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in creator/contributor: "Nygren,R. E."  
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Physics Design Requirements for the National Spherical Torus Experiment Liquid Lithium Divertor

Description: Recent NSTX high power divertor experiments have shown significant and recurring benefits of solid lithium coatings on PFC's to the performance of divertor plasmas in both L- and H- mode confinement regimes heated by high-power neutral beams. The next step in this work is installation of a liquid lithium divertor (LLD) to achieve density control for inductionless current drive capability (e.g., about a 15-25% ne decrease from present highest non-inductionless fraction discharges which often evo… more
Date: September 26, 2008
Creator: Kugel, W.; Bell, M.; Berzak,L.; Brooks, A.; Ellis, R.; Gerhardt, S. et al.
Partner: UNT Libraries Government Documents Department
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Engineering testing requirements in FED/INTOR

Description: The FED/INTOR critical issues activity has addressed three key testing requirements that have the largest impact on the design, operation and cost of FED/INTOR. These are: (1) the total testing time (fluence) during the device lifetime, (2) the minimum number of back-to-back cycles, and (3) the neutron wall load (power density in the first wall/blanket). The testing program activities were structured into three tasks in order to define the benefits, and in some cases, costs and risks of these t… more
Date: October 1, 1982
Creator: Abdou, M. A.; Nygren, R. E.; Morgan, G. D.; Trachsel, C. A.; Wire, G.; Oppermann, E. et al.
Partner: UNT Libraries Government Documents Department
open access

US blanket technology programs. [Directory of current research]

Description: Experimental research in US programs related to blanket technology is described through brief summaries of the objectives, facilities, recent experimental results and principal investigators for the Blanket Technology Program, TRIO-1 Experiment, TSTA, Fusion Hybrid Program and selected activities in the Fusion Materials and Fusion Safety Programs in neutronics research.
Date: January 1, 1985
Creator: Nygren, R.E.
Partner: UNT Libraries Government Documents Department
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Erosion and redeposition behavior of selected net-candidate materials under high-flux hydrogen, deuterium plasma bombardment in PISCES

Description: Plasma erosion and redeposition behavior of selected candidate materials for plasma-facing components in the NET-machine have been investigated using the PISCES-A facility. Materials studied include SiC-impregnated graphite, 2D graphite weaves with and without CVD- SiC coatings, and isotropic graphite. These specimens were exposed to continuous hydrogen or deuterium plasmas under the following conditions: electron temperature range from 5 to 35eV; plasma density range from 5 x 10/sup 11/ to 1 x… more
Date: June 1, 1988
Creator: Franconi, E.; Hirooka, Y.; Conn, R. W.; Leung, W. K.; LaBombard, B. & Nygren, R. E.
Partner: UNT Libraries Government Documents Department
open access

Deuterium pumping and erosion behavior of selected graphite materials under high flux plasma bombardment in PISCES

Description: Deuterium plasma recycling and chemical erosion behavior of selected graphite materials have been investigated using the PISCES-A facility. These materials include: Pyro-graphite; 2D-graphite weave; 4D-graphite weave; and POCO-graphite. Deuterium plasma bombardment conditions are: fluxes around 7 /times/ 10/sup 17/ ions s/sup /minus/1/cm/sup /minus/2/; exposure time in the range from 10 to 100 s; bombarding energy of 300 eV; and graphite temperatures between 20 and 120/degree/C. To reduce deute… more
Date: June 1, 1988
Creator: Hirooka, Y.; Conn, R. W.; Goebel, D. M.; LaBombard, B.; Lehmer, R.; Leung, W. K. et al.
Partner: UNT Libraries Government Documents Department
open access

First Wall, Blanket, Shield Engineering Technology Program

Description: The First Wall/Blanket/Shield Engineering Technology Program sponsored by the Office of Fusion Energy of DOE has the overall objective of providing engineering data that will define performance parameters for nuclear systems in advanced fusion reactors. The program comprises testing and the development of computational tools in four areas: (1) thermomechanical and thermal-hydraulic performance of first-wall component facsimiles with emphasis on surface heat loads; (2) thermomechanical and therm… more
Date: January 1, 1982
Creator: Nygren, R.E.
Partner: UNT Libraries Government Documents Department
open access

Summary of results from the TEXTOR helium self-pumping experiment

Description: Helium removal experiments were conducted in TEXTOR with a small helium self-pumping module located in a modified ALT-I limiter head. The module contained two heated nickel alloy trapping plates, a nickel deposition filament array, a Langmuir probe, flux probe, and thermocouples. The experiment examined plasma helium removal via trapping of helium ions in the deposited nickel surfaces. Such helium removal was successfully observed, with about 10% of the helium He/D plasma being removed in a {ap… more
Date: March 1, 1992
Creator: Brooks, J. N.; Krauss, A.; Nygren, R. E.; Doyle, B. L.; Dippel, K. H. & Finken, K. H.
Partner: UNT Libraries Government Documents Department
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Qualitative Reliability Issues for In-Vessel Solid and Liquid Wall Fusion Designs

Description: This paper presents the results of a study of the qualitative aspects of plasma facing component (PFC) reliability for actively cooled solid wall and liquid wall concepts for magnetic fusion reactor vessels. These two designs have been analyzed for component failure modes. The most important results of that study are given here. A brief discussion of reliability growth in design is included to illustrate how solid wall designs have begun as workable designs and have evolved over time to become … more
Date: October 1, 2001
Creator: Cadwallader, Lee Charles & Nygren, R. E.
Partner: UNT Libraries Government Documents Department
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Assessing braze quality in the actively cooled Tore Supra Phase III outboard pump limiter

Description: The quality of brazing of pyrolytic graphite armor brazed to copper tubes in Tore Supra`s Phase III Outboard Pump Limiter was assessed through pre-service qualification testing of individual copper/tile assemblies. The evaluation used non-destructive, hot water transient heating tests performed in the high-temperature, high-pressure flow loop at Sandia`s Plasma Materials Test Facility. Surface temperatures of tiles were monitored with an infrared camera as water at 120{degrees}C at about 2.07 M… more
Date: December 31, 1994
Creator: Nygren, R. E.; Lutz, T. L.; Miller, J. D.; McGrath, R. & Dale, G.
Partner: UNT Libraries Government Documents Department
open access

The TEXTOR helium self-pumping experiment: Design, plans, and supporting ion-beam data on helium retention in nickel

Description: A proof-of-principle experiment to demonstrate helium self-pumping in a tokamak is being undertaken in TEXTOR. The experiment will use a helium self-pumping module installed in a modified ALT-I limiter head. The module consists of two, {approximately}25 {times} 25 cm{sup 2} heated nickel alloy trapping plates, a nickel deposition filament array, and associated diagnostics. Between plasma shots a coating of {approximately}50 {angstrom} nickel will be deposited on the two trapping plates. During … more
Date: January 1, 1990
Creator: Brooks, J. N.; Krauss, A.; Mattas, R. F.; Smith, D. L.; Nygren, R. E.; Doyle, B. L. et al.
Partner: UNT Libraries Government Documents Department
open access

In situ spectroscopic measurements of erosion behavior of TFTR-redeposited carbon materials under high-flux plasma bombardment in PISCES-A

Description: The chemical erosion behavior of graphite materials pre-exposed in Tokamak Fusion Test Reactor (TFTR) as the bumper limiter has been investigated spectroscopically under deuterium plasma bombardment in the PISCES-A facility. The deuterium plasma bombardment conditions are: ion bombarding energy of 300 eV; ion flux of 1.7 /times/ 10/sup 18/ ions s/sup /minus/1/ cm/sup /minus/2/; plasma density of 1.4 /times/ 10/sup 12/ cm/sup /minus/3/; electron temperature of 11 eV; and neutral pressure of 3 /t… more
Date: August 1, 1988
Creator: Hirooka, Y.; Pospieszczyk, A.; Conn, R.W.; Labombard, B.; Mills, B.; Nygren, R.E. et al.
Partner: UNT Libraries Government Documents Department
open access

Design Integration of Liquid Surface Divertors

Description: The US Enabling Technology Program in fusion is investigating the use of free flowing liquid surfaces facing the plasma. We have been studying the issues in integrating a liquid surface divertor into a configuration based upon an advanced tokamak, specifically the ARIES-RS configuration. The simplest form of such a divertor is to extend the flow of the liquid first wall into the divertor and thereby avoid introducing additional fluid streams. In this case, one can modify the flow above the dive… more
Date: November 13, 2003
Creator: Nygren, R E; Cowgill, D F; Ulrickson, M A; Nelson, B E; Fogarty, P J; Rognlien, T D et al.
Partner: UNT Libraries Government Documents Department
open access

Liquid Surface Divertor Designs for Fusion Reactors

Description: As part of work in the US on free flowing liquid surfaces facing the plasma, we are studying issues of integrating a liquid surface divertor into a configuration based upon an advanced tokamak (ARIES-RS). The simplest form of such a divertor is to extend the flow of the liquid first wall and avoid introducing any separate fluid streams. A design and some of the issues in design integration are presented for a divertor (and first wall) with the molten salt Flinabe, a mixture of lithium and sodiu… more
Date: November 11, 2003
Creator: Nygren, R. E.; Rognlien, T. D.; Rensink, M. E.; Smolentsev, S.; Nelson, B. E.; Fogarty, P. J. et al.
Partner: UNT Libraries Government Documents Department
open access

A comparison of stresses in armor joints with and without interlayers

Description: Reliable joining of armor to heat sinks for plasma facing components has been a persistent problem in fusion and a concern for the International Thermonuclear Experimental Reactor (ITER). Post-fabrication and operating stresses in heat sinks with a 1mm compliant layer (or no interlayer) between tungsten armor and a CuCrZr channel were analyzed with a 2-D finite element model with temperature dependent properties, generalized plane strain, and strain hardening.
Date: November 1, 1997
Creator: Nygren, R.E.
Partner: UNT Libraries Government Documents Department
open access

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

Description: Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W-5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics desig… more
Date: August 1, 1999
Creator: Wong, C. P. C.; Nygren, R. E.; Baxi, C. B.; Fogarty, P.; Ghoniem, N.; Khater, H. et al.
Partner: UNT Libraries Government Documents Department
open access

A Fusion Reactor Design with a Liquid First Wall and Divertor

Description: Within the magnetic fusion energy program in the US, a program called APEX is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around the plasma. As part of this work, the APEX Team has investigated several possible design implementations and developed a specific engineering concept for a fusion reactor with liquid walls. Our approach has been to utilize an already established design for a future fusion reactor, the ARIES-RS, for the basic chamber g… more
Date: November 13, 2003
Creator: Nygren, R E; Rognlien, T D; Rensink, M E; Smolentsev, S S; Youssef, M E; Sawan, M Z et al.
Partner: UNT Libraries Government Documents Department
open access

A Fusion Chamber Design with a Liquid First Wall and Divertor

Description: The APEX study is investigating the use of free flowing liquid surfaces to form the inner surface of the chamber around a fusion plasma. We present a design for the chamber of a 3840MW fusion reactor based on the configuration for the chamber and magnets from ARIESRS but with a fast flowing molten salt of mixed Be, Li and Na fluorides for the first wall and divertor and molten salt blanket with a ferritic steel structure. Our design analysis includes strong radiation from the core and edge plas… more
Date: November 11, 2003
Creator: Nygren, R. E.; Sze, D. K.; Nelson, B. E.; Fogarty, P. J.; Eberle, C.; Rognlien, T. D. et al.
Partner: UNT Libraries Government Documents Department
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