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Influence of the deuteron energy on the testing volume of IFMIF and its impact on other parameters

Description: The influence of the energy of the deuteron beam on irradiation parameters of IFMIF is analyzed. The main purpose of this paper is to identify possible positive and negative impacts on irradiation parameters that an increase in the deuteron energy of the beam can cause. Several parameters of the facility, such as neutron generation rate, number of neutrons with energy above 20 MeV at the source and in the test assembly, volume with dpa rate above a threshold value, gas production, and gradient of the atomic displacement rate, are analyzed and conclusions are drawn based on the calculated values. It is shown that an increase in the deuteron energy to 40 MeV does not produce a significant negative impact for the elements analyzed, but instead is beneficial in producing nuclear responses more similar to a fusion environment than the lower deuteron energies.
Date: September 1, 1995
Creator: Gomes, I.C. & Smith, D.L.
Partner: UNT Libraries Government Documents Department

High-flux source of fusion neutrons for material and component testing

Description: The inner part of a fusion reactor will have to operate at very high neutron loads. In steady-state reactors the minimum fluence before the scheduled replacement of the reactor core should be at least l0-15 Mw.yr/m<sup>2</sup>. A more frequent replacement of the core is hardly compatible with economic constraints. A most recent summary of the discussions of these issues is presented in Ref. [l]. If and when times come to build a commercial fusion reactor, the availability of information on the behavior of materials and components at such fluences will become mandatory for making a final decision. This makes it necessary an early development and construction of a neutron source for fusion material and component testing. In this paper, we present information on one very attractive concept of such a source: a source based on a so called Gas Dynamic Trap. This neutron source was proposed in the mid 1980s (Ref. [2]; see also a survey [3] with discussion of the early stage of the project). Since then, gradual accumulation of the relevant experimental information on a modest-scale experimental facility GDT at Novosibirsk, together with a continuing design activity, have made initial theoretical considerations much more credible. We believe that such a source can be built within 4 or 5 years. Of course, one should remember that there is a chance for developing steady-state reactors with a liquid (and therefore continuously renewable) first wall [4], which would also serve as a tritium breeder. In this case, the need in the neutron testing will become less pressing. However, it is not clear yet that the concept of the flowing wall will be compatible with all types of steady-state reactors. It seems therefore prudent to be prepared to the need of a quick construction of a neutron source. It should also be ...
Date: January 7, 1999
Creator: Baldwin, D. E.; Hooper, E. B.; Ryutov, D. D. & Thomassen, K. I.
Partner: UNT Libraries Government Documents Department

Magnetic Fustion Reactor Design Studies Program final report, 1 July 1986--30 September 1986

Description: This report presents progress reported during the period, 7/1/86 - 9/30/86 for the Technical Support Services (TSS) for the Magnetic Fusion Reactor Design Studies Program. Tasks reported include: systems studies work plan, normalization of reactor design studies, interpretation of design study activities, research and development plan, conference support, and reports generated.
Date: September 30, 1986
Partner: UNT Libraries Government Documents Department

Fabrication and installation of the DIII-D radiative divertor structures

Description: Phase 1A of the Radiative Divertor Program (RDP) is now installed in the DIII-D tokamak located at General Atomics. This hardware was added to enhance both the Divertor and Advanced Tokamak research elements of the DIII-D program. This installation consists of a divertor baffle enveloping a cryocondensation pump at the upper outer divertor target of DIII-D. The divertor baffle consists of two toroidally continuous Inconel 625 water-cooled rings and a toroidal array of discontinuous radiatively-cooled plates. The water-cooled rings are each comprised of four quadrants, mechanically formed, chem.-milled, and resistance and TIG welded Inconel 625 panels. The supports attaching the panels to the vessel wall are designed to accommodate the differential thermal expansion between the rings and vessel during bake and to react the electromagnetic loads induced during disruptions. They are made from either Inconel 625 or Inconel 718 depending on the stress levels predicted in Finite Element Analysis. Gas seals are designed to limit the leakage from the baffle chamber back to the core plasma to 2,500 {ell}/s and incorporate plasma sprayed alumina to minimize currents flowing through them. The bulk of the water-cooled ring fabrication was performed by a vendor, however, the final machining of penetrations in the conical ring for diagnostic access was performed in-house using a unique machining configuration. This configuration, and the machining of the diagnostic cutouts is described. Graphite tiles were machined from ATJ graphite to form a smooth plasma-facing surface. The installation of all divertor components required only four weeks.
Date: November 1, 1997
Creator: Hollerbach, M.A. & Smith, J.P.
Partner: UNT Libraries Government Documents Department

IFMIF, International Fusion Materials Irradiation Facility conceptual design activity cost report

Description: This report documents the cost estimate for the International Fusion Materials Irradiation Facility (IFMIF) at the completion of the Conceptual Design Activity (CDA). The estimate corresponds to the design documented in the Final IFMIF CDA Report. In order to effectively involve all the collaborating parties in the development of the estimate, a preparatory meeting was held at Oak Ridge National Laboratory in March 1996 to jointly establish guidelines to insure that the estimate was uniformly prepared while still permitting each country to use customary costing techniques. These guidelines are described in Section 4. A preliminary cost estimate was issued in July 1996 based on the results of the Second Design Integration Meeting, May 20--27, 1996 at JAERI, Tokai, Japan. This document served as the basis for the final costing and review efforts culminating in a final review during the Third IFMIF Design Integration Meeting, October 14--25, 1996, ENEA, Frascati, Italy. The present estimate is a baseline cost estimate which does not apply to a specific site. A revised cost estimate will be prepared following the assignment of both the site and all the facility responsibilities.
Date: December 1, 1996
Creator: Rennich, M.J.
Partner: UNT Libraries Government Documents Department

Engineering design of the National Spherical Torus Experiment

Description: NSTX is a proof-of-principle experiment aimed at exploring the physics of the ``spherical torus'' (ST) configuration, which is predicted to exhibit more efficient magnetic confinement than conventional large aspect ratio tokamaks, amongst other advantages. The low aspect ratio (R/a, typically 1.2--2 in ST designs compared to 4--5 in conventional tokamaks) decreases the available cross sectional area through the center of the torus for toroidal and poloidal field coil conductors, vacuum vessel wall, plasma facing components, etc., thus increasing the need to deploy all components within the so-called ``center stack'' in the most efficient manner possible. Several unique design features have been developed for the NSTX center stack, and careful engineering of this region of the machine, utilizing materials up to their engineering allowables, has been key to meeting the desired objectives. The design and construction of the machine has been accomplished in a rapid and cost effective manner thanks to the availability of extensive facilities, a strong experience base from the TFTR era, and good cooperation between institutions.
Date: May 11, 2000
Creator: Neumeyer, C.; Heitzenroeder, P.; J. Spitzer, J. Chrzanowski & al, et
Partner: UNT Libraries Government Documents Department

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

Description: 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{sup 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{sup 2} and surface temperatures up to 690{degrees}C before debonding at 10 MW/m{sup 2}. A third tile debonded after 9200 thermal fatigue cycles at 5 MW/m{sup 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.
Date: December 31, 1994
Creator: Youchison, D.L.; Guiniiatouline, R. & Watson, R.D.
Partner: UNT Libraries Government Documents Department

Modified Rate-Theory Predictions in Comparison to Microstructural Data

Description: Standard rate theory methods have recently been combined with experimental microstructures to successfully reproduce measured swelling behavior in ternary steels around 400 C. Fit parameters have reasonable values except possibly for the recombination radius, R{sub c}, which can be larger than expected. Numerical simulations of void nucleation and growth reveal the importance additional recombination processes at unstable clusters. Such extra recombination may reduce the range of possible values for R{sub c}. A modified rate theory is presented here that includes the effect of these undetectably small defect clusters. The fit values for R{sub c} are not appreciably altered, as the modification has little effect on the model behavior in the late steady state. It slightly improves the predictions for early transient times, when the sink strength of stable voids and dislocations is relatively small. Standard rate theory successfully explains steady swelling behavior in high purity stainless steel.
Date: November 3, 2003
Creator: Surh, M P; Okita, T & Wolfer, W G
Partner: UNT Libraries Government Documents Department

FENIX experimental results of large-scale CICC made of bronze-processed Nb{sub 3}Sn strands

Description: The Fusion ENgineering International eXperiments (FENIX) Test Facility recently has successfully complete the testing of a pair of Nb{sub 3}rSn cable-in-conduit conductors developed by the Japan Atomic Energy Research Institute. These conductors, made of bronze-processed strands, were designed to operate stably with 40-kA transport current at a magnetic field of 13 T. In addition to the measurements of major design parameters such as current-sharing temperature, FENIX provided several experiments specifically designed to provide results urgently needed by magnet designers. Performed experiments include measurements of ramp-rate limit, current-distribution, stability, and joint performance. This paper presents the design and results of these special experiments.
Date: October 13, 1994
Creator: Shen, S.S.; Felker, B.; Moller, J.M.; Parker, J.M.; Isono, T.; Yasukawa, Y. et al.
Partner: UNT Libraries Government Documents Department

Tritium containment in the dust and debris of plasma-facing materials produced during operations

Description: Tritium behavior in plasma-facing components of future tokamak reactors such as ITER is an essential factor in evaluating and choosing a successful candidate for a plasma-facing material (PFM). One important parameter that influence tritium build-up and release in the Generated dust of PFMs is the effect of material porosity on tritium behavior. Diffusion in porous materials, for example, consists of three different diffusion processes: along grain boundaries, along micro-crystallites, and diffusion in pure structure crystallites. A model is developed to evaluate and assess the sensitivity of tritium accumulation and permeation of candidate materials due to porosity. Specific laboratory experiments relevant to reactor conditions, in currently existing and available facilities, are required to help in selecting the best candidate material.
Date: September 1, 1996
Creator: Konkashbaev, I.; Grebenshikov, J. & Hassanein, A.
Partner: UNT Libraries Government Documents Department

IFMIF test cell design issues

Description: The International Fusion Materials Irradiation Facility (IFMIF) project was started in 1994 under sponsorship of the International Energy Agency (IEA). This project, which is in its conceptual design phase, is aimed at fulfilling the need for a high-flux. high-fluence neutron irradiation facility for fusion materials research and development. Initial assessments of the materials testing needs have shown that the testing space provided in the high-flux region of the IFMIF Test Cell is adequate for development of a data base for the engineering design of a demonstration fusion power-producing-reactor (DEMO). Using the proposed set of miniaturized test specimens, compact test module configurations have been developed for several reference loadings. The test modules have been integrated into a test assembly that includes the coolant flow system for the test modules, instrumentation and control equipment, and a shield to minimize activation to ancillary areas of the IFMIF facility. These test assemblies are configured to facilitate installation, removal, and transport to hot cells using remote handling equipment.
Date: December 31, 1995
Creator: Haines, J.R.; Zinkle, S.J.; Williams, D.M. & Gomes, I.
Partner: UNT Libraries Government Documents Department

IFMIF - International Fusion Materials Irradiation Facility Conceptual Design Activity/Interim Report

Description: Environmental acceptability, safety, and economic viability win ultimately be the keys to the widespread introduction of fusion power. This will entail the development of radiation- resistant and low- activation materials. These low-activation materials must also survive exposure to damage from neutrons having an energy spectrum peaked near 14 MeV with annual radiation doses in the range of 20 displacements per atom (dpa). Testing of candidate materials, therefore, requires a high-flux source of high energy neutrons. The problem is that there is currently no high-flux source of neutrons in the energy range above a few MeV. The goal, is therefore, to provide an irradiation facility for use by fusion material scientists in the search for low-activation and damage-resistant materials. An accellerator-based neutron source has been established through a number of international studies and workshops` as an essential step for materials development and testing. The mission of the International Fusion Materials Irradiation Facility (IFMIF) is to provide an accelerator-based, deuterium-lithium (D-Li) neutron source to produce high energy neutrons at sufficient intensity and irradiation volume to test samples of candidate materials up to about a full lifetime of anticipated use in fusion energy reactors. would also provide calibration and validation of data from fission reactor and other accelerator-based irradiation tests. It would generate material- specific activation and radiological properties data, and support the analysis of materials for use in safety, maintenance, recycling, decommissioning, and waste disposal systems.
Date: December 1, 1995
Creator: Rennich, M.J.
Partner: UNT Libraries Government Documents Department

Fabrication of intermetallic coatings for electrical insulation and corrosion resistance on high-temperature alloys

Description: Several intermetallic films were applied to high-temperature alloys (V alloys and 304, 316 stainless steels) to provide electrical insulation and corrosion resistance. Alloy grain growth at 1000 C for the V-5Cr-5Ti alloy was investigated to determine stability of the alloy substrate during coating formation by CVD or metallic vapor processes at 800-850 C. Film layers were examined by optical and scanning electron microscopy and by electron-energy-dispersive and XRD analysis; they were also tested for electrical resistivity and corrosion resistance. Results elucidated the nature of the coatings, which provided both electrical insulation and high-temperature corrosion protection.
Date: November 1, 1996
Creator: Park, J.-H. & Cho, W.D.
Partner: UNT Libraries Government Documents Department

Erosion damage of nearby plasma-facing components during a disruption on the divertor plate

Description: Intense energy flow from the disrupting plasma during, a thermal quench will cause a sudden vapor cloud to form above the exposed divertor area. The vapor-cloud layer has been proved to significantly reduce the subsequent energy flux of plasma particles to the original disruption location. However, most of the incoming plasma energy is quickly converted to intense photon radiation emitted by heating of the vapor cloud. This radiation energy can cause serious erosion damage of nearby components not directly exposed to the disrupting, plasma. The extent of this ``secondary damage`` will depend on the divertor design, disrupting plasma parameters, and design of nearby components. The secondary erosion damage of these components due to intense radiation can exceed that of the original disruption location.
Date: September 1, 1996
Creator: Hassanein, A. & Konkashbaev, I.
Partner: UNT Libraries Government Documents Department

Proceedings of US/Japan workshop, Q219 on high heat flux components and plasma surface interactions for next fusion devices

Description: This report contains the viewgraphs from the proceedings of US/Japan Workshop on High Heat Flux Components and Plasma Surface Interactions for Next Fusion Devices. Some of the general topics covered by this report are: PFC/PSI in tokamak and helical devices; development of high heat flux components; PSIS and plasma facing materials;tritium; and material damage.
Date: December 1, 1996
Creator: Ulrickson, M.A.; Stevens, P.L.; Hino, T. & Hirohata, Y.
Partner: UNT Libraries Government Documents Department

Proceedings of the ninth IEA workshop on radiation effects in ceramic insulators

Description: Several IEA workshops have been held over the past few years to discuss the growing number of experimental studies on the intriguing phenomenon of radiation induced electrical degradation (RIED). In the past year, several new RIED irradiation experiments have been performed which have a significant impact on the understanding of the RIED phenomenon. These experiments include a HFIR neutron irradiation experiment on 12 different grades of single- and poly-crystal alumina (450 C, {approximately}3 dpa, 200 V/mm) and several additional neutron, electron and light ion irradiation experiments. The primary objective of the IEA workshop was to review the available RIED studies on ceramic insulators. Some discussion of recent work in other areas such as loss tangent measurements, mechanical strength, etc. occurred on the final afternoon of the workshop. The IEA workshop was held in conjunction with a US-Japan JUPITER program experimenter`s workshop on dynamic radiation effects in ceramic insulators.
Date: December 31, 1997
Creator: Zinkle, S.J.; Burn, G.L.; Hodgson, E.R. & Shikama, T.
Partner: UNT Libraries Government Documents Department

Suppression of erosion in the DIII-D divertor with detached plasmas

Description: The ability to withstand disruptions makes carbon-based materials attractive for use as plasma-facing components in divertors. However, such materials suffer high erosion rates during attached plasma operation which, in high power long pulse machines, would give short component lifetimes and high tritium inventories. The authors present results from recent experiments in DIII-D, in which the Divertor Materials Evaluation System (DiMES) was used to examine erosion and deposition during short exposures to well defined plasma conditions. These studies show that during operation with detached plasmas, produced by gas injection, net erosion is suppressed everywhere in the divertor. Net deposition of carbon with deuterium was observed at the inner and outer strikepoints and in the private-flux region between strikepoints. For these low temperature plasmas (T{sub e} &lt; 2eV), physical sputtering is eliminated. These results show that with detached plasmas, the location of carbon net erosion and the carbon impurity source, probably lies outside the divertor. Physical or chemical sputtering by charge-exchange neutrals or ions in the main plasma chamber is a probable source of carbon under these plasma conditions.
Date: May 25, 2000
Creator: Wampler, William R.; Bastasz, Robert J.; Whyte, D. G.; Wong, C. P. C. & West, W. P.
Partner: UNT Libraries Government Documents Department

Disruption simulation experiments and extrapolation to reactor conditions

Description: Laboratory experiments to simulate plasma disruptions have contributed significantly in many aspects to the understanding of the physical processes occurring during high-energy deposition on target material surfaces due to plasma instabilities. Laser light, electron beams, and plasma guns have been used worldwide to study disruption effects and erosion damage of candidate divertor materials. The differences among these simulation experiments are examined. The net power flux reaching the originally exposed surface depends on many parameters, such as type of energy deposited, target material, pulse duration, and geometrical factors. Experimental results have been evaluated and compared with theoretical predictions, and the overall relevance of simulation experiments to reactor conditions has been critically examined.
Date: August 1, 1998
Creator: Hassanein, A. & Konkashbaev, I.K.
Partner: UNT Libraries Government Documents Department

Impact of transmutations in fusion environment on Flibe chemistry.

Description: Transmutation rates of Li, Be and F are calculated for a typical flibe blanket. The results concluded that the transmutation rate of F is more than double that of Be. Because of the high destruction rate of fluorine, there will be no free fluorine in the molten salt. Therefore, experimental program to address the chemistry control of flibe does not have to worry about the issues associated with free fluorine. Also, this calculation defines the chemical of flibe after irradiation. This chemical state needs to be simulated closely for the flibe chemistry control experiment.
Date: November 15, 2000
Creator: Sze, D. K.; Sawan, M. E. & Cheng, E. T.
Partner: UNT Libraries Government Documents Department

Overview of design activities for Li/V blankets

Description: Recent fusion power plant design studies in the US have been conducted within the ARIES project. The most recent design of Li/V blankets was conducted as part of the ARIES-RS design. The ARIES-RS fusion power plant design study is based on reversed-shear (RS) physics with a Li/V (lithium breeder and vanadium structure) blanket. The reversed-shear discharge has been documented in many large tokamak experiments. The plasma in the RS mode has a high beta, low current, and low current drive requirement. Therefore, it is an attractive physics regime for a fusion power plant. The blanket system based on a Li/V has high temperature operating capability, good tritium breeding, excellent high heat flux removal capability, long structural life time, low activation, low after heat and good safety characteristics. For these reasons, the ARIES-RS reactor study selected Li/V as the reference blanket. The combination of attractive physics and attractive blanket engineering is expected to result in a superior power plant design.
Date: December 31, 1997
Creator: Sze, D.K. & Mattas, R.F.
Partner: UNT Libraries Government Documents Department

Study of material destruction during powerful plasma exposure.

Description: Results of experimental research during the interaction of powerful plasma flow with aluminum and graphite targets are presented. The experiments were carried out at plasma energy density of up to 5 MJ/m{sup 2} and time duration of up to 600 {micro}s. Strong damage of target materials was observed as a result of the plasma flow interaction as well as the formation and the ejection of macroscopic particles from the target surface. The ejected particle size and spatial distribution were measured as well as the velocity of these particles during the experiments. The method of laser scattering was applied for these measurements.
Date: August 1, 1998
Creator: Belan, V.G.; Levashov, V.F.; Maynashev, V.S.; Muzichenko, A.D.; Podkovirov, V.L.; Zitlukhin, A.M. et al.
Partner: UNT Libraries Government Documents Department

Hydrodynamic effects of eroded materials on response of plasma-facing component during a tokamak disruption

Description: Loss of plasma confinement causes surface and structural damage to plasma-facing materials (PFMs) and remains a major obstacle for tokamak reactors. The deposited plasma energy results in surface erosion and structural failure. The surface erosion consists of vaporization, spallation, and liquid splatter of metallic materials, while the structural damage includes large temperature increases in structural materials and at the interfaces between surface coatings and structural members. Comprehensive models (contained in the HEIGHTS computer simulation package) are being used self-consistently to evaluate material damage. Splashing mechanisms occur as a result of volume bubble boiling and liquid hydrodynamic instabilities and brittle destruction mechanisms of nonmelting materials. The effect of macroscopic erosion on total mass losses and lifetime is evaluated. The macroscopic erosion products may further protect PFMs from severe erosion (via the droplet-shielding effect) in a manner similar to that of the vapor shielding concept.
Date: October 25, 1999
Creator: Hassanein, A. & Konkashbaev, I.
Partner: UNT Libraries Government Documents Department

Molybdenum erosion measurements in Alcator C-Mod

Description: Erosion of molybdenum was measured on a set of 21 tiles after a run campaign of 1,090 shots in the Alcator C-Mod tokamak. The net erosion of molybdenum, was determined from changes in the depth of a thin chromium marker layer measured by Rutherford backscattering. Net Mo erosion was found to be approximately 150 nm near the outer divertor strike point, and much less everywhere else. Gross erosion rates by sputtering were estimated using ion energies and fluxes obtained from Langmuir probe measurements of edge-plasma conditions. Predicted net erosion using calculated gross erosion with prompt redeposition and measured net erosion agree within a factor of 3. Sputtering by boron and molybdenum impurities dominates erosion.
Date: May 1, 1998
Creator: Wampler, W.R.; LaBombard, B.; Lipshultz, B.; Pappas, D.; Pitcher, C.S. & McCracken, G.M.
Partner: UNT Libraries Government Documents Department