Abstract: A fuel material evaluation was made by destructively examining a full-scale experimental fuel element, irradiated in the SRE to a maximum of 850 Mwd/MTU.

Two fused chloride salt eutectics, binary LiCl-KCl and ternary NaCl- KCl- MgCl/sub 2/, were used in fuel processing studies as part of the Liquid Metal Fuel Reactor research and development program. Results of engineering work done at Brookhaven since 1950 are summarized. It was demonstrated that fused chloride salt technology is sufficiently developed so that loops and other experimental equipment can be designed and operated at 500 deg C with a high degree of confidence. The equipment, which was operated for many hours, included a large forced-circulation loop and many thermal-convection loops and tanks. The specifications used for the fabrication, cleaning, and testing of equipment for salt service are described. All welded systems, welded by the usual inert-arc procedures, are preferred, but ring type joint stainless-steel flanged connections were found satisfactory, mainly for connecting melt tanks to experimental equipment and for mounting orifice flowmeters. The surfaces of equipment to be used with fused salts were cleaned satisfactorily prior to assembly by several different methods, but sandblasting was found applicable to all types of equipment. Radiography was used to check all welds in contact with fused salt for flaws and, during operation, to locate and determine the cause of any malfunction. …

Answers by ICPP personnel to questions asked by Eurochemic are given on a variety of subjects, including off-gas sampling systems, off-gas filtration, iodine evolution during fuel dissolution, and process cell contamination. Data are presented on the% of I/sup 131/ retained in waste solutions in the absence and presence of mercury. (D.L.C.)

The Blue Print File groups provided by HAPO for Lurochemic were attached to cover pages, which listed the BPF number and the information contained. These cover pages were combined and an ORNL-CF number assigned. (M.C.G.)

From foreword: This report investigates the economics and performance of alternate cycles for heavy water moderated power reactors.

In efforts to improve the neutron economy and lower the capital costs of sodium graphite reactors, several methods of separating the sodium and graphite were investigated including the calandria, the thimble, and the canned moderator reactors. An analysis including nuclear, heat transfer, and economic comparisons was made of these SGR concepts. Based upon neutron economy and feasibility of core fabrication, the calandria concept appears to offer the greatest potential for improvement in 8GR design. The thimble concept provides some improvement in neutron economy but introduced numerous problems requiring developmental work. (auth)

A practical, prototype silicon p-n junction fast-neutron dosimeter, sensitive in the same range as human tissue, was developed, together with sn associated read-out circuit to facilitate the accurate measurement of accumulated dose. From both theoretical and experimental considerations, it was demonstrated that the dosimeter is essentially insensitive to the gamma and thermal components of a uranium fission spectrum. It was shown that accumulated damage effects appear to be environmentally stable up to an ambient temperature of 100 C. A rather raarked reversible temperature dependence of the read-out parameters requires either control of the read-out temperature or temperature compensation in the read-out device. A high degree of reproducibility of dosimeter characteristics from one device to another was not achieved. The lack of reproducibility was attributed to uncontrolled variables in the bulk silicon from which the devices are fabricated, and in the production procedure. (auth)

An experimental investigation into the gas dynamics of a jet-driven vortex tube for application of a cavity nuclear reactor to rocket propulsion has shown that viscous retardation of the vortex motion is severe, because of a high level of turbulence near the periphery. Based on the experience gained in this study, it is estimated that the achievement of vortex strengths sufficient for practical application will require the use of small diameter tubes with appreciable expenditure of power for recirculation of the gas. The effect of the high degree of turbulence on the separation process near the periphery remains to be determined. The independent variables which were found to influence the vortex strength significantly for a given gas and temperature condition are the tube diameter, the mass flow rate per unit tube length, the injection velocity, and the wall pressure. Estimates of the degree of turbulence in vortex flow have been made from data on the variation in tangential velocity with radius. Virtual (total) viscosities near the periphery ranged from 30 to 700 times the molecular viscosity for tangential Reynolds numbers of from 4 x 10/sup 4/ to 2 x l0/sup 6/. Measurements of the position of the mole- fraction peak …

A controlled radiation experiment was performed to determine the stability of fuel containing 28 wt.% UN dispersed in and clad with Type 318 stainless steel as compared with fuel containing 30 wt.% UO/sub 2/ dispersed in the same material. The specimens were prepared by hot rolling the fuel coupons in Type 318 stainless steel using the picture-frame technique for initial bonding and reduction. Final dimensions were obtained by cold rolling. A special radiation capsule was designed which contained heat control and enough thermocouples to ensure a good continuoustemperature history throughout the test. This capsule and the method by which the specimens were loaded are discussed in some detail. Because of the capsule instrumentatton. the known MTR position into which the capsule was placed. and the dostmeters placed tn the capsule it was possible to obtain a complete flux and temperature history of the capsule during the irradiation. When it was estimated that the specimen burnup was about 7.2 at.% of uranium-235 the capsule was removed from the reactor and returned to the Battelle Hot-Cell Facility. The postirradiation examination which consisted of fission-gas analysis, density and dimensional determinations, radiochemical and isotropic burnup analysis, and metallography is discussed completely in the report. …

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)

An analytical approach is proposed which allows the bunnup (by fission) of uranium required to cause failure in a uranium dioxide-stainicss steel dispersion fuel element to be calculated. The analysis is developed by assuming the matrix of the fuel eicment to be made up of a uniform, close-packed array of spherical UO/sub 2/ particles, each surrounded by and associated with a hollow stainless steel sphere. Equations are then written for the amount of fission gas released into the stainless steel cavity in terms of the UO/sub 2/ particle size and density and the burnup. The release mechanism is by recoil only, since diffusion is unimportant for the particle sizes and temperatures (<1000 tained F) of interest. The gas atoms recoiled from the UO/sub 2/ particle are assumed to diffuse from the stuinless steel shell into the caviiy. The pressure thus exerted in-side the stuinless steel sphere is computed by the application of a real gas law. A suitable failure criterion for an internally pressunized, heavy-walled metal sphere appears to be when the sphere becomes entirely plastic. An equation for the pressure at failure and displacements of the sphere is written in terms of the UO/sub 2/ loading and the yield …

The following report is the final safety analysis report for the Task 2 Radioisotope Powered Thermoelectric Generator prepared by The Martin Company. It presents analyses, tests and evaluation of the operational safety criteria for the generator.

The safety aspects involved in utilizing the Task 2 radioisotope-powered thermoelectric generator in a terrestrial satellite are described. It is based upon a generalized satellite mission having a 600-day orbital lifetime. A description of the basic design of the generator is presented in order to establish the analytical model. This includes the generator design, radiocerium fuel properties, and the fuel core. The transport of the generator to the launch site is examined, including the shipping cask, shipping procedures, and shipping hazards. A description of ground handling and vehicle integration is presented including preparation for fuel transfer, transfer, mating of generators to final stage, mating final stage to booster, and auxiliary support equipment. The flight vehicle is presented to complete the analytical model. Contained in this chapter are descriptions of the booster-sustainer, final stage, propellants, and built-in safety systems. The typical missile range is examined with respect to the launch complex and range safety characteristics. The shielding of the fuel is discussed and includes both dose rates and shield thicknesses required. The bare core, shielded generator, fuel transfer operation and dose rates for accidental conditions are treated. mechanism of re-entry from the successful mission is covered. Radiocerium inventories with respect to …

From summary: An analysis has been performed to determine the ability of the fuel container to withstand the various thermal, mechanical and chemical forces imposed upon the generator by vehicle failures. Where theoretical analysis was impossible, and where experimental evidence was desired, capsules and generators were tested under simulated missile-failure conditions. Thus, the safety limits of SNAP III in a satellite application were defined.

The SNAP-III thermoelectric generator procedures power from the decay heat of 2100 curies of Po/sup 210/. This generator is to be used as a source of auxiliary power in a terrestrial satellite. For purposes of analysis, the satellite system postulated is launched from the Pacific Missile Bange into a 275- statute mile polar orbit with an orbital lifetime of about 1 year. Po/sup 210/ is an alpha emitter having a half life of 138 days and alpha and gamma decay energies of 5.3 and 0.8 mev, respectively. It is a natural component of the earth's crust, as a member of the uranium disintegration series. Sampling of polonium in the biosphere was conducted specifically for this program to determine background radiation levels. Since the fuel is primarily an alpha emitter, there is no direct radiation problem. An analysis was performed to determine the ability of the fuel container to withstand the various thermal, mechanical, and chemical forces imposed upon the generator by vehicle failures. Where theoretical analysis was impossible and experimental evidence was desired, capsules and generators were tested under simulated missile-failure conditions, Thus, the safety limits of SNAP-III in a satellite application were defined. SNAP-III is designed to be aerothermodynamically …

The objective of the study which is being carried out under contract AT(30-1)-2477 is the exploration of all aspects of research in water resources and supply to determine the potential for using radioisotope technology in this research. Problem areas in the application of tracers in this research are being investigated through the evaluation of past experimentation with radioisotopic techniques and through discussions with those who are active in this work. A series of suggestions relating to these techniques will de drawn up to indicate which techniques should be developed further in order that more extensive applications may be found for them.

Angular distribution and kinetic-energy spectra of fragments, and cross sections for fission of gold with 68- to 124-Mev C{sup 12} ions have been obtained by observation of the fragments in two types of detectors, gas scintillation chambers and silicon p-n junctions. From the parameters used to fit the angular distributions to the theoretical curves of Halpern and Strutinski, we have obtained the average excitation energy of the fissioning nucleus at the time of fission. This quantity is approximately 25 Mev, which is nearly independent of bombarding energy, suggesting that fission is preceded by the emission of several particles from the compound nucleus. The fission cross section increases from a value of 100 mb at 68 Mev to 1.28 b. at 124 Mev. Over this range of bombarding energies, the total fragment kinetic-energy release rises from 142 {+-} 6 to 146 {+-} 6 Mev. At all bombarding energies, the variation of laboratory-system kinetic energy of the fragments with laboratory-system angle indicates full momentum transfer by the bombarding particle to the fissioning system.

Ex-reactor high temperature, high pressure recirculating water loop defect-tests have been made on SPR size coextruded fuel tubes. The behavior in terms of fuel corrosion loss and fuel shape distortion for ``pin-hole`` type defects has been determined. Effects of annular spacing on the defect-test behavior and the influence on the mechanical and metallurgical conditions of adjacent components has been observed.

At the present time, high-radioactivity-level wastes at Hanford are neutralized and stored as liquid in underground tanks lined with mild steel. This method of storage is relatively inexpensive and is satisfactory on a short-term basis. However, on a long term basis, liquid storage is less desirable than solid storage because of the greater mobility of the liquid. In addition, storage as aa solid would significantly reduce the volume of waste stored. Consequently, various research and development studies have been undertaken in an attempt to develop a practical waste solidification.

S>Studies of the separation process known as zone melting were enclosed in 5 to 10 mm glass tubes and pulled through a stationary heater, which generated a liquid zone. The separation increased as the zone travel rate decreased, as the size of the tube increased, and as the difference in liquid density between the belk solid and the freezing interface increased. It was also found that, for vertical tubes, the separation was much greater when the fluid of lower density Between the buld zone and the freezing interface) was on the bottom the When it was on the top. Insertion of an axial thbe or rod of metal or glass into the zone also increased the separation. A correlation was developed which enables estimation of the separation for various - sithations in zone melting, Equations and principles were developed which enable estimation of the thermal requirements for zone melting and a theoretical study of pure diffusional mass transfer in some melting was also made. A general expression for concentration profiles was derived for materials with a constant distribution coefficient and a method for the rapid estimation of these concentration profiles was developed, Numerical results for eutectic-forming systems were obtained, and …

Free energy functions for gaseous monoxides were calculated from presently available spectroscopic results. However, the electronic contributions to the free energy functions were estimated. A simple ionic model was assumed since the molecular electronic states for most of these oxides were not known. In some instances where experimental data were insufficient to calculate the interatomic distances and the equilibrium frequencies of vibrations, they were estimated. The results of these calculations were tabulated for 500 tained intervals from room temperature up to 3000 tained K. (auth)

Reactivity and fuel burnup studies were performed for a 255 Mw(e) sodium- graphite reactor of the advanced calandria core type. This reactor is briefly described. Initial criticality calculations and flux distributions were obtained, using two-group theory for enrichments between 2.0 at.% U/sup 325/ and 4.0 at.% U235. A four-group burnup study was performed for enrichments between 2.5 at.% Uisup nd 3.25 at.% U/sup 235/. Core lifetime, changes in isotopic fuel composition, variations in radial power distribution, and fuel cross sec tions are presented. Reactivity during core lifetime was assumed to be controlled by the presence of a homogeneous poison which simulated the effects of control rcds. The results presentad are useful in determining initial enrichment selection in fuel programming and fuel cost studies. (auth)

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An exploratory program on subsurface coatin8s for graphite fuel elements is summarized. A number of coatings with various melting points which could be located beneath the surface of a fueled graphite sphere were investigated. Of the materials with lower melting points. nickel and a special glass compound appeared to form continuous coatings when a hot-pressing technique was employed. Several materials with high melting points. such as Ti, Cr, and MoSi/sub 2/, showed some promise, even though present equipment limitations prevented these specimens from being hot-pressed at the melting point of the coating. (W.L.H.)