The potential-current density relationships for 304 stainless steel dissolution in a nitrate system were studied as a function of solution composition and temperature in order to optimize the conditions for electrolytic dissolution of ihis material. In the nitrate system, the anodic dissolution of steel takes place in the transpassive region. Under some conditions, deviations from Tafel behavior are observed which depend greatly on the nitrate and hydrogen ion concentration, and on temperature. A discussion of passivity, transpassivity, secondary passivation, the limiting current density, and the effect of alloy composition on the dissolution behavior is given. It was found that at temperatures above 60 deg C efficient dissolver operation should be possible over a wide range of solution compositions and at current densities up to 2 amp/cm/sup 2/. (auth)

Mixed Fission Products (MFP) for use as a heat source for thermoelectric generators will become increasingly available in the coming years. The Atomic Energy Conamission sponsored program on solidification of nuclear wastes is now entering the hot-bench scale test phase. During this phase approximately 5000 thermal watts of two year old MFP could be produced monthly. Two different types of hot calcination pilot plants are planned for installation at the Hanford National Laboratories in the 1964 to 1966 time period. Each of these plants should be able to produce 160,000 thermal watts of two year MFP and 16,000 thermal watts of ten year MFP on a monthly basis. During this phase, MFP costs should be less than 15 per ihermal watt for two year MFP and 50 for ten year MFP. This cost includes operation of the plant solely to obtain heat sources and sealing the MFP into fuel containers. A full scale plant for a 15,000 Mw(e) nuclear economy is estimated to produce four to five times as much MFP as either of the pilot plants. Costs will be dependent upon AEC policy in effect at the time the plant is operating. lf the policy indicates that the full …

Thoria compacts containing refractory metal fibers in quantities as low as 5 wt% showed significantly better resistance to thermal shock spalling than thoria alone. Of the metals and alloys evaluated, molybdenum and niobium gave the best results. Values are presented for room- and elevated-temperature properties of fiber-reinforced thoria. Measured properties included compressive strength, modulus of rupture, impact strength, thermal shock resistance, thermal conductivity, thermal expansion, and oxidation resistance. The widespread presence of microcracks in these compacts resulted in significantly lower strengths and elastic moduli than those of thoria. Fibers improve resistance to thermal spalling by suppressing and limiting crack propagation and by structurally reinforcing the cracked body. Room-temperature thermal conductivity of reinforced thoria is slightly higher than that of thoria, but at 1600 deg C it is 3 times greater. Oxidation resistance of molybdenum-reinforced thoria is best of all combinations investigated and improves with increased specimen density. However, because of the presence of microcracks, even the densest specimens are severely attacked after 24 hr in air at 1000 deg C. Thoria-urania compacts reinforced with molybdenum fibers were bonded together by means of conventional brazing techniques. Irradiation of metal-reinforced thoria-urania specimens indicated that molybdenum fibers aided the heat transfer from the …

The state of the pulse-height analyzing art is reviewed with particular emphasis on the developments of the past two years. The discussion includes consideration of multidimensional instruments, calibration-stabilizing techniques, and some of the auxiliary features that are becoming increasingly available on commercial instruments. Possible future developments with respect to resolving time and memory organization is discussed briefly. (auth)

This program of the GEORGE Assembly Routine (GAR) will accept Fortran- like statements from paper tape and create the GAR language program on tape. This includes the needed calls for common subroutines and the reservations for the named variables and temporaries. The original statements in Fortran are carried along as remarks. The GAR language program may then be processed in the usual way by the GEORGE Assembly Routine, giving machine-language code. The level of sophistication of the source language is roughly equal to that of Fortransit or SALT. (auth)

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The purpose of these tests is to provide necessary laboratory data in support of an R,PEO program in determining the energy loss associated with various hardware size combinations on the rear face of the B-D-F reactors. The original method used to check for critical flow was determined to be faulty. A revised method demonstrated critical flow did occur in the 5/8-inch inconel connector and combination 1 fittings. The remaining fitting combinations with the 5/8-inch inconel and 3/4-inch aluminum connector were not rechecked because of the reaming of the I.D. to permit the continuation of the original tests. During test number 6, audible cavitation was heard with the highest severity at a point midway between pressure points 3 and 4 on the connector. This condition appeared again in tests 6A, 7, and 7A, with incipient cavitation at approximately 40 gpm in each test, regardless of the rear header pressure and/or temperature.

In 1959 the U. S. Atomic Energy Commission contracted with the General Electric Company to carry out a Sodium Mass Transfer investigation with the objective of evaluating three prospective steel construction materials for utilization in liquid sodium-cooled reactor systems operated in the temperature range of 600 to 1300 F. Six liquid sodium test loops were designed and built to implement the study of the three selected materials: AISI 316 stainless steel (316SS); 2 1/4 percent Chromium- 1 percent Molybdenum steel (2 1/4 Cr-1Mo), and 5 percent Chromium - 1/2 percent Molybdenum- 1/2 percent Titanium steel (5Cr-1/2Mo-Ti). This document reports and interprets extensive observations of the metallurgical effects noted after exposure of the materials to liquid sodium. This exposure was made under dynamic conditions in test runs of up to 3000 hours from the initial startup through December 1961.

Twelve 0.22-in.-diameter fuel specimens containing a longitudinal central vent and clad with 0.010 in. of Type 304 stainless steel were irradiated to evaluate the effect of restraint and a central vent on fuel element stability. The cladding of 10 of the specimens contained porous end plugs to vent any released fission gas and thus to minimize the buildup of gas pressure within the stainless steel cladding. The specimens consisted of a 20% enriched uranium--2 wt% zirconium alloy core surrounded by a natural uranium--2 wt% zirconium alloy sleeve. Eight of the specimens were irradiated to burnups of the enriched core of 6.9 to 12.8% of all atoms (1.2 to 2.2 at.% of the duplex assembly) at maximum fuel temperatures ranging from 280 to 760 deg C. Most of the clad specimens exhibited negligible volume increases as a result of irradiation. Two specimens containing central vents but unclad were irradiated together with the clad specimens in an attempt to differentiate between the effects due to a central vent and the effects due to cladding. The central vent in itself did not appear to reduce the swelling characteristics of the alloy. Mechanical restraint appeared to have extended the useful operating temperatures of the …

A total of 11,117 enriched uranium-5 wt.% fissium alloy fuel pins was manufactured for EBR-II, Core I. These were made from a synthetic fission product alloy of nonradioactive elements, natural uranium, and enriched uranium. The material was supplied as precast billets. The manufacturing methods were developed for the EBR-III Fuel Cycle Facility. Experimental refabrication equipment was used to production test both the methods and the equipment. The billets were induction melted and pressure cast into precision-bore, high-silica glass molds in batches of 90 to 160. The number of molds used was adjusted according to batch weight. After casting, the molds were broken away and the castings were fed into a pin-process and inspection machine. Both ends were sheared from the castings to produce finished pins measuring 0.144 in. in diameter by 14.22 in. long. The pins were inspected for diameter, porosity, weight, and length. Rejected pins and sheared ends were broken into short lengths and returned for consolidation. Acceptable fuel pins were sealed and sodium bonded in stainless steel jackets, and assembled into Core-I fuel elements. (auth)

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The Darex process developed for ihe recovery of U from stainless steel- containing reactor fuels consists of dissolution of the fuel material in dilute aqua regia, removal of chloride from the solution to prevent corrosion of downstream stainless steel process equipment, and adjustment of the nitrate solution to solvent extraction feed conditions. Each step can be either continuous, semi-continuous, or batch with continuous operation showing much higher throughput for comparable equipment. The preferred dissolvent is 5 M HNO/ sub 3/-2 M HCl, since dissolution rates and metal loadings are near maximum. Nitric acid from 60 to 95 wt% can be used in decreasing ihe chloride concentration to <350 ppm; ihe higher strength acids have process advantages. Excess nitric acid is recovered and recycled during produciion of a concentrated metal-salt solution, which is diluted io Purex solvent extraction feed acidity, 2- 3 M HNO/sub 3/. Titanium is a satisfactory material of construction, wiih corrosion rates <l mil/mo in all process environments and over-all heat transfer coefficients comparable to those of stainless steel. (auth)

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Two methods for measuring target efficiencies are briefly discussed. The second method puts an upper bound on the efficiency and permits observation of instantaneous efficiency, thereby aiding location of losses. Measurements agree well with conventional radiochemical values. (D.C.W.)

Use of Pu/sup 239/ -- U/sup 235/-fueled fast critical assemblies to estimate properties of Pu/sup 239/-fueled assemblies is of interest because of safety considerations and limited plutonium availability. Bare and reflected homogeneous cores and reflected two-region cores are considered. The fuel, 5% by volume, is assumed to be Pu/sup 239/ and U/sup 235/ of various fuel composition ratios for the homogeneous cores. For the tworegion cores the 5% fuel volume is Pu/sup 235/ in the central region and U/sup 235/ in the outer core region. Core diluents, simulating fertile, structural, and coolant materials, are assumed identical in all cases. it is estimated that construction of the reflected two- region core with ratio of central core region volume to total core volume of 0.1 will theoretically decrease the calculated error in prediction of the critical size of a corresponding solely Pu/sup 239/-fueled assembly by a factor of about 10 to 20. (auth)

The summary of this report is that a substance isolated from Chlorella Pyrenoidosa metabolizing {sup 14}CO{sub 2} in the light, previously believed to be a diphosphate ester of a 2-carboxy-4-pentulose, has now been shown to be a disphosphate of 2-keto-L-gulonic acid. The phosphate groups appear to be attached to two of the carbon atoms 3-6. Evidence is presented suggesting that this compound arises from glucose, or a glucose phosphate, which is not in rapid equilibrium with photosynthetically produced glucose derivatives.

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From Purpose: "This Monograph is to assist scientists and engineers in the solution of problems of protection from ionizing radiation, particularly radiation from fallout."

Flow tests which were requested have been completed and the results of these tests are reported. Data which allow the determination of the normal operating flow rates for process tubes equipped with the various combinations of fittings tested is presented. Losses through these fittings, both with and without vaporization allowed, are shown and the relative contribution of the different fittings to the total pressure drop across the outlet assembly is also shown.

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The need for a source of pulses randomly distributed in time according to the Poisson distribution is discussed. A theoretical analysis showed that these pulses may be generated by employing an electrical white noise'' source. Calculations indicated the feasibility of a generator capable of producing Poisson pulse at a maximum average rate of one megacycle. The results of laboratory experiments support the calculations and demonstrate the nadequacy of using periodic pulses to determine instrument response to random pulses. A design of a nuclear signal simulator incorporating the Poisson pulse generator is suggested. (auth)