This document details activities of the Fuels Preparation Department during the month of March 1959. (FI)

This document details activities of the Fuels Preparation Department during the month of December 1958. (FI)

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This document details activities of the Fuels Preparation Department during the month of January 1959. (FI)

This document details activities of the Fuels Preparation Department during the month of August 1959. (FI)

Two Zircaloy-2 jacketed seven-rod cluster elements will be irradiated in the 3674 KE front-to-rear test hole to an exposure of 1000 MWD/T and two elements will be irradiated in the 3674 KW front-to-rear test hole to an exposure of 2000 MWD/T. After irradiation, the elements will be sent to the ETR where they will be ruptured during reactor operation to determine the failure characteristics of co-extruded Zircaloy-2 jacketed cluster elements.

This document details activities of the Fuels Preparation Department during the month of September 1959. (FI)

Purpose of this study was to determine the {sup 237}Np input to the Purex process. The {sup 237}Np content of 639 g Pu/ton U irradiated fuel was found to be 1.78 {plus_minus} .09 g/ton of uranium at the 95% confidence level. Standard recovery for the chemical method was 96.7%, 98.0% for the sampling.

Significant production gains could be made by improving metal rupture performance. A means of achieving this is by depeaking the axial flux distribution, thus reducing the maximum specific power for a given tube-power. The flux traverse program has provided some knowledge of front-to-rear flux, and was instrumental in prompting a change in the equilibrium HCR pattern during February, 1958. A ``before and after`` comparison of in-pile flux conditions is presented, and suggestions for further improvement are given.

Five eighteen-inch enriched tube-and tube-elements will be irradiated in the KER Loops in pH 10 water to an exposure of 3500 MWD/T, (to see if a pitting corrosion problem exists under NPR conditions).

This document is a HAPO report dated January 23, 1959, and describes the plutonium tail-end anion exchange system, installed in Purex as a prototype unit. Although some modifications, those considered most needed, were made to the unit, additional changes and refinements were still needed to convert the prototype to a fully acceptable manufacturing facility. This document covers the scope design of these modifications. The purpose of this document is to provide scope design criteria for a project to convert the plutonium tail-end Purex Anion Exchange Prototype to a manufacturing facility.

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.

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The casting procedures, mechanical working characteristics, composite fuel plate fabrication, and fuel element assembly methods for a nominal 48 wt.% U- 3 wt.% Si -AP alloy are examined. It was found that localized clad thinning, resulting from dogboning'' of the fuel alloy, is essentially eliminated in composite Plates when this alloy is used in combination with 5154 Al frames and 1100 Al covers. However, fuel component dimensions are difficult to maintain within commonly accepted tolersnces in the brazing prccess because of the marked differences in the thermal expansion of the fuel alloy and the containment materials. The mechanical properties, irradiation studies, corrosfon testing, pneumatic pressure testing, and chemical reprocessing of the Si-modified alloy are also discussed. (auth)

The preliminary process design of a PRFR pilot leaching plant, the proposed location of which is in Cell B of Building 3026 at ORNL, is considered. Chemical, physical, and nuclear parameters are investigated to assure safe leaching operations. Nitric acid solvents are used for leaching the uranium and/ or thorium from the sheared spent fuel elements, and the dissolved fuel is sent through a shielded pipeline to the extraction plant for further processing. Recommended materials of construction are 304L stainless steel and 3O9SCb stainless steel, and maintenance is by direct procedures. (auth)

A theoretical method of determining the time, temperature, and pressure required for pressure bonding was proposed and then evaluated experimentally. The analysis considered only the mechanical deformation of the bonding surfaces induced by the external pressure. It was postulated that such deformations obey the power law for steady creep. The applicaiion of the law to the closure of thick-walled aluminum cylinders under gas pressure was investigated. The behavior was compared with ihat predicted from creep data obtained from short- time torsion creep tests. Only fair agreement was establtshed. The creep law was then applied to the calculation of the closure of voids at the interface between bonding surfaces. It was assumed that the voids exist as isolated spherical or cylindrical holes. The law failed to predict the total closure which occurred. Apparently, the creep law does not hold in the almost microscopic domain of the interfacial voids, although the disagreement may have been due to the oversimplification of the geometric representation. Resolution of the ambiguities probably will be attained only when sufficient experimental data to establish an empirical relationship of closure with bonding conditions are available. (auth) l8092 To develop fatigue-design information for the ORNL reactor programs, a study was …

The detonation of a contained or partially contained nuclear explosion is accompanied by the deposition of a large fraction of the energy in the form of high-temperature, high-pressure regions. The nature of the surrounding medium, the time-temperature history, and the time of cavity collapse or venting determine the extent to which undesirable nuclides such as Sr/sup 90/ and Cs/sup 137/ will appear outside a fused insoluble matrix and be available to ground water or to the atmosphere. The movement of these undesirable radioactivities relative to the ground water movement can be predicted on the basis of measured K/ sub D/'s (distribution coefficients) for the radioactivities in the medium. The induced radioactivities are a 20 to 25% contribution to the fission product radioactivity at times the order of one day, a 1% contribution at about 1 week, decreasing to 0. 1% at about 45 days, increasing to about 2% because of the Co/ sup 60/ for a period of 3 to 15 years. (auth)

The Turco 4501 process is recommended for the decontamination of the PRFR pilot leaching plant equipment. The caustic-tartrate-nitric acid process is recommended for the decontamination of the cell and the equipment exterior. (auth)

The reaction of dispersed Uo/sub 2/ in an aluminum matrix in pressed compacts and fuel plates was studied at 500 to 600 deg C. Correlation of reaction with the accompanying volume increases was rmde and a growth mechanism proposed. Reaction rates and preducts are given as functions of UO/sub 2/ concentration temperature, and time. (W.D.M.)

Presented at the 1958 Winter Meeting of the American Nuclear Society, December 9, 1958. Experimental and theoretical work on the selflimiting response of reactors to step and ramp insertions of reactivity is discussed along with the general characteristics of self-limiting power bursts. The static characteristics of the cores investigated are presented and the techniques of measurement are discussed. Data from step and ramp tests are presented and compared with the predictions of a theoretical formulation of reactor self- shutdown in terms of energy release. The discussion includes an evaluation of some postulated shutdown mechanisms in the light of experimental results. Some results of detailed calculations of shutdown effects due to specific mechanisms which are believed to contribute significantly to reactor self-shutdown are presented. (auth)

Angular distributions in SIGMA /sup +/ decay are analyzed for informatlon on SIGMA spin and parity conservation in SIGMA production. No evidence for a SIGMA spin > 1/2 or parity nonconservation in SIGMA prodiction is found. (auth)

A complete algebraic analysts was obtained for the variation of the steady state ion density n+ with injected current I in an OGRA-tane fusion device (i.e., a device based on trapping of ions by breakup of energetic molecular ions on collision with either the background gas or trapped ions). The most general variation of n+ with I is shown to be an s-curve with at most three rots of n+ for a given input I. A physical interpretation of these three roots is given. In addition algebraic expressions are obtained for the two currents at which the bends in the s-curve occur. It will be necessary to attain the larger current in order to build up a high density plasma when the density ia being increased from below. On the other hand, once the high density has been achieved it may be maintained by steady injection of a current larger than the lower value. Parameters corresponding roughly io the specifications of OGRA are used io obtaam some numerical results. The previously published formulas for burnout in DCX are extended to include effects of neutral backstreaming from the input beam and "ion- pumping." (aath)

A simplified model for solids deposition in the pot calcination of waste was analyzed, and numerical calculations were made. In long calcination pots of 10 to 12 in. diameter, calcination times should not exceed 24 hours and might be as low as three hours if the pot is kept full. If the pots are fed at a constant rate, the cake might form with a steady state V'' when viewed in vertical section which would progress from bottom to top. Cake deposition rates appear to be independent of pot radius. Several advantages to using larger diameter pots are discussed. (auth)

The existence of one or more long-lived nonmetastable excited states of argon in the 15 to 15.5 ev range has been indicated. The increased ionization observed when certain impurities are added to argon is attributed to this long lived excited staate rather than to molecular argon or subexcitation electrons. A competitive process between optical decay and ionizing collisions was evidenced by spectral analysis of the gas mixtures. From these experiments it was not possible to determine the exact energy levels of the suggested long-lived excited states. (auth)