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Analysis of the ATR fuel element swaging process

Description: This report documents a detailed evaluation of the swaging process used to connect fuel plates to side plates in Advanced Test Reactor (ATR) fuel elements. The swaging is a mechanical process that begins with fitting a fuel plate into grooves in the side plates. Once a fuel plate is positioned, a lip on each of two side plate grooves is pressed into the fuel plate using swaging wheels to form the joints. Each connection must have a specified strength (measured in terms, of a pullout force capacity) to assure that these joints do not fail during reactor operation. The purpose of this study is to analyze the swaging process and associated procedural controls, and to provide recommendations to assure that the manufacturing process produces swaged connections that meet the minimum strength requirement. The current fuel element manufacturer, Babcock and Wilcox (B&W) of Lynchburg, Virginia, follows established procedures that include quality inspections and process controls in swaging these connections. The procedures have been approved by Lockheed Martin Idaho Technologies and are designed to assure repeatability of the process and structural integrity of each joint. Prior to July 1994, ATR fuel elements were placed in the Hydraulic Test Facility (HTF) at the Idaho National Engineering Laboratory (AGNAIL), Test Reactor Area (TRA) for application of Boehmite (an aluminum oxide) film and for checking structural integrity before placement of the elements into the ATR. The results presented in this report demonstrate that the pullout strength of the swaged connections is assured by the current manufacturing process (with several recommended enhancements) without the need for- testing each element in the HTF.
Date: December 1, 1995
Creator: Richins, W.D. & Miller, G.K.
Partner: UNT Libraries Government Documents Department


Description: Production methods for producting homogeneous uranium-10 wt.% molybdenum fuel-alloy pin with a uniform zirconium cladding have been studied. The effect of various impurities has been investigated and the nature of and methods for ellminating cracking in swaged fuel pins have been examined. A major portion of the results is based on the study of material produced from 25-lb ingot castings. Small expertmental-scale ingots were also employed in impurity, heat-treatment, and cracking studies. On the basis of the data obtained recommendattons concerning casting, fabrication, and heat-treatment techniques necessary to produce a fuel pin of satisfactory integrity are presented. Specifications as to allowable carbon, chromium, iron, nickel, oxygen, and zirconium content are recommended. (auth)
Date: October 27, 1958
Creator: Fox, J.B.; Cheney, D.M.; Bauer, A.A. & Dickerson, R.F.
Partner: UNT Libraries Government Documents Department

Coextrusion and rotary swaging in gun tube fabrication

Description: The development of the processes was performed in the Hanford facilities of the United States Department of Energy at Richland, WA by the Pacific Northwest Laboratories which are operated by Battelle Memorial Institute. The impetus for the work was the desire to increase the life of rapid fire weapon barrels in order to decrease the overall systems cost. Loss of accuracy is the primary measure of gun barrel failure, which in turn is caused by barrel droop or whip, and loss of bore configuration or dimensions.
Date: March 1, 1978
Creator: Allison, G.S.
Partner: UNT Libraries Government Documents Department


Description: Full-size U10Mo foils are being developed for use in high density LEU monolithic fuel plates. The application of a zirconium barrier layer too the foil is applied using a hot co-rolling process. Aluminum clad fuel plates are fabricated using Hot Isostatic Pressing (HIP) or a Friction Bonding (FB) process. An overview is provided of ongoing technology development activities, including: the co-rolling process, foil shearing/slitting and polishing, cladding bonding processes, plate forming, plate-assembly swaging, and fuel plate characterization. Characterization techniques being employed include, Ultrasonic Testing (UT), radiography, and microscopy.
Date: March 1, 2010
Creator: Moore, G. A.; Jue, J-F; Rabin, B. H. & Nilles, M. J.
Partner: UNT Libraries Government Documents Department

The Fabrication of a Plutonium Helix for a Doppler Experiment. Work Completed: September 1956

Description: A helix constructed of plutonium was made to test the Doppler temperature effect in ZPR-III. The helix. 1 in. in diameter and 6 1/4 in. long. contained 240 g of deltaphase plutonium alloy encapsulated in titanium tubing. rour piutonium rods were extruded, joined together, and pushed into a titanium tube. This tube was swaged tightly over the piutonium rod. and the assembly was wound into a coil. Electrical leads to the coil were made by swaging copper tubing over the ends of the coil. The helix was tested by cycling about 500 times between 50 and 190 deg C. The coil was heated with a current of 130 amperes and cooled with a blast of chilled helium. Several helices of uranium were cycled during the same tests. Despite the severity of the thermal cycles. the helices were undamaged. (auth)
Date: February 1, 1958
Creator: Dunworth, R. J.; Rhude, H. V. & Kelman, L. R.
Partner: UNT Libraries Government Documents Department


Description: ABS>The fuel material specified for the Mark III core of EBR-I was uranium-2 wt. % zirconium alloy coextruded with Zircaloy-2 cladding. From previous work on swaged or rolled uranium-2 wt% zirconium alloy, it was anticipated that the extruded alloy would be dimensionally unstable under irradiation unless stabilized by suitable heat treatment. In order to determine an effective heat treatment, irradiation studies were made on both clad and unclad extruded uranium-2 wt.% zirconium alloy specimens at irradiation temperature estimated at 200 to 750 deg C. The irradiation specimens included material with three different heat treatinents, selected on the basis of previous studies, and material transient melted in its cladding. For unclad specimens, it was found that the irradiation temperature strongly influenced the various irradiation growth rates resulting from different heat treatments. Growth rates of the clad specimens were relatively insensitive to either irradiation temperature or prior heat treatment An exception was the transient-melted material, which shortened under irradiation. The cladding had only limited ability to restrain the swelling rates of specimens irradiated at the more elevated temperatures. Clad transient-melted material was found to be most resistant to high-temperature swelling under irradiation. The results of the present study combined with observations in earlier investigations resulted in a recommendation that the reference heat treatment for the core consist of gamma solution at 800 deg C followed by isothermal transformation at 690 deg C. (auth)
Date: October 1, 1959
Creator: Kittel, J.H.
Partner: UNT Libraries Government Documents Department

Materials considerations in accelerator targets

Description: Future nuclear materials production and/or the burn-up of long lived radioisotopes may be accomplished through the capture of spallation produced neutrons in accelerators. Aluminum clad-lead and/or lead alloys has been proposed as a spallation target. Aluminum was the cladding choice because of the low neutron absorption cross section, fast radioactivity decay, high thermal conductivity, and excellent fabricability. Metallic lead and lead oxide powders were considered for the target core with the fabrication options being casting or powder metallurgy (PM). Scoping tests to evaluate gravity casting, squeeze casting, and casting and swaging processes showed that, based on fabricability and heat transfer considerations, squeeze casting was the preferred option for manufacture of targets with initial core cladding contact. Thousands of aluminum clad aluminum-lithium alloy core targets and control rods for tritium production have been fabricated by coextrusion processes and successfully irradiated in the SRS reactors. Tritium retention in, and release from the coextruded product was modeled from experimental and operational data. Newly produced tritium atoms were trapped by lithium atoms to form a lithium tritide. The effective tritium pressure required for trap or tritide stability was the equilibrium decomposition pressure of tritium over a lithium tritide-aluminum mixture. The temperature dependence of tritium release was determined by the permeability of the cladding to tritium and the local equilibrium at the trap sites. The model can be used to calculate tritium release from aluminum clad, aluminum-lithium alloy targets during postulated accelerator operational and accident conditions. This paper describes the manufacturing technologies evaluated and presents the model for tritium retention in aluminum clad, aluminum-lithium alloy tritium production targets.
Date: August 1, 1994
Creator: Peacock, H. B. Jr.; Iyer, N. C. & Louthan, M. R. Jr.
Partner: UNT Libraries Government Documents Department

DEVELOPMENT OF PLUTONIUM-BEARING FUEL MATERIALS. Progress Report for Period January 1 through March 31, 1962

Description: During this reporting period, particular effort was of aced on powder blending and pellet sintering studies prior to irradiation sample fabrication, and, subsequently, the production and characterization of the pellets slated for irradiation. Also, PuO/sub 2/ and UO/sub 2/-PuO/sub 2/ characterization studies were continued, and new techniques are being developed. Specifically, dynamic moisture pickup determinations on PuO/sub 2/ were made in moist air, N, and CO/ sub 2/ atmospheres using a recording thermogravimetric balance; the Sharples Micromerograph was committed to Pu, and powder particle size distributions were measured and compared with previous determinations made with air-permeability equipment; and the suitability and reliability of analytical chemistry assaying procedures such as x-ray-fluorescence and gamma spectrometry are being evaluated. Prototype work on UO/sub 2/ for the direct precipitation of PuO/sub 2/ and PuO/ sub 2/-UO/sub 2/ feed materials for swaging, vibratory compaction, and dispersion fabrication was also continued. In addition, investigation of PuO/sub 2/ spherical particle formation by mechanical buildup and by plasma torch fusion was extended. Associated reactor physics studies were concentrated on the further comparison of Pu and U/sup 235/ in near-thermal converter reactors. In preparation for the fabrication of irradiation test specimens to be prepared by the mechanical blending of individuaI PuO/sub 2/ and UO/sub 2/ powders, bIending studies were initiated to develop methods required for the attainment of desired homogeneity. Sintering studies were carried out on PuOs/sub 2/ to study the effects of compaction pressure, firing temperature, firing time, and firing atmosphere. It was determined that 1400 to 1500 deg C is the best firing temperature to obtain maximum pellet density, and that sintering in air yields higher densities than sintering in a N/sub 2/--H/sub 2/ atmosphere. Further, it was noted that the degree of Pu/sub 2/O/sub 3/ formation while sintering in an N/ sub 2/--H/sub 2/ atm osphere ...
Date: October 31, 1962
Partner: UNT Libraries Government Documents Department