Nuclear fuel cost of 1.25 mills/kwh would make nuclear power competitive with conventional power in lowcost coal areas if capital and operating costs can be brought to within about 10 percent of those of coal-fired plants. Substantial decreases in fuel fabrication cost are anticipated by 1970: other costs in the fuel cycle are expccted to remain about the same as at present. Unit costs and irradiation levels that would be needed to give a fuel cost of 1.25 mills/kwh are believed to be attainable by 1970. (auth)

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In the fabrication of the steam generator on APPR-1A it was considered necessary to roll the Type 430M tubes into carbon steel tubesheets to ASTM Specification A350-Grade LF-1, modifled with 1.66% nickel; and weld the tube ends to the stainless steel overlay previously applied to the tubesheet. The rolled joint was a necessary precaution to prevent secondary water, that might contain chlorides, from contacting the stainless steel weld joining the tubes to the tubesheets. The welded joint provided the mechanical strength for attaching the tubes to the tubesheets. A laboratory program was conducted, therefore, to develop practicable procedures for welding the Type 430M tubes to the stainless steel overlay; as well as to assure that the tubes could satisfactorily be rolled to the tubesheets. Automatic and manual tungstenare welding procedures were developed that were capable of consistently providing an austenitic weld having a strength exceeding that of the heat affected zone or the unaffected tube itself. Type 430M tubes in the asreceived, and softened conditions were rolled into prototype test units under various conditions of rolling. It was concluded that the Type 450M tubes in the as-received condition could be satisfactorily rolled into the A360Grade LF-1 tubesheet and be tlght …

A study was made of hydraulic characteristics of Yankee-type fuel rod assemblies using experimental and analytical methods. Two scale model fuel assemblies utilizing both ferrule and strap type arrangements were constructed and tested at atmospheric pressure and room temperature. Analytical methods using semiempirical relationships are substantiated by experimental results for both the fuel assembly having strap-type spacers and the fuel assembly having cylindrical ferruletype spacers. The experimental pressure drop across the assembly model using either straps or ferrules correlated within 5% of the value calculated by means of equations based on the equivalent diameter concept for flow inside pipes. The individual frictional drops along the rods and across the end plates and straps correlated within 15% of the predicted pressure drops. The indlvidual pressure drops across both the staggered ferrule sections and the full ferrule section correlated to within 17% of the predicted pressure drops. Comparison of the ferrule and the strap pressure drops indicates that the pressure drop across a level of straps was more than four times the pressure drop across a full ferruled section. It is concluded that the analytical methods based on the equivalent diametcr concept can be satisfactorlly used to calculate pressure drops for flow …

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(((Abstract unscannable)))<><DSN>13:014503<ABS>Thc Nb-O equilibrium system was determined by metallographic examination of arc-cast alloys made ot electron-gun-refined Nb metal and special purity Nb/sub 2/O/sub 5/. Two intermediate oxides. NbO and NbO/sub 2/, melt without decomposition at 1945 C and 1915 C, respectively. Eutectic reactions exist between Nb and NbO at 1915 C and between NbO and NbO/sub 2/ at 1810 C . Experimental evidence supports a peritectic reaction between NbO/sub 2/ and Nb/sub 2/O/sub 5/ at 1510 C. The maxinium solid solubility of 0 in Nb metal is 0.72 wt.%. (auth)

Following-the K-pile start-ups in early 1955, a program of planned power raises was begun. The operating level had reached 1700--2000 MW by late 1955, and a severe operational control problem became apparent; the power distribution in the reactor was difficult to control and appeared inherently unstable. A study of available data led to the initiation of a production test so that a more detailed study of the phenomena could be made. This report describes the measures taken which led to an improvement in the operating characteristics of the K-piles; the current status and future outlook are also discussed in a general way.

A fuel-element assembly from the first loading of the OMRE was examined in detail after experiencing an average uranium burnup of between 1 and 2 at.%. The rate of decay heat generation was evaluated by temperature monitoring of the shipping-cask coolant. Temperatures of the fuel-element-box assembly and the fuel plates were measured with thermocouples and tempilstiks. Structurally, the fuel-element assembly was affected very little by either radiation or the organic coolant-moderator. Although there was some distortion in the side and end plates of the assembly, the coolant channels between the fuel plates were free from major fouling and obstructions. The channel cross sections were reduced at specific points less than 5 per cent. The plates studied were subjected to complete gamma scanning. Specimens removed from selected areas of the scanned plates were radiochemically analyzed for burnup and the results correlated with the gamma-scan data. Burnup profiles were constructed for each of the scanned plates. The gamma-scan data were also utilized to determine the average plate burnup. (auth)

BS>Critical concentration calculations were made for several D/sub 2/O-H/ sub 2/O moderated HRT-type reactors with 30- and 28-in. core diameters and pressure vessel diameters of 60 and 54 in. A core temperature of 300 C was assumed for all cases while the blanket temperatures assumed the values 250, 280, and 300 C. The assumed moderator compositions were 80, 90, and 100% D/sub 2/O. (auth)

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Information concerning graphite contraction applicable to high- temperature, graphite-moderatored reactors is presented. The scope includes relevant data from all available sources, interpretation and extrapolations as can reasonably be made, and a discussion of the effects observed in terms of current radiation damage theory. Limits of accuracy and a discussion of experimental techniques are presented. (auth)

Thermal-conductivity measurements are in progress on an unirradiated, unclad, natural U specimen. Data are presented on thermal conductivity measurements performed on UO/sub 2/. The creep properties of annealed and of 15% cold-worked Zircaloy-2 are being studied. A program was initiated to evaluate loss-of-coolant incidents in the PRTR by means of simulation on a digital computer. Research on the casting of hollow Al-35 wt. extrusion billets is reported. Further refinement of the method developed for the analysis of Mg in cement is in progress. The infrared and gaschromatography analysis of irradiated dodecane, decane, cetane. and octane, and their urea complexes, were continued. The manner in which U metal solidifies in cylindrical graphite molds is under study. Work has continued on development of a stabilized hightemperature nuclear fuel capable of operation in either oxidizing or reducing atmospheres. Progress in the stud of potential fueled moderators has continued with the determination of hydrogen-absorption isotherms for the Zr-25 wt. alloy. The effect of fast-neutron flux on the mechanical properties of AISI Tvpe 347 stainless steel are being determined and evaluated. The forging of Nb-U alloys is reported. Thorium-uranium alloys are being studied for the purpose of developing improved corrosion resistance and irradiation stability of …

The equations and nomographs most frequently used intended calculating behavior of stack effluents are given and explained. (T. R. H.)

A radiological safety analysis is presented for the SNAP-III thermoelectric generator. Since the fuel of the device is polonium-210, a toxic radioisotope, certain safety measures have been designed into the device and its shipping container to prevent a release of the contaminant into any environment during normal operation or a catastrophic accident. Once containment is assured, the direct radiation problem is considered. It has been shown that the direct radiation from the thermal source is kept within tolerance limits by surrounding materials and spatial and temporal factors. It must be emphasized that this device should not be deliberately abused or mishandled since this would serve to increase the probability of accident. The device has been evaluated with respect to internal forces such as heat and helium pressure and external forces such as impact and chemical attack. The mechanical thermal and chemical integrity of the thermoelectric generator is shown to be quite reliable. The basic physical, chemical, thermal, atomic and nuclear characteristics of polonium-210 have been presented. Potential internal and external radiation hazards have been set forth. (auth)

The reaction between formaldehyde and nitric acid, in which the acid is destroyed with the production of predominantly gaseous products, has been recognized as of great potential value in the processing of radioactive fuels, particularly during waste treatment. Laboratory studies of the reaction at Harwell and at Hanford have shown that a major fraction of the nitric acid can be readily removed from an acidic solution containing nitrates by the addition of formaldehyde. The process possesses the advantages of low chemical cost; recoverability of nitric acid; and, in the case of waste treatment, the production of a solution relatively low in inert salt concentration suitable for fission product recovery or ultimate disposal. The primary purpose of the study was to confirm and extend existing information on the formaldehyde reaction to the destruction of nitric acid in Purex type waste (1WW) through operation of pilot plant scale apparatus. Operational behavior, formaldehyde utilization efficiency, and safety considerations were particular subjects of study. In addition, destruction of nitric acid in a Darex-type dissolver solution was investigated.

The energy distribution of ions and electrons in DCX are being studied by means of the Fokker-Planck approximation to the Boltzmann equation. An IBM- 704 code, called FOPP, was constructed to solve simultaneously the coupled Fokker-Planck equations for each of the two species of particles. This report discusses the difference scheme employed and derives the boundary conditions necessary in order that this difference scheme conserve energy and particles in the absence of sources and sinks. In particular, detailed discussion is given of problems arising from the use of two grid sizes, which proved advantageous on account of the great difference in the mass of ions and electrons. (auth)

The possibility of measuring the electromagnetic form factor of the pion by extrapolation of the cross section for e/sup -/ + p 1100 deg C are n + The effects of /sup +/ + e/sup -/ was investigated. The method is based on the existence of a pole in the electropionproduction scattering amplitude as a function of the invariant momentum-transfer of the nucleon. The residue of this pole is the pion form factor multiplied by a known coefficient. Since the pole lies slightly outside the physical region of the invariant momentum transfer, an extrapolation of the experimental data is required. An approximate calculation of the electropion-production cross section was made in order to estimate the experimental accuracy necessary for a significant extrapolation. Accuracy is required which is an order of magnitude better than that achieved at present in similar experiments. (auth)

To determine the safety of the array in the storage vault for the SM-2 experimental fuel plates, two criticality criteria were applied. A maximum of 18 fuel plates was stored in sthainless steel tubes and the tubes belted to a frame on the wall to prevent movement. No tube could go critical by itseIf. The vauit was then assumed completely flooded by water. In the first calculation, the fuel array was assumed to be distributed uniformly over the wall forming a large slab. This method indicated the array might be critical if the steel tube and cadmium lining were neglected. In the second method, a conservative calculation, wnich included the steel tube and cadmium lining was made. This method indicataed the array was subcritical. Calculations were then made of the criticalty of the SM-2 vault without the steel--cadmium tubes and wcoden blocks. The multiplication factor of the vault was also calculated. In order to determine the accuracy of these calculations, an ORNL critical experimental array was calculated applying the same analytical techniques. (M.C.G.)

A hydrostatic journal bearing mounted near the impeller of an overhung vertical shaft centrifugal pump was subjected to water testing as a part of the molten salt lubrication investigation at ORNL. Three tests were performed with bearings having radial clearances of 0.003 in., 0.0075 in., and 0.005 in. The first journal and bearing (0.003 in. radial clearance) were found to be heavily scored after testing. Only faint localized scratches were found on tho second journal (0.0075 in. radial clearance) and these may have been caused by the many test starts and stops. Localized scratches, somewhat deeper than those on the second journal, were observed on the third journal, but no measurable wear had occurred from testing. An apparent inconsistency was noted in that at the same pump operating condition the bearing load as computed from pocket pressure data increased by a factor of 1.2 to 1.7 as bearing radial clearance was increased from 0.005 in. to 0.0075 in. The configuration of the submerged hydrostatic bearing used in these tests appears to be satisfactory for use as a lower journal bearing in this size and type of centrifugal pump, at least insofar as operation in water is concerned. (auth)

This document details activities of the irradiation processing department during the month of January 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering operation; Employee Relations Operation; and Financial Operation.

A fuel-cycle economic study was made for a 315-Mw(e) graphite-moderated U/sup 235/-Th-fueled fused-salt reactor. Fuel cycle costs of approximately 1.3 mills/kwh may be possible for such reactors when reprocessed for U/sup 233/ and U/ sup 235/ recover y at the end of a 9-year cycle. Continuous removal of fission products during the reactor cycle does not appear to offer any great economic advantage for the converter reactor considered. (auth)

The two predominate methods of dispersing uranium in graphite are reviewed and evaluated. This study indicated that the most feasible method of dispersing uranium in graphite would be to fabricate a mixture of graphite and U/ sub 3/O/sub 8/ bonded with a thermosetting resin. A commercial type graphite was developed through independent research, and this fabrication procedure was adapted for the manufacture of the TREAT fuel matrix. (auth)

This report for January 1959, from the Chemical Processing Department at HAPO, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance: Financial operations; facilities engineering; research; and employee relations.

Several functions used in the calculation of x-ray crystallographic intensities are tabulated over large ranges. These tabulations include Lorentz- polarization factors as a function of Bragg angle, the Debye function as a function of THETA /T, and the Debye-Waller temperature factor as a function of B for selected sin theta / lambda values. (auth)