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The AFIP-6 (ATR Full-size-plate In center flux trap Position) Characterization Summary Report outlines the fresh fuel characterization efforts performed during the AFIP-6 experiment. The AFIP-6 experiment was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a scale prototypic of Advanced Test Reactor (ATR) fuel plates (45-inches long). The AFIP-6 test was the first test with plates that were swaged into the rails of the assembly. This test served to examine the effects of a plate in a swaged condition with longer fuel zones (22.5-inches long), that were centered in the plate. AFIP-6 test plates employed a zirconium interlayer that was co-rolled with the fuel foil. Previous mini-plate and AFIP irradiation experiments performed in ATR have demonstrated the stable behavior of the interface between the U-Mo fuel and the zirconium interlayer.

The Advanced Test Reactor (ATR) Full size plate In center flux trap Position (AFIP) experiment AFIP-6 was designed to evaluate the performance of monolithic uranium-molybdenum (U-Mo) fuels at a length prototypic to that of the ATR fuel plates (45 inches in length). The AFIP-6 test was the first test with plates in a swaged condition with longer fuel zones of approximately 22.5 inches in length1,2. The following report summarizes the life of the AFIP-6 experiment through end of irradiation, including a brief description of the safety analysis, as-run neutronic analysis results, hydraulic testing results, and thermal analysis results.

Deformation processed metal-metal composites (DMMC’s) are composites formed by mechanical working (i.e., rolling, swaging, or wire drawing) of two-phase, ductile metal mixtures. Since both the matrix and reinforcing phase are ductile metals, the composites can be heavily deformed to reduce the thickness and spacing of the two phases. Recent studies have shown that heavily drawn DMMCs can achieve anomalously high strength and outstanding combinations of strength and conductivity. In this study, Al-Fe wire composite with 0.07, 0.1, and 0.2 volume fractions of Fe filaments and Al-Ca wire composite with 0.03, 0.06, and 0.09 volume fractions of Ca filaments were produced in situ, and their mechanical properties were measured as a function of deformation true strain. The Al-Fe composites displayed limited deformation of the Fe phase even at high true strains, resulting in little strengthening effect in those composites. Al-9vol%Ca wire was deformed to a deformation true strain of 13.76. The resulting Ca second-phase filaments were deformed to thicknesses on the order of one micrometer. The ultimate tensile strength increased exponentially with increasing deformation true strain, reaching a value of 197 MPa at a true strain of 13.76. This value is 2.5 times higher than the value predicted by the rule of mixtures. A quantitative relationship between UTS and deformation true strain was determined. X-ray diffraction data on transformation of Al + Ca microstructures to Al + various Al-Ca intermetallic compounds were obtained at the Advanced Photon Source at Argonne National Laboratory. Electrical conductivity was measured over a range of true strains and post-deformation heat treatment schedules.

A joint University of Idaho (UI) and Idaho National Laboratory (INL) University Nuclear Research Initiative (UNERI) was to initiated to extend initial INL efforts to develop doped molybdenum/niobium alloy High Temperature Irradiation Resistant Thermocouples (HTIR-TCs). The overall objective of this UNERI was to develop recommendations for an optimized thermocouple design for high temperature, long duration, in-pile testing by expanding upon results from initial INL efforts. Tasks to quantify the impact of candidate enhancements, such as alternate alloys, alternate geometries, and alternate thermocouple fabrication techniques, on thermocouple performance were completed at INL's High Temperature Test Laboratory (HTTL), a state of the art facility equipped with specialized equipment and trained staff in the area of high temperature instrumentation development and evaluation. Key results of these evaluations, which are documented in this report, are as follows. The doped molybdenum and Nb-1%Zr, which were proposed in the initial INL HTIR-TC design, were found to retain ductility better than the developmental molybdenum-low niobium alloys and the niobium-low molybdenum alloys evaluated. Hence, the performance and lower cost of the commercially available KW-Mo makes a thermocouple containing KW-Mo and Nb-1%Zr the best option at this time. HTIR-TCs containing larger diameter wires offer the potential to increase HTIR-TC stability and reliability at higher temperatures. HTIR-TC heat treatment temperatures and times should be limited to not more than 100 C above the proposed operating temperatures and to durations of at least 4 to 5 hours. Preliminary investigations suggest that the performance of swaged and loose assembly HTIR-TC designs is similar. However, the swaged designs are less expensive and easier to construct. In addition to optimizing HTIR-TC performance, This UNERI project provided unique opportunities to several University of Idaho students, allowing them to become familiar with the techniques and equipment used for specialized high temperature instrumentation fabrication and evaluation and to author/coauthor several key conference papers and journal articles.

Fabrication was well under way for the JOYO biaxial creep and tensile specimens when the NR Space program was canceled. Tubes of FS-85, ASTAR-811C, and T-111 for biaxial creep specimens had been drawn at True Tube (Paso Robles, CA), while tubes of Mo-47.5 Re were being drawn at Rhenium Alloys (Cleveland, OH). The Mo-47.5 Re tubes are now approximately 95% complete. Their fabrication and the quantities produced will be documented at a later date. End cap material for FS-85, ASTAR-811C, and T-111 had been swaged at Pittsburgh Materials Technology, Inc. (PMTI) (Large, PA) and machined at Vangura (Clairton, PA). Cutting of tubes, pickling, annealing, and laser engraving were in process at PMTI. Several biaxial creep specimen sets of FS-85, ASTAR-811C, and T-111 had already been sent to Pacific Northwest National Laboratory (PNNL) for weld development. In addition, tensile specimens of FS-85, ASTAR-811C, T-111, and Mo-47.5 Re had been machined at Kin-Tech (North Huntington, PA). Actual machining of the other specimen types had not been initiated. Flowcharts 1-3 detail the major processing steps each piece of material has experienced. A more detailed description of processing will be provided in a separate document [B-MT(SRME)-51]. Table 1 lists the in-process materials and finishedmore » specimens. Also included are current metallurgical condition of these materials and specimens. The available chemical analyses for these alloys at various points in the process are provided in Table 2.« less

The 93% W-5.6% Ni-1.4% Fe and 93.1% W-4.7% Ni-2.2% Co alloys (WHA) provided by Army Research Laboratory (ARL), Aberdeen are characterized to determine the effects of matrix alloying and swaging on the microstructure and fracture behavior. The W particles are oblong with respect to the swaging direction. The microstructure of the W-Ni-Fe alloy reveals good cohesive bonding between W particles, but there is W-matrix interface separation and matrix alloy cracking. The microstructure of the W-Ni-Co alloy reveals regions of good cohesive bonding between W particles, but also regions where some wetting has not occurred by the liquid. No evidence was observed of matrix alloy cracking. The fracture characteristic of WHA is dominantly cleavage of W particles.

The successful development of deformation-processed metal-metal composites (DMMC) offers the potential for ductile, high-strength structural materials with high-temperature stability. An infiltration casting process was used to permeate steel wool preforms with molten magnesium-lithium (Mg-Li) alloys. The selected matrix alloys were hexagonal close packed (HCP) Mg-4wt%Li or body centered cubic (BCC) Mg-12wt%Li; the low carbon steel wool fibers were predominantly BCC ferrite. These cast HCP/BCC and BCC/BCC composites were deformed by rolling or by extrusion and swaging. Mechanical properties, microstructure, and texture development of the composites were characterized at various levels of deformation. The HCP/BCC composites had limited formability at temperatures up to 400 C while the BCC/BCC composites had excellent formability during sheet rolling at room temperature but limited formability during swaging at room temperature. The tensile strengths of these HCP/BCC and BCC/BCC composite materials increased moderately with deformation, though less than predicted from rule of mixtures (ROM) calculations. The microstructure was characterized to correlate the filament size to the deformation strain and mechanical properties of the composite material. Stereological measurements of the filament size were used to adjust ROM calculations to reflect the actual deformation strain in the fibers. However, the experimental strengths of these composite materials were stillmore » less than ROM predictions, possibly due to the presence of considerably large fibers. Of the many models used to describe the strengthening observed in DMMC materials, the Hall-Petch relationship adequately described the experimental data. Texture development was also characterized to explain the deformation characteristics of the composite materials. Chapters 2, 3 and 4 are not included here. They are being processed separately.« less

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.

An investigation of various closure techniques was performed in an effort to meet requirements for closure of evacuated waste cans to be used at Westinghouse Idaho Nuclear Company`s Idaho Chemical Processing Plant. Although other sealing techniques are available, welding was considered to be the best for sealing the cans. For various reasons. techniques other than welding are not suitable for cans that are subject to the Hot Isostatic Press (HIP) process. For example. elastomeric seals. solders. and brazing would,not withstand the temperature associated (approximately 1000 degrees centigrade) with the HIP process. Mechanical joining techniques such as threading, crimping, and swaging could result in the joint opening as the can wall is deformed during HIP process. Unlike the above joint methods, welding results in physical joining of the lid or plug to the can itself, thus the wall would have to be ruptured before leakage would occur. This document investigates welding techniques for application to the can closure.

This report is a memorandum requesting that two swaged thorium rods being held a MIT be sent to the Savannah River Plant for examination. The memorandum is dated October 15, 1952.

Compacted UO/sub 2/ powder specimens with purposely defected stainless steel cladding were erosion flow tested under simulated boiling-water reactor operating conditions, excluding neutron flux. The type of clad defects were 0.026inch wide slits, completely through the clad, varying in length from 0.25 to 1 inch. Maximum erosion flow test exposure was 6000 hours in 7 ft/sec, 536 deg F flowing water. The results of post test examination indicated that defected stainiess-clad swaged fuel with UO/sub 2/ densities greater than 90% T.D. is comparable to pellet type fuel with respect to erosion resistance. Within the test parameters, the stresses and imperfections introduced into the stainless cladding as a result of swaging did not adversely affect the corrosion characteristics of stainless steel in a boiling water reactor coolant environment. As a result of the test data, two swaged fuel rods containing 0.50-inch long slits were inserted into the VBWR reactor. The defected rods are performing satisfactorily after accumulating 1000 Mwd/t exposure, with activity due to fission gas release below 20 mu c/sec. (auth)

Tantalum fabrication methods primarily for use as a container for molten plutonium are described. One method involved the rolling of a Ta billet into sheet of the desired age, cutting a circular blank, and deep drawing the tantalum blank through successive stages into the desimed shape. Another method used for the fabrication of Ta containers consisted of a combination of impact extrusion and ironing. This method involved the extrusion of the cast billet into rod, swaging the rod to a suitable diameter, and cutting it into slugs of the desired length. The slugs were then impact rod, swaging the rid ti a suitable diameter, and cutting it into slugs of the desired length. The slugs were then impact extruded into heavy walled containers. These starting containers were then taken through successive dies where the wall thickness was ironed down to the desired gage. (auth)

The fabrication of UO/sub 2/ fuel elements clad in metallic sheaths by swaging, rolling, and coextrusion was investigated. The effects of the type of UO/sub 2/ and of the materials and dimensions of the sheath were investigated. Fused UO/sub 2/ swaged in stairless-steel tubing reached a maximum density of 93% of theoretical. (auth)

Uranium --molybdenum alloy rods containing from 10 to 15 wt% Mo and 1/16- in. in diameter were successfully fabricated by hot rotary swaging, followed by machining to remove the protective sheathing (Inconel with molybdenum barrier). Structurally strong rods with densities greater than 95% of theoretical were produced from both calciumreduced uranium mixed with hydrogen-reduced molybdenum and acid-cleaned, prealloyed shot when reduced in area about 55% at 1050 or 1100 deg C. Alloy homogeneity was good with prealloyed powders; however, traces of molybdenum -rich, gamma phase persisted in the elemental uranium -molybdenum material after swaging at 1100 deg C. Swagings embodying hydride uranium or oxide- contaminated prealloyed shot were unsatisfactory because of insufficient consolidation or poor interparticle bonding. (auth)

Progress is reviewed on the fabrication techniques for rare earth metals including the processing into ingots, extrusion, forging, hot and cold rolling, swaging, and machining. (J.E.D.)

ABS>High-Power Density Fuel Development. High-power density VBWR fuel assemblies were irradiated to approximately 400 Mwd/t and found to be in good condition after irradiation. Fabrication and calibration of two instrumented VBWR fuel assemblies are described. Fuel Fabrication Development. A tandem- rolled fuel bundle was produced. Swaging studies were carried out for stainless steel-clad fuel rods. Inspection of defected clad powder-compacted UO/sub 2/ specimens after 570 hr of erosion flow testing in water or steam at 535 to 545 deg F indicated that only small amounts of U0/sub 2/ had been eroded. A study of density-particle size relationships was made for vibratory-compacted binary and ternary fused UO/sub 2/ powder mixes. Design characteristics are given for 10 special HPD-VBWR assemblies. Vibration tests on a 36-rod Consunaers-type element are reported. Stability, Heat Transfer, and Fluid Flow. The results of a stability analysis of the 50-Mw Big Rock Reactor are presented. Physics Development. The effects of enrichment variations in the fuel rods and of inserted control rods on the power distribution were studied. Studies of gross radial power distribution for a 300-Mw reactor indicate that proper radial cycling of partially burned fuel can produce desirable radial power shapes. The physics analysis of the reactormore » core for a laine 300-Mwe conceptual design is summarized. The incentives and functions of a computer to be used for maximizing fuel burnup are discussed. (D.L.C.)« less

Samples of 304 stainless steel cladding from unirradiated fuel rods fabricated by cold swaging and tandem rolling were investigated to determine the effects of cold working. These fabrication techniques were found to produce a structure containing a large amount of slip bands. Results of tensile tests and fracture behavior observations indicate that this cladding should be adequate for boiling reactor fuel rods. (J.R.D.)

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)

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)

Data from a survey of the literature and other available information on zirconium--uranium alloys have been reviewed for the purpose of obtaining a coherent picture of current knowledge about the properties and behavior of zirconium--uranium alloys. The results of the survey were used to revise and extend the presentation of material gathered earlier in a similar study and reported in BMI1030 in August 1955. The constitution of zirconium-uranium alloys is discussed, and a constitutional diagram for the system is presented. The effects of oxygen and nitrogen, which are present in these alloys as contaminants, on alloy constitution ars shownin the form of ternary diagrams and in terms of their quantitative effects on the phases present. The transformation kinetics and the nature of the transformation of the high-temperature body- centered-cubic gamma phase to the phases stable at room temperature are described. Two regions are discussed: in the 20 to 70 wi. % uranium composition range, gamma, which is retained on quenching, transforms isothermally to the intermediate epsilon-phase structure by a diffusion-controlled nucleation-and- growth process; in alloys containing less than 20 wt.% uranium, gamma transforms martensitically to a strained alpha-zirconium structure on quenching, with the diffusion-controlled transformation of gamma to either epsilonmore » or alpha proceeding above the M/sub s/ temperature. Typical zirconium -- uranium alloy microstructures produced by working and by various heat treatments are shown, and a correlation of the structure with the manner in which it was produced and the resultant mechanical properties is attempted. Hardness, tensile, and creep data for zirconium--uranium alloys are tabulated and evaluated in terms of alloy constitution and transformation-kinetics behavior. Physical-property data for zirconium-uranium alloys are presented in tabular and graphical form. The best fabrication techniques for forging, hot and cold rolling, extruding, and swaging are detailed. The corrosion resistance of zirconium-uranium alloys in wafer at various temperatures and corrosion behavior as affected by alloy constitution are discussed. The effect of ternary additions on corrosion resistance is discussed, and pertinent data are tabulated. The influence of zircocium -uranium alloy constitution and properties on neutronirradiation behavior is discussed, and the general effects of irradiation upon these alloys are summarized. (auth)« less

The lifetime of a B/sub 4/C-in-tubes'' control rod may be limited by either nuclear worth depreciation or internal gas pressure buildup as helium is formed and released from the B/sub 4/C particles. Nuclear life is enhanced by thin tube walls for any given O.D. of tubing and by high B/sub 4/C density, but the time to failure from gas pressure buildup is increased by the opposite. The optimum B/sub 4/C density and tube wall thickness were calculated for maximum control rod life, using several release fractions of generated helium and a number of allowable cladding stress values. It was estimated that lifetimes of l0 or more years for B/sub 4/C-in-tubes control systems are achievable through the proper selection of tubing material and B/sub 4/C density values. The B/sub 4/C densities required for maximum lifetimes are in a range that can be obtained by vibratory compaction followed by swaging. (auth)

Reliable line disconnects are necessary to permit the remote replacement of process equipment in the Transuranium Processing Facility. Two disconnect designs were subjected to several severe tests in order to select the most suitable design. These performance tests demonstrated the superiority of the surface-seal concept over that of the line seal-concept and the reliability of this seal when subjected to misalignment forces, corrosive solutions, rough handling and interchange of mating parts. Tests of Hastelloy C disconnects after heat treatment demonstrated the necessity for joining the disconnect to the process tubing by means other than welding. Several leak-tight joints were made by external mechanical swaging, by external electrical swaging (by the Magniform'' process developed by the General Dynamics Corporation) and by internal rolling. Hastelloy C disconnect samples machined to reasonable machine tolerance established the economic feasibility of the design. (auth)

The objective of this supplement is to authorize a change in the panellit trip range from 25--75 psi to 5--95 psi. The test hole facility consists of two concentric aluminum tubes which extend from the front face to the rear face of the reactor. The ID of the inner tube is 2--7/8 inch. Water from one crossheader supplies the annulus, water from another crossheader supplies the inner tube. The three-foot-long, .570 inch OD fuel element is centered in a 40-inch long aluminum holder which has an ID of 1.380 inch and an OD of 2.800 inch. The panellit gage which monitors the flow to the inner tube fluctuates to such an extent during start-up that on two occasions the reactor was scrammed. During equilibrium operation the panellit gage reading remains stable. A possible explanation of this behavior is that during start-up aluminum spacers which are in the inner tube as part of the test charge chatter and cause variations in the water path through the tube. It is further surmised that at equilibrium operation the pressure drop across the column in the tube is sufficient to suppress the chattering. It is concluded that extending the trip range to 5--95 psimore » will reduce the possibility of scramming without increasing any hazards to the reactor.« less