, bare-film radiography, secondary ion mass spectroscopy, electron microprobe and physical testing were used to examine beryllium specimens exhibiting a stratified, pitted, pattern after chemical milling. The objective was to find the cause of this pattern. Specimens were found to have voids in excess of density specification allowances. These voids are attributed, at least in part, to the sublimation of beryllium fluoride during the vacuum hot pressing operation. The origin of the pattern is attributed to these voids and etching out of fines and associated impurities. Hot isostatic pressing with a subsequent heat treatment close residual porosity and dispersed impurities enough to correct the problem

  15. Bioenvironmental Engineering Guide to Beryllium

    Science.gov (United States)

    2017-07-26

    Dermal contact with beryllium can result in dermatitis resembling first- or second-degree burns and skin granulomas [7]. Beryllium dust, fume...minute short-term exposure limit (STEL) of 2.0 µg/m3 [§1910.1024(c)(2) & §1926.1124(c)(2)], and added provisions to prevent skin contact [§1910.1024(b...document you want more information, contact the Environmental, Safety, and Occupational Health (ESOH) Service Center at DSN 798-3764, 1-888-232-ESOH (3764

  16. In-pile thermocycling testing and post-test analysis of beryllium divertor mockups

    Energy Technology Data Exchange (ETDEWEB)

    Giniatulin, R.; Mazul, I. [Efremov Inst., St. Petersburg (Russian Federation); Melder, R.; Pokrovsky, A.; Sandakov, V.; Shiuchkin, A.

    1998-01-01

    The main damaging factors which impact the ITER divertor components are neutron irradiation, cyclic surface heat loads and hydrogen environment. One of the important questions in divertor mockups development is the reliability of beryllium/copper joints and the beryllium resistance under neutron irradiation and thermal cycling. This work presents the experiment, where neutron irradiation and thermocyclic heat loads were applied simultaneously for two beryllium/copper divertor mockups in a nuclear reactor channel to simulate divertor operational conditions. Two mockups with different beryllium grades were mounted facing each other with the tantalum heater placed between them. This device was installed in the active zone of the nuclear reactor SM-2 (Dimitrovgrad, Russia) and the tantalum block was heated by neutron irradiation up to a high temperature. The main part of the heat flux from the tantalum surface was transported to the beryllium surface through hydrogen, as a result the heat flux loaded two mockups simultaneously. The mockups were cooled by reactor water. The device was lowered to the active zone so as to obtain the heating regime and to provide cooling lifted. This experiment was performed under the following conditions: tantalum heater temperature - 1950degC; hydrogen environment -1000 Pa; surface heat flux density -3.2 MW/m{sup 2}; number of thermal cycles (lowering and lifting) -101; load time in each cycle - 200-5000 s; dwell time (no heat flux, no neutrons) - 300-2000 s; cooling water parameters: v - 1 m/s, Tin - 50degC, Pin - 5 MPa; neutron fluence -2.5 x 10{sup 20} cm{sup -2} ({approx}8 years of ITER divertor operation from the start up). The metallographic analysis was performed after experiment to investigate the beryllium and beryllium/copper joint structures, the results are presented in the paper. (author)

  17. Defense programs beryllium good practice guide

    International Nuclear Information System (INIS)

    Herr, M.

    1997-07-01

    Within the DOE, it has recently become apparent that some contractor employees who have worked (or are currently working) with and around beryllium have developed chronic beryllium disease (CBD), an occupational granulomatous lung disorder. Respiratory exposure to aerosolized beryllium, in susceptible individuals, causes an immunological reaction that can result in granulomatous scarring of the lung parenchyma, shortness of breath, cough, fatigue, weight loss, and, ultimately, respiratory failure. Beryllium disease was originally identified in the 1940s, largely in the fluorescent light industry. In 1950, the Atomic Energy Commission (AEC) introduced strict exposure standards that generally curtailed both the acute and chronic forms of the disease. Beginning in 1984, with the identification of a CBD case in a DOE contractor worker, there was increased scrutiny of both industrial hygiene practices and individuals in this workforce. To date, over 100 additional cases of beryllium-specific sensitization and/or CBD have been identified. Thus, a disease previously thought to be largely eliminated by the adoption of permissible exposure standards 45 years ago is still a health risk in certain workforces. This good practice guide forms the basis of an acceptable program for controlling workplace exposure to beryllium. It provides (1) Guidance for minimizing worker exposure to beryllium in Defense Programs facilities during all phases of beryllium-related work, including the decontamination and decommissioning (D ampersand D) of facilities. (2) Recommended controls to be applied to the handling of metallic beryllium and beryllium alloys, beryllium oxide, and other beryllium compounds. (3) Recommendations for medical monitoring and surveillance of workers exposed (or potentially exposed) to beryllium, based on the best current understanding of beryllium disease and medical diagnostic tests available. (4) Site-specific safety procedures for all processes of beryllium that is

  18. Inhalation hazards from machining beryllium metal

    International Nuclear Information System (INIS)

    Hoover, M.D.; Finch, G.L.; Mewhinney, J.A.; Eidson, A.F.

    1987-01-01

    Beryllium metal has special nuclear and structural properties that make it useful for applications in fission and fusion reactor designs. Unfortunately, concerns for its toxicity have made designers wary of using beryllium metal. The work being reported here was undertaken to characterize the aerosols produced by two very common operations performed during preparation or modification of components for use in reactor systems: sawing and milling of beryllium metal. The study also covered beryllium metal alloys to allow comparison. Information from this study is to enable better assessments of the risk of using beryllium metal in reactor designs

  19. Optimization of beryllium for fusion blanket applications

    International Nuclear Information System (INIS)

    Billone, M.C.

    1993-01-01

    The primary function of beryllium in a fusion reactor blanket is neutron multiplication to enhance tritium breeding. However, because heat, tritium and helium will be generated in and/or transported through beryllium and because the beryllium is in contact with other blanket materials, the thermal, mechanical, tritium/helium and compatibility properties of beryllium are important in blanket design. In particular, tritium retention during normal operation and release during overheating events are safety concerns. Accommodating beryllium thermal expansion and helium-induced swelling are important issues in ensuring adequate lifetime of the structural components adjacent to the beryllium. Likewise, chemical/metallurgical interactions between beryllium and structural components need to be considered in lifetime analysis. Under accident conditions the chemical interaction between beryllium and coolant and breeding materials may also become important. The performance of beryllium in fusion blanket applications depends on fabrication variables and operational parameters. First the properties database is reviewed to determine the state of knowledge of beryllium performance as a function of these variables. Several design calculations are then performed to indicate ranges of fabrication and operation variables that lead to optimum beryllium performance. Finally, areas for database expansion and improvement are highlighted based on the properties survey and the design sensitivity studies

  20. Galvanic corrosion of beryllium welds

    International Nuclear Information System (INIS)

    Hill, M.A.; Butt, D.P.; Lillard, R.S.

    1997-01-01

    Beryllium is difficult to weld because it is highly susceptible to cracking. The most commonly used filler metal in beryllium welds is Al-12 wt.% Si. Beryllium has been successfully welded using Al-Si filler metal with more than 30 wt.% Al. This filler creates an aluminum-rich fusion zone with a low melting point that tends to backfill cracks. Drawbacks to adding a filler metal include a reduction in service temperature, a lowering of the tensile strength of the weld, and the possibility for galvanic corrosion to occur at the weld. To evaluate the degree of interaction between Be and Al-Si in an actual weld, sections from a mock beryllium weldment were exposed to 0.1 M Cl - solution. Results indicate that the galvanic couple between Be and the Al-Si weld material results in the cathodic protection of the weld and of the anodic dissolution of the bulk Be material. While the cathodic protection of Al is generally inefficient, the high anodic dissolution rate of the bulk Be during pitting corrosion combined with the insulating properties of the Be oxide afford some protection of the Al-Si weld material. Although dissolution of the Be precipitate in the weld material does occur, no corrosion of the Al-Si matrix was observed

  1. Worker Environment Beryllium Characterization Study

    International Nuclear Information System (INIS)

    2009-01-01

    This report summarizes the conclusion of regular monitoring of occupied buildings at the Nevada Test Site and North Las Vegas facility to determine the extent of beryllium (Be) contamination in accordance with Judgment of Needs 6 of the August 14, 2003, 'Minnema Report.'

  2. Worker Environment Beryllium Characterization Study

    Energy Technology Data Exchange (ETDEWEB)

    NSTec Environment, Safety, Health & Quality

    2009-12-28

    This report summarizes the conclusion of regular monitoring of occupied buildings at the Nevada Test Site and North Las Vegas facility to determine the extent of beryllium (Be) contamination in accordance with Judgment of Needs 6 of the August 14, 2003, “Minnema Report.”

  3. Post irradiation characterization of beryllium and beryllides after high temperature irradiation up to 3000 appm helium production in HIDOBE-01

    Energy Technology Data Exchange (ETDEWEB)

    Fedorov, A.V., E-mail: fedorov@nrg.eu [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Til, S. van; Stijkel, M.P. [Nuclear Research and Consultancy Group, Westerduinweg 3, Postbus 25, Petten, 1755 ZG (Netherlands); Nakamichi, M. [Japan Atomic Energy Agency, Rokkasho (Japan); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/ Josep Pla, n° 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

    2016-01-15

    Titanium beryllides are considered as advanced candidate material for neutron multiplier for the helium cooled pebble bed (HCPB) and/or the water cooled ceramic breeder (WCCB) breeder blankets. In the HIDOBE-01 (HIgh DOse irradiation of BEryllium) experiment, beryllium and beryllide pellets with 5 at% and 7 at% Ti are irradiated at four different target temperatures (T{sub irr}): 425 °C, 525 °C, 650 °C and 750 °C up to the dose corresponding to 3000 appm He production in beryllium. The pellets were supplied by JAEA. During post irradiation examinations the critical properties of volumetric swelling and tritium retention were studied. Both titanium beryllide grades show significantly less swelling than the beryllium grade, with the difference increasing with the irradiation temperature. The irradiation induced swelling was studied by using direct dimensions. Both beryllide grades showed much less swelling as compare to the reference beryllium grade. Densities of the grades were studied by Archimedean immersion and by He-pycnometry, giving indications of porosity formation. While both beryllide grades show no significant reduction in density at all irradiation temperatures, the beryllium density falls steeply at higher T{sub irr}. Finally, the tritium release and retention were studied by temperature programmed desorption (TPD). Beryllium shows the same strong tritium retention as earlier observed in studies on beryllium pebbles, while the tritium inventory of the beryllides is significantly less, already at the lowest T{sub irr} of 425 °C.

  4. Prediction for disruption erosion of ITER plasma facing components; a comparison of experimental and numerical results

    International Nuclear Information System (INIS)

    Laan, J.G. van der; Akiba, M.; Seki, M.; Hassanein, A.; Tanchuk, V.

    1991-01-01

    An evaluation is given for the prediction for disruption erosion in the International Thermonuclear Engineering Reactor (ITER). At first, a description is given of the relation between plasma operating paramters and system dimensions to the predictions of loading parameters of Plasma Facing Components (PFC) in off-normal events. Numerical results from ITER parties on the prediction of disruption erosion are compared for a few typical cases and discussed. Apart from some differences in the codes, the observed discrepancies can be ascribed to different input data of material properties and boundary conditions. Some physical models for vapour shielding and their effects on numerical results are mentioned. Experimental results from ITER parties, obtained with electron and laser beams, are also compared. Erosion rates for the candidate ITER PFC materials are shown to depend very strongly on the energy deposition parameters, which are based on plasma physics considerations, and on the assumed material loss mechanisms. Lifetimes estimates for divertor plate and first wall armour are given for carbon, tungsten and beryllium, based on the erosion in the thermal quench phase. (orig.)

  5. Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

    International Nuclear Information System (INIS)

    Hirose, T.; Ogiwara, H.; Enoeda, M.; Akiba, M.

    2007-01-01

    Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 μm and 10 μm were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

  6. Interfacial properties of HIP joint between beryllium and reduced activation ferritic/martensitic steel

    Energy Technology Data Exchange (ETDEWEB)

    Hirose, T. [Blanket Engineering Group, Japan Atomic Energy Agency, Naka, Ibaraki (Japan); Ogiwara, H. [Japan Atomic Energy Agency, Tokai-mura, Naga-gun, Ibaraki-ken (Japan); Enoeda, M. [Naka Fusion Research Establishment, J.A.E.R.I., Japan Atomic Energy Research Institute, Naka-gun, Ibaraki-ken (Japan); Akiba, M. [Naka Fusion Institute, Japan Atomic Energy Agency, Naka, Ibaraki (Japan)

    2007-07-01

    Full text of publication follows: ITER test blanket module is the most important components to validate energy production and fuel breeding process for future demonstration reactor. Reduced activation ferritic / martensitic steel is recognized as a promising structural material for breeding blanket systems. And Beryllium must be used as plasma facing materials for ITER in vessel components. In this work, interfacial properties of beryllium/reduced activation ferritic/martensitic steel (RAF/Ms) joint were investigated for a first wall of ITER test blanket module (TBM). The starting materials were ITER grade Beryllium, S65C and a Japanese RAF/M, F82H. The joint was produced by solid state hot isostatic pressing (HIP) method. Chromium layer with the thickness of 1 {mu}m and 10 {mu}m were formed by plasma vapor deposition on the beryllium surface as a diffusion barrier. The HIP was carried out at 1023 K and 1233 K which are determined by standard normalizing and tempering temperature of F82H. The joint made at 1233 K was followed by tempering at 1033 K. The bonding interface was characterized by electron probe microanalysis (EPMA). The bonding strength was also investigated by isometric four point bending tests at ambient temperature. EPMA showed chromium layer effectively worked as a diffusion barrier at 1023 K. However, the beryllium rich layer was formed in F82H after HIP at 1233 K followed by tempering. Bending tests revealed that thin chromium layer and low temperature HIP is preferable. The high temperature HIP introduce brittle BeFe inter metallic compounds along bonding interface. On the other hand, joint with thick chromium layer suffer from brittleness of chromium itself. (authors)

  7. OVERVIEW OF BERYLLIUM SAMPLING AND ANALYSIS

    Energy Technology Data Exchange (ETDEWEB)

    Brisson, M

    2009-04-01

    Because of its unique properties as a lightweight metal with high tensile strength, beryllium is widely used in applications including cell phones, golf clubs, aerospace, and nuclear weapons. Beryllium is also encountered in industries such as aluminium manufacturing, and in environmental remediation projects. Workplace exposure to beryllium particulates is a growing concern, as exposure to minute quantities of anthropogenic forms of beryllium may lead to sensitization and to chronic beryllium disease, which can be fatal and for which no cure is currently known. Furthermore, there is no known exposure-response relationship with which to establish a 'safe' maximum level of beryllium exposure. As a result, the current trend is toward ever lower occupational exposure limits, which in turn make exposure assessment, both in terms of sampling and analysis, more challenging. The problems are exacerbated by difficulties in sample preparation for refractory forms of beryllium, such as beryllium oxide, and by indications that some beryllium forms may be more toxic than others. This chapter provides an overview of sources and uses of beryllium, health risks, and occupational exposure limits. It also provides a general overview of sampling, analysis, and data evaluation issues that will be explored in greater depth in the remaining chapters. The goal of this book is to provide a comprehensive resource to aid personnel in a wide variety of disciplines in selecting sampling and analysis methods that will facilitate informed decision-making in workplace and environmental settings.

  8. Experimental and thermodynamic studies of beryllium replacement materials for CANDU brazed joints

    Energy Technology Data Exchange (ETDEWEB)

    Potter, K.N.; Ferrier, G.A.; Corcoran, E.C., E-mail: Kieran.Potter@rmc.ca [Royal Military College of Canada, Kingston, ON (Canada)

    2015-07-01

    Currently, appendages are joined to CANDU fuel elements via a brazing process, which uses beryllium as the filler material. A potential reduction in the occupational limit on airborne beryllium particulates has motivated research into alternative brazing materials. To this end, the Canadian nuclear industry has funded an initiative to identify and evaluate the suitability of several candidate materials. This work describes contributions toward the assessment of alternative brazing materials from the Royal Military College of Canada. Thermodynamic modelling was performed to predict the aqueous behaviour of each candidate material in CANDU coolant conditions characteristic of reactor shutdown, and experiments are underway to support modelling predictions. These results will assist in selecting a suitable replacement material for beryllium. (author)

  9. Mechanisms of hydrogen retention in metallic beryllium and beryllium oxide and properties of ion-induced beryllium nitride

    International Nuclear Information System (INIS)

    Oberkofler, Martin

    2011-01-01

    In the framework of this thesis laboratory experiments on atomically clean beryllium surfaces were performed. They aim at a basic understanding of the mechanisms occurring upon interaction of a fusion plasma with a beryllium first wall. The retention and the temperature dependent release of implanted deuterium ions are investigated. An atomistic description is developed through simulations and through the comparison with calculations based on density functional theory. The results of these investigations are compared to the behaviour of hydrogen upon implantation into thermally grown beryllium oxide layers. Furthermore, beryllium nitride is produced by implantation of nitrogen into metallic beryllium and its properties are investigated. The results are interpreted with regard to the use of beryllium in a fusion reactor. (orig.)

  10. Investigations of the ternary system beryllium-carbon-tungsten and analyses of beryllium on carbon surfaces; Untersuchung des ternaeren Systems Beryllium-Kohlenstoff-Wolfram und Betrachtungen von Beryllium auf Kohlenstoffoberflaechen

    Energy Technology Data Exchange (ETDEWEB)

    Kost, Florian

    2009-05-25

    Beryllium, carbon and tungsten are planned to be used as first wall materials in the future fusion reactor ITER. The aim of this work is a characterization of mixed material formation induced by thermal load. To this end, model systems (layers) were prepared and investigated, which give insight into the basic physical and chemical concepts. Before investigating ternary systems, the first step was to analyze the binary systems Be/C and Be/W (bottom-up approach), where the differences between the substrates PG (pyrolytic graphite) and HOPG (highly oriented pyrolytic graphite) were of special interest. Particularly X-ray photoelectron spectroscopy (XPS), low energy ion scattering (ISS) and Rutherford backscattering spectroscopy (RBS) were used as analysis methods. Beryllium evaporated on carbon shows an island growth mode, whereas a closed layer can be assumed for layer thicknesses above 0.7 nm. Annealing of the Be/C system induces Be{sub 2}C island formation for T{>=}770 K. At high temperatures (T{>=}1170 K), beryllium carbide dissociates, resulting in (metallic) beryllium desorption. For HOPG, carbide formation starts at higher temperatures compared to PG. Activation energies for the diffusion processes were determined by analyzing the decreasing beryllium amount versus annealing time. Surface morphologies were characterized using angle-resolved XPS (ARXPS) and atomic force microscopy (AFM). Experiments were performed to study processes in the Be/W system in the temperature range from 570 to 1270 K. Be{sub 2}W formation starts at 670 K, a complete loss of Be{sub 2}W is observed at 1170 K due to dissociation (and subsequent beryllium desorption). Regarding ternary systems, particularly Be/C/W and C/Be/W were investigated, attaching importance to layer thickness (reservoir) variations. At room temperature, Be{sub 2}C, W{sub 2}C, WC and Be{sub 2}W formation at the respective interfaces was observed. Further Be{sub 2}C is forming with increasing annealing temperatures

  11. Technical aspects of the joint JET-ISX-B beryllium limiter experiment

    International Nuclear Information System (INIS)

    Edmonds, P.H.; Dietz, K.J.; Mioduszewski, P.K.; Watson, R.D.; Emerson, L.C.; Gabbard, W.A.; Goodall, D.; Simpkins, J.E.; Yarber, J.L.

    1985-01-01

    An experiment has been performed on the Impurity Study Experiment (ISX-B) tokamak to test beryllium as a limiter material. Beryllium is an attractive candidate for a limiter and has been proposed for use in the Joint European Torus (JET) experiment. A temperature-controlled, segmented, beryllium top-rail limiter was located inside the plasma radius described by the existing titanium limiters. An extended set of diagnostics was added for measurement of scrapeoff and limiter parameters. These included visible and infrared monitoring systems, probes, and surface analysis experiments. Tokamak experiments included parameter surveys of both Ohmically heated and neutral-beam-heated plasmas and an extended fluence test of the limiter. The most significant effect of operation with a beryllium limiter was the reduction in low-Z impurities caused by gettering action of beryllium deposited on the liner walls. The experiment required the design and implementation of contamination control apparatus and work procedures to prevent the accidental dispersion of beryllium dust

  12. Defense programs beryllium good practice guide

    Energy Technology Data Exchange (ETDEWEB)

    Herr, M.

    1997-07-01

    Within the DOE, it has recently become apparent that some contractor employees who have worked (or are currently working) with and around beryllium have developed chronic beryllium disease (CBD), an occupational granulomatous lung disorder. Respiratory exposure to aerosolized beryllium, in susceptible individuals, causes an immunological reaction that can result in granulomatous scarring of the lung parenchyma, shortness of breath, cough, fatigue, weight loss, and, ultimately, respiratory failure. Beryllium disease was originally identified in the 1940s, largely in the fluorescent light industry. In 1950, the Atomic Energy Commission (AEC) introduced strict exposure standards that generally curtailed both the acute and chronic forms of the disease. Beginning in 1984, with the identification of a CBD case in a DOE contractor worker, there was increased scrutiny of both industrial hygiene practices and individuals in this workforce. To date, over 100 additional cases of beryllium-specific sensitization and/or CBD have been identified. Thus, a disease previously thought to be largely eliminated by the adoption of permissible exposure standards 45 years ago is still a health risk in certain workforces. This good practice guide forms the basis of an acceptable program for controlling workplace exposure to beryllium. It provides (1) Guidance for minimizing worker exposure to beryllium in Defense Programs facilities during all phases of beryllium-related work, including the decontamination and decommissioning (D&D) of facilities. (2) Recommended controls to be applied to the handling of metallic beryllium and beryllium alloys, beryllium oxide, and other beryllium compounds. (3) Recommendations for medical monitoring and surveillance of workers exposed (or potentially exposed) to beryllium, based on the best current understanding of beryllium disease and medical diagnostic tests available. (4) Site-specific safety procedures for all processes of beryllium that is likely to

  13. Belgian research on fusion beryllium waste

    International Nuclear Information System (INIS)

    Druyts, F.; Mallants, D.; Sillen, X.; Iseghem, P. Van

    2004-01-01

    Future fusion power plants will generate important quantities of neutron irradiated beryllium. Although recycling is the preferred management option for this waste, this may not be technically feasible for all of the beryllium, because of its radiological characteristics. Therefore, at SCK·CEN, we initiated a research programme aimed at studying aspects of the disposal of fusion beryllium, including waste characterisation, waste acceptance criteria, conditioning methods, and performance assessment. One of the main issues to be resolved is the development of fusion-specific waste acceptance criteria for surface or deep geological disposal, in particular with regard to the tritium content. In case disposal is the only solution, critical nuclides can be immobilised by conditioning the waste. As a first approach to immobilising beryllium waste, we investigated the vitrification of beryllium. Corrosion tests were performed on both metallic and vitrified beryllium to provide source data for performance assessment. Finally, a first step in performance assessment was undertaken. (author)

  14. Beryllium minerals - demand strong for miniaturisation

    International Nuclear Information System (INIS)

    Griffiths, J.

    1985-01-01

    Beryllium is an essential constituent of over 40 minerals of which two are exploited commercially. Beryl is largely produced in the USSR and China and bertrandite in the U.S.A. Phenacite, from Canada, is also under investigation. The largest extraction plant for the recovery of beryllium in the western world is in Utah, U.S.A. and the company also produces beryllium oxide used in the manufacture of ceramics widely used in the electronics industry and for refractory articles. Beryllium-copper alloys in strip, rod and tube form are produced in the U.S.A., Germany and the U.K. Beryllium ceramics are important because of their high thermal conductivity, electrical insulation, strength and rigidity. The alloys, used as electric connectors, microswitch contacts are important for their high suitability for miniaturisation. The future growth potential for the beryllium industry is in the automotive industries in Europe and Japan. (U.K.)

  15. Method for hot pressing beryllium oxide articles

    Science.gov (United States)

    Ballard, Ambrose H.; Godfrey, Jr., Thomas G.; Mowery, Erb H.

    1988-01-01

    The hot pressing of beryllium oxide powder into high density compacts with little or no density gradients is achieved by employing a homogeneous blend of beryllium oxide powder with a lithium oxide sintering agent. The lithium oxide sintering agent is uniformly dispersed throughout the beryllium oxide powder by mixing lithium hydroxide in an aqueous solution with beryllium oxide powder. The lithium hydroxide is converted in situ to lithium carbonate by contacting or flooding the beryllium oxide-lithium hydroxide blend with a stream of carbon dioxide. The lithium carbonate is converted to lithium oxide while remaining fixed to the beryllium oxide particles during the hot pressing step to assure uniform density throughout the compact.

  16. Analysis of a beryllium-copper diffusion joint after HHF test

    International Nuclear Information System (INIS)

    Guiniatouline, R.N.; Mazul, I.V.; Gorodetsky, A.E.; Zalavutdinov, R.Kh.; Rybakov, S.Yu.; Savenko, V.I.

    1996-01-01

    The development of beryllium-copper joints which can withstand relevant ITER divertor conditions is one of the important tasks at the present time. One of the main problems associated with these joints is the intermetallic layers. The strength and life of these joints significantly depend on the width and contents of the intermetallic layers. The objective of this work is to study the diffusion joint of TGP-56 beryllium to OFHC copper after thermal response and thermocyclic tests with a beryllium-copper mock-up. The HHF test was performed on the e-beam facility (EBTS, SNLA). The following methods were used for analysis: roentgenographic analysis, X-ray spectrum analysis and fracture analysis. During the investigation the following studies were undertaken: the analysis of the diffusion boundary layer, which was obtained at the cross-section of one of the tiles, the analysis of the debonded surfaces of several beryllium tiles and corresponding copper parts and the analysis of the upper surface of one of the tiles after HHF tests. The joint roentgenographic and element analyses revealed the following phases in the diffusion zone: Cu 2 Be (∝170 μm), CuBe (∝30 μm), CuBe 2 (∝1 μm) and a solid solution of copper in beryllium. The phases Cu 2 Be, CuBe and the solid solution of copper in beryllium were detected by the quantitative microanalysis and the phases CuBe, CuBe 2 and CuBe, by the roentgenographic analysis. The fracture (origin) is located in the central part of the tiles. This crack was caused by residual stresses and thermal fatigue testing. This analysis provides important data on the joint quality and may be used for all types of joints used for ITER applications. (orig.)

  17. MEASUREMENTS OF THE PROPERTIES OF BERYLLIUM FOIL

    International Nuclear Information System (INIS)

    ZHAO, Y.; WANG, H.

    2000-01-01

    The electrical conductivity of beryllium at radio frequency (800 MHz) and liquid nitrogen temperature were investigated and measured. This summary addresses a collection of beryllium properties in the literature, an analysis of the anomalous skin effect, the test model, the experimental setup and improvements, MAFIA simulations, the measurement results and data analyses. The final results show that the conductivity of beryllium is not as good as indicated by the handbook, yet very close to copper at liquid nitrogen temperature

  18. New audio applications of beryllium metal

    International Nuclear Information System (INIS)

    Sato, M.

    1977-01-01

    The major applications of beryllium metal in the field of audio appliances are for the vibrating cones for the two types of speakers 'TWITTER' for high range sound and 'SQUAWKER' for mid range sound, and also for beryllium cantilever tube assembled in stereo cartridge. These new applications are based on the characteristic property of beryllium having high ratio of modulus of elasticity to specific gravity. The production of these audio parts is described, and the audio response is shown. (author)

  19. Thermal expansion of beryllium oxide

    International Nuclear Information System (INIS)

    Solodukhin, A.V.; Kruzhalov, A.V.; Mazurenko, V.G.; Maslov, V.A.; Medvedev, V.A.; Polupanova, T.I.

    1987-01-01

    Precise measurements of temperature dependence of the coefficient of linear expansion in the 22-320 K temperature range on beryllium oxide monocrystals are conducted. A model of thermal expansion is suggested; the range of temperature dependence minimum of the coefficient of thermal expansion is well described within the frames of this model. The results of the experiment may be used for investigation of thermal stresses in crystals

  20. Intoxication experiments with beryllium sulphate

    International Nuclear Information System (INIS)

    Bucurescu, I.; Stan, T.

    1990-01-01

    The changes in the particular number of animals in two groups of 40 rats each subjected to intoxication experiments with beryllium sulphate was investigated. The two investigations had very different characteristics. In the case of chronic intoxication there was a marked lethality over given time intervals. In the case of subacute intoxication the number of animals decreased with time. It was found empirically that this change can be described by an exponential relationship which lends itself to statistical interpretation. (author)

  1. Thermodynamic properties of beryllium hydroxide

    International Nuclear Information System (INIS)

    Baur, A.; Lecocq, A.

    1964-01-01

    The study of the hydro-thermal decomposition of beryllium hydroxide has made it possible to determine the free energy of formation and the entropy. The results obtained are in good agreement with the theoretical values calculated from the solubility product of this substance. They give furthermore the possibility of acquiring a better understanding of the BeO-H 2 O-Be (OH) 2 system between 20 and 1500 C. (authors) [fr

  2. Thermal Properties of Beryllium Metal

    International Nuclear Information System (INIS)

    Cho, Tae Won; Baek, Je Kyun; Jeong, Gwan Yoon; Kim, Ji Hyeon; Sohn, Dong Seong

    2013-01-01

    It is known that the presence of as-fabricated porosity largely affect thermal conductivity of beryllium. Therefore, in this paper we will suggest a new thermal conductivity equation which consider volume fraction and discuss how this can be applied to irradiation induced degradation of thermal conductivity later. This study was performed to develop a new correlation of thermal conductivity of Beryllium. Although there are many factors like BeO contents, impurity level, grain size, and porosity, we assumed porosity will be the dominant factor for thermal conductivity. Therefore, a new correlation which consider volume fraction by applying Maxwell-Eucken equation is developed and this is consistent to some degrees. However, increasing impurity level and decreasing grain size will decrease thermal conductivity. Therefore, we need to consider their effects although we assume BeO contents, impurity, and grain size do not make noticeable effects in the future. Furthermore, thermal conductivity degradation by neutron irradiation should be considered afterward. There are two main factors for the thermal conductivity degradation: the one is defects formed by neutron collisions and the other is helium generated by transmutation of Be. It is known that they make a considerable degradation of conductivity. Beryllium is known there are considerable volume increases by helium accumulation. Therefore, we anticipate our suggested model can be applicable if it has been developed furthermore considering irradiation induced swelling

  3. (Beryllium). Internal Report No. 137, Jan. 15, 1958

    International Nuclear Information System (INIS)

    Mouret, P.; Rigaud, A.

    1959-01-01

    After a brief summary of the physical and chemical properties of beryllium, the various chemical treatments which can be applied to beryllium minerals either directly or after a physical enrichment are discussed. These various treatments give either the hydroxide or beryllium salts, from which either beryllium oxide or metallic beryllium can easily be obtained. The purification, analysis and uses of beryllium are also briefly discussed. (author)

  4. Status of the European R and D on beryllium as multiplier material for breeder blankets

    International Nuclear Information System (INIS)

    Moeslang, A.; Boccaccini, L.V.; Rabaglino, E.; Piazza, G.; Cardella, A.; Sannen, L.; Scibetta, M.; Laan, J. van der; Hegeman, J.B.J.W.

    2004-01-01

    Within the international fusion community a variety of breeding blanket concepts are being considered, ranging from more conservative concepts to higher-risk concepts for fusion power reactors. In Europe, the Helium Cooled Pebble Bed (HCPB) blanket is one of the two reference concepts which will also be tested as Test Blanket Module (TBM) in ITER. In addition to the R and D for structural parts of the HCPB blanket, a considerable effort is devoted to the production and qualification of ceramic breeder and neutron multiplier (beryllium or beryllide) pebble beds. Since in the HCPB blanket pebbles made of lithium ceramics are foreseen, a high volume fraction of beryllium as a neutron multiplier to Li-based ceramic of about 4: l is needed. The typical loading conditions for beryllium are, with a neutron wall load of ∼12.5 MWa/m 2 and in ∼5 years lifetime: T min ∼300degC, T max ∼600-900degC, displacement damage ∼80 dpa, peak 4 He production ∼26000 appm and peak 3 H production ∼700 appm at the End-Of-Life. The behaviour of beryllium under irradiation is considered to be a key issue of the HCPB blanket, because of swelling due to helium bubbles and tritium retention. A large R and D programme on beryllium has been implemented in Europe, aimed at characterising and predicting the material behaviour before and under irradiation. An overview on experimental and modelling activities performed during the past 2 years is given with typical results on non-irradiated and irradiated Beryllium materials and pebble beds and the relevance of major results on future beryllium R and D is addressed. (author)

  5. Beryllium dust generation resulting from plasma bombardment

    International Nuclear Information System (INIS)

    Doerner, R.; Mays, C.

    1997-01-01

    The beryllium dust resulting from erosion of beryllium samples subjected to plasma bombardment has been measured in PISCES-B. Loose surface dust was found to be uniformly distributed throughout the device and accounts for 3% of the eroded material. A size distribution measurement of the loose surface dust shows an increasing number of particles with decreasing diameter. Beryllium coatings on surfaces with a line of sight view of the target interaction region account for an additional 33% of the eroded beryllium material. Flaking of these surface layers is observed and is thought to play a significant role in dust generation inside the vacuum vessel. (orig.)

  6. Microplasticity in hot-pressed beryllium

    International Nuclear Information System (INIS)

    Plane, D.C.; Bonfield, W.

    1977-01-01

    Closed hysteresis loops measured in the microstrain region of hot pressed, commercially pure, polycrystalline beryllium are correlated with a dislocation - impurity atom, energy dissipating mechanism. (author)

  7. Beryllium thin films for resistor applications

    Science.gov (United States)

    Fiet, O.

    1972-01-01

    Beryllium thin films have a protective oxidation resistant property at high temperature and high recrystallization temperature. However, the experimental film has very low temperature coefficient of resistance.

  8. Final Report on ITER Task Agreement 81-08

    Energy Technology Data Exchange (ETDEWEB)

    Richard L. Moore

    2008-03-01

    As part of an ITER Implementing Task Agreement (ITA) between the ITER US Participant Team (PT) and the ITER International Team (IT), the INL Fusion Safety Program was tasked to provide the ITER IT with upgrades to the fusion version of the MELCOR 1.8.5 code including a beryllium dust oxidation model. The purpose of this model is to allow the ITER IT to investigate hydrogen production from beryllium dust layers on hot surfaces inside the ITER vacuum vessel (VV) during in-vessel loss-of-cooling accidents (LOCAs). Also included in the ITER ITA was a task to construct a RELAP5/ATHENA model of the ITER divertor cooling loop to model the draining of the loop during a large ex-vessel pipe break followed by an in-vessel divertor break and compare the results to a simular MELCOR model developed by the ITER IT. This report, which is the final report for this agreement, documents the completion of the work scope under this ITER TA, designated as TA 81-08.

  9. Qualification Test for Korean Mockups of ITER Blanket First Wall

    International Nuclear Information System (INIS)

    Kim, S. K.; Lee, D. W.; Bae, Y. D.; Hong, B. G.; Jung, H. K.; Jung, Y. I.; Park, J. Y.; Jeong, Y. H.; Choi, B. K.; Kim, B. Y.

    2009-01-01

    ITER First Wall (FW) includes the beryllium armor tiles joined to CuCrZr heat sink with stainless steel cooling tubes. This first wall panels are one of the critical components in the ITER machine with the surface heat flux of 0.5 MW/m 2 or above. So qualification program needs to be performed with the goal to qualify the joining technologies required for the ITER First Wall. Based on the results of tests, the acceptance of the developed joining technologies will be established. The results of this qualification test will affect the final selection of the manufacturers for the ITER First Wall

  10. Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; Longhurst, G.R. [Lockheed Martin Idaho Technol. Co., Idaho Falls, ID (United States). Idaho Nat. Eng. and Environ. Lab.; Causey, R.A.; Wampler, W.R.; Wilson, K.L. [Sandia National Laboratories, Livermore, CA (United States)]|[Sandia National Labs., Albuquerque, NM (United States); Davis, J.W.; Haasz, A.A. [Institute for Aerospace Studies, University of Toronto, Toronto (Canada); Doerner, R.P. [California Univ., San Diego, La Jolla, CA (United States). Center for Magnetic Recording Research; Federici, G. [ITER JWS Garching Co-center, Garching (Germany)

    1999-06-01

    Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions, show a difference of several orders of magnitude in both inventory and permeation rate to the coolant. (orig.) 78 refs.

  11. Joining technologies for the plasma facing components of ITER

    International Nuclear Information System (INIS)

    Barabash, V.; Kalinin, G.; Matera, R.

    1998-01-01

    An extensive R and D program on the development of the joining technologies between armour (beryllium, tungsten and carbon fibre composites)/copper alloys heat sink and copper alloys/ stainless steel has been carried out by ITER Home Teams. A brief review of this R and D program is presented in this paper. Based on the results, reference technologies for use in ITER have been selected and recommended for further development. (author)

  12. Lifetime analysis of the ITER first wall under steady-state and off-normal loads

    International Nuclear Information System (INIS)

    Mitteau, R; Sugihara, M; Raffray, R; Carpentier-Chouchana, S; Merola, M; Pitts, R A; Labidi, H; Stangeby, P

    2011-01-01

    The lifetime of the beryllium armor of the ITER first wall is evaluated for normal and off-normal operation. For the individual events considered, the lifetime spans between 930 and 35×10 6 discharges. The discrepancy between low and high estimates is caused by uncertainties about the behavior of the melt layer during off-normal events, variable plasma operation parameters and variability of the sputtering yields. These large uncertainties in beryllium armor loss estimates are a good example of the experimental nature of the ITER project and will not be truly resolved until ITER begins burning plasma operation.

  13. The benchmark experiment on slab beryllium with D–T neutrons for validation of evaluated nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Nie, Y., E-mail: nieyb@ciae.ac.cn [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); Ren, J.; Ruan, X.; Bao, J. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); Han, R. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Zhang, S. [Institute of Modern Physics, Chinese Academy of Sciences, Lanzhou 730000 (China); Inner Mongolia University for the Nationalities, Inner Mongolia, Tongliao 028000 (China); Huang, H.; Li, X. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); Ding, Y. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China); School of Nuclear Science and Technology, Lanzhou University, Lanzhou 730000 (China); Wu, H.; Liu, P.; Zhou, Z. [Science and Technology on Nuclear Data Laboratory, China Institute of Atomic Energy, Beijing 102413 (China)

    2016-04-15

    Highlights: • Evaluated data for beryllium are validated by a high precision benchmark experiment. • Leakage neutron spectra from pure beryllium slab are measured at 61° and 121° using time-of-flight method. • The experimental results are compared with the MCNP-4B calculations with the evaluated data from different libraries. - Abstract: Beryllium is the most favored neutron multiplier candidate for solid breeder blankets of future fusion power reactors. However, beryllium nuclear data are differently presented in modern nuclear data evaluations. In order to validate the evaluated nuclear data on beryllium, in the present study, a benchmark experiment has been performed at China Institution of Atomic Energy (CIAE). Neutron leakage spectra from pure beryllium slab samples were measured at 61° and 121° using time-of-flight method. The experimental results were compared with the calculated ones by MCNP-4B simulation, using the evaluated data of beryllium from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 libraries. From the comparison between the measured and the calculated spectra, it was found that the calculation results based on CENDL-3.1 caused overestimation in the energy range from about 3–12 MeV at 61°, while at 121°, all the libraries led to underestimation below 3 MeV.

  14. The benchmark experiment on slab beryllium with D–T neutrons for validation of evaluated nuclear data

    International Nuclear Information System (INIS)

    Nie, Y.; Ren, J.; Ruan, X.; Bao, J.; Han, R.; Zhang, S.; Huang, H.; Li, X.; Ding, Y.; Wu, H.; Liu, P.; Zhou, Z.

    2016-01-01

    Highlights: • Evaluated data for beryllium are validated by a high precision benchmark experiment. • Leakage neutron spectra from pure beryllium slab are measured at 61° and 121° using time-of-flight method. • The experimental results are compared with the MCNP-4B calculations with the evaluated data from different libraries. - Abstract: Beryllium is the most favored neutron multiplier candidate for solid breeder blankets of future fusion power reactors. However, beryllium nuclear data are differently presented in modern nuclear data evaluations. In order to validate the evaluated nuclear data on beryllium, in the present study, a benchmark experiment has been performed at China Institution of Atomic Energy (CIAE). Neutron leakage spectra from pure beryllium slab samples were measured at 61° and 121° using time-of-flight method. The experimental results were compared with the calculated ones by MCNP-4B simulation, using the evaluated data of beryllium from the CENDL-3.1, ENDF/B-VII.1 and JENDL-4.0 libraries. From the comparison between the measured and the calculated spectra, it was found that the calculation results based on CENDL-3.1 caused overestimation in the energy range from about 3–12 MeV at 61°, while at 121°, all the libraries led to underestimation below 3 MeV.

  15. The impact of transient thermal loads on beryllium as plasma facing material

    Energy Technology Data Exchange (ETDEWEB)

    Spilker, Benjamin Christof

    2017-01-24

    The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO{sub 2} free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m{sup -2} range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby

  16. The impact of transient thermal loads on beryllium as plasma facing material

    International Nuclear Information System (INIS)

    Spilker, Benjamin Christof

    2017-01-01

    The rising global energy consumption requires a broad research and development approach in the field of energy technology. Besides renewables, nuclear fusion promises an efficient, CO_2 free, no long-term radioactive waste producing, and safe energy source using only deuterium and lithium as primary resources, which are widely abundant. However, several technical challenges have to be overcome before a nuclear fusion power plant can be built. For this purpose, the experimental reactor ITER is currently under construction in France. ITER is intended to demonstrate the scientific and technological feasibility of net energy generation via nuclear fusion. The most heavily loaded components inside a fusion reactor, which are directly facing the fusion plasma, have to be armoured with well suited materials, which need to be able to withstand the high thermal and particle loads for an economically reasonable lifetime. For ITER, beryllium is chosen as plasma facing material for the largest fraction of the inner vacuum vessel, the so called first wall. Tungsten will be applied in the bottom region of the vacuum vessel, the so called divertor, which acts as the exhaust system of the machine. The choice of beryllium as plasma facing material was driven by its outstanding advantages, e.g. the low atomic number assures that eroded wall material does not strongly decrease the fusion plasma performance, while it combines a high thermal conductivity with low chemical sputtering characteristics. However, the relatively low melting temperature of beryllium of 1287 C comprises the risk of amour damage by melting during transient plasma events, such as edge localized modes or plasma disruptions. Even when mitigated, these events put tremendous power densities in the GW m"-"2 range with durations in the ms scale onto the plasma facing materials. Hence, the performance of the ITER reference beryllium grade S-65 under transient thermal loads was studied within this work. Thereby, the

  17. Beryllium and growth. II. The effect of beryllium on plant growth

    Energy Technology Data Exchange (ETDEWEB)

    Hoagland, M B

    1952-01-01

    Experiments were undertaken to determine whether beryllium could replace magnesium in a growing organism. This was stimulated by the several known growth effects of beryllium in animals and by the fact that beryllium apparently competes with magnesium for animal alkaline phosphatases. The following findings are noted: (1) beryllium can reduce the magnesium requirement of plants by some 60% within a certain range of magnesium deficiency. (2) The residual obligatory magnesium requirements is probably accounted for by chlorophyll since beryllium appears to have no primary effect on chlorophyll or chlorophyll production. (3) The pH of the nutrient solution is critical: at acid pH's, beryllium is highly toxic, and growth increase due to beryllium only appears at initial pH's above 11.2, although this initial pH rapidly falls to neutrality during the experimental period. 22 references, 4 figures, 1 table.

  18. Measurement of the thermal conductivity and heat transfer coefficient of a binary bed of beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Donne, M.D.; Piazza, G. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Goraieb, A.; Sordon, G.

    1998-01-01

    The four ITER partners propose to use binary beryllium pebble bed as neutron multiplier. Recently this solution has been adopted for the ITER blanket as well. In order to study the heat transfer in the blanket the effective thermal conductivity and the wall heat transfer coefficient of the bed have to be known. Therefore at Forschungszentrum Karlsruhe heat transfer experiments have been performed with a binary bed of beryllium pebbles and the results have been correlated expressing thermal conductivity and wall heat transfer coefficients as a function of temperature in the bed and of the difference between the thermal expansion of the bed and of that of the confinement walls. The comparison of the obtained correlations with the data available from the literature show a quite good agreement. (author)

  19. Beryllium in aircraft brakes - a summary

    International Nuclear Information System (INIS)

    Zenczak, S.

    1977-01-01

    Beryllium has been in use in aircraft brakes for ten years. During the original design phases of the several aircraft programs using beryllium a number of problems requiring solution confronted the designers. In actual service the solution to these problems performed much better than had been anticipated. A summary is presented. (author)

  20. ICT diagnostic method of beryllium welding quality

    International Nuclear Information System (INIS)

    Sun Lingxia; Wei Kentang; Ye Yunchang

    2002-01-01

    To avoid the interference of high density material for the quality assay of beryllium welding line, a slice by slice scanning method was proposed based upon the research results of the Industrial Computerized Tomography (ICT) diagnostics for weld penetration, weld width, off-centered deviation and weld defects of beryllium-ring welding seam with high density material inside

  1. Investigation of beryllium/steam interaction

    Energy Technology Data Exchange (ETDEWEB)

    Chekhonadskikh, A.M.; Vurim, A.D.; Vasilyev, Yu.S.; Pivovarov, O.S. [Inst. of Atomic Energy National Nuclear Center of the Republic of Kazakstan Semipalatinsk (Kazakhstan); Shestakov, V.P.; Tazhibayeva, I.L.

    1998-01-01

    In this report program on investigations of beryllium emissivity and transient processes on overheated beryllium surface attacked by water steam to be carried out in IAE NNC RK within Task S81 TT 2096-07-16 FR. The experimental facility design is elaborated in this Report. (author)

  2. Modeling of hydrogen interactions with beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Longhurst, G.R. [Lockheed Martin Idaho Technologies Co., Idaho Falls, ID (United States)

    1998-01-01

    In this paper, improved mathematical models are developed for hydrogen interactions with beryllium. This includes the saturation effect observed for high-flux implantation of ions from plasmas and retention of tritium produced from neutronic transmutations in beryllium. Use of the models developed is justified by showing how they can replicated experimental data using the TMAP4 tritium transport code. (author)

  3. Managing health effects of beryllium exposure

    National Research Council Canada - National Science Library

    Committee on Beryllium Alloy Exposures; Committee on Toxicology; National Research Council; Division on Earth and Life Studies; National Research Council

    2008-01-01

    ... to its occurrence in exposed people. Despite reduced workplace exposure, chronic beryllium disease continues to occur. In addition, beryllium has been classified as a likely human carcinogen by several agencies, such as the International Agency for Research on Cancer, the National Toxicology Program, and the U.S. Environmental Protection Agency. Thos...

  4. Structure investigations of some beryllium materials

    Energy Technology Data Exchange (ETDEWEB)

    Faeldt, I; Lagerberg, G

    1960-05-15

    Metallographic structure, microhardness and texture have been studied on various types of beryllium metal including hot pressed powder, a rolled strip and an extruded tube It was found that beryllium exhibits its highest hardness in directions perpendicular to the basal plane. Good ideas of the prevailing textures were obtained with an ordinary X-ray diffractometer.

  5. Structure investigations of some beryllium materials

    International Nuclear Information System (INIS)

    Faeldt, I.; Lagerberg, G.

    1960-05-01

    Metallographic structure, microhardness and texture have been studied on various types of beryllium metal including hot pressed powder, a rolled strip and an extruded tube It was found that beryllium exhibits its highest hardness in directions perpendicular to the basal plane. Good ideas of the prevailing textures were obtained with an ordinary X-ray diffractometer

  6. Iterating skeletons

    DEFF Research Database (Denmark)

    Dieterle, Mischa; Horstmeyer, Thomas; Berthold, Jost

    2012-01-01

    a particular skeleton ad-hoc for repeated execution turns out to be considerably complicated, and raises general questions about introducing state into a stateless parallel computation. In addition, one would strongly prefer an approach which leaves the original skeleton intact, and only uses it as a building...... block inside a bigger structure. In this work, we present a general framework for skeleton iteration and discuss requirements and variations of iteration control and iteration body. Skeleton iteration is expressed by synchronising a parallel iteration body skeleton with a (likewise parallel) state......Skeleton-based programming is an area of increasing relevance with upcoming highly parallel hardware, since it substantially facilitates parallel programming and separates concerns. When parallel algorithms expressed by skeletons involve iterations – applying the same algorithm repeatedly...

  7. BERYLLIUM MEASUREMENT IN COMMERCIALLY AVAILABLE WET WIPES

    Energy Technology Data Exchange (ETDEWEB)

    Youmans-Mcdonald, L.

    2011-02-18

    Analysis for beryllium by fluorescence is now an established method which is used in many government-run laboratories and commercial facilities. This study investigates the use of this technique using commercially available wet wipes. The fluorescence method is widely documented and has been approved as a standard test method by ASTM International and the National Institute for Occupational Safety and Health (NIOSH). The procedure involves dissolution of samples in aqueous ammonium bifluoride solution and then adding a small aliquot to a basic hydroxybenzoquinoline sulfonate fluorescent dye (Berylliant{trademark} Inc. Detection Solution Part No. CH-2) , and measuring the fluorescence. This method is specific to beryllium. This work explores the use of three different commercial wipes spiked with beryllium, as beryllium acetate or as beryllium oxide and subsequent analysis by optical fluorescence. The effect of possible interfering metals such as Fe, Ti and Pu in the wipe medium is also examined.

  8. Beryllium Measurement In Commercially Available Wet Wipes

    International Nuclear Information System (INIS)

    Youmans-Mcdonald, L.

    2011-01-01

    Analysis for beryllium by fluorescence is now an established method which is used in many government-run laboratories and commercial facilities. This study investigates the use of this technique using commercially available wet wipes. The fluorescence method is widely documented and has been approved as a standard test method by ASTM International and the National Institute for Occupational Safety and Health (NIOSH). The procedure involves dissolution of samples in aqueous ammonium bifluoride solution and then adding a small aliquot to a basic hydroxybenzoquinoline sulfonate fluorescent dye (Berylliant(trademark) Inc. Detection Solution Part No. CH-2) , and measuring the fluorescence. This method is specific to beryllium. This work explores the use of three different commercial wipes spiked with beryllium, as beryllium acetate or as beryllium oxide and subsequent analysis by optical fluorescence. The effect of possible interfering metals such as Fe, Ti and Pu in the wipe medium is also examined.

  9. Assessment of LANL beryllium waste management documentation

    International Nuclear Information System (INIS)

    Danna, J.G.; Jennrich, E.A.; Lund, D.M.; Davis, K.D.; Hoevemeyer, S.S.

    1991-04-01

    The objective of this report is to determine present status of the preparation and implementation of the various high priority documents required to properly manage the beryllium waste generated at the Laboratory. The documents being assessed are: Waste Acceptance Criteria, Waste Characterization Plan, Waste Certification Plan, Waste Acceptance Procedures, Waste Characterization Procedures, Waste Certification Procedures, Waste Training Procedures and Waste Recordkeeping Procedures. Beryllium is regulated (as a dust) under 40 CFR 261.33 as ''Discarded commercial chemical products, off specification species, container residues and spill residues thereof.'' Beryllium is also identified in the 3rd thirds ruling of June 1, 1990 as being restricted from land disposal (as a dust). The beryllium waste generated at the Laboratory is handled separately because beryllium has been identified as a highly toxic carcinogenic material

  10. Postirradiation examination of beryllium pebbles

    International Nuclear Information System (INIS)

    Gelles, D.S.

    1998-01-01

    Postirradiation examinations of COBRA-1A beryllium pebbles irradiated in the EBR-II fast reactor at neutron fluences which generated 2700--3700 appm helium have been performed. Measurements included density change, optical microscopy, scanning electron microscopy, and transmission electron microscopy. The major change in microstructure is development of unusually shaped helium bubbles forming as highly non-equiaxed thin platelet-like cavities on the basal plane. Measurement of the swelling due to cavity formation was in good agreement with density change measurements

  11. Characteristics of microstructure and tritium release properties of different kinds of beryllium pebbles for application in tritium breeding modules

    Energy Technology Data Exchange (ETDEWEB)

    Kurinskiy, P., E-mail: petr.kurinskiy@kit.edu [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Vladimirov, P.; Moeslang, A. [Karlsruhe Institute of Technology, Institute for Applied Materials – Applied Materials Physics (IAM-AWP), P.O. Box 3640, Karlsruhe 76021 (Germany); Rolli, R. [Karlsruhe Institute of Technology, Institute for Applied Materials – Materials and Biomechanics (IAM-WBM), P.O. Box 3640, Karlsruhe 76021 (Germany); Zmitko, M. [The European Joint Undertaking for ITER and the Development of Fusion Energy, c/Josep Pla, no. 2, Torres Diagonal Litoral, Edificio B3, Barcelona 08019 (Spain)

    2014-10-15

    Highlights: • Tritium release properties and characteristics of microstructure of beryllium pebbles having different sizes of grains were studied. • Fine-grained beryllium pebbles showed the best ability to release tritium compared to pebbles from another charges. • Be pebbles with the grain sizes exceeding 100 μm contain a great number of small pores and inclusions presumably referring to the history of material fabrication. • The sizes of grains are one of a key characteristic of microstructure which influences the parameters of tritium release. - Abstract: Beryllium pebbles with diameters of 1 mm are considered to be perspective material for the use as neutron multiplier in tritium breeding modules of fusion reactors. Up to now, the design of helium-cooled breeding blanket in ITER project foresees the use of 1 mm beryllium pebbles fabricated by NGK Insulators Ltd., Japan. It is notable that beryllium pebbles from Russian Federation and USA are also available and the possibility of their large-scale fabrication is under study. Presented work is dedicated to a study of characteristics of microstructure and parameters of tritium release of beryllium pebbles produced by Bochvar Institute, Russian Federation, and Materion Corporation, USA.

  12. Implementation of technical stage of ITER project in Kazakstan

    International Nuclear Information System (INIS)

    Shkol'nik, V.S.; Shemanskij, V.A.; Cherepnin, Yu.S.; Skoz, E.Ya.; Nazarenko, L.I.; Shestakov, V.P.; Tazhibaeva, I.L.

    1996-01-01

    The main executors of ITER project in Kazakstan are: National nuclear Centre; National Joint-stock Co. of the enterprises on atomic engineering and industry (KATEP) and Science and Research Inst. of Experimental and Theoretical Physics of the Kazakh National Univ. Principle directions of Kazakhstan Republic works in 1996/97 under the ITER project is possible to formulate as follows: - irradiation test of materials on the BN-350 reactor; -continuation of in-pile experiments on determination of deuterium interaction parameters with beryllium samples, manufactured by Ulba Metallurgical Plant (KATEP), using Research Water Heterogeneous Reactor; - conducting of in-pile experiments on determination of hydrogen isotope permeation parameters through vanadium alloys using Research Water Heterogeneous Reactor. Conducting of out-pile experiments with beryllium, and vanadium alloys; - investigation of beryllium structure and properties in respect to hydrogen in condition, simulating gas impurities inventory, under thermal and neutron influence; -simulation of plasma disruption on divertor; -simulation of water steam interaction with beryllium; - determination of hydrogen permeation parameters for various materials with 'in situ' surface by electronic spectroscopy; -development of manufacturing technology of beryllium connections with copper alloys

  13. Technical issues for beryllium use in fusion blanket applications

    International Nuclear Information System (INIS)

    McCarville, T.J.; Berwald, D.H.; Wolfer, W.; Fulton, F.J.; Lee, J.D.; Maninger, R.C.; Moir, R.W.; Beeston, J.M.; Miller, L.G.

    1985-01-01

    Beryllium is an excellent non-fissioning neutron multiplier for fusion breeder and fusion electric blanket applications. This report is a compilation of information related to the use of beryllium with primary emphasis on the fusion breeder application. Beryllium resources, production, fabrication, properties, radiation damage and activation are discussed. A new theoretical model for beryllium swelling is presented

  14. Production of beryllium oxide of nuclear purity from beryl

    Energy Technology Data Exchange (ETDEWEB)

    Copat, A; Sood, S P

    1984-01-01

    Production of beryllium oxide from beryl by the fluoride process was optimized in this study. Optimum results were obtained using a mixture of sodium hexafluorsilicate and sodium hexafluorferrate as flux and calcinating at 740/sup 0/C for 2 hours. The beryllium concentrate produced was further purified by crystallization as beryllium sulfate to obtain nuclear grade beryllium oxide

  15. Production of beryllium oxide of nuclear purity from beryl

    International Nuclear Information System (INIS)

    Copat, A.; Sood, S.P.

    1983-01-01

    Production of beryllium oxide from beryl by the fluoride process was optimized in this study. Optimum results were obtained using a mixture of sodium hexafluorsilicate and sodium hexafluorferrate as flux and calcinating at 740 0 C for 2 hours. The beryllium concentrate produced was further purified by crystallization as beryllium sulfate to obtain nuclear grade beryllium oxide (Author) [pt

  16. ITER safety

    International Nuclear Information System (INIS)

    Raeder, J.; Piet, S.; Buende, R.

    1991-01-01

    As part of the series of publications by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this document describes the ITER safety analyses. It contains an assessment of normal operation effluents, accident scenarios, plasma chamber safety, tritium system safety, magnet system safety, external loss of coolant and coolant flow problems, and a waste management assessment, while it describes the implementation of the safety approach for ITER. The document ends with a list of major conclusions, a set of topical remarks on technical safety issues, and recommendations for the Engineering Design Activities, safety considerations for siting ITER, and recommendations with regard to the safety issues for the R and D for ITER. Refs, figs and tabs

  17. Fluorimetric method for determination of Beryllium

    International Nuclear Information System (INIS)

    Sparacino, N.; Sabbioneda, S.

    1996-10-01

    The old fluorimetric method for the determination of Beryllium, based essentially on the fluorescence of the Beryllium-Morine complex in a strongly alkaline solution, is still competitive and stands the comparison with more modern methods or at least three reasons: in the presence of solid or gaseous samples (powders), the times necessary to finalize an analytic determination are comparable since the stage of the process which lasts the longest is the mineralization of the solid particles containing Beryllium, the cost of a good fluorimeter is by far Inferior to the cost, e. g., of an Emission Spectrophotometer provided with ICP torch and magnets for exploiting the Zeeman effect and of an Atomic absorption Spectrophotometer provided with Graphite furnace; it is possible to determine, fluorimetrically, rather small Beryllium levels (about 30 ng of Beryllium/sample), this potentiality is more than sufficient to guarantee the respect of all the work safety and hygiene rules now in force. The study which is the subject of this publication is designed to the analysis procedure which allows one to reach good results in the determination of Beryllium, chiefly through the control and measurement of the interference effect due to the presence of some metals which might accompany the environmental samples of workshops and laboratories where Beryllium is handled, either at the pure state or in its alloys. The results obtained satisfactorily point out the merits and limits of this analytic procedure

  18. Research of flaw assessment methods for beryllium reflector elements

    International Nuclear Information System (INIS)

    Shibata, Akira; Ito, Masayasu; Takemoto, Noriyuki; Tanimoto, Masataka; Tsuchiya, Kunihiko; Nakatsuka, Masafumi; Ohara, Hiroshi; Kodama, Mitsuhiro

    2012-02-01

    Reflector elements made from metal beryllium is widely used as neutron reflectors to increase neutron flux in test reactors. When beryllium reflector elements are irradiated by neutron, bending of reflector elements caused by swelling occurs, and beryllium reflector elements must be replaced in several years. In this report, literature search and investigation for non-destructive inspection of Beryllium and experiments for Preliminary inspection to establish post irradiation examination method for research of characteristics of metal beryllium under neutron irradiation were reported. (author)

  19. Preliminary proposal for a beryllium technology program for fusion applications

    International Nuclear Information System (INIS)

    1985-02-01

    The program was designed to provide the answers to the critical issues of beryllium technology needed in fusion blanket designs. The four tasks are as follows: (1) Beryllium property measurements needed for fusion data base. (2) Beryllium stress relaxation and creep measurements for lifetime modelling calculations. (3) Simplified recycle technique development for irradiated beryllium. (4) Beryllium neutron multiplier measurements using manganese bath absolute calibration techniques

  20. Beryllium and growth. III. The effect of beryllium on plant phosphatase

    Energy Technology Data Exchange (ETDEWEB)

    Hoagland, M B

    1952-01-01

    The purpose of the investigations was to correlate the apparent ability of beryllium to substitute for magnesium in plant growth with a specific biochemical effect of the metal. Through association with earlier work on beryllium inhibition of animal alkaline phosphatase, a study was made of the effect of beryllium and other metals upon the activity of a phosphatase derived from tomato leaves. Although only indirect evidence is available that this enzyme system was magnesium-activated, beryllium was found to inhibit reversibly the splitting of GP and ATP. Other metals were also found to be inhibitory but the ATP-ase inhibition - and especially the ratio of P split from GP to P split from ATP - was higher for beryllium than for any other metal studied. The significance of this finding in relation to energy metabolism, growth, and beryllium toxicity is discussed. 12 references, 5 figures, 2 tables.

  1. Spectrographic measurement of beryllium in the atmosphere; Dosage spectrographique du beryllium dans l'atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Artaud, J; Cittanova, J [Commissariat a l' Energie Atomique, Service d' Analyses et Recherches Chimiques Appliquees, Saclay (France). Centre d' Etudes Nucleaires; Crehange, G; Frequelin, S [Commissariat a l' Energie Atomique, Dir. des Applications Militaires, Service Chimie, Saclay (France). Centre d' Etudes Nucleaires; Baudin, G [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1961-07-01

    We describe here a method for the spectrographic determination of beryllium on filters which is valid for amounts varying between 0,01 and 30 {mu}g of beryllium and which is independent of the nature of the beryllium compound involved. This is a flux method (graphite-lithium carbonate mixture), the excitation being by a direct current arc. (author) [French] Nous decrivons ici, une methode de dosage spectrographique de beryllium sur filtre, valable pour des teneurs comprises entre 0,01 et 30 {mu}g de beryllium et independante de la nature du compose de beryllium a doser. C'est une methode de 'flux' (melange graphite-carbonate de lithium) l'excitation etant un arc a courant continu. (auteur)

  2. Laser cleaning of ITER's diagnostic mirrors

    Science.gov (United States)

    Skinner, C. H.; Gentile, C. A.; Doerner, R.

    2012-10-01

    Practical methods to clean ITER's diagnostic mirrors and restore reflectivity will be critical to ITER's plasma operations. We report on laser cleaning of single crystal molybdenum mirrors coated with either carbon or beryllium films 150 - 420 nm thick. A 1.06 μm Nd laser system provided 220 ns pulses at 8 kHz with typical power densities of 1-2 J/cm^2. The laser beam was fiber optically coupled to a scanner suitable for tokamak applications. The efficacy of mirror cleaning was assessed with a new technique that combines microscopic imaging and reflectivity measurements [1]. The method is suitable for hazardous materials such as beryllium as the mirrors remain sealed in a vacuum chamber. Excellent restoration of reflectivity for the carbon coated Mo mirrors was observed after laser scanning under vacuum conditions. For the beryllium coated mirrors restoration of reflectivity has so far been incomplete and modeling indicates that a shorter duration laser pulse is needed. No damage of the molybdenum mirror substrates was observed.[4pt][1] C.H. Skinner et al., Rev. Sci. Instrum. at press.

  3. Metallurgical viewpoints on the brittleness of beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Lagerberg, G

    1960-02-15

    At present the development and use of beryllium metal for structural applications is severely hampered by its brittleness. Reasons for this lack of ductility are reviewed in discussing the deformation behaviour of beryllium in relation to other hexagonal metals. The ease of fracturing in beryllium is assumed to be a consequence of a limited number of deformation modes in combination with high deformation resistance. Models for the nucleation of fracture are suggested. The relation of ductility to elastic constants as well as to grain size, texture and alloying additions is discussed.

  4. THE IDAHO NATIONAL LABORATORY BERYLLIUM TECHNOLOGY UPDATE

    International Nuclear Information System (INIS)

    Glen R. Longhurst

    2007-01-01

    A Beryllium Technology Update meeting was held at the Idaho National Laboratory on July 18, 2007. Participants came from the U.S., Japan, and Russia. There were two main objectives of this meeting. One was a discussion of current technologies for beryllium in fission reactors, particularly the Advanced Test Reactor and the Japan Materials Test Reactor, and prospects for material availability in the coming years. The second objective of the meeting was a discussion of a project of the International Science and Technology Center regarding treatment of irradiated beryllium for disposal. This paper highlights discussions held during that meeting and major conclusions reached

  5. Metallurgical viewpoints on the brittleness of beryllium

    International Nuclear Information System (INIS)

    Lagerberg, G.

    1960-02-01

    At present the development and use of beryllium metal for structural applications is severely hampered by its brittleness. Reasons for this lack of ductility are reviewed in discussing the deformation behaviour of beryllium in relation to other hexagonal metals. The ease of fracturing in beryllium is assumed to be a consequence of a limited number of deformation modes in combination with high deformation resistance. Models for the nucleation of fracture are suggested. The relation of ductility to elastic constants as well as to grain size, texture and alloying additions is discussed

  6. Influence of stray light for divertor spectroscopy in ITER

    International Nuclear Information System (INIS)

    Kajita, Shin; Veshchev, Evgeny; Lisgo, Steve; Barnsley, Robin; Morgan, Philip; Walsh, Michael; Ogawa, Hiroaki; Sugie, Tatsuo; Itami, Kiyoshi

    2015-01-01

    The influence of stray light in the divertor spectroscopy system in ITER is quantitatively investigated using a ray tracing simulation. Simulation results show that the stray light is negligible at positions in the divertor where the plasma emission is strong. However, it is also shown that the stray light can be significantly greater than the real signal if the plasma intensity is low. Deuterium and beryllium emissions are used for the assessment; for beryllium cases in particular, since the emission profile may be non-uniform in the divertor region, the influence of stray light can be non-negligible at some positions, e.g., above the divertor dome

  7. Beryllium health effects in the era of the beryllium lymphocyte proliferation test.

    Science.gov (United States)

    Maier, L A

    2001-05-01

    The beryllium lymphocyte proliferation test (BeLPT) has revolutionized our approach to the diagnosis, screening, and surveillance of beryllium health effects. Based on the development of a beryllium-specific cell-mediated immune response, the BeLPT has allowed us to define early health effects of beryllium, including beryllium sensitization (BeS), and chronic beryllium disease (CBD) at a subclinical stage. The use of this test as a screening tool has improved our understanding of these health effects. From a number of studies it is apparent that BeS precedes CBD and develops after as little as 9 weeks of beryllium exposure. CBD occurs within 3 months and up to 30 years after initial beryllium exposure. Exposure-response variables have been associated with BeS/CBD, including work as a machinist, chemical or metallurgical operator, laboratory technician, work in ceramics or beryllium metal production, and years of beryllium exposure. Recent studies have found BeS and CBD in workplaces in which the majority of exposures were below the 2 microg/m3 OSHA time-weighted average (TWA). Ideally, the BeLPT would be used in surveillance aimed at defining other risk-related processes, determining exposure variables which predict BeS and CBD, and defining the exposure level below which beryllium health effects do not occur. Unfortunately, the BeLPT can result in false negative tests and still requires an invasive procedure, a bronchoscopy, for the definitive diagnosis of CBD. Thus, research is needed to establish new tests to be used alone or in conjunction with the BeLPT to improve our ability to detect early beryllium health effects.

  8. The analysis of beryllium-copper diffusion joint after HHF test

    International Nuclear Information System (INIS)

    Guiniatouline, R.N.; Mazul, I.V.; Rubkin, S.Y.

    1995-01-01

    The development of beryllium-copper joints which can withstand to relevant ITER divertor conditions is one of the important tasks at present time. One of the main problem for beryllium-copperjoints, is the inter-metallic layers, the strength and life time of joints significantly depends from the width and contents of the intermetallic layers. The objective of this work is to study the diffusion joint of TGP-56 beryllium to OFHC copper after thermal response and thermocyclic tests with beryllium-copper mockup. The BEY test were performed at e-beam facility (EBTS, SNLA). The following methods were used for analyses: the roentgenographic analysis; X-ray spectrum analysis; the fracture graphic analysis. During the investigation the followed studies were done: the analysis of diffusion boundary Be-Cu, which was obtained at the crossection of one of the tiles, the analysis of the debonded surfaces of a few beryllium tiles and corresponding copper parts; the analysis of upper surface of one of the tiles after HHF tests. The results of this work have showed that: the joint roentgenographic and elements analyses indicated the following phases in the diffusion zone: Cu 2 Be (∼170 μm), CuBe (∼30μm), CuBe 2 (∼1 μm) and solid solution of copper in beryllium. The phases Cu 2 Be, CuBe and solid solution of copper in beryllium were indicated using quantitative microanalysis and phases CuBe, CuBe 2 , Cu, Be - by roentgenographic analysis; the source of fracture (initial crack) is located in the central part of the tiles, the crack caused by the influence of residual stresses during cooling of a mock-up after fabrication and developed under the conditions of slow elastic-plastic growing during the process of thermal fatigue testing. The analysis gives the important data about joint's quality and also may be used for any type of joints and its comparison for ITER applications

  9. MD simulation: determination of the physical properties and surface vaporization analysis of beryllium armours

    International Nuclear Information System (INIS)

    Prinzio, M. Di; Aquaro, D.

    2006-01-01

    The erosion of the divertor and of the first wall determined on the base of the anticipated operating conditions, is a critical issue that could affect the performance and the operating schedule of the nuclear fusion reactor ITER. This paper deals with the analysis of beryllium thermal properties by means of MD simulations, in order to better predict thermal behaviour of beryllium armoured PFCs in fusion devices. The importance of this analysis is clearly connected to thermal response evaluation of PFCs to high heat flux exposure, during off-normal events and Edge Localized Modes. The ensuing strong over-heating, in fact, produces material ablation through vaporization of surface material layers and possible loss of melting material. The overall PFCs erosion has bearings on plasma contamination, due to eroded material transport, and components lifetime, due to armour thickness reduction. An important feature of beryllium is its high vapour pressure. During thermal transients the strong vaporization keeps surface temperature relatively low but eroded thickness results high as well. Small changes in beryllium vapour pressure produce not negligible differences in thermal analyses results. On the basis of available force fields, classical Molecular Dynamics simulations have been carried out in order to better understand surface vaporization in tokamak conditions and to evaluate the effect of beryllium oxides formation. This effect has been successfully modelled by MD simulation, carried out with Moldy code. Morse stretching and bending potential for Be-O bond simulation have been used, and partial charges method, accounting for molecular polarity, has been employed. Since during short thermal transients, such as ELMs, only a few microns of Be armour will be overheated and reach melting threshold, the effective thermal conductivity is very important in determining the temperature evolution of surface layers and the ensuing erosion. Thermal conductivity can be evaluated

  10. An assessment of disruption erosion in ITER environment

    International Nuclear Information System (INIS)

    Hassanein, A.; Konkashbaev, I.

    1994-01-01

    The behavior of divertor materials during a major disruption in ITER is very important for the successful and reliable operation of the reactor. Erosion of material surfaces due to the thermal energy dump can severely limit the lifetime of the plasma facing components therefore degrading reactor economic feasibility. A comprehensive numerical model recently developed is used in this analysis in which all major physical processes taking place during plasma-material interactions are included. Models to account for material thermal evolution, plasma-vapor interaction physics, and models for hydrodynamic radiation transport in the developed vapor cloud are implemented in a self-consistent manner to realistically assess the disruption damage. The extent of the self-protection from the developed vapor cloud in front of the incoming plasma particles is critically important in determining the overall disruption lifetime. The aim of this study is to estimate the divertor lifetime for a range of reactor conditions. Candidate materials such as beryllium and graphite are both considered in this analysis. The dependence of the divertor disruption lifetime on the characteristics of plasma-vapor interaction zone for incident plasma ions and electrons is analyzed and discussed. The effect of uncertainties in reactor disruption conditions on the overall divertor erosion lifetime is also analyzed

  11. Combined aging of beryllium bronze

    International Nuclear Information System (INIS)

    Duraev, P.P.; Kaplun, Yu.A.; Pastukhova, Zh.P.; Rakhshtadt, A.G.

    1986-01-01

    This article evaluates the possibility of increasing the resistance of beryllium bronze to small plastic deformations as a result of the application of stepped aging under stress. Low-temperature aging under conditions of bending under a stress of about 100 MPa was applied to alloy BrBNT1, 9Mg at 150, 180, and 210 0 C, high-temperature aging at 300 and 340 0 C under stress and without stress. As a result of applying stepped aging under stress, the elastic limit of the alloy BrBNT1, 9Mg was raised to 900 MPa. Stepped aging under stress has a substantial effect on the relaxation stability of the alloy. The procedure suggested in the article for aging may be used efficiently for treating elastic elements made of other brands of bronze as well

  12. Status of the beryllium replacement project

    Energy Technology Data Exchange (ETDEWEB)

    Dimayuga, I. [Atomic Energy of Canada, Chalk River, Ontario (Canada); Corcoran, E. [Royal Military College of Canada, Kingston, Ontario (Canada); Daniels, T. [Ontario Power Generation, Pickering, Ontario (Canada); Harmsen, J. [Cameco Fuel Manufacturing Inc., Port Hope, Ontario (Canada); Lu, E. [Bruce Power, Tiverton, Ontario (Canada); Onderwater, T.; Palleck, S. [General Electric- Hitachi, Wilmington, North Carolina (United States); Pant, A. [Cameco Fuel Manufacturing Inc., Port Hope, Ontario (Canada)

    2013-07-01

    Currently, beryllium (Be) is used as the filler metal for brazing appendages on the sheaths of CANDU® fuel elements. Because of its toxicity, occupational exposure limits for Be are being reduced to very low levels, resulting in significant challenges to CANDU® fuel fabricators. The CANDU® Owners Group (COG) initiated a test program to identify a filler material to replace Be and confirm that the brazed joints meet the established technical requirements for CANDU® fuel. Together with eliminating health risks associated with the use of Be, the industry needs to be assured that continuation of fuel supply remains unaffected and that fuel fabrication processes continue to comply with health and safety standards. A literature survey of studies on brazing and joining of Zircaloy identified potential filler materials that can meet or exceed existing design requirements of the brazed joint, including the required mechanical, microstructural, corrosion resistance, and irradiation properties equivalent to those obtained with Be as braze material. Candidate materials were evaluated against several criteria, including manufacturability, melting point, wettability, mechanical properties, corrosion resistance, effect on neutron economy, potential activation products, and interaction with fuel channels and other related disciplines. This exercise resulted in a list of promising candidate materials that were recommended for the first phase of testing. These materials include stainless steel (304 or 316), Al-Si, Ni-P, and Zr-Mn alloys. To allow a CANDU® utility have sufficient confidence in considering implementation of a different braze filler material, a Be Replacement Test Program, involving out-reactor and in-reactor tests, is being undertaken as a collaborative endeavour by the Canadian nuclear industry. The out-reactor tests consist of: a constructability assessment to determine the material’s suitability with current fuel manufacturing methods; evaluation of

  13. ITER overview

    International Nuclear Information System (INIS)

    Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.R.

    2001-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  14. ITER Overview

    International Nuclear Information System (INIS)

    Shimomura, Y.; Aymar, R.; Chuyanov, V.; Huguet, M.; Parker, R.

    1999-01-01

    This report summarizes technical works of six years done by the ITER Joint Central Team and Home Teams under terms of Agreement of the ITER Engineering Design Activities. The major products are as follows: complete and detailed engineering design with supporting assessments, industrial-based cost estimates and schedule, non-site specific comprehensive safety and environmental assessment, and technology R and D to validate and qualify design including proof of technologies and industrial manufacture and testing of full size or scalable models of key components. The ITER design is at an advanced stage of maturity and contains sufficient technical information for a construction decision. The operation of ITER will demonstrate the availability of a new energy source, fusion. (author)

  15. Manufacturing and thermomechanical testing of actively cooled all beryllium high heat flux test pieces

    International Nuclear Information System (INIS)

    Vasiliev, N.N.; Sokolov, Yu.A.; Shatalov, G.E.

    1995-01-01

    One of the problems affiliated to ITER high heat flux elements development is a problem of interface of beryllium protection with heat sink routinely made of copper alloys. To get rid of this problem all beryllium elements could be used as heat receivers in places of enhanced thermal loads. In accordance with this objectives four beryllium test pieces of two types have been manufactured in open-quotes Institute of Berylliumclose quotes for succeeding thermomechanical testing. Two of them were manufactured in accordance with JET team design; they are round open-quotes hypervapotron typeclose quotes test pieces. Another two ones are rectangular test sections with a twisted tape installed inside of the circular channel. Preliminary stress-strain analysis have been performed for both type of the test pieces. Hypervapotrons have been shipped to JET where they were tested on JET test bed. Thermomechanical testing of pieces of the type of open-quotes swirl tape inside of tubeclose quotes have been performed on Kurchatov Institute test bed. Chosen beryllium grade properties, some details of manufacturing, results of preliminary stress-strain analysis and thermomechanical testing of the test pieces open-quotes swirl tape inside of tubeclose quotes type are given in this report

  16. The adhesive bonding of beryllium structural components

    International Nuclear Information System (INIS)

    Fullerton-Batten, R.C.

    1977-01-01

    Where service conditions permit, adhesive bonding is a highly recommendable, reliable means of joining beryllium structural parts. Several important programs have successfully used adhesive bonding for joining structural and non-structural beryllium components. Adhesive bonding minimizes stress concentrations associated with other joining techniques and considerably improves fatigue resistance. In addition, no degradation of base metal properties occur. In many instances, structural joints can be fabricated more cheaply by adhesive bonding or in combination with adhesive bonding than by any other method used alone. An evaluation program on structural adhesive bonding of beryllium sheet components is described. A suitable surface pretreatment for beryllium adherends prior to bonding is given. Tensile shear strength and fatigue properties of FM 1000 and FM 123-5 adhesive bonded joints are reviewed and compared with data obtained from riveted joints of similar geometry. (author)

  17. Chronic Beryllium Disease Prevention Program Report

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S

    2012-03-29

    This document describes how Lawrence Livermore National Laboratory (LLNL) meets the requirements and management practices of federal regulation 10 CFR 850, 'Chronic Beryllium Disease Prevention Program (CBDPP).' This revision of the LLNL CBDPP incorporates clarification and editorial changes based on lessons learned from employee discussions, observations and reviews of Department of Energy (DOE) Complex and commercial industry beryllium (Be) safety programs. The information is used to strengthen beryllium safety practices at LLNL, particularly in the areas of: (1) Management of small parts and components; and (2) Communication of program status to employees. Future changes to LLNL beryllium activities and on-going operating experience will be incorporated into the program as described in Section S, 'Performance Feedback.'

  18. Beryllium-aluminum alloys for investment castings

    International Nuclear Information System (INIS)

    Nachtrab, W.T.; Levoy, N.

    1997-01-01

    Beryllium-aluminum alloys containing greater than 60 wt % beryllium are very favorable materials for applications requiring light weight and high stiffness. However, when produced by traditional powder metallurgical methods, these alloys are expensive and have limited applications. To reduce the cost of making beryllium-aluminum components, Nuclear Metals Inc. (NMI) and Lockheed Martin Electronics and Missiles have recently developed a family of patented beryllium-aluminum alloys that can be investment cast. Designated Beralcast, the alloys can achieve substantial weight savings because of their high specific strength and stiffness. In some cases, weight has been reduced by up to 50% over aluminum investment casting. Beralcast is now being used to make thin wall precision investment castings for several advanced aerospace applications, such as the RAH-66 Comanche helicopter and F-22 jet fighter. This article discusses alloy compositions, properties, casting method, and the effects of cobalt additions on strength

  19. Experiments on tritium behavior in beryllium, (2)

    International Nuclear Information System (INIS)

    Ishitsuka, Etsuo; Kawamura, Hiroshi; Nakata, Hirokatsu; Sugai, Hiroyuki; Tanase, Masakazu.

    1990-02-01

    Beryllium has been used as the neutron reflector of material testing reactor and as the neutron multiplier for the fusion reactor lately. To study the tritium behavior in beryllium, we conducted the experiments, i.e., tritium release by recoil or diffusion by using the hot-pressed beryllium which had been produced both tritium and helium by neutron irradiation. From our experiments, we found that (1) amount of tritium production per one cycle irradiation (lasting 22 days) of JMTR is 10 mCi/g, (2) amount of tritium per surface area of hot-pressed beryllium released by recoil is 4 μCi/cm 2 , (3) diffusion coefficient of tritium in a temperature range of 800 ∼1180degC can be expressed with the following equation; D = 8.7 x 10 4 exp(-2.9x10 5 /R/T) cm 2 /s. (author)

  20. Beryllium concentration in pharyngeal tonsils in children

    Directory of Open Access Journals (Sweden)

    Ewa Nogaj

    2014-06-01

    Full Text Available Power plant dust is believed to be the main source of the increased presence of the element beryllium in the environment which has been detected in the atmospheric air, surface waters, groundwater, soil, food, and cigarette smoke. In humans, beryllium absorption occurs mainly via the respiratory system. The pharyngeal tonsils are located on the roof of the nasopharynx and are in direct contact with dust particles in inhaled air. As a result, the concentration levels of beryllium in the pharyngeal tonsils are likely to be a good indicator of concentration levels in the air. The presented study had two primary aims: to investigate the beryllium concentration in pharyngeal tonsils in children living in southern Poland, and the appropriate reference range for this element in children’s pharyngeal tonsils. Pharyngeal tonsils were extracted from a total of 379 children (age 2–17 years, mean 6.2 ± 2.7 years living in southern Poland. Tonsil samples were mineralized in a closed cycle in a pressure mineralizer PDS 6, using 65% spectrally pure nitric acid. Beryllium concentration was determined using the ICP-AES method with a Perkin Elmer Optima 5300DVTM. The software Statistica v. 9 was used for the statistical analysis. It was found that girls had a significantly greater beryllium concentration in their pharyngeal tonsils than boys. Beryllium concentration varies greatly, mostly according to the place of residence. Based on the study results, the reference value for beryllium in pharyngeal tonsils of children is recommended to be determined at 0.02–0.04 µg/g.

  1. Tritium release from neutron irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik

    1998-01-01

    One of the most important open issues related to beryllium for fusion applications refers to the kinetics of the tritium release as a function of neutron fluence and temperature. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating the beryllium response under neutron irradiation. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from the above mentioned irradiation experiments, trying to elucidate the tritium release controlling processes. In agreement with previous studies it has been found that release starts at about 500-550degC and achieves a maximum at about 700-750degC. The observed release at about 500-550degC is probably due to tritium escaping from chemical traps, while the maximum release at about 700-750degC is due to tritium escaping from physical traps. The consequences of a direct contact between beryllium and ceramics during irradiation, causing tritium implanting in a surface layer of beryllium up to a depth of about 40 mm and leading to an additional inventory which is usually several times larger than the neutron-produced one, are also presented and the effects on the tritium release are discussed. (author)

  2. Selection of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Ulrickson, M.; Barabash, V.; Chiocchio, S.

    1996-01-01

    ITER will be the first tokamak having long pulse operation using deuterium-tritium fuel. The problem of designing heat removal structures for steady state in a neutron environment is a major technical goal for the ITER Engineering Design Activity (EDA). The steady state heat flux specified for divertor components is 5 MW/m 2 for normal operation with transients to 15 MW/m 2 for up to 10 s. The selection of materials for plasma facing components is one of the major research activities. Three materials are being considered for the divertor; carbon fiber composites, beryllium, and tungsten. This paper discusses the relative advantages and disadvantages of these materials. The final section of plasma facing materials for the ITER divertor will not be made until the end of the EDA

  3. Beryllium electrodeposition on aluminium cathode from chloride melts

    International Nuclear Information System (INIS)

    Nichkov, I.F.; Novikov, E.A.; Serebryakov, G.A.; Kanashin, Yu.P.; Sardyko, G.N.

    1980-01-01

    Cathodic processes during beryllium deposition on liquid and solid aluminium cathodes are investigated. Mixture of sodium, potassium and beryllium chloride melts served as an lectrolyte. Beryllium ion discharge at the expense of alloy formation takes place at more positive potentials than on an indifferent cathode at low current densities ( in the case of liquid aluminium cathode). Metallographic analysis and measurements of microhardness have shown, that the cathodic product includes two phases: beryllium solid solution in aluminium and metallic beryllium. It is concluded, that aluminium-beryllium alloys with high cathodic yield by current can be obtained by the electrolytic method

  4. Sanitary-hygienic and ecological aspects of beryllium production

    Energy Technology Data Exchange (ETDEWEB)

    Dvinskykh, E.M.; Savchuk, V.V.; Sidorov, V.L.; Slobodin, D.B.; Tuzov, Y.V. [Ulba Metallurgical Plant, Ust-Kamenogorsk (Kazakhstan)

    1998-01-01

    The Report describes An organization of sanitary-hygienic and ecological control of beryllium production at Ulba metallurgical plant. It involves: (1) the consideration of main methods for protection of beryllium production personnel from unhealthy effect of beryllium, (2) main kinds of filters, used in gas purification systems at different process areas, (3) data on beryllium monitoring in water, soil, on equipment. This Report also outlines problems connected with designing devices for a rapid analysis of beryllium in air as well as problems of beryllium production on ecological situation in the town. (author)

  5. Beryllium satellite thrust cone design, manufacture and test

    International Nuclear Information System (INIS)

    Schneiter, H.; Chandler, D.

    1977-01-01

    Pre-formed beryllium sheet material has been used in the design, manufacturing and test of a satellite thrust cone structure. Adhesive bonding was used for attachment of aluminium flanges and conical segment lap strips. Difficulties in beryllium structure design such as incompatibilities with aluminium and handling problems are discussed. Testing to optimize beryllium-beryllium and beryllium-aluminium adhesive bonds is described. The completed thrust cone assembly has been subjected to static load testing and the results are presented. A summary of the relative merits of the use of beryllium in satellite structures is given with recommendations for future users. (author)

  6. Beryllium-rich intermediate phases in beryllium alloys

    International Nuclear Information System (INIS)

    Raynor, G.V.

    1977-01-01

    The results of a survey of the factors affecting the formation of phases of stoichiometry MBe 5 , M 2 Be 17 , MBe 12 and MBe 13 are presented. Using published information it is shown that the structures adopted at the higher Be:M ratios involve different characteristics from those adopted at lower Be:M ratios. In the ThMn 12 and NaZn 13 structures adopted in the former case dsub(M-Be) > (rsub(M) + rsub(Be)), (where the atomic radii refer to coordination number 12) and the dsub(Be-Be) distances are contracted. In the CaCu 5 structure adopted at composition MBe 5 , dsub(M-Be) < (rsub(M) + rsub(Be)) and interactions between unlike atoms are significant. The structural characteristics, occurrence, and the stabilities of these phases, and of the others mentioned above, are discussed in terms of atomic radius ratios, the position of the M component in the periodic table, and the value of the univalent ionic radius of the M component, taken as a measure of the extension in space of the hard incompressible ionic core. Though the compound-forming characteristics of beryllium are largely dictated by its small atomic diameter, other factors such as the nature of the bonding which can be exerted by the partner atoms are also significant. In particular, the proportions of the volumes of the atoms which are occupied by the hard incompressible ionic cores assume importance. (author)

  7. Thermomechanical testing of beryllium for the JET/ISX-B beryllium limiter experiment

    International Nuclear Information System (INIS)

    Watson, R.D.; Smith, M.F.; Whitley, J.B.; McDonald, J.M.

    1984-01-01

    Materials testing of S-65-B beryllium has been conducted in support of the beryllium limiter experiment on the ISX-B tokamak. The S-65-B grade of hot-pressed beryllium was chosen over S-200-E because of its superior strength and ductility at elevated temperatures. The testing has included measurement of tensile and yield strength, ductility, Young's Modulus, thermal conductivity, and specific heat from 50 0 C to 700 0 C. Thermal fatigue testing of a 2.5 cm beryllium cube was conducted using an electron beam to apply a heat flux of 2.5 kw/cm 2 for 0.3 second pulses for 1500 cycles. Results from the tests are compared to elastic-plastic finite element stress calculations. The testing indicates that the ISX-B beryllium limiter should survive the tokamak environment without serious structural failure, although some surface cracking is expected to occur. (author)

  8. Influence of neutron irradiation on the tritium retention in beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Rolli, R.; Ruebel, S.; Werle, H. [Forschungszentrum Karlsruhe, Inst. fuer Neutronenphysik und Reaktortechnik, Karlsruhe (Germany); Wu, C.H.

    1998-01-01

    Carbon-based materials and beryllium are the candidates for protective layers on the components of fusion reactors facing plasma. In contact with D-T plasma, these materials absorb tritium, and it is anticipated that tritium retention increases with the neutron damage due to neutron-induced traps. Because of the poor data base for beryllium, the work was concentrated on it. Tritium was loaded into the samples from stagnant T{sub 2}/H{sub 2} atmosphere, and afterwards, the quantity of the loaded tritium was determined by purged thermal annealing. The specification of the samples is shown. The samples were analyzed by SEM before and after irradiation. The loading and the annealing equipments are contained in two different glove boxes with N{sub 2} inert atmosphere. The methods of loading and annealing are explained. The separation of neutron-produced and loaded tritium and the determination of loaded tritium in irradiated samples are reported. Also the determination of loaded tritium in unirradiated samples is reported. It is evident that irradiated samples contained much more loaded tritium than unirradiated samples. The main results of this investigation are summarized in the table. (K.I.)

  9. Impact of the surface quality on the thermal shock performance of beryllium armor tiles for first wall applications

    Energy Technology Data Exchange (ETDEWEB)

    Spilker, B., E-mail: b.spilker@fz-juelich.de; Linke, J.; Pintsuk, G.; Wirtz, M.

    2016-11-01

    Highlights: • Different surface qualities of S-65 beryllium are tested under high heat flux conditions. • After 1000 thermal shocks, the loaded area exhibits a crucial destruction. • Stress accelerated grain boundary oxidation/dynamic embrittlement effects are linked to the thermal shock performance of beryllium. • Thermally induced cracks form between 1 and 10 pulses and grow wider and deeper between 10 and 100 pulses. • Thermally induced cracks form and propagate independently from surface grooves and the surface quality. - Abstract: Beryllium will be applied as first wall armor material in ITER. The armor has to sustain high steady state and transient power fluxes. For transient events like edge localized modes, these transient power fluxes rise up to 1.0 GW m{sup −2} with a duration of 0.5–0.75 ms in the divertor region and a significant fraction of this power flux is deposited on the first wall as well. In the present work, the reference beryllium grade for the ITER first wall application S-65 was prepared with various surface conditions and subjected to transient power fluxes (thermal shocks) with ITER relevant loading parameters. After 1000 thermal shocks, a crucial destruction of the entire loaded area was observed and linked to the stress accelerated grain boundary oxidation (SAGBO)/dynamic embrittlement (DE) effect. Furthermore, the study revealed that the majority of the thermally induced cracks formed between 1 and 10 pulses and then grew wider and deeper with increasing pulse number. The surface quality did not influence the cracking behavior of beryllium in any detectable way. However, the polished surface demonstrated the highest resistance against the observed crucial destruction mechanism.

  10. ITER licensing

    International Nuclear Information System (INIS)

    Gordon, C.W.

    2005-01-01

    ITER was fortunate to have four countries interested in ITER siting to the point where licensing discussions were initiated. This experience uncovered the challenges of licensing a first of a kind, fusion machine under different licensing regimes and helped prepare the way for the site specific licensing process. These initial steps in licensing ITER have allowed for refining the safety case and provide confidence that the design and safety approach will be licensable. With site-specific licensing underway, the necessary regulatory submissions have been defined and are well on the way to being completed. Of course, there is still work to be done and details to be sorted out. However, the informal international discussions to bring both the proponent and regulatory authority up to a common level of understanding have laid the foundation for a licensing process that should proceed smoothly. This paper provides observations from the perspective of the International Team. (author)

  11. Response of beryllium to severe thermal shocks -simulation of disruption and vertical displacement events in future thermonuclear devices

    Energy Technology Data Exchange (ETDEWEB)

    Linke, J.; Duwe, R.; Roedig, M.; Schuster, A. [Association Euratom-Forschungszentrum Juelich GmbH (Germany); Merola, M.; Qian, R.H.

    1998-01-01

    Beryllium will play an important role for plasma facing components in next step thermonuclear fusion devices such as ITER. In particular for the first wall beryllium will be used with an armor thickness of several millimeters. However, during plasma instabilities they will experience severe thermal shocks. Here plasma disruptions with deposited energy densities of several ten MJm{sup -2} are the most essential damaging mechanism. However, a signifant fraction of the incident energy will be absorbed by a dense cloud of ablation vapor, hence reducing the effective energy density at the beryllium surface to values in the order of 10 MJm{sup -2}. To investigate the material response to all these plasma instabilities thermal shock tests on small scale test coupons (disruption effects) and on actively cooled divertor modules (VDEs) have been performed in the electron beam test facility JUDITH at ITER relevant surface heat loads. These tests have been performed on different bulk beryllium grades and on plasma sprayed coatings; the influence of pulse duration, power density, and temperature effects has been investigated experimentally. Detailed in-situ diagnostics (for beam characterization, optical pyrometry etc.) and post mortem analyses (profilometry, metallography, optical and electron microscopy) have been applied to quantify the resulting material damage. 1D- and 2D models have developed to verify the experimental results obtained in the electron beam simulation experiments. (J.P.N.)

  12. Proceedings of the sixth IEA international workshop on beryllium technology for fusion

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hiroshi [Japan Atomic Energy Research Inst., Kashiwa, Chiba (Japan); Tanaka, Satoru [Tokyo Univ., Tokyo (Japan); Ishitsuka, Etsuo [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

    2004-03-01

    This report is the Proceedings of the Sixth International Energy Agency International Workshop on Beryllium Technology for Fusion. The workshop was held on December 2-5, 2003, at SEAGAIA in Miyazaki City, Japan with 69 participants who attended from Europe, the Russian Federation, Kazakhstan, Ukraine, China, the United States and Japan. The topics for papers were arranged into nine sessions; Status of beryllium study, Plasma and tritium interactions, ITER oriented issues, Neutron irradiation effects, Beryllide application, Disposal and recycling, Molten salt, Health and safety issues and Panel discussion. In the Panel discussion, the international collaboration for three topics, i.e., Neutron irradiation effects, Beryllide application, Recycling and Disposal, were discussed, and necessary items for the international collaboration were proposed. The 46 of the presented papers are indexed individually. (J.P.N.)

  13. Tritium retention in S-65 beryllium after 100 eV plasma exposure

    Energy Technology Data Exchange (ETDEWEB)

    Causey, R.A. [Sandia National Labs., Livermore, CA (United States); Longhurst, G.R. [Idaho National Engineering Laboratories, Idaho Falls, 83415 (United States); Harbin, W. [Los Alamos National Laboratories, Los Alamos, NM 87545 (United States)

    1997-02-01

    The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 x 10{sup 17} (D+T)/cm{sup 2} s up to about 3 x 10{sup 18} (D+T)/cm{sup 2} s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H{sub 2} gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 x 10{sup 17} (D+T)/cm{sup 2} at 373 K to a low of 1 x 10{sup 16} (D+T)/cm{sup 2} at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C=0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters. (orig.).

  14. Tritium retention in S-65 beryllium after 100 eV plasma exposure

    Science.gov (United States)

    Causey, Rion A.; Longhurst, Glen R.; Harbin, Wally

    1997-02-01

    The tritium plasma experiment (TPE) has been used to measure the retention of tritium in S-65 beryllium under conditions similar to that expected for the international thermonuclear experimental reactor (ITER). Beryllium samples 2 mm thick and 50 mm in diameter were exposed to a plasma of tritium and deuterium. The particle flux striking the samples was varied from approximately 1 × 10 17 ( D + T)/ cm2s up to about 3 × 10 18 ( D + T)/ cm2s. The beryllium samples were negatively biased to elevate the energy of the impinging ions to 100 eV. The temperature of the samples was varied from 373 K to 973 K. Exposure times of 1 h were used. Subsequent to the plasma exposure, the samples were outgassed in a separate system where 99% He and 1% H 2 gas was swept over the samples during heating. The sweep gas along with the released tritium was sent through an ionization chamber, through a copper oxide catalyst bed, and into a series of glycol bubblers. The amount of released tritium was determined both by the ionization chamber and by liquid scintillation counting of the glycol. Tritium retention in the beryllium disks varied from a high of 2.4 × 10 17 ( D + T)/ cm2 at 373 K to a low of 1 × 10 16 ( D + T)/ cm2 at 573 K. For almost every case, the tritium retention in the beryllium was less than that calculated using the C = 0 boundary condition at the plasma facing surface. It is believed that this lower than expected retention is due to rapid release of tritium from the large specific surface area created in the implant zone due to the production of voids, bubbles, and blisters.

  15. Beryllium. Health hazards and their control. Pt. 2

    International Nuclear Information System (INIS)

    Lires, O.A.; Delfino, C.A.; Botbol, J.

    1991-01-01

    In this work (continuation of 'Beryllium' series) health hazards, toxic effects, limits of permissible atmospheric contamination and safe exposure to beryllium are described. Guidelines to the design, control operations and hygienic precautions of the working facilities are given. (Author) [es

  16. 75 FR 80734 - Chronic Beryllium Disease Prevention Program

    Science.gov (United States)

    2010-12-23

    ... are used in nuclear weapons as nuclear reactor moderators or reflectors and as nuclear reactor fuel...), grinding, and machine tooling of parts. Inhalation of beryllium particles may cause chronic beryllium...

  17. Mechanical performance of irradiated beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Scaffidi-Argentina, F.; Dalle-Donne, M.; Werle, H. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1998-01-01

    For the Helium Cooled Pebble Bed (HCPB) Blanket, which is one of the two reference concepts studied within the European Fusion Technology Programme, the neutron multiplier consists of a mixed bed of about 2 and 0.1-0.2 mm diameter beryllium pebbles. Beryllium has no structural function in the blanket, however microstructural and mechanical properties are important, as they might influence the material behavior under neutron irradiation. The EXOTIC-7 as well as the `Beryllium` experiments carried out in the HFR reactor in Petten are considered as the most detailed and significant tests for investigating it. This paper reviews the present status of beryllium post-irradiation examinations performed at the Forschungszentrum Karlsruhe with samples from these irradiation experiments, emphasizing the effects of irradiation of essential material properties and trying to elucidate the processes controlling the property changes. The microstructure, the porosity distribution, the impurity content, the behavior under compression loads and the compatibility of the beryllium pebbles with lithium orthosilicate (Li{sub 4}SiO{sub 4}) during the in-pile irradiation are presented and critically discussed. Qualitative information on ductility and creep obtained by hardness-type measurements are also supplied. (author)

  18. Experiments on tritium behavior in beryllium, (1)

    International Nuclear Information System (INIS)

    Kawamura, Hiroshi; Ishizuka, Etsuo; Matsumoto, Mikio; Inada, Seiji; Sezaki, Katsuji; Saito, Minoru; Kato, Mineo.

    1989-06-01

    In JMTR, it was observed that the tritium concentration of the primary coolant increases with the reactor operation at 50 MW. As one of the tritium generation sources, we paid attention to a neutron reflector made of beryllium because the tritium generation rate in the beryllium is bigger than other components in the reactor core. On the other hand, the irradiation test of blanket materials (i.e. tritium breeding materials and neutron multipling materials) are planned for development of the fusion reactor in JMTR and the beryllium will be also irradiated as a neutron multiplier with tritium breeding materials. Therefore, as the irradiated specimens, we used a hot-pressed beryllium disk fabricated by the same method as the neutron reflector or the neutron multiplier and conducted the irradiation tests in JMTR. The purpose of these tests are to clarify the tritium behavior in the hot-pressed beryllium. In this paper, from a viewpoint of the fabrication of capsules for neutron irradiation, the specifications of the irradiated specimens and capsules are summarized. Additionally, the results on the puncture test of the container of the irradiation specimens are described. (author)

  19. Study of wall conditioning in tokamaks with application to ITER

    International Nuclear Information System (INIS)

    Kogut, Dmitri

    2014-01-01

    Thesis is devoted to studies of performance and efficiency of wall conditioning techniques in fusion reactors, such as ITER. Conditioning is necessary to control the state of the surface of plasma facing components to ensure plasma initiation and performance. Conditioning and operation of the JET tokamak with ITER-relevant material mix is extensively studied. A 2D model of glow conditioning discharges is developed and validated; it predicts reasonably uniform discharges in ITER. In the nuclear phase of ITER operation conditioning will be needed to control tritium inventory. It is shown here that isotopic exchange is an efficient mean to eliminate tritium from the walls by replacing it with deuterium. Extrapolations for tritium removal are comparable with expected retention per a nominal plasma pulse in ITER. A 1D model of hydrogen isotopic exchange in beryllium is developed and validated. It shows that fluence and temperature of the surface influence efficiency of the isotopic exchange. (author) [fr

  20. Comparison of plastic, high density carbon, and beryllium as indirect drive NIF ablators

    Science.gov (United States)

    Kritcher, A. L.; Clark, D.; Haan, S.; Yi, S. A.; Zylstra, A. B.; Callahan, D. A.; Hinkel, D. E.; Berzak Hopkins, L. F.; Hurricane, O. A.; Landen, O. L.; MacLaren, S. A.; Meezan, N. B.; Patel, P. K.; Ralph, J.; Thomas, C. A.; Town, R.; Edwards, M. J.

    2018-05-01

    Detailed radiation hydrodynamic simulations calibrated to experimental data have been used to compare the relative strengths and weaknesses of three candidate indirect drive ablator materials now tested at the NIF: plastic, high density carbon or diamond, and beryllium. We apply a common simulation methodology to several currently fielded ablator platforms to benchmark the model and extrapolate designs to the full NIF envelope to compare on a more equal footing. This paper focuses on modeling of the hohlraum energetics which accurately reproduced measured changes in symmetry when changes to the hohlraum environment were made within a given platform. Calculations suggest that all three ablator materials can achieve a symmetric implosion at a capsule outer radius of ˜1100 μm, a laser energy of 1.8 MJ, and a DT ice mass of 185 μg. However, there is more uncertainty in the symmetry predictions for the plastic and beryllium designs. Scaled diamond designs had the most calculated margin for achieving symmetry and the highest fuel absorbed energy at the same scale compared to plastic or beryllium. A comparison of the relative hydrodynamic stability was made using ultra-high resolution capsule simulations and the two dimensional radiation fluxes described in this work [Clark et al., Phys. Plasmas 25, 032703 (2018)]. These simulations, which include low and high mode perturbations, suggest that diamond is currently the most promising for achieving higher yields in the near future followed by plastic, and more data are required to understand beryllium.

  1. Effect of machining damage on tensile properties of beryllium

    International Nuclear Information System (INIS)

    Hanafee, J.E.

    1976-01-01

    It is well established that damage introduced at the surface of beryllium during machining operations can lower its mechanical properties. Tensile tests were conducted to illustrate this on beryllium presently being used for parts in the W79 program and similar to the new powder-processed beryllium specified for production (tentative specification MEL 76-001319). The objective of this study is to quantitatively illuminate the importance of controlling machining damage in this particular grade of powder-processed beryllium

  2. New concurrent iterative methods with monotonic convergence

    Energy Technology Data Exchange (ETDEWEB)

    Yao, Qingchuan [Michigan State Univ., East Lansing, MI (United States)

    1996-12-31

    This paper proposes the new concurrent iterative methods without using any derivatives for finding all zeros of polynomials simultaneously. The new methods are of monotonic convergence for both simple and multiple real-zeros of polynomials and are quadratically convergent. The corresponding accelerated concurrent iterative methods are obtained too. The new methods are good candidates for the application in solving symmetric eigenproblems.

  3. Selection, development and characterisation of plasma facing materials for ITER

    International Nuclear Information System (INIS)

    Barabash, V.; Akiba, M.; Ulrickson, M.; Vieider, G.

    1996-01-01

    The current status of the selection of the armour materials for first wall, limiters and divertor are presented. The candidate armour materials are beryllium, tungsten and carbon base materials (mainly carbon fiber composites). The selection of the references grades from these material classes is discussed and the candidate grades are described. The main reasons for the selection of the reference grades are also discussed. The urgent materials R and D needs for the development of the design are described briefly. (orig.)

  4. Occupational and non-occupational allergic contact dermatitis from beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Vilaplana, J; Romaguera, C; Grimalt, F [Allergy Department, Dermatological Service Hospital Clinico, Barcelona (Spain)

    1992-01-01

    There are various references to sensitization to beryllium in the literature. Since introducing a patch testing series for patients with suspected sensitization to metals, we have found 3 cases of sensitization to beryllium. Of these 3 cases, we regard the first 2 as having relevant sensitization. Beryllium chloride (1% pet.) was positive in 3 patients and negative in 150 controls. (au).

  5. Occupational and non-occupational allergic contact dermatitis from beryllium

    International Nuclear Information System (INIS)

    Vilaplana, J.; Romaguera, C.; Grimalt, F.

    1992-01-01

    There are various references to sensitization to beryllium in the literature. Since introducing a patch testing series for patients with suspected sensitization to metals, we have found 3 cases of sensitization to beryllium. Of these 3 cases, we regard the first 2 as having relevant sensitization. Beryllium chloride (1% pet.) was positive in 3 patients and negative in 150 controls. (au)

  6. Beryllium phonon spectrum from cold neutron measurements

    International Nuclear Information System (INIS)

    Bulat, I.A.

    1979-01-01

    The inelastic coherent scattering of neutrons with the initial energy E 0 =4.65 MeV on the spectrometer according to the time of flight is studied in polycrystalline beryllium. The measurements are made for the scattering angles THETA=15, 30, 45, 60, 75 and 90 deg at 293 K. The phonon spectrum of beryllium, i-e. g(w) is reestablished from the experimental data. The data obtained are compared with the data of model calculations. It is pointed out that the phonon spectrum of beryllium has a bit excessive state density in the energy range from 10 to 30 MeV. It is caused by the insufficient statistical accuracy of the experiment at low energy transfer

  7. Spectrographic determination of impurities in beryllium oxide

    International Nuclear Information System (INIS)

    Paula Reino, L.C. de; Lordello, A.R.; Pereira, A.S.A.

    1986-03-01

    A method for the spectrographic determination of Al, B, Cd, Co, Cu, Cr, Fe, Mg, NaNi, Si and Zn in nuclear grade beryllium oxide has been developed. The determination of Co, Al, Na and Zn is besed upon a carrier distillation technique. Better results were obtained with 2% Ga 2 O 3 as carrier in beryllium oxide. For the elements B, Cd, Cu, Fe, Cr, Mg, Ni and Si the sample is loaded in a Scribner-Mullin shallow cup electrode, covered with graphite powder and excited in DC arc. The relative standard deviation values for different elements are in the range of 10 to 20%. The method fulfills requirements of precision and sensitivity for specification analysis of nuclear grade beryllium oxide.(Author) [pt

  8. Beryllium research on FFHR molten salt blanket

    International Nuclear Information System (INIS)

    Terai, T.; Tanaka, S.; Sze, D.-K.

    2000-01-01

    Force-free helical reactor, FFHR, is a demo-relevant heliotron-type D-T fusion reactor based on the great amount of R and D results obtained in the LHD project. Since 1993, collaboration works have made great progress in design studies of FFHR with standing on the major advantage of current-less steady operation with no dangerous plasma disruptions. There are two types of reference designs, FFHR-1 and FFHR-2, where molten Flibe (LiF-BeF2) is utilized as tritium breeder and coolant. In this paper, we present the outline of FFHR blanket design and some related R and D topics focusing on Be utilization. Beryllium is used as a neutron multiplier in the design and Be pebbles are placed in the front part of the tritium breeding zone. In a Flibe blanket, HF (TF) generated due to nuclear transmutation will be a problem because of its corrosive property. Though nickel-based alloys are thought to be intact in such a corrosive environment, FFHR blanket design does not adopt the alloys because of their induced radioactivity. The present candidate materials for the structure are low-activated ferritic steel (JLF-1), V-4Cr-4Ti, etc. They are capable to be corroded by HF in the operation condition, and Be is expected to work as a reducing agent in the system as well. Whether Be pebbles placed in a Flibe flow can work well or not is a very important matter. From this point, Be solubility in Flibe, reaction rate of the Redox reaction with TF in the liquid and on the surface of Be pebbles under irradiation, flowing behavior of Flibe through a Be pebble bed, etc. should be investigated. In 1997, in order to establish more practical and new data bases for advanced design works, we started a collaboration work of R and D on blanket engineering, where the Be research above mentioned is included. Preliminary dipping-test of Be sheets and in-situ tritium release experiment from Flibe with Be sheets have got started. (orig.)

  9. Method of beryllium implantation in germanium substrate

    International Nuclear Information System (INIS)

    Kagawa, S.; Baba, Y.; Kaneda, T.; Shirai, T.

    1983-01-01

    A semiconductor device is disclosed, as well as a method for manufacturing it in which ions of beryllium are implanted into a germanium substrate to form a layer containing p-type impurity material. There after the substrate is heated at a temperature in the range of 400 0 C. to 700 0 C. to diffuse the beryllium ions into the substrate so that the concentration of beryllium at the surface of the impurity layer is in the order of 10 17 cm- 3 or more. In one embodiment, a p-type channel stopper is formed locally in a p-type germanium substrate and an n-type active layer is formed in a region surrounded by, and isolated from, the channel stopper region. In another embodiment, a relatively shallow p-type active layer is formed at one part of an n-type germanium substrate and p-type guard ring regions are formed surrounding, and partly overlapping said p-type active layer. In a further embodiment, a p-type island region is formed at one part of an n-type germanium substrate, and an n-type region is formed within said p-type region. In these embodiments, the p-type channel stopper region, p-type guard ring regions and the p-type island region are all formed by implanting ions of beryllium into the germanium substrate

  10. ESR investigations of gamma irradiated beryllium ceramics

    International Nuclear Information System (INIS)

    Ryabikin, Yu.A.; Polyakov, A.I.; Petukhov, Yu.V.; Bitenbaev, M.I.; Zashkvara, O.V.

    2000-01-01

    In this report the result of ESR- investigation of kinetics of radiation paramagnetic defects accumulated in beryllium ceramics under gamma irradiation are presented. The data on quantum yield and destruction rate constants of these defects under ionizing irradiation are obtained. (orig.)

  11. ESR investigations of gamma irradiated beryllium ceramics

    Energy Technology Data Exchange (ETDEWEB)

    Ryabikin, Yu A; Polyakov, A I; Petukhov, Yu V; Bitenbaev, M I; Zashkvara, O V [Physical-Technical Inst., Almaty (Kazakhstan)

    2000-04-01

    In this report the result of ESR- investigation of kinetics of radiation paramagnetic defects accumulated in beryllium ceramics under gamma irradiation are presented. The data on quantum yield and destruction rate constants of these defects under ionizing irradiation are obtained. (orig.)

  12. Beryllium, a material of great promise

    International Nuclear Information System (INIS)

    Le Fauconnier, J.P.; Nomine, A.M.

    1992-01-01

    Beryllium is one of the lightest metals. It also owns an outstanding combination of physical, mechanical and nuclear properties which gives it a favorable position, compared to more usual materials, in various fields of applications. Constant technological advancements in the elaboration and working up have induced a significant improvement of its ductility and a reduction of the production costs. (Author). 12 refs., 7 figs

  13. Thermophysical properties of solid and liquid beryllium

    International Nuclear Information System (INIS)

    Boivineau, M.; Arles, L.; Vermeulen, J.M.; Thevenin, Th.

    1993-01-01

    A submillisecond resistive heating technique under high pressure (0.12 GPa) has been used to measure selected thermophysical properties of both solid and liquid beryllium. Data have been obtained between room temperature and 2900 K. Results on enthalpy, volume expansion, electrical resistivity, and sound velocity measurements are presented

  14. Critical parameters controlling irradiation swelling in beryllium

    International Nuclear Information System (INIS)

    Dubinko, V.I.

    1995-01-01

    Radiation effects in beryllium can hardly be explained within a framework of the conventional theory based on the bias concept due to elastic interaction difference (EID) between vacancies and self-interstitial atoms (SIAs) since beryllium belongs to hexagonal close-packed metals where diffusion has been shown to be anisotropic. Diffusional anisotropy difference (DAD) between point defects changes the cavity bias for their absorption and leads to dependence of the dislocation bias on the distribution of dislocations over crystallographic directions. On the other hand, the elastic interaction between point defects and cavities gives rise to the size and gas pressure dependencies of the cavity bias, resulting in new critical quantities for bubble-void transition effects at low temperature irradiation. In the present paper, we develop the concept of the critical parameters controlling irradiation swelling with account of both DAD and EID, and take care of thermal effects as well since they are of major importance for beryllium which has an anomalously low self-diffusion activation energy. Experimental data on beryllium swelling are analyzed on the basis of the present theory. (orig.)

  15. Potential exposures and risks from beryllium-containing products.

    Science.gov (United States)

    Willis, Henry H; Florig, H Keith

    2002-10-01

    Beryllium is the strongest of the lightweight metals. Used primarily in military applications prior to the end of the Cold War, beryllium is finding new applications in many commercial products, including computers, telecommunication equipment, and consumer and automotive electronics. The use of beryllium in nondefense consumer applications is of concern because beryllium is toxic. Inhalation of beryllium dust or vapor causes a chronic lung disease in some individuals at concentrations as low as 0.01 microg/m3 in air. As beryllium enters wider commerce, it is prudent to ask what risks this might present to the general public and to workers downstream of the beryllium materials industry. We address this question by evaluating the potential for beryllium exposure from the manufacturing, use, recycle, and disposal of beryllium-containing products. Combining a market study with a qualitative exposure analysis, we determine which beryllium applications and life cycle phases have the largest exposure potential. Our analysis suggests that use and maintenance of the most common types of beryllium-containing products do not result in any obvious exposures of concern, and that maintenance activities result in greater exposures than product use. Product disposal has potential to present significant individual risks, but uncertainties concerning current and future routes of product disposal make it difficult to be definitive. Overall, additional exposure and dose-response data are needed to evaluate both the health significance of many exposure scenarios, and the adequacy of existing regulations to protect workers and the public. Although public exposures to beryllium and public awareness and concern regarding beryllium risks are currently low, beryllium risks have psychometric qualities that may lead to rapidly heightened public concern.

  16. Experimental investigation of the energy and temperature dependence of beryllium self sputtering

    International Nuclear Information System (INIS)

    Korshunov, S.N.; Guseva, M.I.; Stolijarova, V.G.

    1995-01-01

    The low-Z metal beryllium is considered as plasma facing material (PFM) for the ITER. It is expected that operation temperature range of beryllium PFM will be (670 - 1070) K. While experimental Be-sputtering data bases exist for H + , D + and He + -ions, the self-sputtering yields of Be have only been estimated by computer simulation. In this paper we report the experimental results on the energy and temperature dependence of the beryllium self-sputtering yield (S). The energy dependence of S s in the energy range (0.5 - 10.0) keV was measured at 670 K. The self-sputtering yield of Be attains its maximal value at the ion energy of 1.5 keV, being equal to 0.32 ± at./ion. Comparison of the experimental results and theoretical prediction shows a good agreement for energy dependence of S s . The temperature dependence of S s in the temperature range (370-1070)K was obtained for 0.9keV Be + -ions. The value of S s is not changed up to 870 K. It sharply increases at the temperatures above 870 attaining the value of 0.75 at./ion at 1070 K

  17. Optimization of the JET Beryllium tile profile for power handling

    International Nuclear Information System (INIS)

    Nunes, I.; Vries, P. de; Lomas, P.J.; Loarte, A.

    2006-01-01

    The primary objective of the ITER-like wall project is to install a beryllium main wall and a tungsten divertor. From the point of view of plasma operations, the power handling properties of the new Be tiles may affect the operational space. The tiles design has to be such that it allows routine plasma operation for ITER relevant scenarios, i.e., 3-5 MA ELMy H-modes with high power input (P in > 30 MW) for lengths of time of ∼ 10 s. Due to the constrains imposed by heat conductivity, eddy current and stress torques on a Be tile, a single Be tile must be an assembly of castellated slices [Thompson V. et al, this conference]. From the point of view of plasma operations, the power handling properties of the new Be tiles can restrict the operational space of JET, if considerable melting of the tiles is to be avoided. This paper describes the power handling studies for the beryllium wall tiles and the optimisation of their design to achieve the operation goal described above. The melting temperature for Be is 1289 o C, corresponding to a energy limit of 60 MJ/m 2 for 10 s [Thompson V. et al, this conference]. For low field line angles, the power density on the toroidally facing surfaces is several times higher than the power density on the tile face requiring these to be shadowed. Furthermore the poloidally facing surfaces also have to be shadowed from assembly to assembly due to the large gap between assemblies. The tiles have been designed taking into account these limits and with a geometrical design such as to avoid exposed surfaces at high angles to the magnetic field being melted due to the expected loads. This has been achieved after detailed studies of the power handling of the various limiters and protections, including the effect of the curvature of the flux surfaces, shadowing and tolerance to misalignment. The surface of the tiles is defined such that, when possible, there is an even distribution of power density over the entire tile surface, and that

  18. Thermo-mechanical screening tests to qualify beryllium pebble beds with non-spherical pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Reimann, Joerg, E-mail: joerg.reimann@partner.kit.edu [IKET, Karlsruhe Institute of Technology, Karlsruhe (Germany); Fretz, Benjamin [KBHF GmbH, Eggenstein-Leopoldshafen (Germany); Pupeschi, Simone [IAM, Karlsruhe Institute of Technology, Karlsruhe (Germany)

    2015-10-15

    Highlights: • In present ceramic breeder blankets, pebble-shaped beryllium is used as a neutron multiplier. • Spherical pebbles are considered as the candidate material, however, non-spherical particles are of economic interest. • Thermo-mechanical pebble bed data do merely exist for non-spherical beryllium grades. • Uniaxial compression tests (UCTs), combined with the Hot Wire Technique (HWT) were used to measure the stress–strain relations and the thermal conductivity. • A small experimental set-up had to be used and a detailed 3D modelling was of prime importance. • Compared to spherical pebble beds, non-spherical pebble beds are generally softer and mainly the thermal conductivity is lower. - Abstract: In present ceramic breeder blankets, pebble-shaped beryllium is used as a neutron multiplier. Fairly spherical pebbles are considered as a candidate material, however, non-spherical particles are of economic interest because production costs are much lower. Yet, thermo-mechanical pebble bed data do merely exist for these beryllium grades, and the blanket relevant potential of these grades cannot be judged. Screening experiments were performed with three different grades of non-spherical beryllium pebbles, produced by different companies, accompanied by experiments with the reference beryllium pebble beds. Uniaxial compression tests (UCTs), combined with the Hot Wire Technique (HWT), were performed to measure both the stress–strain relation and the thermal conductivity, k, at different stress levels. Because of the limited amounts of the non-spherical materials, the experimental set-ups were small and a detailed 3D modelling was of prime importance in order to prove that the used design was appropriate. Compared to the pebble beds consisting of spherical pebbles, non-spherical pebble beds are generally softer (smaller stress for a given strain), and, mainly as a consequence of this, for a given strain value, the thermal conductivity is lower. This

  19. Control of beryllium powder at a DOE facility

    International Nuclear Information System (INIS)

    Langner, G.C.; Creek, K.L.; Castro, R.G.

    1997-01-01

    Beryllium is contained in a number of domestic and national defense items. Although many items might contain beryllium in some manner, few people need worry about the adverse effects caused by exposure to beryllium because it is the inhalable form of beryllium that is most toxic. Chronic beryllium disease (CBD), a granulomas and fibrotic lung disease with long latency, can be developed after inhalation exposures to beryllium. It is a progressive, debilitating lung disease. Its occurrence in those exposed to beryllium has been difficult to predict because some people seem to react to low concentration exposures whereas others do not react to high concentration exposures. Onset of the disease frequently occurs between 15 to 20 years after exposure begins. Some people develop the disease after many years of low concentration exposures but others do not develop CBD even though beryllium is shown to be present in lungs and urine. Conclusions based on these experiences are that their is some immunological dependence of developing CBD in about 3--4% of the exposed population, but the exact mechanism involved has not yet been identified. Acute beryllium disease can occur after a single exposure to a concentration of greater than 0.100 mg/m3 (inhalation exposure); it is characterized by the development of chemical pneumoconiosis, a respiratory disease. The acute effect of skin contact is a dermatitis characterized by itching and reddened, elevated, or fluid-accumulated lesions which appear particularly on the exposed surfaces of the body, especially the face, neck, arms, and hands. Small particles of beryllium that enter breaks in the skin can lead to the development of granulomas and/or open sores that do not heal until the beryllium has been removed. Our interest is only airborne beryllium, which is found in areas that machine or produce beryllium

  20. ITER council proceedings: 2001

    International Nuclear Information System (INIS)

    2001-01-01

    Continuing the ITER EDA, two further ITER Council Meetings were held since the publication of ITER EDA documentation series no, 20, namely the ITER Council Meeting on 27-28 February 2001 in Toronto, and the ITER Council Meeting on 18-19 July in Vienna. That Meeting was the last one during the ITER EDA. This volume contains records of these Meetings, including: Records of decisions; List of attendees; ITER EDA status report; ITER EDA technical activities report; MAC report and advice; Final report of ITER EDA; and Press release

  1. Toxicokinetics of beryllium following inhalation of beryllium oxide by Beagle dogs. III

    International Nuclear Information System (INIS)

    Finch, G.L.; Haley, P.J.; Hoover, M.D.; Mewhinney, J.A.; Bice, D.E.; Eidson, A.F.

    1988-01-01

    Young adult Beagle dogs inhaled radiolabeled beryllium oxide aerosols ( 7 BeO) prepared at either 500 deg. or 1000 deg. C to achieve one of two initial lung burdens (ILBs) of BeO. After exposure, animals were monitored by whole body counting for 7 Be, and excreta, clinical, and radiographic data were collected. One group of dogs was assigned for serial sacrifice for quantitation of beryllium clearance from lung, translocation to other organs, and histopathologic analysis of lung and lymph nodes. A second group of dogs was subjected to periodic bronchopulmonary lavage for analysis of lymphocyte responsiveness to beryllium. These latter dogs were subsequently re-exposed to the high ILB level of BeO prepared t 500 deg. C. ILBs following the second exposure were higher than that after the first exposure (74 vs. 42 μg BeO/kg, respectively). Except for one dog that exhibited enhanced beryllium retention after the second exposure, patterns of whole body clearance were similar to those observed after the initial exposures to the 500 deg. C-BeO. Analysis of lymphocyte responsiveness to beryllium in the second group of dogs is continuing. (author)

  2. Toxicokinetics of beryllium following inhalation of beryllium oxide by Beagle dogs. III

    Energy Technology Data Exchange (ETDEWEB)

    Finch, G L; Haley, P J; Hoover, M D; Mewhinney, J A; Bice, D E; Eidson, A F

    1988-12-01

    Young adult Beagle dogs inhaled radiolabeled beryllium oxide aerosols ({sup 7}BeO) prepared at either 500 deg. or 1000 deg. C to achieve one of two initial lung burdens (ILBs) of BeO. After exposure, animals were monitored by whole body counting for {sup 7}Be, and excreta, clinical, and radiographic data were collected. One group of dogs was assigned for serial sacrifice for quantitation of beryllium clearance from lung, translocation to other organs, and histopathologic analysis of lung and lymph nodes. A second group of dogs was subjected to periodic bronchopulmonary lavage for analysis of lymphocyte responsiveness to beryllium. These latter dogs were subsequently re-exposed to the high ILB level of BeO prepared t 500 deg. C. ILBs following the second exposure were higher than that after the first exposure (74 vs. 42 {mu}g BeO/kg, respectively). Except for one dog that exhibited enhanced beryllium retention after the second exposure, patterns of whole body clearance were similar to those observed after the initial exposures to the 500 deg. C-BeO. Analysis of lymphocyte responsiveness to beryllium in the second group of dogs is continuing. (author)

  3. Status of material development for lifetime expansion of beryllium reflector

    Energy Technology Data Exchange (ETDEWEB)

    Dorn, C [Materion Brush Beryllium and Composites, California (United States); Tsuchiya, Kunihiko; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Hatano, Y [Univ. of Toyama, Toyama (Japan); Chakrov, P [INP-KNNC, Almaty (Kazakhstan); Kodama, M [Nippon Nuclear Fuel Development Co., Ltd., Oarai, Ibaraki (Japan)

    2012-03-15

    Beryllium has been used as the reflector element material in the reactor, specifically S-200F structural grade beryllium manufactured by Materion Brush Beryllium and Composites (former, Brush Wellman Inc.). As a part of the reactor upgrade, the Japan Atomic Energy Agency (JAEA) also has carried out the cooperation experiments to extend the operating lifetime of the beryllium reflector elements. It will first be necessary to determine which of the material's physical, mechanical and chemical properties will be the most influential on that choice. The irradiation testing plans to evaluate the various beryllium grades are also briefly considered and prepared. In this paper, material selection, irradiation test plan and PEI development for lifetime expansion of beryllium are described for material testing reactors. (author)

  4. Mechanisms of hydrogen retention in metallic beryllium and beryllium oxide and properties of ion-induced beryllium nitride; Rueckhaltemechanismen fuer Wasserstoff in metallischem Beryllium und Berylliumoxid sowie Eigenschaften von ioneninduziertem Berylliumnitrid

    Energy Technology Data Exchange (ETDEWEB)

    Oberkofler, Martin

    2011-09-22

    In the framework of this thesis laboratory experiments on atomically clean beryllium surfaces were performed. They aim at a basic understanding of the mechanisms occurring upon interaction of a fusion plasma with a beryllium first wall. The retention and the temperature dependent release of implanted deuterium ions are investigated. An atomistic description is developed through simulations and through the comparison with calculations based on density functional theory. The results of these investigations are compared to the behaviour of hydrogen upon implantation into thermally grown beryllium oxide layers. Furthermore, beryllium nitride is produced by implantation of nitrogen into metallic beryllium and its properties are investigated. The results are interpreted with regard to the use of beryllium in a fusion reactor. (orig.)

  5. The material system beryllium-nitrogen-deuterium: review of recent results

    Energy Technology Data Exchange (ETDEWEB)

    Dittmar, Timo; Koeppen, M.; Kreter, A.; Linsmeier, C. [Institut fuer Energie- und Klimaforschung - Plasmaphysik, Forschungszentrum Juelich (Germany); Doerner, R.P. [Center for Energy Research, UC San Diego, CA (United States); Oberkofler, M.; Schwarz-Selinger, T. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2014-07-01

    Beryllium has been chosen as first-wall amour material in JET-ILW and ITER. Due to the usage of N seeding in current tokamaks and a possible use of N in ITER, the properties of the Be-N-D material system became of greater interest: Ion beam experiments showed the formation of thermally stable Be{sub 3}N{sub 2}. Beryllium samples exposed to D-plasma in the linear plasma device PISCES-B showed a significant reduction in Be erosion rate with N seeding. Measured N/Be and D/Be ratios in codeposits were independent of the N seeding rate (0.8% - 10% N seeding). Gas balance measurements for N seeded L-mode discharges in JET indicate strong N retention and show a potential increase in D retention, possibly due to ND{sub 3} formation. This paper will re-visit and evaluate the available data from the different experiments in order to build a consistent model of the processes relevant for plasma-material interactions in the Be-N-D material system.

  6. Measuring device for bending of beryllium reflector

    International Nuclear Information System (INIS)

    Nishida, Seiri; Sakamoto, Naoki.

    1994-01-01

    The device of the present invention can measure bending of a beryllium reflector formed in a reactor core of a nuclear reactor by a relatively easy operation. Namely, a sensor portion comprises a long-support that can be inserted to a fuel element-insertion hole disposed in the reactor and a plurality of distance sensors disposed in a longitudinal direction of the support. A supersonic wave sensor which is advantageous in the heat resistance, the size and the accuracy and can conduct measurement in water relatively easily is used as the distance sensors. However, other sensors, instead of the sensor described above, may also be used. The plurality of distance sensors detect the bending amount of the beryllium reflector in the longitudinal direction by such an easy operation of inserting such a sensor portion to the fuel element-insertion hole upon exchange of fuel elements. (I.S.)

  7. Interaction of hydrogen and its isotopes with irradiated beryllium

    International Nuclear Information System (INIS)

    Tazhibaeva, I.L.; Shestakov, V.P.; Klepikov, A.Kh.; Pomanenko, O.G.; Chikhraj, E.V.; Kenzhin, E.A.; Zverev, V.V.; Kolbanenkov, A.N.

    2000-01-01

    In the article the results of experiments on hydrogen and its isotopes accumulation and gas-release from irradiated beryllium are presented. The irradiation was conducted at different media and temperatures in the RA and IVG.1M reactors. The measurements were carried out by thermal desorption method. Hydrogen release from beryllium samples saturated at different conditions were calculated. Dependence of hydrogen confinement character in beryllium from grain orientation in the sample, temperature and irradiation rate was revealed

  8. Premelting hcp to bcc Transition in Beryllium

    Science.gov (United States)

    Lu, Y.; Sun, T.; Zhang, Ping; Zhang, P.; Zhang, D.-B.; Wentzcovitch, R. M.

    2017-04-01

    Beryllium (Be) is an important material with wide applications ranging from aerospace components to x-ray equipment. Yet a precise understanding of its phase diagram remains elusive. We have investigated the phase stability of Be using a recently developed hybrid free energy computation method that accounts for anharmonic effects by invoking phonon quasiparticles. We find that the hcp → bcc transition occurs near the melting curve at 0 materials.

  9. Analysis of surface contaminants on beryllium windows

    International Nuclear Information System (INIS)

    Gmur, N.F.

    1986-12-01

    It is known that various crystalline and liquid compounds form on the downstream surfaces of beryllium windows exposed to air. It is also known that the integrity of such windows may be compromised resulting in leaks through the window. The purpose of this report is to document the occurrences described as they pertain to the NSLS and to analyze, where possible, the various substances formed

  10. Preparation and properties of beryllium diphosphide

    International Nuclear Information System (INIS)

    Brice, J.F.; Gerardin, R.; Zanne, M.; Gleitzer, C.; Aubry, J.

    1975-01-01

    The compound BeP 2 can be obtained by direct action of phosphor on beryllium metal at 800-1000 0 C, and by removal of arsenic with phosphor in the diarsenide BeAs 2 . BeP 2 is a non hygroscopic brown-red powder. The X rays diffraction provide evidence for a quadratic cell with a = 7.08 A and c = 15.06 A. The atomic stacking is diamond type

  11. Behavior of beryllium pebbles under irradiation

    Energy Technology Data Exchange (ETDEWEB)

    Dalle-Donne, M.; Scaffidi-Argentina, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reactortechnik; Baldwin, D.L.; Gelles, D.S.; Greenwood, L.R.; Kawamura, H.; Oliver, B.M.

    1998-01-01

    Beryllium pebbles are being considered in fusion reactor blanket designs as neutron multiplier. An example is the European `Helium Cooled Pebble Bed Blanket.` Several forms of beryllium pebbles are commercially available but little is known about these forms in response to fast neutron irradiation. Commercially available beryllium pebbles have been irradiated to approximately 1.3 x 10{sup 22} n/cm{sup 2} (E>1 MeV) at 390degC. Pebbles 1-mm in diameter manufactured by Brush Wellman, USA and by Nippon Gaishi Company, Japan, and 3-mm pebbles manufactured by Brush Wellman were included. All were irradiated in the below-core area of the Experimental Breeder Reactor-II in Idaho Falls, USA, in molybdenum alloy capsules containing helium. Post-irradiation results are presented on density change measurements, tritium release by assay, stepped-temperature anneal, and thermal ramp desorption tests, and helium release by assay and stepped-temperature anneal measurements, for Be pebbles from two manufacturing methods, and with two specimen diameters. The experimental results on density change and tritium and helium release are compared with the predictions of the code ANFIBE. (author)

  12. Dynamic behaviour of S200F beryllium

    International Nuclear Information System (INIS)

    Montoya, Dominique

    1991-01-01

    Compression tests have been made on a large scale of strain, strain rate (up to 2000 s -1 ) and temperature (between 20 C and 300 C). From these experiences, we have calculated a constitutive model for beryllium S200F, which can be used by computer codes. Its formulation is not far from Steinberg, Cochran and Guinan's. But in our case, the influences of temperature and strain rate appear clearly within the expression. To validate our equation, we have used it in a computer code. Its extrapolation for higher strain rates is in good agreement with experiments such as Taylor impact tests or plate impact tests (strain rates greater than 10 4 s -1 ). With micrography, we could settle a link between the main strain mode within the material, and the variation of one parameter of the model. Beside the constitutive model, we have shown that shock loaded beryllium behaves in two different ways. If the strain rate is lower than 5.10 6 s -1 , then it is proportional to the squared shock pressure. Beyond, it is a linear function of shock pressure to the power of four. By a spall study on beryllium, we have confirmed that it is excessively fragile. Its fracture is sudden, at a strength near 1 GPa. (author) [fr

  13. Quantitative method of determining beryllium or a compound thereof in a sample

    Science.gov (United States)

    McCleskey, T. Mark; Ehler, Deborah S.; John, Kevin D.; Burrell, Anthony K.; Collis, Gavin E.; Minogue, Edel M.; Warner, Benjamin P.

    2010-08-24

    A method of determining beryllium or a beryllium compound thereof in a sample, includes providing a sample suspected of comprising beryllium or a compound thereof, extracting beryllium or a compound thereof from the sample by dissolving in a solution, adding a fluorescent indicator to the solution to thereby bind any beryllium or a compound thereof to the fluorescent indicator, and determining the presence or amount of any beryllium or a compound thereof in the sample by measuring fluorescence.

  14. The Cryogenic Properties of Several Aluminum-Beryllium Alloys and a Beryllium Oxide Material

    Science.gov (United States)

    Gamwell, Wayne R.; McGill, Preston B.

    2003-01-01

    Performance related mechanical properties for two aluminum-beryllium (Al-Be) alloys and one beryllium-oxide (BeO) material were developed at cryogenic temperatures. Basic mechanical properties (Le., ultimate tensile strength, yield strength, percent elongation, and elastic modulus were obtained for the aluminum-beryllium alloy, AlBeMetl62 at cryogenic [-195.5"C (-320 F) and -252.8"C (-423"F)I temperatures. Basic mechanical properties for the Be0 material were obtained at cyrogenic [- 252.8"C (-423"F)] temperatures. Fracture properties were obtained for the investment cast alloy Beralcast 363 at cryogenic [-252.8"C (-423"F)] temperatures. The AlBeMetl62 material was extruded, the Be0 material was hot isostatic pressing (HIP) consolidated, and the Beralcast 363 material was investment cast.

  15. The Rocky Flats Environmental Technology Site beryllium characterization project

    International Nuclear Information System (INIS)

    Morrell, D.M.; Miller, J.R.; Allen, D.F.

    1999-01-01

    A site beryllium characterization project was completed at the Rocky Flats Environmental Technology Site (RFETS) in 1997. Information from historical reviews, previous sampling surveys, and a new sampling survey were used to establish a more comprehensive understanding of the locations and levels of beryllium contamination in 35 buildings. A feature of the sampling strategy was to test if process knowledge was a good predictor of where beryllium contamination could be found. Results revealed that this technique was effective at identifying where surface contamination levels might exceed the RFETS smear control level but that it was not effective in identifying where low concentrations of beryllium might be found

  16. Formation of cellular structure in beryllium at plastic working

    International Nuclear Information System (INIS)

    Papirov, I.I.; Nikolaenko, A.A.; Shokurov, V.S.; Pikalov, A.I.

    2013-01-01

    Conditions of cellular structure formation are investigated at various kinds of deformation and heat treatment of beryllium ingots. It is shown that the cellular structure plays the important role in formation of complex of physical mechanical properties of beryllium. Influence of impurity, various conditions of deformation (temperature, squeezing degree) and heat treatments on substructure, texture and mechanical properties of metal is investigated. Optimum conditions of rolling and heat treatments of beryllium are defined. The way of sign-variable cyclic deformation of beryllium ingots is offered for reception quasi-isotropic fine-grained metal. Physical-mechanical properties of ultra fine-grained metal are studied

  17. The JET ITER-like wall experiment: First results and lessons for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Horton, Lorne, E-mail: Lorne.Horton@jet.efda.org [EFDA-CSU Culham, Culham Science Centre, Abingdon, OX14 3DB (United Kingdom); European Commission, B-1049 Brussels (Belgium)

    2013-10-15

    Highlights: ► JET has recently completed the installation of an ITER-like wall. ► Important operational aspects have changed with the new wall. ► Initial experiments have confirmed the expected low fuel retention. ► Disruption dynamics have change dramatically. ► Development of wall-compatible, ITER-relevant regimes of operation has begun. -- Abstract: The JET programme is strongly focused on preparations for ITER construction and exploitation. To this end, a major programme of machine enhancements has recently been completed, including a new ITER-like wall, in which the plasma-facing armour in the main vacuum chamber is beryllium while that in the divertor is tungsten—the same combination of plasma-facing materials foreseen for ITER. The goal of the initial experimental campaigns is to fully characterise operation with the new wall, concentrating in particular on plasma-material interactions, and to make direct comparisons of plasma performance with the previous, carbon wall. This is being done in a progressive manner, with the input power and plasma performance being increased in combination with the commissioning of a comprehensive new real-time protection system. Progress achieved during the first set of experimental campaigns with the new wall, which took place from September 2011 to July 2012, is reported.

  18. Detailed mechanical design and manufacturing study for the ITER reference breeding blanket

    International Nuclear Information System (INIS)

    Zacchia, F.; Daenner, W.; Stefanis, L. de; Ferrari, M.; Gerber, A.; Mustoe, J.

    1998-01-01

    This papers relates on the detailed mechanical design, manufacturing feasibility and assembly analysis of a water-cooled solid breeding blanket concept, selected as the ITER reference design. This breeding blanket design is characterised by: i) pressurised water flowing inside flat steel panels for cooling of the internals; each panel is welded along its contour onto the first wall structure and to the rear shield plate after closure of the module (last assembly step). ii) Beryllium (neutronic multiplier) in the form of micro-spheres filling the volume between parallel flat coolant panels. iii) Breeder pebbles enclosed in rods, which form bundles and are themselves embedded inside the Beryllium micro-spheres. (authors)

  19. Source term evaluation for accident transients in the experimental fusion facility ITER

    Energy Technology Data Exchange (ETDEWEB)

    Virot, F.; Barrachin, M.; Cousin, F. [IRSN, BP3-13115, Saint Paul lez Durance (France)

    2015-03-15

    We have studied the transport and chemical speciation of radio-toxic and toxic species for an event of water ingress in the vacuum vessel of experimental fusion facility ITER with the ASTEC code. In particular our evaluation takes into account an assessed thermodynamic data for the beryllium gaseous species. This study shows that deposited beryllium dusts of atomic Be and Be(OH){sub 2} are formed. It also shows that Be(OT){sub 2} could exist in some conditions in the drain tank. (authors)

  20. Neutron induced displacement damage in beryllium in the blanket of a (d,t)-fusion reactor

    International Nuclear Information System (INIS)

    Hermanutz, D.

    1995-09-01

    Beryllium is a favoured candidate for a neutron multiplier in solid breeder blankets of fusion reactors. This is mainly due to its low (n, 2n)-reaction threshold and because of its good thermal and mechanical properties. Its behaviour under intense neutron irradiation, however, is a crucial issue for its use in future fusion reactors. Displacement damage in beryllium so far has been calculated both with data related and methodological deficiencies. First of all, there is a need to have accurate cross-section data in order to obtain reliable spectra of primary knock-on atoms (PKA's). Furthermore, there are principal restrictions of the NRT-model in general used to calculate secondary displacements initiated by PKA's. The underlying theory of damage-energy (part of kinetic energy of PKA transferred elastically to matrix atoms) according to Lindhard is strictly valid only for medium and heavy mass ions with moderate energies in targets of the same element. In this work improved damage cross-sections and displacement rates (dpa/s) in beryllium have been calculated based on cross-section data from ENDF/B-VI (with a significantly improved (n, 2n)-evaluation) and on an appropriate treatment of damage-energy that is suitable for fusion relevant damage of light mass materials. ''This work has been performed in the framework of the Nuclear Fusion Project of the Forschungszentrum Karlsruhe and is supported by the European Communities within the European Fusion Technology Program''. (orig.)

  1. Corrosion of beryllium oxide; Corrosion de l'oxyde de beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Elston, J; Caillat, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

    Data are reported on the volatilization rate of beryllium oxide in moist air depending on temperature and water vapour concentration. They are concerned with powder samples or sintered shapes of various densities. For sintered samples, the volatilization rate is very low under the following conditions: - temperature: 1300 deg. C, - water vapour concentration in moist air: 25 g/m{sup 3}, - flow rate: 12 I/hour corresponding to a speed of 40 m/hour on the surface of the sample. For calcinated powders (1300 deg. C), grain growth has been observed under a stream of moist air at 1100 deg. C. For instance, grain size changes from 0,5 to at least 2 microns after 500 hours of exposure at this temperature. Furthermore, results data are reported on corrosion of sintered beryllium oxide in pressurized water. At 250 deg. C, under a pressure of 40 kg/cm{sup 2} water is very slightly corrosive; however, internal strains are revealed. Finally, some features on the corrosion in liquid sodium are exposed. (author)Fren. [French] La volatilisation de l'oxyde de beryllium dans l'air humide est etudiee en fonction de la temperature pour differentes teneurs de vapeur d'eau. Les essais decrits portent sur de l'oxyde de beryllium en poudre ou sur des echantillons d'oxyde de beryllium fritte de differentes densites. Avec un debit d'air de 12 I/h contenant 25 g de vapeur par m{sup 3} correspondant a une vitesse de 40 m/h sur la surface de l'echantillon, la volatilisation des frittes a 1300 deg. C reste tres faible. Sur de la poudre d'oxyde de beryllium calcinee initialement a 1300 deg. C, on observe un grossissement de la taille des grains sous l'influence de l'air humide a 1100 deg. C. Par exemple, elle passe de 0,5 a au moins 2 microns apres 500 heures d'exposition a cette temperature. On donne d'autre part les resultats d'une etude de la corrosion de frittes d'oxyde de beryllium par l'eau, en autoclave. A 250 deg. C, sous une pression de 40 kg/cm{sup 2}, l'action de l'eau reste tres

  2. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    Energy Technology Data Exchange (ETDEWEB)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li{sub 2}O) and lithium zirconate (Li{sub 2}ZrO{sub 3}) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers.

  3. U.S. technical report for the ITER blanket/shield: A. blanket: Topical report, July 1990--November 1990

    International Nuclear Information System (INIS)

    1995-01-01

    Three solid-breeder water-cooled blanket concepts have been developed for ITER based on a multilayer configuration. The primary difference among the concepts is in the fabricated form of breeder and multiplier. All the concepts have beryllium for neutron multiplication and solid-breeder temperature control. The blanket design does not use helium gaps or insulator material to control the solid breeder temperature. Lithium oxide (Li 2 O) and lithium zirconate (Li 2 ZrO 3 ) are the primary and the backup breeder materials, respectively. The lithium-6 enrichment is 95%. The use of high lithium-6 enrichment reduces the solid breeder volume required in the blanket and consequently the total tritium inventory in the solid breeder material. Also, it increases the blanket capability to accommodate power variation. The multilayer blanket configuration can accommodate up to a factor of two change in the neutron wall loading without violating the different design guidelines. The blanket material forms are sintered products and packed bed of small pebbles. The first concept has a sintered product material (blocks) for both the beryllium multiplier and the solid breeder. The second concept, the common ITER blanket, uses a packed bed breeder and beryllium blocks. The last concept is similar to the first except for the first and the last beryllium zones. Two small layers of beryllium pebbles are located behind the first wall and the back of the last beryllium zone to reduce the total inventory of the beryllium material and to improve the blanket performance. The design philosophy adopted for the blanket is to produce the necessary tritium required for the ITER operation and to operate at power reactor conditions as much as possible. Also, the reliability and the safety aspects of the blanket are enhanced by using low-pressure water coolant and the separation of the tritium purge flow from the coolant system by several barriers

  4. Sintering of beryllium oxide; Frittage de l'oxyde de beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Caillat, R; Pointud, R [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    This study had for origin to find a process permitting to manufacture bricks of beryllium oxide of pure nuclear grade, with a density as elevated as possible and with standardized shape. The sintering under load was the technique kept for the manufacture of the bricks. Because of the important toxicity of the beryllium oxide, the general features for the preliminary study of the sintering, have been determined while using alumina. The obtained results will be able to act as general indication for ulterior studies with sintering under load. (M.B.) [French] Cette etude a eu pour origine la recherche d'un procede permettant de fabriquer industriellement des briques d'oxyde de beryllium nucleaireraent pures, de densite aussi elevee que possible et de forme standardisee. Le frittage sous charge fut la technique retenue pour la fabrication des briques. En raison de la grande toxicite de l'oxyde de beryllium, les caracteristiques generales du frittage, pour l'etude preliminaire, ont ete determine en utilisant de l'alumine. Les resultats obtenus pourront servir d'indication generale pour des etudes ulterieurs avec frittage sous charge. (M.B.)

  5. Beryllium-stimulated neopterin as a diagnostic adjunct in chronic beryllium disease.

    Science.gov (United States)

    Maier, Lisa A; Kittle, Lori A; Mroz, Margaret M; Newman, Lee S

    2003-06-01

    The diagnosis of chronic beryllium disease (CBD) relies on the beryllium lymphocyte proliferation test (BeLPT) to demonstrate a Be specific immune response. This test has improved early diagnosis, but cannot discriminate beryllium sensitization (BeS) from CBD. We previously found high neopterin levels in CBD patients' serum and questioned whether Be-stimulated neopterin production by peripheral blood cells in vitro might be useful in the diagnosis of CBD. CBD, BeS, Be exposed workers without disease (Be-exp) normal controls and sarcoidosis subjects were enrolled. Peripheral blood mononuclear cells (PBMN) were cultured in the presence and absence of beryllium sulfate. Neopterin levels were determined from cell supernatants by enzyme linked immunosorbent assay (ELISA). Clinical evaluation of CBD subjects included chest radiography, pulmonary function testing, exercise testing, and the BeLPT. CBD patients produced higher levels of neopterin in both unstimulated and Be-stimulated conditions compared to all other subjects (P workplace screening. Copyright 2003 Wiley-Liss, Inc.

  6. Implanted deuterium retention and release in carbon-coated beryllium

    International Nuclear Information System (INIS)

    Anderl, R.A.; Longhurst, G.R.; Pawelko, R.J.; Oates, M.A.

    1997-01-01

    Deuterium implantation experiments have been conducted on samples of clean and carbon-coated beryllium. These studies entailed preparation and characterization of beryllium samples coated with carbon thicknesses of 100, 500, and 1000 angstrom. Heat treatment of a beryllium sample coated with carbon to a thickness of approximately 100 angstrom revealed that exposure to a temperature of 400 degrees C under high vacuum conditions was sufficient to cause substantial diffusion of beryllium through the carbon layer, resulting in more beryllium than carbon at the surface. Comparable concentrations of carbon and beryllium were observed in the bulk of the coating layer. Higher than expected oxygen levels were observed throughout the coating layer as well. Samples were exposed to deuterium implantation followed by thermal desorption without exposure to air. Differences were observed in deuterium retention and postimplantation release behavior in the carbon-coated samples as compared with bare samples. For comparable implantation conditions (sample temperature of 400 degrees C and an incident deuterium flux of approximately 6 X 10 19 D/m 2 sec), the quantity of deuterium retained in the bare sample was less than that retained in the carbon-coated samples. Further, the release of the deuterium took place at lower temperatures for the bare beryllium surfaces than for carbon-coated beryllium samples. 4 refs., 8 figs., 1 tab

  7. Implanted Deuterium Retention and Release in Carbon-Coated Beryllium

    Science.gov (United States)

    Anderl, R. A.; Longhurst, G. R.; Pawelko, R. J.; Oates, M. A.

    1997-06-01

    Deuterium implantation experiments have been conducted on samples of clean and carbon-coated beryllium. These studies entailed preparation and characterization of beryllium samples coated with carbon thicknesses of 100, 500, and 1000 Å. Heat treatment of a beryllium sample coated with carbon to a thickness of approximately 100 Å revealed that exposure to a temperature of 400°C under high vacuum conditions was sufficient to cause substantial diffusion of beryllium through the carbon layer, resulting in more beryllium than carbon at the surface. Comparable concentrations of carbon and beryllium were observed in the bulk of the coating layer. Higher than expected oxygen levels were observed throughout the coating layer as well. Samples were exposed to deuterium implantation followed by thermal desorption without exposure to air. Differences were observed in deuterium retention and postimplantation release behavior in the carbon-coated samples as compared with bare samples. For comparable implantation conditions (sample temperature of 400°C and an incident deuterium flux of approximately 6 × 1019 D/m2-s), the quantity of deuterium retained in the bare sample was less than that retained in the carbon-coated samples. Further, the release of the deuterium took place at lower temperatures for the bare beryllium surfaces than for carbon-coated beryllium samples.

  8. 10 CFR 850.20 - Baseline beryllium inventory.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 4 2010-01-01 2010-01-01 false Baseline beryllium inventory. 850.20 Section 850.20 Energy... Baseline beryllium inventory. (a) The responsible employer must develop a baseline inventory of the... inventory, the responsible employer must: (1) Review current and historical records; (2) Interview workers...

  9. Atlas of hot isostatic beryllium powder pressing diagrams

    International Nuclear Information System (INIS)

    Stoev, P.I.; Papirov, I.I.; Tikhinskij, G.F.; Vasil'ev, A.A.

    1995-01-01

    Diagrams of hot isotopic pressing (HIP) of beryllium powder with different grain size in a wide range of pressing parameters are built by mathematical modeling methods. The HIP diagrams presented are divided into 3 groups: parametric dependencies D=f(P,T); technological HIP diagrams; compacting mechanisms. The created data bank permits to optimise beryllium powder HIP with changing parameters. 4 refs., 23 figs

  10. Reactivity effects due to beryllium poisoning of BR2

    International Nuclear Information System (INIS)

    Kalcheva, S.; Ponsard, B.; Koonen, E.

    2004-01-01

    This paper illustrates the impact of the poisoning of the beryllium reflector on reactivity variations of the Belgian MTR BR2 in SCK.CEN. Detailed calculations by MCNP-4C of reactivity effects caused by strong neutron absorbers 3 He and 6 Li during reactor operation history are presented. The importance of beryllium poisoning for the accuracy of reactivity predictions is discussed. (authors)

  11. Progress report of preliminary studies of beryllium toxicity

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, H.C.

    1947-09-01

    This document was prepared in connection with a symposium of beryllium poisoning held at the Saranac Laboratories and describes progress made and a research program aimed at characterizing the toxicity of beryllium. Seven individual papers in this document are separately indexed and cataloged for the database.

  12. Joining of beryllium by braze welding technique: preliminary results

    Energy Technology Data Exchange (ETDEWEB)

    Banaim, P.; Abramov, E. [Ben-Gurion Univ. of the Negev, Beersheba (Israel); Zalkind, S.; Eden, S.

    1998-01-01

    Within the framework of some applications, there is a need to join beryllium parts to each other. Gas Tungsten Arc Braze Welds were made in beryllium using 0.3 mm commercially Aluminum (1100) shim preplaced at the joint. The welds exhibited a tendency to form microcracks in the Fusion Zone and Heat Affected Zone. All the microcracks were backfilled with Aluminum. (author)

  13. Laser welding of a beryllium/tantalum collimator

    International Nuclear Information System (INIS)

    Lingenfelter, A.C.; Anglin, C.D.

    1985-01-01

    This report describes the methods utilized in the fabrication of a collimator from 0.001 inch thick beryllium and tantalum foil. The laser welding process proved to be an acceptable method for joining the beryllium in a standing edge joint configuration

  14. Protection of air in premises and environment against beryllium aerosols

    Energy Technology Data Exchange (ETDEWEB)

    Bitkolov, N.Z.; Vishnevsky, E.P.; Krupkin, A.V. [Research Inst. of Industrial and Marine Medicine, St. Petersburg (Russian Federation)

    1998-01-01

    First and foremost, the danger of beryllium aerosols concerns a possibility of their inhalation. The situation is aggravated with high biological activity of the beryllium in a human lung. The small allowable beryllium aerosols` concentration in air poses a rather complex and expensive problem of the pollution prevention and clearing up of air. The delivery and transportation of beryllium aerosols from sites of their formation are defined by the circuit of ventilation, that forms aerodynamics of air flows in premises, and aerodynamic links between premises. The causes of aerosols release in air of premises from hoods, isolated and hermetically sealed vessels can be vibrations, as well as pulses of temperature and pressure. Furthermore, it is possible the redispersion of aerosols from dirty surfaces. The effective protection of air against beryllium aerosols at industrial plants is provided by a complex of hygienic measures: from individual means of breath protection up to collective means of the prevention of air pollution. (J.P.N.)

  15. Hydrodynamic instabilities in beryllium targets for the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Yi, S. A., E-mail: austinyi@lanl.gov; Simakov, A. N.; Wilson, D. C.; Olson, R. E.; Kline, J. L.; Batha, S. H. [Los Alamos National Laboratory, P.O. Box 1663, Los Alamos, New Mexico 87545 (United States); Clark, D. S.; Hammel, B. A.; Milovich, J. L.; Salmonson, J. D.; Kozioziemski, B. J. [Lawrence Livermore National Laboratory, P.O. Box 808, Livermore, California 94551 (United States)

    2014-09-15

    Beryllium ablators offer higher ablation velocity, rate, and pressure than their carbon-based counterparts, with the potential to increase the probability of achieving ignition at the National Ignition Facility (NIF) [E. I. Moses et al., Phys. Plasmas 16, 041006 (2009)]. We present here a detailed hydrodynamic stability analysis of low (NIF Revision 6.1) and high adiabat NIF beryllium target designs. Our targets are optimized to fully utilize the advantages of beryllium in order to suppress the growth of hydrodynamic instabilities. This results in an implosion that resists breakup of the capsule, and simultaneously minimizes the amount of ablator material mixed into the fuel. We quantify the improvement in stability of beryllium targets relative to plastic ones, and show that a low adiabat beryllium capsule can be at least as stable at the ablation front as a high adiabat plastic target.

  16. Preparation and characterization of beryllium doped organic plasma polymer coatings

    International Nuclear Information System (INIS)

    Brusasco, R.; Letts, S.; Miller, P.; Saculla, M.; Cook, R.

    1995-01-01

    We report the formation of beryllium doped plasma polymerized coatings derived from a helical resonator deposition apparatus, using diethylberyllium as the organometaric source. These coatings had an appearance not unlike plain plasma polymer and were relatively stable to ambient exposure. The coatings were characterized by Inductively Coupled Plasma Mass Spectrometry and X-Ray Photoelectron Spectroscopy. Coating rates approaching 0.7 μm hr -1 were obtained with a beryllium-to-carbon ratio of 1:1.3. There is also a significant oxygen presence in the coating as well which is attributed to oxidation upon exposure of the coating to air. The XPS data show only one peak for beryllium with the preponderance of the XPS data suggesting that the beryllium exists as BeO. Diethylberyllium was found to be inadequate as a source for beryllium doped plasma polymer, due to thermal decomposition and low vapor recovery rates

  17. ZIRCONIUM-TITANIUM-BERYLLIUM BRAZING ALLOY

    Science.gov (United States)

    Gilliland, R.G.; Patriarca, P.; Slaughter, G.M.; Williams, L.C.

    1962-06-12

    A new and improved ternary alloy is described which is of particular utility in braze-bonding parts made of a refractory metal selected from Group IV, V, and VI of the periodic table and alloys containing said metal as a predominating alloying ingredient. The brazing alloy contains, by weight, 40 to 50 per cent zirconium, 40 to 50 per cent titanium, and the balance beryllium in amounts ranging from 1 to 20 per cent, said alloy having a melting point in the range 950 to 1400 deg C. (AEC)

  18. Moessbauer study of iron diffusion in beryllium

    International Nuclear Information System (INIS)

    Sepiol, B.; Ruebenbauer, K.; Miczko, B.; Birchall, T.

    1991-01-01

    The broadening of the 14.41 keV Moessbauer line of 57 Fe dure to diffusion of iron atoms in polycrystalline beryllium has been investigated in the temperature range 1123 to 1423 K. The observed broadenings obey the Arrhenius law with activation energy 1.66(10) eV, i.e., lower than that obtained from the corresponding polycrystalline tracer data. The variations of the resonant fraction, second order Doppler shift and quadrupole splitting versus temperature are reported. An average diffusion coefficient has been calculated from the obtained broadenings and compared with the tracer results. (orig.)

  19. ITER council proceedings: 1998

    International Nuclear Information System (INIS)

    1999-01-01

    This volume contains documents of the 13th and the 14th ITER council meeting as well as of the 1st extraordinary ITER council meeting. Documents of the ITER meetings held in Vienna and Yokohama during 1998 are also included. The contents include an outline of the ITER objectives, the ITER parameters and design overview as well as operating scenarios and plasma performance. Furthermore, design features, safety and environmental characteristics are given

  20. Microstructure Analysis on Beryllium Reflector Blocks of Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Suk Hoon; Jang, Jin Sung; Jeong, Yong Hwan; Han, Chang Hee; Jung, Yang Il; Kim, Tae Kyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Choi, Yong Seok; Oh, Kyu Hwan [Seoul National University, Seoul (Korea, Republic of)

    2012-05-15

    A pure beryllium has a very low mass absorption coefficient: it has been used as the reflector element material in research reactors. The lifetime of beryllium reflector elements usually determined by the swelling: the swelling leads to dimensional change in the reflector frame, which results in bending or cracking of the parts. The mechanical interference in between parts should be avoided; the anisotropy of beryllium also needs to be considered. A beryllium has hexagonal close-pack (HCP) crystal structure, which is inherently anisotropic. It has virtually no ductility in one direction. There are two main aspects in the manufacturing of beryllium which will affect its isotropy, and those are the powder morphology and the consolidation process. Powder metallurgy permits the material to be produced in isotropic and fine-grained form, which overcomes the crystal structure problem by distributing loads in low ductility oriented grains to high ductility oriented grains. There are three representative consolidating methods to make beryllium reflector blocks. Traditionally, most powder-derived grades of beryllium have been consolidated by vacuum hot-pressing (VHP). A column of loose beryllium powder is compacted under vacuum by the pressure of the opposed upper and lower punches, bringing the billet to final density. The VHP process is directional in nature: it contributes to the anisotropy of the material properties. Another consolidating method for beryllium powder is hot isostatic pressing (HIPing), which will enhance its isotropy. During HIPing, The argon gas exerts pressure uniformly in all directions on the can containing the beryllium powder. The HIP process is effective to improve the isotropy of the resulting material as well as refinement of grain sizes. The last consolidating method is hot extrusion (HE). A roughly close packed beryllium is subjected to severe plastic defomation, the grains are refined and the tensile strength is enhanced. Since the material

  1. Copper-beryllium alloys for technical applications

    International Nuclear Information System (INIS)

    Heller, W.

    1976-01-01

    Data of physical properties are compiled for the most commonly used copper-beryllium alloys (CuBe 2, CuBe 1.7, CuCoBe, and CuCoAgBe), with emphasis on their temperature dependence and their variation with particular annealing and hardening treatments. The purpose is to provide a reference source and to indicate the versatility of these materials with respect to other copper alloys and to pure copper. The special features of CuBe alloys include high mechanical strength with reasonably high electrical conductivity, as well as good wear and corrosion resistance. For example, CuBe 2 has a yield strength of up to 1200 N/mm 2 , about three times that of pure copper, whilst the electrical conductivity of CuCoBe can be as high as 28 MS/m, nearly half that of pure copper. Typical applications are springs and electrical contacts. The importance of a proper heat treatment is discussed in some detail, notably the metallurgy and effects of low-temperature annealing (precipitation-hardening). A chapter on manufacturing processes covers machining, brazing, welding, and cleaning. This is followed by some remarks on safety precautions against beryllium poisoning. CuBe alloys are commercially available in the form of wires, strips, rods, and bars. Typical dimensions, specifications, a brief cost estimate, and addresses of suppliers are listed. (Author)

  2. Development of all-beryllium riveted structures

    International Nuclear Information System (INIS)

    Floyd, D.R.; Leslie, W.W.; Miley, D.V.; Nokes, R.W.

    1976-01-01

    Results are presented of a development program aimed at making a full-scale, all-beryllium frustrum by riveted assembly methods. Included are descriptions of the sheet-metal fabrication practices and assembly plans. Results of extensive mechanical testing of both ingot- and powder-source beryllium products that are presented include tensile, notch-tensile, bearing, and shear tests. Although the full-size structure has not been built, examples are given of several conical and cylindrical structures that were made. The largest of these is a 20-in. diameter, 15-in. long cylinder that was roll-formed from one 0.050-in. thick ingot sheet and assembled with 60 countersunk rivets. Tensile testing of riveted flat coupons is also reported as is bulge testing of riveted cylindrical shells. A cost comparison of riveted deep-drawn and powder-source cylinders is made. Results show that when strength and dimensional tolerance requirements are not severe, a riveted assembly approach is warranted. 33 figures, 8 tables

  3. Charge-density study of crystalline beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Stewart, R F [Carnegie-Mellon Univ., Pittsburgh, Pa. (USA). Dept. of Chemistry

    1977-01-01

    The X-ray structure factors for crystalline beryllium measured by Brown (Phil. Mag. (1972), 26, 1377) have been analyzed with multipole deformation functions for charge-density information. Single exponential radial functions were used for the valence charge density. A valence monopole plus the three harmonics, P/sup 3//sub 5/(cos theta) sin 3phi, P/sub 6/(cos theta) and P/sup 3//sub 7/(cos theta) sin 3phi, provide a least-squares fit to the data with Rsub(w)=0.0081. The superposition of these density functions describes a bonding charge density between Be atoms along the c axis through the tetrahedral vacancy. The results reported here are in qualitative agreement with a recent pseudo-potential calculation of metallic beryllium. The final residuals in the analysis are largest at high sin theta/lambda values. This suggests that core charge deformation is present and/or anharmonic motion of the nuclei is appreciable.

  4. Physical properties of beryllium oxide - Irradiation effects

    International Nuclear Information System (INIS)

    Elston, J.; Caillat, R.

    1958-01-01

    This work has been carried out in view of determining several physical properties of hot-pressed beryllium oxide under various conditions and the change of these properties after irradiation. Special attention has been paid on to the measurement of the thermal conductivity coefficient and thermal diffusivity coefficient. Several designs for the measurement of the thermal conductivity coefficient have been achieved. They permit its determination between 50 and 300 deg. C, between 400 and 800 deg. C. Some measurements have been made above 1000 deg. C. In order to measure the thermal diffusivity coefficient, we heat a perfectly flat surface of a sample in such a way that the heat flux is modulated (amplitude and frequency being adjustable). The thermal diffusivity coefficient is deduced from the variations of temperature observed on several spots. Tensile strength; compressive strength; expansion coefficient; sound velocity and crystal parameters have been also measured. Some of the measurements have been carried out after neutron irradiation. Some data have been obtained on the change of the properties of beryllium oxide depending on the integrated neutron flux. (author) [fr

  5. ITER Council proceedings: 1993

    International Nuclear Information System (INIS)

    1994-01-01

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ''ITER EDA Agreement and Protocol 2'' (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tasks; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings

  6. ITER council proceedings: 2000

    International Nuclear Information System (INIS)

    2001-01-01

    No ITER Council Meetings were held during 2000. However, two ITER EDA Meetings were held, one in Tokyo, January 19-20, and one in Moscow, June 29-30. The parties participating in these meetings were those that partake in the extended ITER EDA, namely the EU, the Russian Federation, and Japan. This document contains, a/o, the records of these meetings, the list of attendees, the agenda, the ITER EDA Status Reports issued during these meetings, the TAC (Technical Advisory Committee) reports and recommendations, the MAC Reports and Advice (also for the July 1999 Meeting), the ITER-FEAT Outline Design Report, the TAC Reports and Recommendations both meetings), Site requirements and Site Design Assumptions, the Tentative Sequence of technical Activities 2000-2001, Report of the ITER SWG-P2 on Joint Implementation of ITER, EU/ITER Canada Proposal for New ITER Identification

  7. ITER council proceedings: 1993

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    Records of the third ITER Council Meeting (IC-3), held on 21-22 April 1993, in Tokyo, Japan, and the fourth ITER Council Meeting (IC-4) held on 29 September - 1 October 1993 in San Diego, USA, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA), such as the text of the draft of Protocol 2 further elaborated in ``ITER EDA Agreement and Protocol 2`` (ITER EDA Documentation Series No. 5), recommendations on future work programmes: a description of technology R and D tastes; the establishment of a trust fund for the ITER EDA activities; arrangements for Visiting Home Team Personnel; the general framework for the involvement of other countries in the ITER EDA; conditions for the involvement of Canada in the Euratom Contribution to the ITER EDA; and other attachments as parts of the Records of Decision of the aforementioned ITER Council Meetings.

  8. Fundamental hydrogen interactions with beryllium : a magnetic fusion perspective.

    Energy Technology Data Exchange (ETDEWEB)

    Wampler, William R. (Sandia National Laboratories, Albuquerque, NM); Felter, Thomas E.; Whaley, Josh A.; Kolasinski, Robert D.; Bartelt, Norman Charles

    2012-03-01

    Increasingly, basic models such as density functional theory and molecular dynamics are being used to simulate different aspects of hydrogen recycling from plasma facing materials. These models provide valuable insight into hydrogen diffusion, trapping, and recombination from surfaces, but their validation relies on knowledge of the detailed behavior of hydrogen at an atomic scale. Despite being the first wall material for ITER, basic single crystal beryllium surfaces have been studied only sparsely from an experimental standpoint. In prior cases researchers used electron spectroscopy to examine surface reconstruction or adsorption kinetics during exposure to a hydrogen atmosphere. While valuable, these approaches lack the ability to directly detect the positioning of hydrogen on the surface. Ion beam techniques, such as low energy ion scattering (LEIS) and direct recoil spectroscopy (DRS), are two of the only experimental approaches capable of providing this information. In this study, we applied both LEIS and DRS to examine how hydrogen binds to the Be(0001) surface. Our measurements were performed using an angle-resolved ion energy spectrometer (ARIES) to probe the surface with low energy ions (500 eV - 3 keV He{sup +} and Ne{sup +}). We were able to obtain a 'scattering maps' of the crystal surface, providing insight on how low energy ions are focused along open surface channels. Once we completed a characterization of the clean surface, we dosed the sample with atomic hydrogen using a heated tungsten capillary. A distinct signal associated with adsorbed hydrogen emerged that was consistent with hydrogen residing between atom rows. To aid in the interpretation of the experimental results, we developed a computational model to simulate ion scattering at grazing incidence. For this purpose, we incorporated a simplified surface model into the Kalypso molecular dynamics code. This approach allowed us to understand how the incident ions interacted with the

  9. Carbon and tungsten effect on characteristics of sputtered and re-deposited beryllium target layers under deuteron bombardment

    International Nuclear Information System (INIS)

    Danelyan, L.S.; Gureev, V.M.; Elistratov, N.G.

    2004-01-01

    The behavior of the plasma facing Be-elements in the International Thermonuclear Experimental Reactor ITER will be affected by the re-deposition of other eroded plasma facing materials. The effect of carbon- and tungsten-additions on the microstructure, chemical composition and hydrogen isotope accumulation in the sputtered and re-deposited layers of beryllium TGP-56 at its interaction with 200 - 300 eV hydrogen isotope ions was studied in the MAGRAS facility equipped with a magnetron sputtering system. (author)

  10. Fabrication of small mock-ups reflecting the design features of the ITER semi-prototype

    International Nuclear Information System (INIS)

    Jung, Yang-Il; Choi, Byoung-Kwon; Park, Jeong-Yong; Kim, Suk-Kwon; Lee, Dong Won; Kim, Byoung Yoon

    2012-01-01

    The ITER semi-prototype was designed to qualify the manufacturing technology for the ITER blanket first wall. However, its fabrication is expected to face great difficulty due to a design complexity. Even though joining technology for different materials such as beryllium, CuCrZr, and stainless steel (SS) was developed during the first stage of qualification, the joining is still a key issue for the fabrication of the semi-prototype. In this study, small mock-ups (SMU) were fabricated to realize and verify the manufacturing of the semi-prototype reflecting the described design features. The joining of multiple beryllium tiles on the angled CuCrZr surface was confirmed with SMU no. 1. Six beryllium tiles were joined using hot isostatic pressing (HIP), and slitting was then performed to form multiple tiles. In SMU no. 2, HIP was performed two times in order to facilitate the cooling channels at the CuCrZr/SS interface, and to join the beryllium tiles on CuCrZr/SS. The method used to form a pressure boundary for the complex cooling channels was also developed by fabricating the SMU no. 3. The SMUs confirmed the applicability of the HIP for the manufacturing of the semi-prototype.

  11. Ion beam assisted deposition of metal-coatings on beryllium

    International Nuclear Information System (INIS)

    Tashlykov, I.S.; Tul'ev, V.V.

    2015-01-01

    Thin films were applied on beryllium substrates on the basis of metals (Cr, Ti, Cu and W) with method of the ion-assisted deposition in vacuum. Me/Be structures were prepared using 20 kV ions irradiation during deposition on beryllium neutral fraction generated from vacuum arc plasma. Rutherford back scattering and computer simulation RUMP code were applied to investigate the composition of the modified beryllium surface. Researches showed that the superficial structure is formed on beryllium by thickness ~ 50-60 nm. The covering composition includes atoms of the deposited metal (0.5-3.3 at. %), atoms of technological impurity carbon (0.8-1.8 at. %) and oxygen (6.3-9.9 at. %), atoms of beryllium from the substrate. Ion assisted deposition of metals on beryllium substrate is accompanied by radiation enhanced diffusion of metals, oxygen atoms in the substrate, out diffusion of beryllium, carbon atoms in the deposited coating and sputtering film-forming ions assists. (authors)

  12. Historical perspectives on the uses and health risks of beryllium

    International Nuclear Information System (INIS)

    Preuss, O.P.

    1985-01-01

    Unawareness of the health risks of beryllium resulted in a decade of unmitigated exposure of several thousand workers and numerous cases of beryllium disease in employees and nearby residents. Subsequent adoption of exposure limits and their implementation with effective technical controls reduced the occurrence of new cases, which were mainly due to accidental exposures, to a minimum. The fact that continuously growing production and consumption did not alter this trend demonstrates the effectiveness of the present threshold limit value. It shows that the potential health hazard can be well contained and that beryllium can be produced and fabricated without undue risk to employees or the general public

  13. Analysis of surface contaminants on beryllium and aluminum windows

    International Nuclear Information System (INIS)

    Gmur, N.F.

    1987-06-01

    An effort has been made to document the types of contamination which form on beryllium window surfaces due to interaction with a synchrotron radiation beam. Beryllium windows contaminated in a variety of ways (exposure to water and air) exhibited surface powders, gels, crystals and liquid droplets. These contaminants were analyzed by electron diffraction, electron energy loss spectroscopy, energy dispersive x-ray spectroscopy and wet chemical methods. Materials found on window surfaces include beryllium oxide, amorphous carbon, cuprous oxide, metallic copper and nitric acid. Aluminum window surface contaminants were also examined

  14. About kinetics of paramagnetic radiation malformations in beryllium ceramics

    International Nuclear Information System (INIS)

    Polyakov, A.I.; Ryabinkin, Yu.A.; Zashkvara, O.V.; Bitenbaev, M.I.; Petukhov, Yu.V.

    1999-01-01

    This paper [1] specifies that γ-radiation of the beryllium-oxide-based ceramics results in development of paramagnetic radiation malformations emerging the ESR spectrum in form of doublet with the splitting rate of oestrasid Δ∼1.6 and g-factor of 2.008. This report presents evaluation outcomes of dependence of paramagnetic radiation malformations concentration in beryllium ceramics on gamma-radiation dose ( 60 Co) within the range of 0-100 Mrad. Total paramagnetic parameters of beryllium ceramics in the range 0-100 Mrad of gamma-radiation dose varied slightly, and were specified by the first type of paramagnetic radiation malformations

  15. Evaluation of the hazard associated with fabricating beryllium copper alloys

    International Nuclear Information System (INIS)

    Senn, T.J.

    1977-01-01

    Beryllium-copper alloys should be considered toxic materials and proper controls must be used when they are machined, heated, or otherwise fabricated. Air samples should be taken for each type of fabrication to determine the worker's exposure and the effectiveness of the controls in use. It has been shown that aerosols containing beryllium are generated during the four methods of fabrication tested, and that these aerosols can be reduced through local exhaust to undetectable levels. Considering the acute, chronic and possibly carcinogenic effects of exposure to beryllium, effective controls should be required because they are feasible both technologically and economically. The health hazards and control measures are reviewed

  16. Preparation of a sinterable beryllium oxide through decomposition of beryllium hydroxide (1963)

    International Nuclear Information System (INIS)

    Bernier, M.

    1963-01-01

    In the course of the present study, we have attempted to precise the factors which among the ones effective in the course of the preparation of the beryllium hydroxide and oxide and during the sintering have an influence on the final result: the density and homogeneity of the sintered body. Of the several varieties of hydroxides precipitated from a sulfate solution the β-hydroxide only is always contaminated with beryllium sulfate and cannot be purified even by thorough washing. We noticed that those varieties of the hydroxide (gel, α, β) have different decomposition rates; this behaviour is used to identify and even to dose the different species in (α, β) mixtures. The various hydroxides transmit to the resulting oxides the shape they had when precipitated. Accordingly the history of the oxide is revealed by its behaviour during its fabrication and sintering. By comparing the results of the sintering operation with the various measurements performed on the oxide powders we are led to the conclusion that an oxide obtained from beryllium hydroxide is sinterable under vacuum if the following conditions are fulfilled: the particle size must lie between 0.1 and 0.2 μ and the BeSO 4 content of the powder must be less than 0.25 per cent wt (expressed as SO 3 /BeO). The best fitting is obtained with the oxide issued from an α-hydroxide precipitated as very small aggregates and with a low sulfur-content. We have observed that this is also the case for the oxide obtained by direct calcination of beryllium sulfate. (author) [fr

  17. Novel plasma source for safe beryllium spectral line studies in the presence of beryllium dust

    Science.gov (United States)

    Stankov, B. D.; Vinić, M.; Gavrilović Božović, M. R.; Ivković, M.

    2018-05-01

    Plasma source for beryllium spectral line studies in the presence of beryllium dust particles was realised. The guideline during construction was to prevent exposure to formed dust, considering the toxicity of beryllium. Plasma source characterization through determination of optimal working conditions is described. The necessary conditions for Be spectral line appearance and optimal conditions for line shape measurements are found. It is proven experimentally that under these conditions dust appears coincidently with the second current maximum. The electron density measured after discharge current maximum is determined from the peak separation of the hydrogen Balmer beta spectral line, and the electron temperature is determined from the ratios of the relative intensities of Be spectral lines emitted from successive ionized stages of atoms. Maximum values of electron density and temperature are measured to be 9.3 × 1022 m-3 and 16 800 K, respectively. Construction details and testing of the BeO discharge tube in comparison with SiO2 and Al2O3 discharge tubes are also presented in this paper.

  18. Real-time monitoring of airborne beryllium, at OSHA limit levels, by time-resolved laser-induced breakdown spectroscopy

    International Nuclear Information System (INIS)

    Radziemski, L.J.; Loree, T.R.; Cremers, D.A.

    1982-01-01

    Real-time detection of beryllium particulate is being investigated by the new technique of laser-induced breakdown spectroscopy. For beryllium detection we monitor the 313.1-nm feature of once ionized beryllium (Be II). Numerous publications describe the technique, our beryllium results, and other applications. Here we summarize the important points and describe our experiments with beryllium

  19. Tritium inventory in the ITER PFC`s: predictions, uncertainties, R and D status and priority needs

    Energy Technology Data Exchange (ETDEWEB)

    Federici, G. [ITER, Garching (Germany). JWS; Anderl, R.; Longhurst, G. [Idaho National Engineering and Environmental Laboratory, Idaho Falls, Idaho 83415 (United States); Brooks, J.N. [Argonne National Laboratory, 9700 S. Cass Avenue, Argonne, Illinois 60439 (United States); Causey, R.; Cowgill, D.; Wampler, W.; Wilson, K.; Youchison, D. [Sandia National Laboratories, Livermore California and Albuquerque, New Mexico (United States); Coad, J.P.; Peacock, A.; Pick, M. [JET Joint Undertaking, Abingdon, Oxfordshire OX14 3EA (United Kingdom); Doerner, R.; Luckhardt, S. [University of California San Diego, La Jolla, California 92093-0417 (United States); Haasz, A.A. [University of Toronto, Institute for Aerospace Studies, Ontario M3H 5T6 (Canada); Mueller, D.; Skinner, C.H. [Plasma Physics Laboratory, Princeton University, Princeton, New Jersey 08543 (United States); Wong, C. [General Atomics, PO Box 85608, San Diego, California 92186-9784 (United States); Wu, C. [NET Team, Boltzmannstrasse 2, 85748 Garching (Germany)

    1998-09-01

    New data on hydrogen plasma isotopes retention in beryllium and tungsten are now becoming available from various laboratories for conditions similar to those expected in the International Thermonuclear Experimental Reactor (ITER) where previous data were either missing or largely scattered. Together with a significant advancement in understanding, they have warranted a revisitation of the previous estimates of tritium inventory in ITER, with beryllium as the plasma facing material for the first-wall components, and tungsten in the divertor with some carbon-fibre-composites clad areas, near the strike points. Based on these analyses, it is shown that the area of primary concern with, respect to tritium inventory, remains codeposition with carbon and possibly beryllium on the divertor surfaces. Here, modelling of ITER divertor conditions continues to show potentially large codeposition rates which are confirmed by tokamak findings. Contrary to the tritium residing deep in the bulk of materials, this surface tritium represents a safety hazard as it can be easily mobilised in the event of an accident. It could, however, be possibly removed and recovered. It is concluded that active and efficient methods to remove the codeposited layers are needed in ITER and periodic conditioning/cleaning would be required to control the tritium inventory and avoid exhausting the available fuel supply. Some methods which could possibly be used for in-situ cleaning are briefly discussed in conjunction with the research and development work required to extrapolate their applicability to ITER. (orig.) 53 refs.

  20. ITER council proceedings: 1995

    International Nuclear Information System (INIS)

    1996-01-01

    Records of the 8. ITER Council Meeting (IC-8), held on 26-27 July 1995, in San Diego, USA, and the 9. ITER Council Meeting (IC-9) held on 12-13 December 1995, in Garching, Germany, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the ITER Interim Design Report Package and Relevant Documents. Figs, tabs

  1. ITER council proceedings: 1999

    International Nuclear Information System (INIS)

    1999-01-01

    In 1999 the ITER meeting in Cadarache (10-11 March 1999) and the Programme Directors Meeting in Grenoble (28-29 July 1999) took place. Both meetings were exclusively devoted to ITER engineering design activities and their agendas covered all issues important for the development of ITER. This volume presents the documents of these two important meetings

  2. ITER council proceedings: 1996

    International Nuclear Information System (INIS)

    1997-01-01

    Records of the 10. ITER Council Meeting (IC-10), held on 26-27 July 1996, in St. Petersburg, Russia, and the 11. ITER Council Meeting (IC-11) held on 17-18 December 1996, in Tokyo, Japan, are presented, giving essential information on the evolution of the ITER Engineering Design Activities (EDA) and the cost review and safety analysis. Figs, tabs

  3. ITER EDA technical activities

    International Nuclear Information System (INIS)

    Aymar, R.

    1998-01-01

    Six years of technical work under the ITER EDA Agreement have resulted in a design which constitutes a complete description of the ITER device and of its auxiliary systems and facilities. The ITER Council commented that the Final Design Report provides the first comprehensive design of a fusion reactor based on well established physics and technology

  4. ITER radio frequency systems

    International Nuclear Information System (INIS)

    Bosia, G.

    1998-01-01

    Neutral Beam Injection and RF heating are two of the methods for heating and current drive in ITER. The three ITER RF systems, which have been developed during the EDA, offer several complementary services and are able to fulfil ITER operational requirements

  5. Investigation of the ion beryllium surface interaction

    International Nuclear Information System (INIS)

    Guseva, M.I.; Birukov, A.Yu.; Gureev, V.M.

    1995-01-01

    The self -sputtering yield of the Be was measured. The energy dependence of the Be self-sputtering yield agrees well with that calculated by W. Eckstein et. al. Below 770 K the self-sputtering yield is temperature independent; at T irr .> 870 K it increases sharply. Hot-pressed samples at 370 K were implanted with monoenergetic 5 keV hydrogen ions and with a stationary plasma (flux power ∼ 5 MW/m 2 ). The investigation of hydrogen behavior in beryllium shows that at low doses hydrogen is solved, but at doses ≥ 5x10 22 m -2 the bubbles and channels are formed. It results in hydrogen profile shift to the surface and decrease of its concentration. The sputtering results in further concentration decrease at doses > 10 25 m -2

  6. Polarizabilities of the beryllium clock transition

    International Nuclear Information System (INIS)

    Mitroy, J.

    2010-01-01

    The polarizabilities of the three lowest states of the beryllium atom are determined from a large basis configuration interaction calculation. The polarizabilities of the 2s 2 1 S e ground state (37.73a 0 3 ) and the 2s2p 3 P 0 o metastable state (39.04a 0 3 ) are found to be very similar in size and magnitude. This leads to an anomalously small blackbody radiation shift at 300 K of -0.018(4) Hz for the 2s 2 1 S e -2s2p 3 P 0 o clock transition. Magic wavelengths for simultaneous trapping of the ground and metastable states are also computed.

  7. Primordial beryllium as a big bang calorimeter.

    Science.gov (United States)

    Pospelov, Maxim; Pradler, Josef

    2011-03-25

    Many models of new physics including variants of supersymmetry predict metastable long-lived particles that can decay during or after primordial nucleosynthesis, releasing significant amounts of nonthermal energy. The hadronic energy injection in these decays leads to the formation of ⁹Be via the chain of nonequilibrium transformations: Energy(h)→T, ³He→⁶He, ⁶Li→⁹Be. We calculate the efficiency of this transformation and show that if the injection happens at cosmic times of a few hours the release of O(10 MeV) per baryon can be sufficient for obtaining a sizable ⁹Be abundance. The absence of a plateau structure in the ⁹Be/H abundance down to a O(10⁻¹⁴) level allows one to use beryllium as a robust constraint on new physics models with decaying or annihilating particles.

  8. Advances in beryllium powder consolidation simulations

    International Nuclear Information System (INIS)

    Reardon, B.J.

    1998-01-01

    A fuzzy logic based multiobjective genetic algorithm (GA) is introduced and the algorithm is used to optimize micromechanical densification modeling parameters for warm isopressed beryllium powder, HIPed copper powder and CIPed/sintered and HIPed tantalum powder. In addition to optimizing the main model parameters using the experimental data points as objective functions, the GA provides a quantitative measure of the sensitivity of the model to each parameter, estimates the mean particle size of the powder, and determines the smoothing factors for the transition between stage 1 and stage 2 densification. While the GA does not provide a sensitivity analysis in the strictest sense, and is highly stochastic in nature, this method is reliable and reproducible in optimizing parameters given any size data set and determining the impact on the model of slight variations in each parameter

  9. Electron-beam fusion welding of beryllium

    International Nuclear Information System (INIS)

    Campbell, R.P.; Dixon, R.D.; Liby, A.L.

    1978-01-01

    Ingot-sheet beryllium (Be) having three different chemistries and three different thicknesses was fusion-welded by the electron-beam process. Several different preheats were used to obtain 100% penetration and crack-free welds. Cracking susceptability was found to be related to aluminum (Al) content; the higher Al-content material was most susceptable. However, adequate preheat allowed full penetration and crack-free welds to be made in all materials tested. The effect of a post-weld heat treatment on the mechanical properties of these compositions was also determined. The heat treatment produced no significant effect on the ultimate tensile strength. However, the yield strength was decreased and the ductility was increased. These changes are attributed to the formation of AlFeBe 4 and FeBe 11

  10. Preparation of selenium coatings onto beryllium foils

    International Nuclear Information System (INIS)

    Erikson, E.D.; Tassano, P.L.; Reiss, R.H.; Griggs, G.E.

    1984-09-01

    A technique for preparing selenium films onto 50.8 microns thick beryllium foils is described. The selenium was deposited in vacuum from a resistance heated evaporation source. Profilometry measurements of the coatings indicate deposit thicknesses of 5.5, 12.9, 37.5, 49.8 and 74.5 microns. The control of deposition rate and of coating thickness was facilitated using a commercially available closed-loop programmable thin film controller. The x-ray transmission of the coated substrates was measured using a tritiated zirconium source. The transmissivities of the film/substrate combination are presented for the range of energies from 4 to 20 keV. 15 references, 3 figures

  11. People's Republic of China joins ITER

    International Nuclear Information System (INIS)

    Huo Yuping

    2003-01-01

    The People's Republic of China is the largest developing country with a projected population of 1.6 - 2 billion people and an energy consumption growing from the current 1.3 Billion Tons Coal Equivalent (TCE) to more than 4 Billion TCE by 2050. This large demand needs to be accommodated in a sustainable way, requiring energy generation in an environmentally friendly way. Fusion is one of the most promising candidates to solve this important issue. This explains why in the second half of 2002, the ITER Participants' delegations to the ITER Negotiations received expression of interest from the People's Republic of China in the possibility of Chinese participation in ITER, including joining the ongoing Negotiations. The speed with which the Chinese authorities had made their decision to participate in the ITER Negotiations was impressive. The Prime Minister and the State Council had already confirmed their decision to apply to join ITER as soon as possible, and Mr. Xu Guanhua, Chinese Minister of Science and Technology, wrote on behalf of his government, on 10 January 2003, to the four heads of delegation in the ITER Negotiations, requesting that China participate in the present ITER Negotiations, pointing out that China intends to provide a substantial contribution to the Project, comparable to what is currently envisaged by some of the participants in the present Negotiations

  12. Electron microscope study of irradiated beryllium oxide

    International Nuclear Information System (INIS)

    Bisson, A.A.

    1965-06-01

    The beryllium oxide is studied first by fractography, before and after irradiation, using sintered samples. The fractures are examined under different aspects. The higher density sintered samples, with transgranular fractures are the most interesting for a microscopic study. It is possible to mark the difference between the 'pores' left by the sintering process and the 'bubbles' of gases that can be produced by former thermal treatments. After irradiation, the grain boundaries are very much weakened. By annealing, it is possible to observe the evolution of the gases produced by the reaction (n, 2n) and (n. α) and gathered on the grain boundaries. The irradiated beryllium oxide is afterwards studied by transmission. For that, a simple method has been used: little chips of the crushed material are examined. Clusters of point defects produced by neutrons are thus detected in crystals irradiated at the three following doses: 6 x 10 19 , 9 x 10 19 and 2 x 10 20 n f cm -2 at a temperature below 100 deg. C. For the irradiation at 6 x 10 19 n f cm -2 , the defects are merely visible, but at 2 x l0 20 n f cm -2 the crystals an crowded with clusters and the Kikuchi lines have disappeared from the micro-diffraction diagrams. The evolution of the clusters into dislocation loops is studied by a series of annealings. The activation energy (0,37 eV) calculated from the annealing curves suggests that it must be interstitials that condense into dislocation loops. Samples irradiated at high temperatures (650, 900 and 1100 deg. C) are also studied. In those specimens the size of the loops is not the same as the equilibrium size obtained after out of pile annealing at the same temperature. Those former loops are more specifically studied and their Burgers vector is determined by micro-diffraction. (author) [fr

  13. Deuterium permeation and diffusion in high-purity beryllium

    International Nuclear Information System (INIS)

    Abramov, E.; Riehm, M.P.; Thompson, D.A.; Smeltzer, W.W.

    1990-01-01

    The permeation rate of deuterium through high-purity beryllium membranes was measured using the gas-driven permeation technique. The time-dependent and the steady-state deuterium flux data were analyzed and the effective diffusivities of the samples were determined. Using multilayer permeation theory the effects of surface oxide were eliminated and the diffusion coefficients of the bulk beryllium determined. The diffusion parameters obtained for the extra-grade beryllium samples (99.8%) are D 0 =6.7x10 -9 m 2 /s and E D =28.4 kJ/mol. For the high-grade beryllium samples (99%) the parameters are D 0 =8.0x10 -9 m 2 /s and E D =35.1 kJ/mol. (orig.)

  14. Deuterium permeation and diffusion in high purity beryllium

    International Nuclear Information System (INIS)

    Abramov, E.

    1990-05-01

    The permeation rate of deuterium through high-purity beryllium membranes was measured using the gas-driven permeation technique. The time-dependent and the steady-state deuterium flux data were analyzed and the effective diffusivities of the samples were determined. A multilayer permeation theory was used in order to eliminate the surface oxide effects and the diffusion coefficients of the bulk beryllium were determined. The diffusion parameters obtained for the extra-grade beryllium samples (99.8%) are D 0 = 6.7 x 10 -9 [m 2 /s] and E D = 28.4 [KJ/mol]; and for the high-grade beryllium samples (99%) the parameters are D 0 = 8.0 x 10 -9 [m 2 /s] and E D = 35.1 [KJ/mol

  15. Spectrochemical determination of beryllium and lithium in stream sediments

    International Nuclear Information System (INIS)

    Gallimore, D.L.; Hues, A.D.; Palmer, B.A.; Cox, L.E.; Simi, O.R.; Bieniewski, T.M.; Steinhaus, D.W.

    1979-11-01

    A spectrochemical method was developed to analyze 200 or more samples of stream sediments per day for beryllium and lithium. One part of ground stream sediment is mixed with two parts graphite-SiO 2 buffer, packed into a graphite electrode, and excited in a direct-current arc. The resulting emission goes to a 3.4-m, direct-reading, Ebert spectrograph. A desk-top computer system is used to record and process the signals, and to report the beryllium and lithium concentrations. The limits of detection are 0.2 μg/g for beryllium and 0.5 μg/g for lithium. For analyses of prepared reference materials, the relative standard deviations were 16% for determining 0.2 to 100 μg/g of beryllium and 15% for determining 0.5 to 500 μg/g of lithium. A correction is made for vanadium interference

  16. Electropolishing or chemical milling of beryllium to remove machining defects

    International Nuclear Information System (INIS)

    Helms, J.R.

    1975-12-01

    The techniques of electropolishing and chemical milling to remove machine damage from beryllium are compared. Both techniques are found to be effective; chemical milling is recommended because it is easier to use and control

  17. Beryllium brazing considerations in CANDU fuel bundle manufacture

    International Nuclear Information System (INIS)

    Harmsen, J.; Pant, A.; Lewis, B.J.; Thompson, W.T.

    2010-01-01

    'Full text:' Appendages of CANDU fuel bundle elements are currently joined to zircaloy sheaths by vacuum beryllium brazing. Ongoing environmental and workplace concerns about beryllium combined with the continuous efforts by Cameco Fuel Manufacturing in its improvement process, initiated this study to find a substitute for pure beryllium. The presentation will review the necessary functionality of brazing alloy components and short list a series of alloys with the potential to duplicate the performance of pure beryllium. Modifications to current manufacturing processes based on in-plant testing will be discussed in relation to the use of these alloys. The presentation will conclude with a summary of the progress to date and further testing expected to be necessary.

  18. Health effects of beryllium exposure: a literature review

    National Research Council Canada - National Science Library

    Committee on Beryllium Alloy Exposures, Committee on Toxicology, National Research Council

    2007-01-01

    Beryllium is an important metal that is used in a number of industries-including the defense, aerospace, automotive, medical, and electronics industries-because of its exceptional strength, stability...

  19. Phosphorus-containing azo compounds as analytical reagents for beryllium

    International Nuclear Information System (INIS)

    Lisenko, N.F.; Dolzhnikova, E.N.; Petrova, G.S.; Tsvetkov, E.N.; Vsesoyuznyj Nauchno-Issledovatel'skij Inst. Khimicheskikh Reaktivov i Osobo Chistykh Veshchestv, Moscow; AN SSSR, Moscow. Inst. Ehlementoorganicheskikh Soedinenij)

    1979-01-01

    The interaction of beryllium with six new azocompounds based on chromotropic or R-acids and o-aminophenyl-phenylphosphonic acids is studied. A sharp difference in the detection limit for beryllium by the two groups of compounds is found. Azoderivatives based on chromotropic acid are promising agent for beryllium due to sufficiently high selectivity. The introduction of the methyl-group into the o-position of the phosphorus-containing group improves the analytical properties of agents. Techniques are developed for the determination of beryllium in bronze, sewage water and in an artificial mixture using a sodium salt of 1.8-dioxi-2 [2' - (oxi- (o-methylphenyl)-phosphenyl)-phenilazo]-naphtalene-3.6-disulfoacid

  20. Data Needs for Erosion and Tritium Retention in Beryllium Surfaces

    International Nuclear Information System (INIS)

    Braams, B.J.

    2011-07-01

    A Consultants' Meeting was held at IAEA Headquarters 30-31 May 2011 with the aim to provide advice about the scope and aims of a planned IAEA coordinated research project on erosion and tritium retention in beryllium plasma-facing materials and about other activities of the A+M Data Unit in the area of plasma interaction with beryllium surfaces. The present report contains the proceedings, recommendations and conclusions of that Consultants' Meeting. (author)

  1. ITER-FEAT safety

    International Nuclear Information System (INIS)

    Gordon, C.W.; Bartels, H.-W.; Honda, T.; Raeder, J.; Topilski, L.; Iseli, M.; Moshonas, K.; Taylor, N.; Gulden, W.; Kolbasov, B.; Inabe, T.; Tada, E.

    2001-01-01

    Safety has been an integral part of the design process for ITER since the Conceptual Design Activities of the project. The safety approach adopted in the ITER-FEAT design and the complementary assessments underway, to be documented in the Generic Site Safety Report (GSSR), are expected to help demonstrate the attractiveness of fusion and thereby set a good precedent for future fusion power reactors. The assessments address ITER's radiological hazards taking into account fusion's favourable safety characteristics. The expectation that ITER will need regulatory approval has influenced the entire safety design and assessment approach. This paper summarises the ITER-FEAT safety approach and assessments underway. (author)

  2. ITER council proceedings: 1997

    International Nuclear Information System (INIS)

    1997-01-01

    This volume of the ITER EDA Documentation Series presents records of the 12th ITER Council Meeting, IC-12, which took place on 23-24 July, 1997 in Tampere, Finland. The Council received from the Parties (EU, Japan, Russia, US) positive responses on the Detailed Design Report. The Parties stated their willingness to contribute to fulfil their obligations in contributing to the ITER EDA. The summary discussions among the Parties led to the consensus that in July 1998 the ITER activities should proceed for additional three years with a general intent to enable an efficient start of possible, future ITER construction

  3. Impurities effect on the swelling of neutron irradiated beryllium

    International Nuclear Information System (INIS)

    Donne, M.D.; Scaffidi-Argentina, F.

    1995-01-01

    An important factor controlling the swelling behaviour of fast neutron irradiated beryllium is the impurity content which can strongly affect both the surface tension and the creep strength of this material. Being the volume swelling of the old beryllium (early sixties) systematically higher than that of the more modem one (end of the seventies), a sensitivity analysis with the aid of the computer code ANFIBE (ANalysis of Fusion Irradiated BEryllium) to investigate the effect of these material properties on the swelling behaviour of neutron irradiated beryllium has been performed. Two sets of experimental data have been selected: the first one named Western refers to quite recently produced Western beryllium, whilst the second one, named Russian refers to relatively old (early sixties) Russian beryllium containing a higher impurity rate than the Western one. The results obtained with the ANFIBE Code were assessed by comparison with experimental data and the used material properties were compared with the data available in the literature. Good agreement between calculated and measured values has been found

  4. Thermogravimetric analysis of the beryllium/steam reaction

    Energy Technology Data Exchange (ETDEWEB)

    Druyts, Frank E-mail: fdruyts@sckcen.be; Iseghem, Pierre van

    2000-11-01

    In view of the safety assessment of new fusion reactor designs, kinetic data are needed on the beryllium/steam reaction. Therefore, thermogravimetric analysis was used to determine the reactivity of beryllium in steam as a function of temperature, irradiation history and porosity of the samples. To this purpose, reference unirradiated S-200 VHP beryllium samples were compared with specimens irradiated in the BR2 reactor up to fast neutron fluences (E>1 MeV) of respectively 1.6x10{sup 21} n cm{sup -2} (resulting in a helium content of 300 appm He and a theoretical density of 99.9%) and 4x10{sup 22} n cm{sup -2} (21000 appm He, 97.2% theoretical density). Kinetics were parabolic for all tested beryllium types at 600 deg. C. At 700 deg. C, kinetics were parabolic for the unirradiated and irradiated 99.9% dense beryllium, and accelerating/linear for the irradiated 97.2% material. At 800 deg. C, all samples showed accelerating/linear behaviour. There was no influence of porosity on the reaction rate of beryllium in steam within the limited investigated density range, except at 700 deg. C, where the measured reaction rate for the irradiated 97.2% dense samples is an order of magnitude higher than for the irradiated 99.9% dense specimens.

  5. Status of beryllium R and D in Japan

    International Nuclear Information System (INIS)

    Kawamura, H.; Ishida, K.

    2004-01-01

    Recently, several R and D program of beryllium for fusion are being promoted in Japan and community of beryllium study is growing up. In the R and D area of beryllium for solid breeding blanket, major subjects are beryllide application for prototype reactor, lifetime evaluation of neutron multiplier, impurity effect of beryllium and recycling of irradiated beryllium. Especially, the study of beryllide application has significant progress in these two years. The basic properties such as tritium inventory, oxidation behavior, steam interaction for stoichiometric Be 12 Ti fabricated by HIP (Hot Isostatic Pressing) have been studied and some advantages against beryllium were made clear. For manufacturing technology development, phase diagram and ductility improvement have been studied. And, Be 12 Ti pebbles with the improved microstructure were successfully fabricated by Rotating Electrode Process. In order to enhance the R and D activities, the R and D network consisted of industries, universities and laboratories in all Japan have been organized. Many collaboration and information exchange strongly promotes the R and D and some projects for commercial application have been launched form these activities. Also international collaborative project such as IEA and ISTC have been launched or planned. Recent result of R and D in Japan is described on this paper. (author)

  6. Beryllium armoured target for extreme heat and neutron loading conditions

    International Nuclear Information System (INIS)

    Mazul, I.; Gervash, A.; Giniyatulin, R.

    2004-01-01

    Beryllium is a primary candidate as a target material for high-energy protons conversion into neutrons used for different applications. In order to get higher neutron flux the conversion area has to be minimized - in our case the target is limited by 1-2 liter volume. This target generates about 5·10 13 fast neutrons per second and removes of 150 kW thermal power deposited by proton beam (30 mA, 5 MeV), coming from linac. The operational condition of the converter is close to the condition of Be-armored components in fusion reactors: high thermal and neutron fluxes and active cooling. Therefore achievements in development of water-cooled high heat flux components for fusion application can be used for design of Be converter and vice versa. However for medical application the using of high-activated heat sink materials such as Cu and SS is strongly limited. So, new materials (Be, Al, Zr) and new joining technologies in comparison with the achievements in fusion area have to be used for construction of such Be converter. In order to reduce amount of heat sink materials in the target saddle-block geometry for Be armor is suggested and developed. Results of R and D works on the development of water cooled Be target for converter are presented, including data on selected materials, technological trials and mockups high heat flux testing. Preliminary design of Be neutron converter for medical applications based on R and D results is presented. (author)

  7. Status of international collaborative efforts on selected ITER materials

    International Nuclear Information System (INIS)

    Belyakov, V.A.; Fabritsiev, S.A.; Mazul, I.V.; Rowcliffe, A.F.

    2000-01-01

    The paper presents an overview of the performance of refractory metals, beryllium, and copper alloys, for the international thermonuclear experimental reactor (ITER) high heat flux structures. High temperature brazing, hot isostatic pressing (HIP), friction welding, explosive bonding, and other methods were explored to join copper alloys to 316 stainless steel for first wall and limiter applications. It is concluded that the main material problems for the ITER high heat flux components are: (a) degradation of properties after the manufacturing cycle (especially for Be/Cu and Cu/stainless steel (SS) joints); (b) helium embrittlement of Be, and Cu, and; (c) radiation-induced loss of fracture toughness for Be, W, and Cu alloys

  8. Exposure and genetics increase risk of beryllium sensitisation and chronic beryllium disease in the nuclear weapons industry.

    Science.gov (United States)

    Van Dyke, Michael V; Martyny, John W; Mroz, Margaret M; Silveira, Lori J; Strand, Matt; Cragle, Donna L; Tankersley, William G; Wells, Susan M; Newman, Lee S; Maier, Lisa A

    2011-11-01

    Beryllium sensitisation (BeS) and chronic beryllium disease (CBD) are caused by exposure to beryllium with susceptibility affected by at least one well-studied genetic host factor, a glutamic acid residue at position 69 (E69) of the HLA-DPβ chain (DPβE69). However, the nature of the relationship between exposure and carriage of the DPβE69 genotype has not been well studied. The goal of this study was to determine the relationship between DPβE69 and exposure in BeS and CBD. Current and former workers (n=181) from a US nuclear weapons production facility, the Y-12 National Security Complex (Oak Ridge, Tennessee, USA), were enrolled in a case-control study including 35 individuals with BeS and 19 with CBD. HLA-DPB1 genotypes were determined by PCR-SSP. Beryllium exposures were assessed through worker interviews and industrial hygiene assessment of work tasks. After removing the confounding effect of potential beryllium exposure at another facility, multivariate models showed a sixfold (OR 6.06, 95% CI 1.96 to 18.7) increased odds for BeS and CBD combined among DPβE69 carriers and a fourfold (OR 3.98, 95% CI 1.43 to 11.0) increased odds for those exposed over an assigned lifetime-weighted average exposure of 0.1 μg/m(3). Those with both risk factors had higher increased odds (OR 24.1, 95% CI 4.77 to 122). DPβE69 carriage and high exposure to beryllium appear to contribute individually to the development of BeS and CBD. Among workers at a beryllium-using facility, the magnitude of risk associated with either elevated beryllium exposure or carriage of DPβE69 alone appears to be similar.

  9. Beryllium data base for in-pile mockup test on blanket of fusion reactor, (1)

    Energy Technology Data Exchange (ETDEWEB)

    Kawamura, Hiroshi; Ishitsuka, Etsuo (Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment); Sakamoto, Naoki; Kato, Masakazu; Takatsu, Hideyuki.

    1992-11-01

    Beryllium has been used in the fusion blanket designs with ceramic breeder as a neutron multiplier to increase the net tritium breeding ratio (TBR). The properties of beryllium, that is physical properties, chemical properties, thermal properties, mechanical properties, nuclear properties, radiation effects, etc. are necessary for the fusion blanket design. However, the properties of beryllium have not been arranged for the fusion blanket design. Therefore, it is indispensable to check and examine the material data of beryllium reported previously. This paper is the first one of the series of papers on beryllium data base, which summarizes the reported material data of beryllium. (author).

  10. Impact of risk analysis on ITER nuclear buildings layout

    International Nuclear Information System (INIS)

    Sabathe, Laurent; Lignini, Franck; Rodriguez-Rodrigo, Lina; Uzan-Elbez, Joelle; Girard, Jean-Philippe

    2007-01-01

    The internal layout of ITER nuclear buildings (Tokamak, Tritium, Hot cell and Radwaste buildings) is the result of an iterative process. Notwithstanding civil engineering requirements, technical and functional requirements for the processes housed in the buildings, the internal layout must also take into account safety requirements and workers protection requirements in normal, incidental and accidental conditions. Potential hazards and constraints result in the definition of various zonings. Each room is classified depending on the level of the risks induced by normal and incidental or accidental conditions: - radiological exposure to tritium, and activated products → radiological zoning; - exposure to beryllium (vacuum vessel blankets) → beryllium zoning; - exposure to electromagnetic fields during plasma shots → electromagnetic zoning; - fire hazard: fire compartmentalization; ·explosion hazard: anti-blast zoning; - waste treatment → waste zoning. Prevention of common mode failure due to internal hazards (fire, flooding, load drop...) for redundant safety important components is also taken into account by geographical separation or by installation of protections between components (either in the same room or preferably when possible by segregation in different rooms). The designer has also taken into account access requirements to equipments and circulation rules for personal, material and substances in ITER nuclear buildings. Risk analyses are performed in the Preliminary Safety Report in order to demonstrate that the layout of ITER nuclear buildings allows meeting the safety objectives and takes into account the requirements associated with the zonings classifications and complies with the regulation. When necessary, modifications are implemented in the design. For example, recent layout modifications are been studied in the tritium building: utility rooms (electrical supply) were transferred outside the radiological hazard zones. A direct access from

  11. Impact of risk analysis on ITER nuclear buildings internal layout

    International Nuclear Information System (INIS)

    Lignini, F.; Sabathe, L.; Rodriguez-Rodrigo, L.; Uzan-Elbez, J.; Girard, J.-P.

    2006-01-01

    The internal layout of ITER nuclear buildings (Tokamak, Tritium, Hot cell and Radwaste buildings) is the result of an iterative process. Notwithstanding civil engineering requirements, technical and functional requirements for the processes housed in the buildings, the internal layout must also take into account safety requirements and workers protection requirements in normal, incidental and accidental conditions. Potential hazards and constraints result in the definition of various zonings. Each room is classified depending on the level of risk induced by normal and incidental or accidental conditions : - radiological exposure to tritium, and activated products → radiological zoning - exposure to beryllium (vacuum vessel blankets) → beryllium zoning - exposure to electromagnetic fields during plasma shots → electromagnetic zoning - fire hazard : fire compartmentalisation - explosion hazard → anti-blast zoning - waste treatment → waste zoning Prevention of common mode failure due to internal hazards (fire, flooding, load drop ...) for redundant safety important components is also taken into account by geographical separation or by installation of protections between components (either in the same room or preferably when possible by segregation in different rooms). The designer also has to take into account access requirements to equipment and circulation rules for personal, material and substances in ITER nuclear buildings. Risk analyses are performed in the Preliminary Safety Report in order to demonstrate that the layout of ITER nuclear buildings allows to meet the safety objectives and takes into account the requirements associated with the zonings classifications and complies with the regulation. When necessary, modifications are implemented in the design. For example, recent layout modifications were included in the tritium building : utility rooms (electrical supply) were transferred outside the radiological hazard zones. A direct access from the

  12. Advanced scenarios for ITER operation

    Energy Technology Data Exchange (ETDEWEB)

    Sips, A.C.C. [Max-Planck-Institut fuer Plasmaphysik, Garching (Germany)

    2004-07-01

    In thermonuclear fusion research using magnetic confinement, the tokamak is the leading candidate for achieving conditions required for a reactor. An international experiment, ITER is proposed as the next essential and critical step on the path to demonstrating the scientific and technological feasibility of fusion energy. ITER is to produce and study plasmas dominated by self heating. This would give unique opportunities to explore, in reactor relevant conditions, the physics of {alpha}-particle heating, plasma turbulence and turbulent transport, stability limits to the plasma pressure and exhaust of power and particles. Important new results obtained in experiments, theory and modelling, enable an improved understanding of the physical processes occurring in tokamak plasmas and give enhanced confidence in ITER achieving its goals. In particular, progress has been made in research to raise the performance of tokamaks, aimed to extend the discharge pulse length towards steady-state operation (advanced scenarios). Standard tokamak discharges have a current density increasing monotonically towards the centre of the plasma. Advanced scenarios on the other hand use a modified current density profile. Different advanced scenarios range from (i) plasmas that sustain a central region with a flat current density profile (zero magnetic shear), capable of operating stationary at high plasma pressure, to (ii) discharges with an off axis maximum of the current density profile (reversed magnetic shear in the core), able to form internal transport barriers, to increase the confinement of the plasma. The physics of advanced tokamak discharges is described, together with an overview of recent results from different tokamak experiments. International collaboration between experiments aims to provide a better understanding, control and optimisation of these plasmas. The ability to explore advanced scenarios in ITER is very desirable, in order to verify the result obtained in

  13. Proton irradiation effects on beryllium – A macroscopic assessment

    Energy Technology Data Exchange (ETDEWEB)

    Simos, Nikolaos, E-mail: simos@bnl.gov [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Elbakhshwan, Mohamed [Nuclear Sciences & Technology Department, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Zhong, Zhong [Photon Sciences, NSLS II, Brookhaven National Laboratory, Upton, NY, 11973 (United States); Camino, Fernando [Center for Functional Nanomaterials, Brookhaven National Laboratory, Upton, NY, 11973 (United States)

    2016-10-15

    Beryllium, due to its excellent neutron multiplication and moderation properties, in conjunction with its good thermal properties, is under consideration for use as plasma facing material in fusion reactors and as a very effective neutron reflector in fission reactors. While it is characterized by unique combination of structural, chemical, atomic number, and neutron absorption cross section it suffers, however, from irradiation generated transmutation gases such as helium and tritium which exhibit low solubility leading to supersaturation of the Be matrix and tend to precipitate into bubbles that coalesce and induce swelling and embrittlement thus degrading the metal and limiting its lifetime. Utilization of beryllium as a pion production low-Z target in high power proton accelerators has been sought both for its low Z and good thermal properties in an effort to mitigate thermos-mechanical shock that is expected to be induced under the multi-MW power demand. To assess irradiation-induced changes in the thermal and mechanical properties of Beryllium, a study focusing on proton irradiation damage effects has been undertaken using 200 MeV protons from the Brookhaven National Laboratory Linac and followed by a multi-faceted post-irradiation analysis that included the thermal and volumetric stability of irradiated beryllium, the stress-strain behavior and its ductility loss as a function of proton fluence and the effects of proton irradiation on the microstructure using synchrotron X-ray diffraction. The mimicking of high temperature irradiation of Beryllium via high temperature annealing schemes has been conducted as part of the post-irradiation study. This paper focuses on the thermal stability and mechanical property changes of the proton irradiated beryllium and presents results of the macroscopic property changes of Beryllium deduced from thermal and mechanical tests.

  14. Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

    International Nuclear Information System (INIS)

    Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

    2002-01-01

    Kazakhstan takes part in ITER project in partnership with Russian Federation since the year of 1994. At present the technical stage of the project is completed and ITER Council should take a decision on the site for international reactor. Four countries such as Canada, Japan, Spain and France have offered their territories for being used as site for launching ITER construction. ITER partners started preparing new international agreement that will cover activities on construction, operation and decommissioning of ITER. It will also include the list of research and experimental work that is conducted in support of ITER project. Kazakhstan has already made an important contribution into technical stage realization of ITER project due to scientific and technical researches conducted by National Nuclear Center, by Institute of Experimental and Theoretical Physics and by JSC 'Ulba Metallurgical plant' ('UMP'). Research activity carried out for the support of ITER project is performed in accordance with the following main trends: Tritium safety (permeability and retentin of hydrogen isotopes during in-pile irradiation in various structural materials, co-deposed layers and protective coatings); Verification of computer codes (LOCA type) loss of coolant accidents modeling in ITER reactor; Investigation of liquid metal blanket of thermonuclear reactor (tritium production in lithium containing eutectics Li17Pb83 and ceramics Li 2 TiO 3 , study of tritium permeability). At present the working group of ITER project participants started introducing proposals for cost distribution and for placing the orders on reactor construction. Further Kazakhstan participation in ITER project may be in manufacturing high-tech parts and assemblies from commercial grades of beryllium. They will be used for armouring the reactor first wall, for its thermal protection and for protection of superconductor's components for magnetic systems that are at JSC U MP'. Scientific and technical support of

  15. ITER site selection studies in Spain

    International Nuclear Information System (INIS)

    Medrano, M.; Alejaldre, C.; Doncel, J.; Garcia, A.; Ibarra, A.; Jimenez, J.A.; Sanchez de Mora, M.A.; Alcala, F.; Diez, J.E.; Dominguez, M.; Albisu, F.

    2003-01-01

    The studies carried out to evaluate and select a candidate site for International Thermonuclear Experimental Reactor (ITER) construction in Spain are presented in this paper. The ITER design, completed in July 2001, considered a number of technical requirements that must be fulfilled by the selected site. Several assumptions concerning the ITER site were made in order to carry on the design before final site selection. In the studies undertaken for ITER site selection in Spain, the referred technical requirements and assumptions were applied across the whole of Spain and two areas were identified as being preferential. These areas are on the Mediterranean coast and are situated in the Catalan and Valencian regions. A comparative evaluation based on technical characteristics for the concrete plots, proposed within the preferential areas, has been done. The result of these studies was the selection of a site that was deemed to be the most competitive--Vandellos (Tarragona)--and it was proposed to the European Commission for detailed studies in order to be considered as a possible European site for ITER construction. Another key factor for hosting ITER in Spain, is the licensing process. The present status is summarised in this paper

  16. Iteration and accelerator dynamics

    International Nuclear Information System (INIS)

    Peggs, S.

    1987-10-01

    Four examples of iteration in accelerator dynamics are studied in this paper. The first three show how iterations of the simplest maps reproduce most of the significant nonlinear behavior in real accelerators. Each of these examples can be easily reproduced by the reader, at the minimal cost of writing only 20 or 40 lines of code. The fourth example outlines a general way to iteratively solve nonlinear difference equations, analytically or numerically

  17. Future plan of ITER

    International Nuclear Information System (INIS)

    Kitsunezaki, Akio

    1998-01-01

    In cooperation of four countries, Japan, USA, EU and Russia, ITER plan has been proceeding as ''the conceptual design activities'' from 1988 to 1990 and ''the industrial design activities'' since 1992. To construct ITER, the legal and work side of ITER operation has been investigated by four countries. However, their economic conditions have been changed to be wrong. So that, construction of ITER can not begin after end of industrial design activities in 1998. Accordingly, they determined to continue the industrial design activities more three years in order to study low cost options and to test the superconductive model·coil. (S.Y.)

  18. ITER test programme

    International Nuclear Information System (INIS)

    Abdou, M.; Baker, C.; Casini, G.

    1991-01-01

    ITER has been designed to operate in two phases. The first phase which lasts for 6 years, is devoted to machine checkout and physics testing. The second phase lasts for 8 years and is devoted primarily to technology testing. This report describes the technology test program development for ITER, the ancillary equipment outside the torus necessary to support the test modules, the international collaboration aspects of conducting the test program on ITER, the requirements on the machine major parameters and the R and D program required to develop the test modules for testing in ITER. 15 refs, figs and tabs

  19. Vapor pressure and thermodynamics of beryllium carbide

    International Nuclear Information System (INIS)

    Rinehart, G.H.; Behrens, R.G.

    1980-01-01

    The vapor pressure of beryllium carbide has been measured over the temperature range 1388 to 1763 K using Knudsen-effusion mass spectrometry. Vaporization occurs incongruently according to the reaction Be 2 C(s) = 2Be(g) + C(s). The equilibrium vapor pressure above the mixture of Be 2 C and C over the experimental temperature range is (R/J K -1 mol -1 )ln(p/Pa) = -(3.610 +- 0.009) x 10 5 (K/T) + (221.43 +- 1.06). The third-law enthalpy change for the above reaction obtained from the present vapor pressures is ΔH 0 (298.15 K) = (740.5 +- 0.1) kJ mol -1 . The corresponding second-law result is ΔH 0 (298.15 K) = (732.0 +- 1.8) kJ mol -1 . The enthalpy of formation for Be 2 C(s) calculated from the present third-law vaporization enthalpy and the enthalpy of formation of Be(g) is ΔH 0 sub(f)(298.15 K) = -(92.5 +- 15.7) kJ mol -1 . (author)

  20. Microstructural study of hydrogen-implanted beryllium

    International Nuclear Information System (INIS)

    Vagin, S.P.; Chakrov, P.V.; Utkelbayev, B.D.

    1998-01-01

    Hot pressed beryllium (TGP-56) was implanted by 650 keV H + ions to a dose of 6.7 x 10 16 cm -2 at a temperature below 50 C. TEM examinations were performed both at as-irradiated specimens and after post-irradiation annealings at 400-600 C for 15 min. After irradiation, a high density of ''black dot'' defects with a size of about 5 nm is observed in the straggling zone, some of which are resolved as small dislocation loops. During post-irradiation annealing, growth of dislocation loops and oriented gas-filled bubbles are observed in the damaged zone. The bubbles are strongly elongated along the left angle 0001 right angle direction, and their sidelong facts lie along {1-100} planes. These facets have a regular ''toothed'' surface with ''tooth'' facets on {1-100} planes. The size of the ''teeth'' increases with annealing temperature, as well as the total volume of bubbles, with their length growing faster than their width. (orig.)

  1. Steam-chemical reactivity for irradiated beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Anderl, R.A.; McCarthy, K.A.; Oates, M.A.; Petti, D.A.; Pawelko, R.J.; Smolik, G.R. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

    1998-01-01

    This paper reports the results of an experimental investigation to determine the influence of neutron irradiation effects and annealing on the chemical reactivity of beryllium exposed to steam. The work entailed measurements of the H{sub 2} generation rates for unirradiated and irradiated Be and for irradiated Be that had been previously annealed at different temperatures ranging from 450degC to 1200degC. H{sub 2} generation rates were similar for irradiated and unirradiated Be in steam-chemical reactivity experiments at temperatures between 450degC and 600degC. For irradiated Be exposed to steam at 700degC, the chemical reactivity accelerated rapidly and the specimen experienced a temperature excursion. Enhanced chemical reactivity at temperatures between 400degC and 600degC was observed for irradiated Be annealed at temperatures of 700degC and higher. This reactivity enhancement could be accounted for by the increased specific surface area resulting from development of a surface-connected porosity in the irradiated-annealed Be. (author)

  2. The unusual properties of beryllium surfaces

    International Nuclear Information System (INIS)

    Stumpf, R.; Hannon, J.B.

    1994-01-01

    Be is a ''marginal metal.'' The stable phase, hcp-Be, has a low Fermi-level density of states and very anisotropic structural and elastic properties, similar to a semiconductor's. At the Be(0001) surface, surface states drastically increase the Fermi-level density of states. The different nature of bonding in bulk-Be and at the Be(0001) surface explains the large outward relaxation. The presence of surface states causes large surface core-level shifts by inducing a higher electrostatic potential in the surface layers and by improving the screening at the surface. The authors experimental and theoretical investigations of atomic vibrations at the Be(0001) surface demonstrate clearly that Be screening of atomic motion by the surface states makes the surface phonon dispersion fundamentally different from that of the bulk. Properties of Be(0001) are so different from those of the bulk that the surface can be considered a new ''phase'' of beryllium with unique electronic and structural characteristics. For comparison they also study Be(11 bar 20), a very open surface without important surface states. Be(11 bar 20) is the only clean s-p metal surface known to reconstruct (1 x 3 missing row reconstruction)

  3. Beryllium parabolic refractive x-ray lenses

    International Nuclear Information System (INIS)

    Lengeler, B.; Schroer, C.G.; Kuhlmann, M.; Benner, B.; Guenzler, T.F.; Kurapova, O.; Somogyi, A.; Snigirev, A.; Snigireva, I.

    2004-01-01

    Parabolic refractive x-ray lenses are novel optical components for the hard x-ray range from about 5 keV to about 120 keV. They focus in both directions. They are compact, robust, and easy to align and to operate. They can be used like glass lenses are used for visible light, the main difference being that the numerical aperture N.A. is much smaller than 1 (of order 10-4 to 10-3). Their main applications are in micro- and nanofocusing, in imaging by absorption and phase contrast and in fluorescence mode. In combination with tomography they allow for 3-dimensional imaging of opaque media with submicrometer resolution. Finally, they can be used in speckle spectroscopy by means of coherent x-ray scattering. Beryllium as lens material strongly enhances the transmission and the field of view as compared to aluminium. With increased N.A. the lateral resolution is also considerably improved with Be lenses. References to a number of applications are given

  4. Allowable heat load on the edge of the ITER first wall panel beryllium flat tiles

    Directory of Open Access Journals (Sweden)

    R. Mitteau

    2017-08-01

    Full Text Available Plasma facing components are usually qualified to a given heat load density applied at the top face of the armour tiles with normal incidence angle. When employed in tokamak fusion machines, heat loading on the tile sides is possible due to optimised shaping, that doesn't provide edge shadowing for all design situations. An edge heat load may occur both at the tile and component scales. The edge load needs to be controlled and quantified. The adequate control of edge heat loads is especially critical for water cooled components that uses armour tiles which are bonded to the heat sink, for ensuring the long-term integrity of the tile bonding. An edge heat load allowance criterion of 10% of the top heat load is proposed. The 10% criterion is supported by experimental heat flux tests.

  5. Validation of cleaning method for various parts fabricated at a Beryllium facility

    Energy Technology Data Exchange (ETDEWEB)

    Davis, Cynthia M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-12-15

    This study evaluated and documented a cleaning process that is used to clean parts that are fabricated at a beryllium facility at Los Alamos National Laboratory. The purpose of evaluating this cleaning process was to validate and approve it for future use to assure beryllium surface levels are below the Department of Energy’s release limits without the need to sample all parts leaving the facility. Inhaling or coming in contact with beryllium can cause an immune response that can result in an individual becoming sensitized to beryllium, which can then lead to a disease of the lungs called chronic beryllium disease, and possibly lung cancer. Thirty aluminum and thirty stainless steel parts were fabricated on a lathe in the beryllium facility, as well as thirty-two beryllium parts, for the purpose of testing a parts cleaning method that involved the use of ultrasonic cleaners. A cleaning method was created, documented, validated, and approved, to reduce beryllium contamination.

  6. Controlling Beryllium Contaminated Material And Equipment For The Building 9201-5 Legacy Material Disposition Project

    International Nuclear Information System (INIS)

    Reynolds, T.D.; Easterling, S.D.

    2010-01-01

    This position paper addresses the management of beryllium contamination on legacy waste. The goal of the beryllium management program is to protect human health and the environment by preventing the release of beryllium through controlling surface contamination. Studies have shown by controlling beryllium surface contamination, potential airborne contamination is reduced or eliminated. Although there are areas in Building 9201-5 that are contaminated with radioactive materials and mercury, only beryllium contamination is addressed in this management plan. The overall goal of this initiative is the compliant packaging and disposal of beryllium waste from the 9201-5 Legacy Material Removal (LMR) Project to ensure that beryllium surface contamination and any potential airborne release of beryllium is controlled to levels as low as practicable in accordance with 10 CFR 850.25.

  7. Spectrographic determination of beryllium in the atmosphere; Dosage spectrographique du beryllium dans l'atmosphere

    Energy Technology Data Exchange (ETDEWEB)

    Soudain, G; Morawek, T [Commissariat a l' Energie Atomique, Saclay (France).Centre d' Etudes Nucleaires

    1959-07-01

    Since the apparatus for continuous determination of beryllium is not yet perfect, a discontinuous method has been developed. The air to be analysed is filtered, and the dust laden filter is dissolved in a mixture of sulphuric and nitric acid. The pH and the conductivity of the solution obtained were adjusted to standard values, and it was then analysed spectro-graphically by the rotating sector method. Up to 0.01 x 10{sup 6} of Be per cm{sup 3} of solution can be detected. The precision is of the order of 10 per cent. (author) [French] Les appareils de dosage du beryllium en continu n'etant pas encore suffisamment au point, on a elabore une methode discontinue. L'air a analyser est filtre et le filtre charge de poussieres est mis en solution par une attaque sulfo-nitrique. La solution obtenue est normalisee par ajustage de son PH et de sa conductivite puis analysee spectrographiquement par la methode du disque tournant. On peut detecter jusqu'a 0,01.10{sup 6} de Be par cm{sup 3} de solution. La precision est de l'ordre de 10 pour cent. (auteur)

  8. Preparation of copper-beryllium alloys from Indian beryl

    International Nuclear Information System (INIS)

    Paul, C.M.; Sharma, B.P.; Subba Rao, K.S.; Rajadhyaksha, M.G.; Sundaram, C.V.

    1975-01-01

    The report presents the results of laboratory scale investigations on the preparation of copper-beryllium and aluminium-beryllium master alloys starting from Indian beryl and adopting the fluoride process. The flow-sheet involves : (1) conversion of the Be-values in beryl into water soluble sodium beryllium fluoride (2) preparation of beryllium hydroxide by alkali treatment of aqueous Na 2 BeF 4 (3) conversion of Be(OH) 2 to (NH 4 ) 2 BeF 4 by treatment with NH 4 HF 2 (4) thermal decomposition of (NH 4 ) 2 BeF 4 to BeF 2 and (5) magnesium reduction of BeF 2 (with the addition of copper/aluminium) to obtain beryllium alloys. The method has been successfully employed for the preparation of Cu-Be master alloys containing about 8% Be and free of Mg on a 200 gm scale. An overall Be-recovery of about 80% has been achieved. Al-8% Be master alloys have also been prepared by this method. Toxicity and health hazards associated with Be are discussed and the steps taken to ensure safe handling of Be are described. (author)

  9. Compatibility of stainless steels and lithiated ceramics with beryllium

    Science.gov (United States)

    Flament, T.; Fauvet, P.; Sannier, J.

    1988-07-01

    The introduction of beryllium as a neutron multiplier in ceramic blankets of thermonuclear fusion reactors may give rise to the following compatibility problems: (i) oxidation of Be by ceramics (lithium aluminate and silicates) or by water vapour; (ii) interaction between beryllium and austenitic and martensitic steels. The studies were done in contact tests under vacuum and in tests under wet sweeping helium. The contact tests under vacuum have revealed that the interaction of beryllium with ceramics seems to be low up to 700°C, the interaction of beryllium with steels is significant and is characterized by the formation of a diffusion layer and of a brittle Be-Fe-Ni compound. With type 316 L austenitic steel, this interaction appears quite large at 600°C whereas it is noticeable only at 700°C with martensitic steels. The experiments carried out with sweeping wet helium at 600°C have evidenced a slight oxidation of beryllium due to water vapour which can be enhanced in the front of uncompletely dehydrated ceramics.

  10. Preparation of copper-beryllium alloys from Indian beryl

    International Nuclear Information System (INIS)

    Paul, C.M.; Sharma, B.P.; Subba Rao, K.S.; Rajadhyaksha, M.G.; Sundaram, C.V.

    1975-01-01

    The paper presents the results of laboratory-scale investigations on the preparation of copper-beryllium and aluminium beryllium master alloys starting from Indian beryl and adopting the fluoride process. The flowsheet involves: (1) conversion of the Be-values in beryl into water soluble sodium beryllium fluoride, (2) preparation of beryllium hydroxide by alkali treatment of aqueous Na 2 BeF 4 (3) conversion of Be(OH) 2 to (NH 4 ) 2 BeF 4 by treatment with NH 4 HF 2 (4) thermal decomposition of (NH 4 ) 2 BeF 4 to BeF 2 and (5) magnesium reduction of BeF 2 (without/with) the addition of copper/aluminium to obtain beryllium metal/alloys. The method has been successfully employed for the preparation of Cu-Be master alloys containing about 8% Be and free of Mg on a 200 gm scale. A1-80% Be master alloys have also been prepared by this method. Toxicity and health hazards associated with Be are discussed and the steps taken to ensure safe handling of Be are described. (author)

  11. Structure/property relationships in multipass GMA welding of beryllium.

    Energy Technology Data Exchange (ETDEWEB)

    Hochanadel, P. W. (Patrick W.); Hults, W. L. (William L.); Thoma, D. J. (Dan J.); Dave, V. R. (Vivek R.); Kelly, A. M. (Anna Marie); Pappin, P. A. (Pallas A.); Cola, M. J. (Mark J.); Burgardt, P. (Paul)

    2001-01-01

    Beryllium is an interesting metal that has a strength to weight ratio six times that of steel. Because of its unique mechanical properties, beryllium is used in aerospace applications such as satellites. In addition, beryllium is also used in x-ray windows because it is nearly transparent to x-rays. Joining of beryllium has been studied for decades (Ref.l). Typically joining processes include braze-welding (either with gas tungsten arc or gas metal arc), soldering, brazing, and electron beam welding. Cracking which resulted from electron beam welding was recently studied to provide structure/property relationships in autogenous welds (Ref. 2). Braze-welding utilizes a welding arc to melt filler, and only a small amount of base metal is melted and incorporated into the weld pool. Very little has been done to characterize the braze-weld in terms of the structure/property relationships, especially with reference to multipass welding. Thus, this investigation was undertaken to evaluate the effects of multiple passes on microstructure, weld metal composition, and resulting material properties for beryllium welded with aluminum-silicon filler metal.

  12. The results of medical surveillance of beryllium production personnel

    International Nuclear Information System (INIS)

    Koviazin, A.; Urikh, A.; Kovianzina, L.

    2004-01-01

    The report presents results of surveillance of 1836 workers of beryllium production of Ulba Metallurgical Plant JSC with the acute and chronic forms of occupation diseases for 52 years of its operation. The dependence of acute and chronic occupation lesions on the protection degree is shown. It has been found out that, the risk of getting an occupation disease increases sharply at the moments of experimental works and at the time of reconstruction and some other extreme conditions in the production, that is supported by fixed lesions of eye mucous coat, skin and lung lesions. In this case, the readiness of people for their work in deleterious conditions and their personal responsibility for following the regulations of safety occupational standards plays a definite role. Therefore, the issues of protection are of paramount importance in prophylaxis both of acute and chronic exposure to beryllium. An influence of duration of service and occupation on chronic beryllium diseases is shown. A parallel between the lung beryllium disease and skin lesions by insoluble beryllium compounds is drawn for the first time. (author)

  13. Experimental studies and modeling of processes of hydrogen isotopes interaction with beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Tazhibaeva, I.L.; Chikhray, Y.V.; Romanenko, O.G.; Klepikov, A.Kh.; Shestakov, V.P.; Kulsartov, T.V. [Science Research Inst. of Experimental and Theoretical Physics of Kazakh State Univ., Almaty (Kazakhstan); Kenzhin, E.A.

    1998-01-01

    The objective of this work was to clarify the surface beryllium oxide influence on hydrogen-beryllium interaction characteristics. Analysis of experimental data and modeling of processes of hydrogen isotopes accumulation, diffusion and release from neutron irradiated beryllium was used to achieve this purpose as well as the investigations of the changes of beryllium surface element composition being treated by H{sup +} and Ar{sup +} plasma glowing discharge. (author)

  14. United States rejoin ITER

    International Nuclear Information System (INIS)

    Roberts, M.

    2003-01-01

    Upon pressure from the United States Congress, the US Department of Energy had to withdraw from further American participation in the ITER Engineering Design Activities after the end of its commitment to the EDA in July 1998. In the years since that time, changes have taken place in both the ITER activity and the US fusion community's position on burning plasma physics. Reflecting the interest in the United States in pursuing burning plasma physics, the DOE's Office of Science commissioned three studies as part of its examination of the option of entering the Negotiations on the Agreement on the Establishment of the International Fusion Energy Organization for the Joint Implementation of the ITER Project. These were a National Academy Review Panel Report supporting the burning plasma mission; a Fusion Energy Sciences Advisory Committee (FESAC) report confirming the role of ITER in achieving fusion power production, and The Lehman Review of the ITER project costing and project management processes (for the latter one, see ITER CTA Newsletter, no. 15, December 2002). All three studies have endorsed the US return to the ITER activities. This historical decision was announced by DOE Secretary Abraham during his remarks to employees of the Department's Princeton Plasma Physics Laboratory. The United States will be working with the other Participants in the ITER Negotiations on the Agreement and is preparing to participate in the ITA

  15. Lawrence Livermore Laboratory's beryllium control program for high-explosive test firing bunkers and tables

    International Nuclear Information System (INIS)

    Johnson, J.S.

    1978-01-01

    This detailed report on Lawrence Livermore Laboratory's control program to minimize beryllium levels in Laboratory workplaces includes an outline of beryllium surface, soil, and air levels and an 11-y summary of sampling results from two high-use, high-explosive test firing bunkers. These sampling data and other studies demonstrate that the beryllium control program is funcioning effectively

  16. Lawrence Livermore Laboratory's beryllium control program for high-explosive test firing bunkers and tables

    International Nuclear Information System (INIS)

    Johnson, J.S.

    1980-01-01

    This report on the control program to minimize beryllium levels in Laboratory workplaces includes an outline of beryllium surface, soil, and air levels and an 11-y summary of sampling results from two high-use, high-explosive test firing bunkers. These sampling data and other studies demonstrate that the beryllium control program is functioning effectively

  17. The Effect Of Beryllium Interaction With Fast Neutrons On the Reactivity Of ETRR-2 Research Reactor

    International Nuclear Information System (INIS)

    Aziz, M.; El Messiry, A.M.

    2000-01-01

    The effect of beryllium interactions with fast neutrons is studied for Etrr 2 research reactors. Isotope build up inside beryllium blocks is calculated under different irradiation times. a new model for the Etrr 2 research reactor is designed using MCNP code to calculate the reactivity and flux change of the reactor due to beryllium poison

  18. ITER CTA newsletter. No. 3

    International Nuclear Information System (INIS)

    2001-11-01

    This ITER CTA newsletter comprises reports of Dr. P. Barnard, Iter Canada Chairman and CEO, about the progress of the first formal ITER negotiations and about the demonstration of details of Canada's bid on ITER workshops, and Dr. V. Vlasenkov, Project Board Secretary, about the meeting of the ITER CTA project board

  19. ITER at Cadarache; ITER a Cadarache

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-06-15

    This public information document presents the ITER project (International Thermonuclear Experimental Reactor), the definition of the fusion, the international cooperation and the advantages of the project. It presents also the site of Cadarache, an appropriate scientifical and economical environment. The last part of the documentation recalls the historical aspect of the project and the today mobilization of all partners. (A.L.B.)

  20. ITER council proceedings: 1992

    International Nuclear Information System (INIS)

    1994-01-01

    At the signing of the ITER EDA Agreement on July, 1992, each of the Parties presented to the Director General the names of their designated members of the ITER Council. Upon receiving those names, the Director General stated that the ITER Engineering Design Activities were ''ready to begin''. The next step in this process was the convening of the first meeting of the ITER Council. The first meeting of the Council, held in Vienna, was opened by Director General Hans Blix. The second meeting was held in Moscow, the formal seat of the Council. This volume presents records of these first two Council meetings and, together with the previous volumes on the text of the Agreement and Protocol 1 and the preparations for their signing respectively, represents essential information on the evolution of the ITER EDA

  1. The uses and adverse effects of beryllium on health

    DEFF Research Database (Denmark)

    Cooper, Ross G.; Harrison, Adrian Paul

    2009-01-01

    published in sources unobtainable through requests at the British Library, and some had no impact factor and were excluded. Conclusion: Beryllium has some useful but undoubtedly harmful effects on health and well-being. Measures needed to be taken to prevent hazardous exposure to this element, making its......Context: This review describes the health effects of beryllium exposure in the workplace and the environment. Aim: To collate information on the consequences of occupational and environmental exposure to beryllium on physiological function and well being. Materials and Methods: The criteria used...... tabulated. Years 2001-10 gave the greatest match (45.9%) for methodological parameters, followed by 27.71% for 1991-2000. Years 1971-80 and 1981-90 were not significantly different in the information published and available whereas years 1951-1960 showed a lack of suitable articles. Some articles were...

  2. On modeling of beryllium molten depths in simulated plasma disruptions

    International Nuclear Information System (INIS)

    Tsotridis, G.; Rother, H.

    1996-01-01

    Plasma-facing components in tokamak-type fusion reactors are subjected to intense heat loads during plasma disruptions. The influence of high heat fluxes on the depth of heat-affected zones of pure beryllium metal and beryllium containing very low levels of surface active impurities is studied by using a two-dimensional transient computer model that solves the equations of motion and energy. Results are presented for a range of energy densities and disruption times. Under certain conditions, impurities, through their effect on surface tension, create convective flows and hence influence the flow intensities and the resulting depths of the beryllium molten layers during plasma disruptions. The calculated depths of the molten layers are also compared with other mathematical models that are based on the assumption that heat is transported through the material by conduction only. 32 refs., 6 figs., 1 tab

  3. On kinetics of paramagnetic radiation defects accumulation in beryllium ceramics

    International Nuclear Information System (INIS)

    Polyakov, A.I.; Ryabikin, Yu.A.; Zashkvara, O.V.; Bitenbaev, M.I.; Petykhov, Yu.V.

    1999-01-01

    Results of paramagnetic radiation defects concentration dependence study in beryllium ceramics from gamma-irradiation dose ( 60 Co) within interval 0-100 Mrem are cited. Obtained dose dependence has form of accumulation curve with saturation typical of for majority of solids (crystals, different polymers, organic substances and others) , in which under irradiation occur not only formation of paramagnetic radiation defects, but its destruction due to recombination and interaction with radiation fields. Analysis of accumulation curve by the method of distant asymptotics allows to determine that observed in gamma-irradiated beryllium ceramics double line of electron spin resonance is forming of two types of paramagnetic radiation defects. It was defined, that sum paramagnetic characteristics of beryllium ceramics within 1-100 Mrad gamma- irradiation dose field change insignificantly and define from first type of paramagnetic radiation defects

  4. Ab Initio Simulation Beryllium in Solid Molecular Hydrogen: Elastic Constant

    Science.gov (United States)

    Guerrero, Carlo L.; Perlado, Jose M.

    2016-03-01

    In systems of inertial confinement fusion targets Deuterium-Tritium are manufactured with a solid layer, it must have specific properties to increase the efficiency of ignition. Currently there have been some proposals to model the phases of hydrogen isotopes and hence their high pressure, but these works do not allow explaining some of the structures present at the solid phase change effect of increased pressure. By means of simulation with first principles methods and Quantum Molecular Dynamics, we compare the structural difference of solid molecular hydrogen pure and solid molecular hydrogen with beryllium, watching beryllium inclusion in solid hydrogen matrix, we obtain several differences in mechanical properties, in particular elastic constants. For C11 the difference between hydrogen and hydrogen with beryllium is 37.56%. This may produce a non-uniform initial compression and decreased efficiency of ignition.

  5. Extraction of beryllium sulfate by a long chain amine

    International Nuclear Information System (INIS)

    Etaix, E.S.

    1968-01-01

    The extraction of sulfuric acid in aqueous solution by a primary amine in benzene solution, 3-9 (diethyl) - 6-amino tri-decane (D.E.T. ) - i.e., with American nomenclature 1-3 (ethyl-pentyl) - 4-ethyl-octyl amine (E.P.O.) - has made it possible to calculate the formation constants of alkyl-ammonium sulfate and acid sulfate. The formula of the beryllium and alkyl-ammonium sulfate complex formed in benzene has next been determined, for various initial acidity of the aqueous solution. Lastly, evidence has been given of negatively charged complexes of beryllium and sulfate in aqueous solution, through the dependence of the aqueous sulfate ions concentration upon beryllium extraction. The formation constant of these anionic complexes has been evaluated. (author) [fr

  6. Corrosion of beryllium exposed to celotex and water

    International Nuclear Information System (INIS)

    Hill, M.A.; Butt, D.P.; Lillard, R.S.

    1997-01-01

    Celotex is a commercial rigid cellulose fiberboard product primarily used in the building construction industry. Currently celotex is being used as a packing material in AL-R8 containers. Ion chromatography of celotex packing material at Lawrence Livermore National Laboratory (LLNL) has indicated that this material contains aggressive anions, including chloride, which may accelerate corrosion. It is well known that beryllium is susceptible to pitting corrosion when exposed to chloride containing environments. Levy noted pitting in beryllium at the open circuit potential when exposed to 0.1 M NaCl solution. This investigation attempts to evaluate the potential risk of accelerated beryllium corrosion from celotex and water which may occur naturally when celotex dust comes into contact with moisture from the atmosphere

  7. Full Dynamic Analysis of Mooring Solution Candidates - First Iteration

    DEFF Research Database (Denmark)

    Thomsen, Jonas Bjerg; Ferri, Francesco

    This report covers an initial full dynamic analysis of the mooring solutions for the four wave energy converters in the project “Mooring Solutions for Large Wave Energy Converters”. The analysis tends to provide the first understanding of the layouts and provide discussion on what parameters that...

  8. Overview of erosion–deposition diagnostic tools for the ITER-Like Wall in the JET tokamak

    International Nuclear Information System (INIS)

    Rubel, M.; Coad, J.P.; Widdowson, A.; Matthews, G.F.; Esser, H.G.; Hirai, T.; Likonen, J.; Linke, J.; Lungu, C.P.; Mayer, M.; Pedrick, L.; Ruset, C.

    2013-01-01

    This paper presents scientific and technical issues related to the development of erosion–deposition diagnostic tools for JET operated with the ITER-Like Wall: beryllium and tungsten marker tiles and several types of wall probes installed in the main chamber and in the divertor. Markers tiles are the standard limiter and divertor components additionally coated first with a thin sandwich of Ni–Be and Mo–W for, beryllium and tungsten markers, respectively. Both types of markers are embedded in regular arrays of limiter and divertor tiles. Coated W–Be probes are also inserted in the Be-covered Inconel cladding tiles on the central column. Other types of erosion–deposition diagnostic tools are: rotating collectors, deposition traps, louver clips, quartz microbalance and mirrors for the First Mirror Test at JET for ITER. The specific role of these tools is discussed in detail

  9. Preparation of a sinterable beryllium oxide through decomposition of beryllium hydroxide (1963); Preparation d'un oxyde de beryllium frittable par decomposition de l'hydiloxyde (1963)

    Energy Technology Data Exchange (ETDEWEB)

    Bernier, M [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1963-07-01

    In the course of the present study, we have attempted to precise the factors which among the ones effective in the course of the preparation of the beryllium hydroxide and oxide and during the sintering have an influence on the final result: the density and homogeneity of the sintered body. Of the several varieties of hydroxides precipitated from a sulfate solution the {beta}-hydroxide only is always contaminated with beryllium sulfate and cannot be purified even by thorough washing. We noticed that those varieties of the hydroxide (gel, {alpha}, {beta}) have different decomposition rates; this behaviour is used to identify and even to dose the different species in ({alpha}, {beta}) mixtures. The various hydroxides transmit to the resulting oxides the shape they had when precipitated. Accordingly the history of the oxide is revealed by its behaviour during its fabrication and sintering. By comparing the results of the sintering operation with the various measurements performed on the oxide powders we are led to the conclusion that an oxide obtained from beryllium hydroxide is sinterable under vacuum if the following conditions are fulfilled: the particle size must lie between 0.1 and 0.2 {mu} and the BeSO{sub 4} content of the powder must be less than 0.25 per cent wt (expressed as SO{sub 3}/BeO). The best fitting is obtained with the oxide issued from an {alpha}-hydroxide precipitated as very small aggregates and with a low sulfur-content. We have observed that this is also the case for the oxide obtained by direct calcination of beryllium sulfate. (author) [French] Au cours de cette etude, nous avons cherche a preciser les facteurs qui, intervenant tout au long de la preparation de l'hydroxyde, puis de l'oxyde de beryllium et enfin du frittage, peuvent avoir une influence sur le resultat final: la densite et l'homogeneite du fritte. Parmi tous les hydroxydes precipites d'une solution de sulfate, seul l'hydroxyde {beta} est toujours fortement pollue par le sulfate

  10. Characterization of beryllium foil produced by hot rolling

    International Nuclear Information System (INIS)

    Wittenauer, J.; Nieh, T.G.; Waychunas, G.

    1992-01-01

    Beryllium foil is important for a number of aerospace applications including honeycomb structures and metal-matrix composites. In this study, a method of producing beryllium foil directly from powder or flake is demonstrated. A variety of foils were produced in the thickness range 90-300 μm, free from defects such as pinholes and excessive surface roughness, and exhibiting sufficient formability for honeycomb manufacture. Foil produced directly from powder or flake exhibits crystallographic texture, microstructure, and formability equivalent to foil produced from more massive precursors. (Author)

  11. CHAPTER 7. BERYLLIUM ANALYSIS BY NON-PLASMA BASED METHODS

    Energy Technology Data Exchange (ETDEWEB)

    Ekechukwu, A

    2009-04-20

    The most common method of analysis for beryllium is inductively coupled plasma atomic emission spectrometry (ICP-AES). This method, along with inductively coupled plasma mass spectrometry (ICP-MS), is discussed in Chapter 6. However, other methods exist and have been used for different applications. These methods include spectroscopic, chromatographic, colorimetric, and electrochemical. This chapter provides an overview of beryllium analysis methods other than plasma spectrometry (inductively coupled plasma atomic emission spectrometry or mass spectrometry). The basic methods, detection limits and interferences are described. Specific applications from the literature are also presented.

  12. Factors influencing the creep strength of hot pressed beryllium

    International Nuclear Information System (INIS)

    Webster, D.; Crooks, D.D.

    1975-01-01

    The parameters controlling the creep strength of hot pressed beryllium block have been determined. Creep strength was improved by a high initial dislocation density, a coarse grain size, and a low impurity content. The impurities most detrimental to creep strength were found to be aluminum, magnesium, and silicon. A uniform distribution of BeO was found to give creep strength which was inferior to a grain boundary distribution. The creep strength of very high purity, hot isostatically pressed beryllium was found to compare favorably with that of other more commonly used high temperature metals

  13. Method for removal of beryllium contamination from an article

    Science.gov (United States)

    Simandl, Ronald F.; Hollenbeck, Scott M.

    2012-12-25

    A method of removal of beryllium contamination from an article is disclosed. The method typically involves dissolving polyisobutylene in a solvent such as hexane to form a tackifier solution, soaking the substrate in the tackifier to produce a preform, and then drying the preform to produce the cleaning medium. The cleaning media are typically used dry, without any liquid cleaning agent to rub the surface of the article and remove the beryllium contamination below a non-detect level. In some embodiments no detectible residue is transferred from the cleaning wipe to the article as a result of the cleaning process.

  14. Characterization of Beryllium Windows for Coherent X-ray Optics

    International Nuclear Information System (INIS)

    Goto, Shunji; Yabashi, Makina; Tamasaku, Kenji; Ishikawa, Tetsuya

    2007-01-01

    Beryllium foils fabricated by several processes were characterized using spatially coherent x rays at 1-km beamline of SPring-8. By thickness dependence of bright x-ray spot density due to Fresnel diffraction from several-micron deficiencies, we found that speckles (bright x-ray spots) were due to voids with densities 103-104 mm-3 in powder foils and ingot foils. Compared with powder and ingot foils, a polished physical-vapor-deposited (PVD) beryllium foil gave highly uniform beams with no speckles. The PVD process eliminates the internal voids in principle and the PVD foil is the best for coherent x-ray applications

  15. Measurements of fusion neutron multiplication in spherical beryllium shells

    International Nuclear Information System (INIS)

    Giese, H.; Kappler, F.; Tayama, R.; Moellendorff, U. von; Alevra, A.; Klein, H.

    1996-01-01

    New results of spherical-shell transmission measurements with 14-MeV neutrons on pure beryllium shells up to 17 cm thick are reported. The spectral flux above 3 MeV was measured using a liquid scintillation detector. At 17 cm thickness, also the total neutron multiplication was measured using a Bonner sphere system. The results agree well with calculations using beryllium nuclear data from the EFF-1 or the ENDF/B-Vi library. (author). 23 refs, 4 figs, 1 tab

  16. Plasma cleaning of ITER edge Thomson scattering mock-up mirror in the EAST tokamak

    Science.gov (United States)

    Yan, Rong; Moser, Lucas; Wang, Baoguo; Peng, Jiao; Vorpahl, Christian; Leipold, Frank; Reichle, Roger; Ding, Rui; Chen, Junling; Mu, Lei; Steiner, Roland; Meyer, Ernst; Zhao, Mingzhong; Wu, Jinhua; Marot, Laurent

    2018-02-01

    First mirrors are the key element of all optical and laser diagnostics in ITER. Facing the plasma directly, the surface of the first mirrors could be sputtered by energetic particles or deposited with contaminants eroded from the first wall (tungsten and beryllium), which would result in the degradation of the reflectivity. The impurity deposits emphasize the necessity of the first mirror in situ cleaning for ITER. The mock-up first mirror system for ITER edge Thomson scattering diagnostics has been cleaned in EAST for the first time in a tokamak using radio frequency capacitively coupled plasma. The cleaning properties, namely the removal of contaminants and homogeneity of cleaning were investigated with molybdenum mirror insets (25 mm diameter) located at five positions over the mock-up plate (center to edge) on which 10 nm of aluminum oxide, used as beryllium proxy, were deposited. The cleaning efficiency was evaluated using energy dispersive x-ray spectroscopy, reflectivity measurements and x-ray photoelectron spectroscopy. Using argon or neon plasma without magnetic field in the laboratory and with a 1.7 T magnetic field in the EAST tokamak, the aluminum oxide films were homogeneously removed. The full recovery of the mirrors’ reflectivity was attained after cleaning in EAST with the magnetic field, and the cleaning efficiency was about 40 times higher than that without the magnetic field. All these results are promising for the plasma cleaning baseline scenario of ITER.

  17. ITER towards the construction

    International Nuclear Information System (INIS)

    Shimomura, Y.

    2005-01-01

    The ITER Project has been significantly developed in the last few years in preparation for its construction. The ITER Participant's Negotiators have developed the Joint Implementation Agreement (JIA), ready for finalisation following selection of the construction site and nomination of the project's Director General. The ITER International Team and Participant Teams have continued technical and organisational preparations. Construction will be able to start immediately after the international ITER organisation is established, following signature of the JIA. The Project is strongly supported by the governments of the Participants as well as by the scientific community. The real negotiations, including siting and the final details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. Their wish to host ITER construction is too strong to allow convergence to a single site considering the ITER device in isolation. A broader collaboration among the Parties is therefore being contemplated, covering complementary activities to help accelerate fusion development towards a viable power source, and allow the Participants to reach a conclusion on ITER siting. This report reviews these preparations, and the status of negotiations

  18. Some aspects of the control and management of air pollution by beryllium aerosols

    International Nuclear Information System (INIS)

    Noor, B.

    2000-01-01

    The concentration of beryllium aerosols in air of rooms and environment is determined by technological operations which are carried out on ITER and can be rather various. It determines differentiated the approaches to means and methods of the air pollution control. It is expedient to classify concentration Be, arising in air. Criterion of classification can be concentration of an impurity and its exposition. In tests of various duration are determined different on value of concentration Be in air of a working zone. At non-stationary processes of allocation the short-term tests have significant dispersion and allow to reveal both low and high meanings of concentration. This difference in conditions of astable allocation Be in air of a working zone can make the order and more. The long test in many respects smooths possible bursts of concentration, that does not allow to estimate real danger of technological operation or process. The high concentration Be can render decisive toxic astion on a human organism at high levels of pollution. The presence of submicronal particles Be, formed at work ITER presumably should strengthen effect of influence on organism. It is possible to explain it to that alongside with traditional toxic effect of aerosols Be, in addition can impose the mechanism action of a submicronal dust. With the purpose of increase information of tests and opportunity of the forecast of the air pollution levels the mathematical model of logarithmic normal distribution can be used. Submitting data of sampling in logarithmic probable grid, it is possible to carry out the analysis on two parameters: the median and the logarithmic standard. The existing standards and rules of the contents Be with reference to conditions ITER are expedient for differentiating with the account toxicity, exposition, affection on a human organism. The distinctions in allowable concentration Be in air on different sites in rooms allow to create more effective and economic methods

  19. Extraction of beryllium sulfate by a long chain amine; Extraction du sulfate de beryllium par une amine a longue chaine

    Energy Technology Data Exchange (ETDEWEB)

    Etaix, E S [Commissariat a l' Energie Atomique, Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

    1968-06-01

    The extraction of sulfuric acid in aqueous solution by a primary amine in benzene solution, 3-9 (diethyl) - 6-amino tri-decane (D.E.T. ) - i.e., with American nomenclature 1-3 (ethyl-pentyl) - 4-ethyl-octyl amine (E.P.O.) - has made it possible to calculate the formation constants of alkyl-ammonium sulfate and acid sulfate. The formula of the beryllium and alkyl-ammonium sulfate complex formed in benzene has next been determined, for various initial acidity of the aqueous solution. Lastly, evidence has been given of negatively charged complexes of beryllium and sulfate in aqueous solution, through the dependence of the aqueous sulfate ions concentration upon beryllium extraction. The formation constant of these anionic complexes has been evaluated. (author) [French] L'etude de l'extraction de l'acide sulfurique en solution aqueuse par une amine primaire en solution dans le benzene, le diethyl-3,9 amino-6 tridecane (D.E.T.) - autre nom americain 1-3 (ethylpentyl) - 4-ethyloctylamine (E.P.O.) a permis de calculer les constantes de formation du sulfate et de l'hydrogenosulfate d'alkyl-ammonium. La formule du complexe de sulfate de beryllium et d'alkyl-ammonium forme en solution benzenique a ete ensuite determinee pour diverses acidites initiales de la solution aqueuse. Enfin, l'influence de la concentration des ions sulfate de la phase aqueuse sur l'extraction du beryllium a mis en evidence la formation en solution aqueuse de complexes anioniques de sulfate et de beryllium dont la constante de formation a ete evaluee. (auteur)

  20. Extraction of beryllium sulfate by a long chain amine; Extraction du sulfate de beryllium par une amine a longue chaine

    Energy Technology Data Exchange (ETDEWEB)

    Etaix, E.S. [Commissariat a l' Energie Atomique, Fontenay-Aux-Roses (France). Centre d' Etudes Nucleaires

    1968-06-01

    The extraction of sulfuric acid in aqueous solution by a primary amine in benzene solution, 3-9 (diethyl) - 6-amino tri-decane (D.E.T. ) - i.e., with American nomenclature 1-3 (ethyl-pentyl) - 4-ethyl-octyl amine (E.P.O.) - has made it possible to calculate the formation constants of alkyl-ammonium sulfate and acid sulfate. The formula of the beryllium and alkyl-ammonium sulfate complex formed in benzene has next been determined, for various initial acidity of the aqueous solution. Lastly, evidence has been given of negatively charged complexes of beryllium and sulfate in aqueous solution, through the dependence of the aqueous sulfate ions concentration upon beryllium extraction. The formation constant of these anionic complexes has been evaluated. (author) [French] L'etude de l'extraction de l'acide sulfurique en solution aqueuse par une amine primaire en solution dans le benzene, le diethyl-3,9 amino-6 tridecane (D.E.T.) - autre nom americain 1-3 (ethylpentyl) - 4-ethyloctylamine (E.P.O.) a permis de calculer les constantes de formation du sulfate et de l'hydrogenosulfate d'alkyl-ammonium. La formule du complexe de sulfate de beryllium et d'alkyl-ammonium forme en solution benzenique a ete ensuite determinee pour diverses acidites initiales de la solution aqueuse. Enfin, l'influence de la concentration des ions sulfate de la phase aqueuse sur l'extraction du beryllium a mis en evidence la formation en solution aqueuse de complexes anioniques de sulfate et de beryllium dont la constante de formation a ete evaluee. (auteur)

  1. Progress on the Fabrication Methods Development for the Korean Test Blanket Module First Wall in the ITER

    International Nuclear Information System (INIS)

    Lee, Dong Won; Kim, Suk Kwon; Bae, Young Dug; Yoon, Jae Sung; Cho, Seung Yon

    2010-01-01

    A Korean helium cooled molten lithium (HCML) test blanket module (TBM) has been designed to be tested in the International Thermonuclear Experimental Reactor (ITER) TBM and related fabrication methods have been developed especially for the purpose of joining. Since the first wall (FW) of the HCML TBM is composed of a beryllium (Be) as an armor material and a FMS as a structural one, joining with Be to FMS and FMS to FMS should be developed in order to fabricate it

  2. Perl Modules for Constructing Iterators

    Science.gov (United States)

    Tilmes, Curt

    2009-01-01

    The Iterator Perl Module provides a general-purpose framework for constructing iterator objects within Perl, and a standard API for interacting with those objects. Iterators are an object-oriented design pattern where a description of a series of values is used in a constructor. Subsequent queries can request values in that series. These Perl modules build on the standard Iterator framework and provide iterators for some other types of values. Iterator::DateTime constructs iterators from DateTime objects or Date::Parse descriptions and ICal/RFC 2445 style re-currence descriptions. It supports a variety of input parameters, including a start to the sequence, an end to the sequence, an Ical/RFC 2445 recurrence describing the frequency of the values in the series, and a format description that can refine the presentation manner of the DateTime. Iterator::String constructs iterators from string representations. This module is useful in contexts where the API consists of supplying a string and getting back an iterator where the specific iteration desired is opaque to the caller. It is of particular value to the Iterator::Hash module which provides nested iterations. Iterator::Hash constructs iterators from Perl hashes that can include multiple iterators. The constructed iterators will return all the permutations of the iterations of the hash by nested iteration of embedded iterators. A hash simply includes a set of keys mapped to values. It is a very common data structure used throughout Perl programming. The Iterator:: Hash module allows a hash to include strings defining iterators (parsed and dispatched with Iterator::String) that are used to construct an overall series of hash values.

  3. Irradiation effects on aluminium and beryllium

    International Nuclear Information System (INIS)

    Bieth, M.

    1992-01-01

    ductility of 1.6%. Besides, due to the effects of embrittlement and swelling induced by irradiation, the HFR beryllium reflector elements had to be replaced after more than 25 years of operation. Operational and practical experiences with these reflector elements are commented, as well as main engineering features of the new reflector elements: upper-end fittings of both filler element and insert in stainless steel, no radially drilled holes and no roll pins

  4. ITER definition phase

    International Nuclear Information System (INIS)

    1989-01-01

    The International Thermonuclear Experimental Reactor (ITER) is envisioned as a fusion device which would demonstrate the scientific and technological feasibility of fusion power. As a first step towards achieving this goal, the European Community, Japan, the Soviet Union, and the United States of America have entered into joint conceptual design activities under the auspices of the International Atomic Energy Agency. A brief summary of the Definition Phase of ITER activities is contained in this report. Included in this report are the background, objectives, organization, definition phase activities, and research and development plan of this endeavor in international scientific collaboration. A more extended technical summary is contained in the two-volume report, ''ITER Concept Definition,'' IAEA/ITER/DS/3. 2 figs, 2 tabs

  5. Power converters for ITER

    CERN Document Server

    Benfatto, I

    2006-01-01

    The International Thermonuclear Experimental Reactor (ITER) is a thermonuclear fusion experiment designed to provide long deuterium– tritium burning plasma operation. After a short description of ITER objectives, the main design parameters and the construction schedule, the paper describes the electrical characteristics of the French 400 kV grid at Cadarache: the European site proposed for ITER. Moreover, the paper describes the main requirements and features of the power converters designed for the ITER coil and additional heating power supplies, characterized by a total installed power of about 1.8 GVA, modular design with basic units up to 90 MVA continuous duty, dc currents up to 68 kA, and voltages from 1 kV to 1 MV dc.

  6. Material migration patterns and overview of first surface analysis of the JET ITER-like wall

    International Nuclear Information System (INIS)

    Widdowson, A; Ayres, C F; Baron-Wiechec, A; Matthews, G F; Alves, E; Catarino, N; Brezinsek, S; Coad, J P; Likonen, J; Heinola, K; Mayer, M; Rubel, M

    2014-01-01

    Following the first JET ITER-like wall operations a detailed in situ photographic survey of the main chamber and divertor was completed. In addition, a selection of tiles and passive diagnostics were removed from the vessel and made available for post mortem analysis. From the photographic survey and results from initial analysis, the first conclusions regarding erosion, deposition, fuel retention and material transport during divertor and limiter phases have been drawn. The rate of deposition on inner and outer base divertor tiles and remote divertor corners was more than an order of magnitude less than during the preceding carbon wall operations, as was the concomitant deuterium retention. There was however beryllium deposition at the top of the inner divertor. The net beryllium erosion rate from the mid-plane inner limiters was found to be higher than for the previous carbon wall campaign although further analysis is required to determine the overall material balance due to erosion and re-deposition. (paper)

  7. ITER EDA and technology

    International Nuclear Information System (INIS)

    Baker, C.C.

    2001-01-01

    The year 1998 was the culmination of the six-year Engineering Design Activities (EDA) of the International Thermonuclear Experimental Reactor (ITER) Project. The EDA results in design and validating technology R and D, plus the associated effort in voluntary physics research, is a significant achievement and major milestone in the history of magnetic fusion energy development. Consequently, the ITER EDA was a major theme at this Conference, contributing almost 40 papers

  8. Codeposition of deuterium ions with beryllium oxide at elevated temperatures

    CERN Document Server

    Markin, A V; Gorodetsky, A E; Negodaev, M A; Rozhanskii, N V; Scaffidi-Argentina, F; Werle, H; Wu, C H; Zalavutdinov, R K; Zakharov, A P

    2000-01-01

    Deuterium-loaded BeO films were produced by sputtering the beryllium target with 10 keV Ne ions in D sub 2 gas at a pressure of approximately 1 Pa. The sputtered beryllium reacts - on the substrate surface - with the residual oxygen, thus forming a beryllium oxide layer. Biasing the substrate negatively with respect to the target provides the simultaneous bombardment of the growing film surface with D ions formed by Ne-D sub 2 collisions. Substrate potential governs the maximum energy of ions striking the growing film surface while its size governs the flux density. According to X-ray photoelectron spectroscopy (XPS), electron probe microanalysis (EPMA) and reflection high energy electron diffraction (RHEED) data, the beryllium is deposited in the form of polycrystalline hcp-BeO layers with negligible (about 1 at.%) carbon and neon retention. Thermal desorption spectroscopy (TDS) data shows a strong deuterium bonding, with a desorption peak at 950 K, in the films deposited at -50 and -400 V substrate potentia...

  9. Thermal cycling tests of actively cooled beryllium copper joints

    Energy Technology Data Exchange (ETDEWEB)

    Roedig, M.; Duwe, R.; Linke, J.; Schuster, A.; Wiechers, B. [Forschungszentrum Juelich GmbH (Germany)

    1998-01-01

    Screening tests (steady state heating) and thermal fatigue tests with several kinds of beryllium-copper joints have been performed in an electron beam facility. Joining techniques under investigation were brazing with silver containing and silver-free braze materials, hot isostatic pressing (HIP) and diffusion bonding (hot pressing). Best thermal fatigue performance was found for the brazed samples. (author)

  10. Investigation of fracture in pressurized gas metal arc welded beryllium

    International Nuclear Information System (INIS)

    Heiple, C.R.; Merlini, R.J.; Adams, R.O.

    1976-01-01

    Premature failures during proof testing of pressurized-gas-metal-arc (PGMA) welded beryllium assemblies were investigated. The failures were almost entirely within the beryllium (a forming grade, similar to HP-10 or S-240), close to and parallel to the weld interface. The aluminum-silicon weld filler metal deposit was not centered in the weld groove in the failed assemblies, and failure occurred on the side of the weld opposite the bias in the weld deposit. Tensile tests of welded samples demonstrated that the failures were unrelated to residual machining damage from cutting the weld groove, and indicated small lack-of-fusion areas near the weld start to be the most likely origin of the failures. Acoustic emission was monitored during tensile tests of the welds. The majority of acoustic emission was probably from crack propagation through the weld filler metal. Tensile bars cut from the region of the weld start behaved differently; they failed at lower loads and exhibited an acoustic emission behavior believed to be from cracking in the weld metal-beryllium interface. Improvement in the quality of these and similar beryllium welds can therefore most likely be made by centering the weld deposit and reducing the size of the weld start defect. 21 fig

  11. Influence of impurities in Beryllium on tritium breeding ratio

    International Nuclear Information System (INIS)

    Yamauchi, M.; Ochiai, K.; Verzilov, Y.; Ito, M.; Wada, M.; Nishitani, T.

    2004-01-01

    Several neutronics experiments simulating fusion blankets have been conducted with 14 MeV neutron source to assess the reliability of nuclear analysis codes. However, the analyses have not always presented good agreements so far between calculated and measured tritium production rates. One of the reasons was considered as impurities in beryllium which has negligibly small neutron absorption cross section in low energy range. Chemical compositions of beryllium were analyzed by Inductively Coupled Plasma (ICP) method, and a pulsed neutron decay experiment discovered that the macroscopic neutron absorption cross section for beryllium medium may be about 30% larger than the value calculated by the data specified by manufacturing company. The influence of the impurities on the calculations was studied on the basis of the fusion DEMO-reactor blanket design. As a result of the study, it was made clear that the impurities affect the local tritium production rates when the size of beryllium medium is more than 20-30 mean free paths (30-40 cm) in thickness. In case of some blanket designs that meet the above condition, the effect on tritium breeding ratio may become as large as about 4%. (author)

  12. Influence of impurities in Beryllium on tritium breeding ratio

    Energy Technology Data Exchange (ETDEWEB)

    Yamauchi, M; Ochiai, K; Verzilov, Y; Ito, M; Wada, M; Nishitani, T [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2004-03-01

    Several neutronics experiments simulating fusion blankets have been conducted with 14 MeV neutron source to assess the reliability of nuclear analysis codes. However, the analyses have not always presented good agreements so far between calculated and measured tritium production rates. One of the reasons was considered as impurities in beryllium which has negligibly small neutron absorption cross section in low energy range. Chemical compositions of beryllium were analyzed by Inductively Coupled Plasma (ICP) method, and a pulsed neutron decay experiment discovered that the macroscopic neutron absorption cross section for beryllium medium may be about 30% larger than the value calculated by the data specified by manufacturing company. The influence of the impurities on the calculations was studied on the basis of the fusion DEMO-reactor blanket design. As a result of the study, it was made clear that the impurities affect the local tritium production rates when the size of beryllium medium is more than 20-30 mean free paths (30-40 cm) in thickness. In case of some blanket designs that meet the above condition, the effect on tritium breeding ratio may become as large as about 4%. (author)

  13. The experience in production of composite refraction lenses from beryllium

    Energy Technology Data Exchange (ETDEWEB)

    Semenov, A. A.; Zabrodin, A. V.; Gorlevskiy, V. V.; Sheverdyaev, M. S.; Lizunov, A. V.; Brylev, D. A.; Anikin, A. S.; Klykov, S. S.; Kozlova, E. V.; Lesina, I. G.; Nebera, A. L.; Morozov, I. A.; Demin, A. V. [Bochvar High-Technology Research Institute of Inorganic Materials (Russian Federation); Buzmakov, A. V.; Dymshicz, Yu. M.; Volkov, V. V.; Zhigalina, O. M.; Konarev, P. V.; Khmelenin, D. N.; Seregin, A. V. [Russian Academy of Sciences, Shubnikov Institute of Crystallography, Federal Scientific Research Centre “Crystallography and Photonics,” (Russian Federation); and others

    2017-01-15

    The choice of beryllium-based material for the use in X-ray optics has been substantiated based on electron microscopy and X-ray diffraction data. The first results of applying refraction lenses made of this material are reported.

  14. TEM study of impurity segregations in beryllium pebbles

    Energy Technology Data Exchange (ETDEWEB)

    Klimenkov, M., E-mail: michael.klimenkov@kit.edu [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Chakin, V.; Moeslang, A. [Institute for Applied Materials – Applied Materials Physics, Karlsruhe Institute of Technology, Hermann-von-Helmholz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rolli, R. [Institute for Applied Materials – Materials and Biomechanics, Karlsruhe Institute of Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany)

    2014-12-15

    Beryllium is planned to be used as a neutron multiplier in the Helium-cooled Pebble Bed European concept of a breeding blanket of demonstration power reactor DEMO. In order to evaluate the irradiation performance, individual pebbles and constrained pebble beds were neutron-irradiated at temperatures typical of fusion blankets. Beryllium pebbles 1 mm in diameter produced by the rotating electrode method were subjected to a TEM study before and after irradiation at High Flux Reactor, Petten, Netherlands at 861 K. The grain size varied in a wide range from sub-micron size up to several tens of micrometers, which indicated formation bimodal grain size distribution. Based on the application of combined electron energy loss spectroscopy and energy dispersive X-ray spectroscopy methods, we suggest that impurity precipitates play an important role in controlling the mechanical properties of beryllium. The impurity elements were present in beryllium at a sub-percent concentration form beryllide particles of a complex (Fe/Al/Mn/Cr)B composition. These particles are often ordered along dislocations lines, forming several micron-long chains. It can be suggested that fracture surfaces often extended along these chains in irradiated material.

  15. Quantifying design trade-offs of beryllium targets on NIF

    Science.gov (United States)

    Yi, S. A.; Zylstra, A. B.; Kline, J. L.; Loomis, E. N.; Kyrala, G. A.; Shah, R. C.; Perry, T. S.; Kanzleiter, R. J.; Batha, S. H.; MacLaren, S. A.; Ralph, J. E.; Masse, L. P.; Salmonson, J. D.; Tipton, R. E.; Callahan, D. A.; Hurricane, O. A.

    2017-10-01

    An important determinant of target performance is implosion kinetic energy, which scales with the capsule size. The maximum achievable performance for a given laser is thus related to the largest capsule that can be imploded symmetrically, constrained by drive uniformity. A limiting factor for symmetric radiation drive is the ratio of hohlraum to capsule radii, or case-to-capsule ratio (CCR). For a fixed laser energy, a larger hohlraum allows for driving bigger capsules symmetrically at the cost of reduced peak radiation temperature (Tr). Beryllium ablators may thus allow for unique target design trade-offs due to their higher ablation efficiency at lower Tr. By utilizing larger hohlraum sizes than most modern NIF designs, beryllium capsules thus have the potential to operate in unique regions of the target design parameter space. We present design simulations of beryllium targets with a large CCR = 4.3 3.7 . These are scaled surrogates of large hohlraum low Tr beryllium targets, with the goal of quantifying symmetry tunability as a function of CCR. This work performed under the auspices of the U.S. DOE by LANL under contract DE-AC52- 06NA25396, and by LLNL under Contract DE-AC52-07NA27344.

  16. Morphing ab initio potential energy curve of beryllium monohydride

    Czech Academy of Sciences Publication Activity Database

    Špirko, Vladimír

    2016-01-01

    Roč. 330, Dec (2016), s. 89-95 ISSN 0022-2852 Institutional support: RVO:61388963 Keywords : beryllium monohydride * potential energy function * reduced potential * homotopic morphing Subject RIV: CF - Physical ; Theoretical Chemistry Impact factor: 1.618, year: 2016

  17. High dose neutron irradiation damage in beryllium as blanket material

    Energy Technology Data Exchange (ETDEWEB)

    Chakin, V.P. E-mail: fae@niiar.ru; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B. E-mail: vniinm.400@g23.relkom.ru

    2001-11-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10{sup 22} and 8.0x10{sup 22} cm{sup -2} (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10{sup 22} cm{sup -2} (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10{sup 22} cm{sup -2} (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket.

  18. High dose neutron irradiation damage in beryllium as blanket material

    International Nuclear Information System (INIS)

    Chakin, V.P.; Kazakov, V.A.; Teykovtsev, A.A.; Pimenov, V.V.; Shimansky, G.A.; Ostrovsky, Z.E.; Suslov, D.N.; Latypov, R.N.; Belozerov, S.V.; Kupriyanov, I.B.

    2001-01-01

    The paper presents the investigation results of beryllium products that operated in the SM and BOR-60 reactors up to neutron doses of 2.8x10 22 and 8.0x10 22 cm -2 (E>1 MeV), respectively. The calculated and experimental data are given on helium and tritium accumulation, swelling, micro-hardness and thermal conductivity. The microstructural investigation results of irradiated beryllium are also presented. It is shown that the rate of helium and tritium accumulation in beryllium in the SM and BOR-60 reactors is high enough, which is of interest from the viewpoint of modeling the working conditions of the DEMO fusion reactor. Swelling of beryllium at irradiation temperature of 70-150 deg. C and neutron fluence of 2.8x10 22 cm -2 (E>1 MeV) makes up 0.8-1.5%, at 400 deg. C and fluence of 8x10 22 cm -2 (E>1 MeV)-3.2-5.0%. Irradiation hardening and decrease of thermal conductivity strongly depend on the irradiation temperature and are more significant at reduced temperatures. All results presented in the paper were analyzed with due account of the supposed working parameters of the DEMO fusion reactor blanket

  19. Toward construction of ITER

    International Nuclear Information System (INIS)

    Shimomura, Yasuo

    2005-01-01

    The ITER Project has been significantly developed in the past years in preparation for its construction. The ITER Negotiators have developed a draft Joint Implementation Agreement (JIA), ready for completion following the nomination of the Project's Director General (DG). The ITER International Team and Participant Teams have continued technical and organizational preparations. The actual construction will be able to start immediately after the international ITER organization will be established, following signature of the JIA. The Project is now strongly supported by all the participants as well as by the scientific community with the final high-level negotiations, focused on siting and the concluding details of cost sharing, started in December 2003. The EU, with Cadarache, and Japan, with Rokkasho, have both promised large contributions to the project to strongly support their construction site proposals. The extent to which they both wish to host the ITER facility is such that large contributions to a broader collaboration among the Parties are also proposed by them. This covers complementary activities to help accelerate fusion development towards a viable power source, as well as may allow the Participants to reach a conclusion on ITER siting. (author)

  20. ITER Status and Plans

    Science.gov (United States)

    Greenfield, Charles M.

    2017-10-01

    The US Burning Plasma Organization is pleased to welcome Dr. Bernard Bigot, who will give an update on progress in the ITER Project. Dr. Bigot took over as Director General of the ITER Organization in early 2015 following a distinguished career that included serving as Chairman and CEO of the French Alternative Energies and Atomic Energy Commission and as High Commissioner for ITER in France. During his tenure at ITER the project has moved into high gear, with rapid progress evident on the construction site and preparation of a staged schedule and a research plan leading from where we are today through all the way to full DT operation. In an unprecedented international effort, seven partners ``China, the European Union, India, Japan, Korea, Russia and the United States'' have pooled their financial and scientific resources to build the biggest fusion reactor in history. ITER will open the way to the next step: a demonstration fusion power plant. All DPP attendees are welcome to attend this ITER town meeting.

  1. Identification of an abnormal beryllium lymphocyte proliferation test

    International Nuclear Information System (INIS)

    Frome, Edward L.; Newman, Lee S.; Cragle, Donna L.; Colyer, Shirley P.; Wambach, Paul F.

    2003-01-01

    The potential hazards from exposure to beryllium or beryllium compounds in the workplace were first reported in the 1930s. The tritiated thymidine beryllium lymphocyte proliferation test (BeLPT) is an in vitro blood test that is widely used to screen beryllium exposed workers in the nuclear industry for sensitivity to beryllium. The clinical significance of the BeLPT was described and a standard protocol was developed in the late 1980s. Cell proliferation is measured by the incorporation of tritiated thymidine into dividing cells on two culture dates and using three concentrations of beryllium sulfate. Results are expressed as a 'stimulation index' (SI) which is the ratio of the amount of tritiated thymidine (measured by beta counts) in the simulated cells divided by the counts for the unstimulated cells on the same culture day. Several statistical methods for use in the routine analysis of the BeLPT were proposed in the early 1990s. The least absolute values (LAV) method was recommended for routine analysis of the BeLPT. This report further evaluates the LAV method using new data, and proposes a new method for identification of an abnormal or borderline test. This new statistical-biological positive (SBP) method reflects the clinical judgment that: (i) at least two SIs show a 'positive' response to beryllium; and (ii) that the maximum of the six SIs must exceed a cut-point that is determined from a reference data set of normal individuals whose blood has been tested by the same method in the same serum. The new data is from the Y-12 National Security Complex in Oak Ridge (Y-12) and consists of 1080 workers and 33 non-exposed control BeLPTs (all tested in the same serum). Graphical results are presented to explain the statistical method, and the new SBP method is applied to the Y-12 group. The true positive rate and specificity of the new method were estimated to be 86% and 97%, respectively. An electronic notebook that is accessible via the Internet was used in

  2. Hanford Site Beryllium Program: Past, Present, and Future - 12428

    Energy Technology Data Exchange (ETDEWEB)

    Fisher, Mark [CH2M Hill Plateau Remediation Company, Richland, Washington 99354 (United States); Garcia, Pete [U.S. Department of Energy - Richland Office, Richland, Washington 99352 (United States); Goeckner, Julie [U.S. Department of Energy - HQ, EMCBC, Cincinnati, Ohio 45202 (United States); Millikin, Emily [Washington Closure Hanford, Richland, Washington 99354 (United States); Stoner, Mike [Mission Support Alliance, Richland, Washington 99354 (United States)

    2012-07-01

    The U.S. Department of Energy (DOE) has a long history of beryllium use because of the element's broad application to many nuclear operations and processes. At the Hanford Site beryllium alloy was used to fabricate parts for reactors, including fuel rods for the N-Reactor during plutonium production. Because of continued confirmed cases of chronic beryllium disease (CBD), and data suggesting CBD occurs at exposures to low-level concentrations, the DOE decided to issue a rule to further protect federal and contractor workers from hazards associated with exposure to beryllium. When the beryllium rule was issued in 1999, each of the Hanford Site contractors developed a Chronic Beryllium Disease Prevention Program (CBDPP) and initial site wide beryllium inventories. A new site-wide CBDPP, applicable to all Hanford contractors, was issued in May, 2009. In the spring of 2010 the DOE Headquarters Office of Health, Safety, and Security (HSS) conducted an independent inspection to evaluate the status of implementation of the Hanford Site Chronic Beryllium Disease Prevention Program (CBDPP). The report identified four Findings and 12 cross-cutting Opportunities for Improvement (OFIs). A corrective action plan (CAP) was developed to address the Findings and crosscutting OFIs. The DOE directed affected site contractors to identify dedicated resources to participate in development of the CAP, along with involving stakeholders. The CAP included general and contractor-specific recommendations. Following initiation of actions to implement the approved CAP, it became apparent that additional definition of product deliverables was necessary to assure that expectations were adequately addressed and CAP actions could be closed. Consequently, a supplement to the original CAP was prepared and transmitted to DOE-HQ for approval. Development of the supplemental CAP was an eight month effort. From the onset a core group of CAP development members were identified to develop a mechanism

  3. Final case for a stainless steel diagnostic first wall on ITER

    Science.gov (United States)

    Pitts, R. A.; Bazylev, B.; Linke, J.; Landman, I.; Lehnen, M.; Loesser, D.; Loewenhoff, Th.; Merola, M.; Roccella, R.; Saibene, G.; Smith, M.; Udintsev, V. S.

    2015-08-01

    In 2010 the ITER Organization (IO) proposed to eliminate the beryllium armour on the plasma-facing surface of the diagnostic port plugs and instead to use bare stainless steel (SS), simplifying the design and providing significant cost reduction. Transport simulations at the IO confirmed that charge-exchange sputtering of the SS surfaces would not affect burning plasma operation through core impurity contamination, but a second key issue is the potential melt damage/material loss inflicted by the intense photon radiation flashes expected at the thermal quench of disruptions mitigated by massive gas injection. This paper addresses this second issue through a combination of ITER relevant experimental heat load tests and qualitative theoretical arguments of melt layer stability. It demonstrates that SS can be employed as material for the port plug plasma-facing surface and this has now been adopted into the ITER baseline.

  4. Tritium loading in ITER plasma-facing surfaces and its release under accident conditions

    International Nuclear Information System (INIS)

    Longhurst, G.R.; Anderl, R.A.; Pawelko, R.J.

    1996-01-01

    Plasma-facing surfaces of the International Thermonuclear Experimental Reactor (ITER) will take up tritium from the plasma. These surfaces will probably consist of matures of Be, C, and possibly W together with other impurities. Recent experimental results have suggested mechanisms, not previously considered in analyses, by which tritium and other hydrogen isotopes are retained in Be. This warrants revised modeling and estimation of the amount of tritium that will be deposited in ITER beryllium plasma-facing surfaces and the rates at which it can be released under postulated accident scenarios. In this paper we describe improvements in modeling and experiments planned at the Idaho National Engineering Laboratory (INEL) to investigate the tritium uptake and thermal release behavior for mixed plasma- facing materials. TMAP4 calculations were made using recent data to estimate first-wall tritium inventories in ITER. 16 refs., 1 fig

  5. Final case for a stainless steel diagnostic first wall on ITER

    International Nuclear Information System (INIS)

    Pitts, R.A.; Bazylev, B.; Linke, J.; Landman, I.; Lehnen, M.; Loesser, D.; Loewenhoff, Th.; Merola, M.; Roccella, R.; Saibene, G.; Smith, M.; Udintsev, V.S.

    2015-01-01

    In 2010 the ITER Organization (IO) proposed to eliminate the beryllium armour on the plasma-facing surface of the diagnostic port plugs and instead to use bare stainless steel (SS), simplifying the design and providing significant cost reduction. Transport simulations at the IO confirmed that charge-exchange sputtering of the SS surfaces would not affect burning plasma operation through core impurity contamination, but a second key issue is the potential melt damage/material loss inflicted by the intense photon radiation flashes expected at the thermal quench of disruptions mitigated by massive gas injection. This paper addresses this second issue through a combination of ITER relevant experimental heat load tests and qualitative theoretical arguments of melt layer stability. It demonstrates that SS can be employed as material for the port plug plasma-facing surface and this has now been adopted into the ITER baseline

  6. Final case for a stainless steel diagnostic first wall on ITER

    Energy Technology Data Exchange (ETDEWEB)

    Pitts, R.A., E-mail: richard.pitts@iter.org [ITER Organization, Route de Vinon-sur-Verdon, CS 90 04, 613067 St. Paul Lez Durance Cedex (France); Bazylev, B. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Linke, J. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research, 52425 Juelich (Germany); Landman, I. [Karlsruhe Institute of Technology, IHM, P.O. Box 3640, 76021 Karlsruhe (Germany); Lehnen, M. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 04, 613067 St. Paul Lez Durance Cedex (France); Loesser, D. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Loewenhoff, Th. [Forschungszentrum Jülich GmbH, Institute of Energy and Climate Research, 52425 Juelich (Germany); Merola, M.; Roccella, R. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 04, 613067 St. Paul Lez Durance Cedex (France); Saibene, G. [Fusion for Energy Joint Undertaking, Josep Pla no. 2 – T B3 7/01, Barcelona 08019 (Spain); Smith, M. [Princeton Plasma Physics Laboratory, Princeton, NJ 08543 (United States); Udintsev, V.S. [ITER Organization, Route de Vinon-sur-Verdon, CS 90 04, 613067 St. Paul Lez Durance Cedex (France)

    2015-08-15

    In 2010 the ITER Organization (IO) proposed to eliminate the beryllium armour on the plasma-facing surface of the diagnostic port plugs and instead to use bare stainless steel (SS), simplifying the design and providing significant cost reduction. Transport simulations at the IO confirmed that charge-exchange sputtering of the SS surfaces would not affect burning plasma operation through core impurity contamination, but a second key issue is the potential melt damage/material loss inflicted by the intense photon radiation flashes expected at the thermal quench of disruptions mitigated by massive gas injection. This paper addresses this second issue through a combination of ITER relevant experimental heat load tests and qualitative theoretical arguments of melt layer stability. It demonstrates that SS can be employed as material for the port plug plasma-facing surface and this has now been adopted into the ITER baseline.

  7. Manufacturing study of beryllium bonded structures

    International Nuclear Information System (INIS)

    Onozuka, M.; Hirai, S.; Kikuchi, K.; Oda, Y.; Shimizu, K.

    2004-01-01

    Manufacturing study has been conducted on Be-bonded structures employed in the first-wall panel of the blanket system for the ITER. For Be tiles bonded to the Cu-Cr-Zr alloy heat sink with stainless-steel cooling pipes, a one-axis hot press with two heating process has been used to bond the three materials. First, Cu-alloy and SS materials are bonded diffusively. Then, Be tiles are bonded to the pre-bonded structure under 20 MPa and at 560 degree C. An Al-Si base interlayer has been used to bond Be to the Cu-Alloy. Because of the limited heat processes using a conventional hot press, the manufacturing cost can be minimized. Using the above bonding techniques, a partial mockup of a blanket first-wall panel with 16 Be tiles (with 50 mm in size) has been successfully manufactured. (author)

  8. ITER CTA newsletter. No. 6

    International Nuclear Information System (INIS)

    2002-01-01

    This ITER CTA Newsletter issue comprises information about the following ITER Meetings: The second negotiation meeting on the joint implementation of ITER, held in Tokyo(Japan) on 22-23 January 2002, and an international ITER symposium on burning plasma science and technology, held the day later after the second negotiation meeting at the same place

  9. ITER CTA newsletter. No. 2

    International Nuclear Information System (INIS)

    2001-10-01

    This ITER CTA newsletter contains results of the ITER toroidal field model coil project presented by ITER EU Home Team (Garching) and an article in commemoration of the late Dr. Charles Maisonnier, one of the former leaders of ITER who made significant contributions to its development

  10. Armour Materials for the ITER Plasma Facing Components

    Science.gov (United States)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A. R.; ITER Home Teams,

    The selection of the armour materials for the Plasma Facing Components (PFCs) of the International Thermonuclear Experimental Reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-à-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R&D.

  11. Armour materials for the ITER plasma facing components

    International Nuclear Information System (INIS)

    Barabash, V.; Federici, G.; Matera, R.; Raffray, A.R.

    1999-01-01

    The selection of the armour materials for the plasma facing components (PFCs) of the international thermonuclear experimental reactor (ITER) is a trade-off between multiple requirements derived from the unique features of a burning fusion plasma environment. The factors that affect the selection come primarily from the requirements of plasma performance (e.g., minimise impurity contamination in the confined plasma), engineering integrity, component lifetime (e.g., withstand thermal stresses, acceptable erosion, etc.) and safety (minimise tritium and radioactive dust inventories). The current selection in ITER is to use beryllium on the first-wall, upper baffle and on the port limiter surfaces, carbon fibre composites near the strike points of the divertor vertical target and tungsten elsewhere in the divertor and lower baffle modules. This paper provides the background for this selection vis-a-vis the operating parameters expected during normal and off-normal conditions. The reasons for the selection of the specific grades of armour materials are also described. The effects of the neutron irradiation on the properties of Be, W and carbon fibre composites at the expected ITER conditions are briefly reviewed. Critical issues are discussed together with the necessary future R and D. (orig.)

  12. ITER tokamak device

    International Nuclear Information System (INIS)

    Doggett, J.; Salpietro, E.; Shatalov, G.

    1991-01-01

    The results of the Conceptual Design Activities for the International Thermonuclear Experimental Reactor (ITER) are summarized. These activities, carried out between April 1988 and December 1990, produced a consistent set of technical characteristics and preliminary plans for co-ordinated research and development support of ITER; and a conceptual design, a description of design requirements and a preliminary construction schedule and cost estimate. After a description of the design basis, an overview is given of the tokamak device, its auxiliary systems, facility and maintenance. The interrelation and integration of the various subsystems that form the ITER tokamak concept are discussed. The 16 ITER equatorial port allocations, used for nuclear testing, diagnostics, fuelling, maintenance, and heating and current drive, are given, as well as a layout of the reactor building. Finally, brief descriptions are given of the major ITER sub-systems, i.e., (i) magnet systems (toroidal and poloidal field coils and cryogenic systems), (ii) containment structures (vacuum and cryostat vessels, machine gravity supports, attaching locks, passive loops and active coils), (iii) first wall, (iv) divertor plate (design and materials, performance and lifetime, a.o.), (v) blanket/shield system, (vi) maintenance equipment, (vii) current drive and heating, (viii) fuel cycle system, and (ix) diagnostics. 11 refs, figs and tabs

  13. ITER-FEAT operation

    International Nuclear Information System (INIS)

    Shimomura, Y.; Huguet, M.; Mizoguchi, T.; Murakami, Y.; Polevoi, A.R.; Shimada, M.; Aymar, R.; Chuyanov, V.A.; Matsumoto, H.

    2001-01-01

    ITER is planned to be the first fusion experimental reactor in the world operating for research in physics and engineering. The first ten years of operation will be devoted primarily to physics issues at low neutron fluence and the following ten years of operation to engineering testing at higher fluence. ITER can accommodate various plasma configurations and plasma operation modes, such as inductive high Q modes, long pulse hybrid modes and non-inductive steady state modes, with large ranges of plasma current, density, beta and fusion power, and with various heating and current drive methods. This flexibility will provide an advantage for coping with uncertainties in the physics database, in studying burning plasmas, in introducing advanced features and in optimizing the plasma performance for the different programme objectives. Remote sites will be able to participate in the ITER experiment. This concept will provide an advantage not only in operating ITER for 24 hours a day but also in involving the worldwide fusion community and in promoting scientific competition among the ITER Parties. (author)

  14. ECRH-assisted start-up in ITER

    International Nuclear Information System (INIS)

    Lloyd, B.; Carolan, P.G.; Warrick, C.D.

    1996-07-01

    In ITER, the electric field applied for ionisation and to ramp up the plasma current may be limited to ∼ 0.3 V/m. In this case, based on established theories of the avalanche process, it is shown that ohmic breakdown in ITER is only possible over a narrow range of pressure and magnetic error field. Therefore, ECRH may be necessary to provide robust and reliable start-up. ECRH can ensure prompt breakdown over a wide range of prefill pressure and error field and can also give control over the initial time and location of breakdown. For ECRH-assisted start-up in ITER, the power and pulse length requirements are essentially determined by the need to ensure burnthrough, i.e. complete ionisation of hydrogen and the transition to high ionisation states of impurities. A 0-D code (with inclusion of some 1-D effects) has been developed to analyse burnthrough in ITER. The 0-D simulations indicate that control of the deuterium density is the key factor for ensuring successful start-up in ITER, where the effects of neutral screening and dynamic fuelling by the ex-plasma volume are also crucial. It is concluded that without ECRH, successful start-up will only be possible over a very restricted range of parameters but 3MW of absorbed ECRH power will ensure reasonably robust start-up for a broad range of conditions with beryllium impurity. In the case of carbon impurity, even with an absorbed ECRH power of 5MW one may be restricted to low prefill pressure and/or low carbon concentration for successful start-up. (Author)

  15. The ITER activity

    International Nuclear Information System (INIS)

    Glass, A.J.

    1991-01-01

    The International Thermonuclear Experimental Reactor (ITER) project is a collaboration among four parties, the United States, the Soviet Union, Japan, and the European Communities, to demonstrate the scientific and technological feasibility of fusion power for peaceful purposes. ITER will demonstrate this through the construction of a tokamak fusion reactor capable of generating 1000 megawatts of fusion power. The ITER project has three missions, as follows: (1) Physics mission -- to demonstrate ignition and controlled burn, with pulse durations from 200 to 1000 S; (2) Technology mission -- to demonstrate the technologies essential to a reactor in an integrated system, operating with high reliability and availability in pulsed operation, with steady-state operation as the ultimate goal; and (3) Testing mission -- to test nuclear and high-heat-flux components at flux levels for 1 mw/m 2 , and fluences of order 1 mw-yr/m 2

  16. Tritium inventory and permeation in the ITER breeding blanket

    International Nuclear Information System (INIS)

    Violante, V.; Tosti, S.; Sibilia, C.; Felli, F.; Casadio, S.; Alvani, C.

    2000-01-01

    A model has allowed us to perform the analysis of the tritium inventory and permeation in the international thermonuclear experimental reactor (ITER) breeding blanket under the hypothesis of steady state conditions. Li 2 ZrO 3 (reference) and Li 2 TiO 3 (alternative) have been studied as breeding materials. The total breeder inventory assessed is 7.64 g for the Li 2 ZrO 3 at reference temperature. The model has also been used for a parametric analysis of the tritium permeation. At reference temperature and purge helium velocity of 0.01 m/s, the HT partial pressure is ranging from 10 to 30 Pa in the breeder and 1.5x10 -3 Pa in the beryllium. At 0.1 m/s of purge helium velocity, the HT partial pressure is reduced of one order by magnitude in the breeder and becomes 5x10 -5 Pa in the beryllium. The tritium permeation into the coolant for the whole blanket is ranging from 100 to 250 mCi per day for purge helium velocity of 0.01 m/s. The analysis of the tritium inventory and permeation for the alternative Li 2 TiO 3 breeding material has been carried out too. The tritium inventory in the breeder is in the range from 6 to 375 g larger than in Li 2 ZrO 3 by about a factor 5; the tritium permeation into coolant is comparable to the Li 2 ZrO 3 one. This analysis provides indications on the influence of the operating parameters on the tritium control in the ITER breeding blanket; particularly the control of the tritium inventory by the temperature and the tritium permeation by the purge gas velocity

  17. Earthly sun called ITER

    International Nuclear Information System (INIS)

    Pozdeyev, Mikhail

    2002-01-01

    Full text: Participating in the film are Academicians Velikhov and Glukhikh, Mr. Filatof, ITER Director from Russia, Mr. Sannikov from Kurchatov Institute. The film tells about the starting point of the project (Mr. Lavrentyev), the pioneers of the project (Academicians Tamme, Sakharov, Artsimovich) and about the situation the project is standing now. Participating in [ITER now are the US, Russia, Japan and the European Union. There are two associated members as well - Kazakhstan and Canada. By now the engineering design phase has been finished. Computer animation used in the video gives us the idea how the first thermonuclear reactor based on famous Russian TOKOMAK works. (author)

  18. ITER plant systems

    International Nuclear Information System (INIS)

    Kolbasov, B.; Barnes, C.; Blevins, J.

    1991-01-01

    As part of a series of documents published by the IAEA that summarize the results of the Conceptual Design Activities for the ITER project, this publication describes the conceptual design of the ITER plant systems, in particular (i) the heat transport system, (ii) the electrical distribution system, (iii) the requirements for radioactive equipment handling, the hot cell, and waste management, (iv) the supply system for fluids and operational chemicals, (v) the qualitative analyses of failure scenarios and methods of burn stability control and emergency shutdown control, (vi) analyses of tokamak building functions and design requirements, (vii) a plant layout, and (viii) site requirements. Refs, figs and tabs

  19. Iterated multidimensional wave conversion

    International Nuclear Information System (INIS)

    Brizard, A. J.; Tracy, E. R.; Johnston, D.; Kaufman, A. N.; Richardson, A. S.; Zobin, N.

    2011-01-01

    Mode conversion can occur repeatedly in a two-dimensional cavity (e.g., the poloidal cross section of an axisymmetric tokamak). We report on two novel concepts that allow for a complete and global visualization of the ray evolution under iterated conversions. First, iterated conversion is discussed in terms of ray-induced maps from the two-dimensional conversion surface to itself (which can be visualized in terms of three-dimensional rooms). Second, the two-dimensional conversion surface is shown to possess a symplectic structure derived from Dirac constraints associated with the two dispersion surfaces of the interacting waves.

  20. Physics fundamentals for ITER

    International Nuclear Information System (INIS)

    Rosenbluth, M.N.

    1999-01-01

    The design of an experimental thermonuclear reactor requires both cutting-edge technology and physics predictions precise enough to carry forward the design. The past few years of worldwide physics studies have seen great progress in understanding, innovation and integration. We will discuss this progress and the remaining issues in several key physics areas. (1) Transport and plasma confinement. A worldwide database has led to an 'empirical scaling law' for tokamaks which predicts adequate confinement for the ITER fusion mission, albeit with considerable but acceptable uncertainty. The ongoing revolution in computer capabilities has given rise to new gyrofluid and gyrokinetic simulations of microphysics which may be expected in the near future to attain predictive accuracy. Important databases on H-mode characteristics and helium retention have also been assembled. (2) Divertors, heat removal and fuelling. A novel concept for heat removal - the radiative, baffled, partially detached divertor - has been designed for ITER. Extensive two-dimensional (2D) calculations have been performed and agree qualitatively with recent experiments. Preliminary studies of the interaction of this configuration with core confinement are encouraging and the success of inside pellet launch provides an attractive alternative fuelling method. (3) Macrostability. The ITER mission can be accomplished well within ideal magnetohydrodynamic (MHD) stability limits, except for internal kink modes. Comparisons with JET, as well as a theoretical model including kinetic effects, predict such sawteeth will be benign in ITER. Alternative scenarios involving delayed current penetration or off-axis current drive may be employed if required. The recent discovery of neoclassical beta limits well below ideal MHD limits poses a threat to performance. Extrapolation to reactor scale is as yet unclear. In theory such modes are controllable by current drive profile control or feedback and experiments should