A series of feasibility studies which werc conducted to determine the capabilities and limitatnnons of several nondestructive testing methods as applnned to the Core B fuel element of the Enrico Fermi Fast Breeder Reactor are discussed. An eddy-current technique is demonstrated to be capable of measuring the fuel plate-clad thickness wtth an accuracy of plus or minus 0.001 in. Discussion is presented of a through-transmission ultrasonic technique which is capable of detecting core-clad interface nonbonds which are nn/6 in. in diameter or larger. A radiographic technnnque is described which can be used to detect UO/ concentration inhomogeneities in excess of the specified plus or minus 5% tolerance. Description is made also of an eddy-current system which will measure coolant channel spacing between fuel plates with an accuracy of 0.0005 in. Details of each of the test systems are presented. (auth)

Four runs were conducted in the Zirflex-Sulfex headend hot cell equipment. Prototype PWR blanket rods, Zircaloy2 clad UO/sub 2/, irradiated from 159 to 356 Mwd/t and decayed 2 years, were declad in boiling 6 M NH/sub 4/F-l M NH/sub 4/NO/sub 3/, terminating with a F/Zr mol ratio of 7 in the spent decladding solution. Average decladding time was 1.5 hr, leaving end cap residues of about 5 g per pin. At the end of the decladding, maximum loss of uranium and plutonium to the decladding solution was 0.04 and 0.37%, respectively. The core pellets were largely shattered with less than 0.5 wt% smaller than 10 mesh. Core dissolution was complete in 5 M HNO/sub 3/ in about 40 minutes, yielding a solvent extraction feed containing 4 M HNO/sub 3/ and 100 g U/l. Solid residue from the decladding and core dissolution was less than 0.001% of the initial weight and consisted of traces of Ca, Fe, Cr, and Sn; uranium and plutonium were not detected. (auth)

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Results are presented of a study to determine if a circumferential rupture in the EGCR hot-leg piping just downstream of the block valve, accompanied by complete lateral displacement of the pipe ends such that the exhaust flows do not oppose each other, will result in excessive pressure differentials across the reactor core. Conclusions and recommendations are included. (J.R.D.)

Abstract: This technical report contains the description and results of an experimental program to evaluate the effect of utilizing Type 3 (SM-2) replacement cores in existing Army field plants SM-1, SM-1A and PM-2A. This program, conducted at the Alco Products Critical Facility, employed SM-2 mockup fuel elements similar in composition to Type 3 fuel elements to determine start-up characteristics of Type 3 cores in SM-1, SM-1A and PM-2A core configurations. measurements include comprehensive power distribution, temperature coefficients, initial critical bank positions, control rod calibrations, critical rod configuration and material coefficients, all obtained under cold, clean, core conditions. The 45 element SM-1 and SM-1A configuration with SM-2 mockup fuel elements contain 36.4 Kg U-235 and an estimated 67.9 gm B-10, while the 37 element PM-2A configuration with SM-2 mockup elements contains 30.0 Kb U-235 and an estimated 56 gm B-10.

This report describes Purex Plant operations at Hanford for January 1962--December 1962. Solvent extraction and ion exchange processes and data are discussed.

This report addresses two inspection systems that have been constructed which use strain gage pressure transducers to pick up changes in static pressure due to change in the hole diameter.

A probe for measurement of the space between fuel plates in high-burnup MTR-type fuel elements by eddycurrent techniques is described/su The sensing section is supported near the end of a flattened stainless steel tube into which a rectangular hole was cut to receive it/su The sensing section consists essentially of two slightly bowed springs, the midsections of which protrude slightly above the upper and lower surfaces of the flattened tube. Each spring- carries a section of a ball bearing as the contact point/su It is these two slightly bowed springs that serve as expanding calipers sensing variations in the spacing between adjacent plates. Essentially the rest of the probe is a long handle to permit probing and a conduit for the electrical leads. A two-wire shielded cable is firmly attached to the unflattened end of the probe and then attached to a modified Wheatstone bridge, from which the output is fed into a chart recorder. (M.C.G.)

Abstract: "This quarterly report, ARF 1194-10, covers the work performed in the time period December 15, 1961 to March 14, 1962, on Contract No. AT(11-1)-578. In this period, Phase II, the detailed evaluation of various fields of radioisotope research, has been extended. Preliminary results on the use of radioisotopes in the industrial applications, apparatus and techniques, and metallurgy categories are presented. These preliminary studies indicate the quantity of research, growth rates, and specialized areas of interest in various countries of the world. More detailed evaluations of these categories will be presented in the final report."

The purpose of this report is to formally document interim power level recommendations for the old reactors consistent with the current tube replacement program. These recommendations were informally presented to Processing Operations by Process Engineering.

The id Thorex process, in which nitric acid is the salting'' agent in the solvent extraction of Th and U from an acid eficient feed with TBP in Amsco, was demonstrated in laboratory equipment for the recovery of synthetic Co solidated Edison Thorium Reactor fuel containing trace fission products. The acid was removed from solution of the declad fuel pellets to produce an acid deficient feed by steam stripping, and the adjusted feed was treated with bisulfite to decrease the extractability of fission products. The Th and U were extracted with 30% TBP in Amsco, and pregnant organic was scrubbed with dilute nitric acid to improve decontamination. Additional acid was added to the extraction section to increase the distribution oefficient of the Th, The Th and U could then be stripped ther simultaneously or separately. No difficulties were ound with either excessive reflux of acid or <0.3%. The co centrated aqueous waste was about 0.2 liter per kilog am of Th processed, about 1/10 of that from the aluminum n trate-salted process. Engineering studies showed that the stage height in the extraction column with the acid Thorex process was 2.1 ft compared with 4 ft with the aluminum-salted Thorex process. …

A review of radiation effects on graphite is presented. Included are discussions of the general relations of these effects with original structural properties, and details of radioinduced changes. Other discussions are devoted to stored energy, annealing, and future problems in the use of nuclear graphite. Data and illustrations concerning structure and radioinduced changes are included. (J.R.D.)

Separate abstracts were prepared for thirty-one of the thirty-three sections. Of the sections not abstracted, the one entitled Fission Product Transport'' contained no information, the other, Transport of Noble Gases in Graphite'' is available in a more complete form as ORNLTM-I35 (NSA 16: 9209) (J.R.D.)

The primary effort to date in connection with this study has been directed toward obtaining source data which indicates (1) the functions performed during reactor outages and the distribution of time required to accomplish these corrective functions, (2) the groups of crafts associated with each of the recovery functions performed, and (3) the radiation exposures experienced during these activities. The first phase of preliminary analysis has been based on the ``time accountability`` report data originated by the various reactor analysts. The attached computer tabulation is one of the analyses performed considering the time and date a reactor was shut down, the ``cause`` for which it went down and the time and date the reactor was considered back on-line. The report summarizes these accountability data into the following summaries in the order presented below: (1) Total hours down per reactor per cause. (April, 1961 to April, 1962) (2) Number of records indicating experience of outages per reactor per cause. (3) The average and standard deviation; same relationship. (4) Outage summary; total hours down, percentage contribution to the department total outage, and time operating efficiency. (5) Department summary (self explanatory). (6) through (21). Interval between like outages by cause. These reports illustrate …

This report presents the technical observations and conclusions to date of the full-reactor demonstration E-N loading for the combined production of plutonium and tritium. A major incentive for such a loading is the increase in conversion ratio resulting from three factors: (1) A ``blacker`` lattice,, in which a combination of ``E`` metal fuel elements (uranium enriched to 0.95% U{sup 235}) and ``N`` metal target elements (enriched lithium-aluminum alloy) approximately matches the reactivity properties of a natural uranium lattice, results in a smaller proportion of parasitic capture of neutrons in the moderator and other reactor components. (2) The capture of neutrons in fringe target material results in the formation of useful product by a significant fraction of the neutrons otherwise lost by leakage. (3) The burnout of product atoms (plutonium and tritium) is reduced in the blacker E-N lattice, thereby increasing the net yield of product atoms per MWD of heat generation compared to the natural loading. Potential advantages for plant-wide E-N production include the routine processing of a single metal stream for Hanford, increased product quality at reduced throughput, and increased production and recovery of higher isotopes in recycled streams.

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"Results of tests indicate that the nucerite-bearium seal system is compatible. This combination was operated as long as 9 hours without serious damage. In other work, the test rig was modified to eliminate pressure fluctuations and extraneous dirt, and Type IIA seal parts were designed."

The detailed stress analysis of the SM-1 steam generator tube sheet showed it to be safe from strain cycling damage. However, the pressure stresses were greater than the yield strength during the hydrostatic test. The differential between pressure stresses and yield strength indicates that some initial deformation may have taken place in the tube sheet. (auth)

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From introduction: "In order to locate systematic trends of states with a certain angular momentum and parity Jπ in nuclei having constant neutron excess, both the ground state and the level energies will be considered simultaneously: we plot the total nuclear binding energy or nuclidic mass excess of excited states, whereby the ground states appear as a special case. Such a natural correlation of level schemes has been employed before Inglis already in 1953."

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A feasibility study was conducted on the application of activation analysis to forensic chemistry. Test portions of automobile paint finish, soot, and soil samples were irradiated in the ORNL Graphite Reactor and the induced gamma radioactivity of each specimen measured with a gamma scintillation spectrometer. Spectral decay curves of the radionuclides in the various specimens classified the materials as to origin. This nondestructive technique is particularly suitable for the analysis of evidence samples that must be maintained in its as-is'' state. (auth)

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At the present, 16- and 23-energy-group calculations are impractical for 2-dimensional-diffusion or transport calculation or for 1-dimensional transport survey calculations. For this reason a set of 4-group cross sections were obtained for cores with spectral characteristics similar to SNAP 2 or SNAP 8. The results of calculations using these cross sections are compared with the results using a 23-group set. The comparison is satisfactory for the values predicted for k eff and for the power distributions. The method for obtaining the cross sections is also presented and because of the ease of obtaining new sets, it is expected that many refinements in the present set and computation of new sets for cases requiring a different spectrum will be made at the discretion of the user. (auth)