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Acceptance sampling plans

Description: A compilation is presented of attributes and variables acceptance sampling plans and procedures which may be used as a ready reference to provide protection to both supplier and purchaser whenever sampling inspection is involved. Some of the basic and underlying principles involved in these plans and procedures are discussed. The plans and procedures provide for selection of adequate sample quantities and evaluation of inspection data for the purpose of determining conformance to the specified lot tolerance percent defective (LTPD) and BETA error acceptance criteria. (auth)
Date: November 1, 1973
Partner: UNT Libraries Government Documents Department

Heat pipe thermal control of irradiation capsules

Description: From 1st international heat pipe conference; Stuttgart, F.R. Germany (15 Oct 1973). The use of heat pipes to control the temperature of irradiation capsules containing fast breeder reactor structural materials is discussed. (TFD)
Date: April 30, 1974
Creator: Deverall, J.E.
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

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

Reactor Material Program Fracture Toughness of Type 304 Stainless Steel

Description: This report describes the experimental procedure for Type 304 Stainless Steel fracture toughness measurements and the application of results. Typical toughness values are given based on the completed test program for the Reactor Materials Program (RMP). Test specimen size effects and limitations of the applicability in the fracture mechanics methodology are outlined as well as a brief discussion on irradiation effects.
Date: March 28, 2001
Creator: Awadalla, N. G.
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

Fusion Materials Semiannual Progress Report for Period Ending December 31, 1998

Description: This is the twenty-fifth in a series of semiannual technical progress reports on fusion materials. This report combines the full spectrum of research and development activities on both metallic and non-metallic materials with primary emphasis on the effects of the neutronic and chemical environment on the properties and performance of materials for in-vessel components. This effort forms one element of the materials program being conducted in support of the Fusion Energy Sciences Program of the U.S. Department of Energy. The other major element of the program is concerned with the interactions between reactor materials and the plasma and is reported separately.
Date: April 1, 1999
Creator: Rowcliff, A.F. & Burn, G.
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