Calculated data and graphs describing the effects of continuous thermal-neutron irradiation of thorium, the usual method of operations of homogeneous reactors, are presented.The buildup and decay of U^233, Pa^233, other heavy isotopes, and fission products are considered on the basis of best available cross-section and fission-yield data. The effects of the heavy isotopes and fission products on neutron economy are discussed.

The Spectroscopy Research Laboratory is concerned with research and development in the fields of nuclear magnetic resonance, microwaves, infrared and optical spectroscopy, spectrochemistry, and x rays. Research is directed toward fruitful methods of isotope analysis; new element and compound analytical methods having application to immediate Laboratory or long-range commission needs; and fundamental research on isotopes, elements, and compounds. The work is reported on a project basis to give a more complete picture of the purpose, activity, and status of each program. More detailed information on reported or inactive projects may be obtained from the previous semiannual report.

The flux through a given region is proportional to the total lengths of the neutron flight paths that intersect that region. The analytical Monte Carlo procedure manufactured neutron flight paths and totaled the lengths of all paths intercepted by the regions illustrated in Figure 1. The procedure was designed to utilize the various symmetries in the lattice. / Consider a portion of the lattice whose planar cross-section is shown in Figure 5. If R is the region in which the flux is to be estimated and F the fuel rod in which the neutron originated, then flight path P results in an intercepted length whose reflection in the plane is L. On the other hand flight path P' intercepts R' with length L'. R' is not the region to be studied, but a translation of the flight path P' to F' would result in the neutron intercepting R. The origin in P was arbitrary. For each neutron originating in P another could, with equal probability, have originated in P' with parallel paths. Hence consulting L' in R' towards the total flux is equivalent to starting a neutron at P'. Thus consideration of all regions symmetric to R with respect to …

The enthalpies and heat capacities of seventeen fluoride mixtures in the liquid state have been determined using Bunsen Ice Calorimeters and copper block calorimeters. The fluoride mixtures were composed of the fluorides of two or more of the following metals: lithium, sodium, potassium, beryllium, zirconium, and uranium. The enthalpies and heat capacities of most of these mixtures were studied in the solid state also. Estimates of the heat of fusion have been made. General empirical equations have been developed which represent the enthalpies and heat capacities of the fluoride mixtures in the liquid and in the solid state.

The countercurrent gas-liquid system BF3(g)—anisole·BF3(l) for the concentration of boron isotopes has been studied. The single-storage separation factor varies from 1.039 at 0°C to 1.029 at 30°C. Rate of exchange is rapid, and, with efficient contacting equipment, complete exchange may be obtained in less than 15 sec. A total separation of 1.525 has been realized in laboratory equipment. The critical-product reflux reaction is quite efficient. Only about 55 moles of BF3 remain in each million moles of effluent solvent under laboratory conditions. The vapor pressure of BF3 over the complex rises sharply as the temperature is increased. At 0°C the pressure is 150 mm Hg, and at 40°C the pressure has risen to 1800 mm Hg. From vapor-pressure measurements, an approximate upper limit of ΔH= -12kcal per mole of complex was calculated for the reaction [equation not transcribed]. Qualitative tests indicate good resistance of anisole to decomposition by BF3 under plant conditions. The uncatalyzed exchange of boron between BF3 and BCl3 was found to be too slow to be exploited in a countercurrent system. The single-stage, equilibrium separation factor for the Nitrox system is a function of acid concentration. At 26°C the factor ranges from 1.064 with 1 M acid …

Consideration has been given to various forms of fissionable material for use in atomic reactors, including the pure metals, their alloys and compounds. Of particular interest is the dioxide of uranium which is refractory and corrosion resistant in some environments.The oxide is useful in both granular and bulk forms. Small grains of uranium oxide can be mixed with other materials to form matrix type elements where they serve either as a convenient or necessary form of fuel or fertile material. For other applications the oxide may be fabricated in bulk form such as pellets, rods, plates, or blocks.There is a need for knowledge of the properties of the properties of this oxide, particularly as it affects fabrication in the various forms required. This knowledge is also required by reactor designers and engineers. There is in addition a challenging field for basic studies of sintering rates, oxidation behavior and other phenomena. Fabrication techniques have been developed to produce uranium oxide in various forms with consideration given to the economy of production. The continued application of basic knowledge of these materials has led to simpler. more practical means of fabrication and has thus widened the scope of their use in atomic reactors.

Observations of the effects of neutron damage to zirconium oxides led to the conclusion that the cubic form of ZrO2 is more stable to such damage than the monoclinic form. It has been reported that zirconium corrodes more rapidly in certain liquids when exposure is made under radiation (neutrons and fission products). It is well known that on heating monoclinic ZrO2 a transformation, monoclinic to tetragonal (very similar to cubic), occurs at about 1500°C. The transformation involves sufficient atomic rearrangement that pieces of ZrO2 normally crack and crumble. It is suggested that the effects of neutrons on monoclinic ZrO2 may be similar so that a protective oxide film on the metal would be destroyed soon after its formation. It might be possible, therefore, that the protective oxide film on zirconium metal which is normally monoclinic might be less resistant to corrosion under radiation damage than a similar film which was cubic.

The hydrogen evolution method for the determination of uranium trifluoride which was developed by Manning, Miller and Rowan has been used for the determination of trivalent uranium in this laboratory for the past three years. The method has been applied to many different sample types supposedly pure UF3, mixtures of UF3 and UF4 and the large variety of mixtures of fluoride salts that have been investigated as possible nuclear fuels. These mixtures contained alkali metal, beryllium and zirconium fluorides. Several modifications have been made that have substantially improved the performance and ease of operation of the method. These improvements include the use of (1) an inexpensive, long-lasting source of pure carbon dioxide, (2) vacuum to assist in purging the system of gases that are insoluble in potassium hydroxide solution, (3) deaerated acid that has an extremely low quantity of non-absorbable gases, (4) slower flow rates of purging gas, (5) a sampling technique to minimize contamination, and (6) more dilute absorber solution to reduce film error. It is the purpose of this report to show the effect of these modifications and the applicability of the method various sample types that contain uranium trifluoride.

This report is presented in order to provide a satisfactory system of symbols and identifications for process-instrumentation equipment and to promote a uniformity of practice that will simplify and expedite instrumentation work. It is intended that the systems presented here should be capable of designating and identifying the multitude of instrumentation items which are used for control and operation of conventional processes, as well as for specialized work peculiar to ORNL. Instrument Society of America standards have been adhered to whenever practical.

To determine the maximum permissible concentration (MPC) of a radionuclide which may be ingested or inhaled with exceeding accepted levels of exposure tot he human body, it is necessary to know the average normal concentration of the natural element in the various tissues of the body. Any element may be ingested or inhaled as the element or as a compound. Although nearly all elements have been observed in living tissue, few quantitative observations had been made before 1950. Preliminary studies showed that individual variations in concentration of all elements were very wide and that geographical variations for certain elements occurred. Collection of tissues from autopsies in 9 cities has been done as of February 1957.

All tissues included in this report were received from Dallas, Texas and were collected, ashed and analyzed according to the methods described in ORNL-CF-57-2-2. The average value of the concentration for each element in a tissue is the average of the concentrations in only those samples of the tissue which contained the element. For example, if, of 25 lives analyzed, only 10 contained lead, the average for lead in liver is the average of those 10 values. For those tissues in which an element does not appear in every sample, the lowest value is reported as less than the limit of sensitivity of the method for that element in that tissue. For the elements, lithium and rubidium, the ash was a composited ash sample for each particular tissue. As an example, in lieu of analyzing 20 individual aortas, only on composited aorta as was analyzed.

All tissues included in this report were received from Miami, Florida and the majority are from Jackson memorial Hospital. They were collected, ashed and analyzed according to the methods described in ORNL-CF-57-2-2. The average value of the concentration for each element in a tissue is the average of the concentrations in only those samples of the tissue which contained the element. For example, if, of 25 lives analyzed, only 10 contained lead, the average for lead in liver is the average of those 10 values. For those tissues in which an element does not appear in every sample, the lowest value is reported as less than the limit of sensitivity of the method for that element in that tissue. For the elements, cesium, lithium and rubidium, the ash was a composited ash sample for each particular tissue. As an example, in lieu of analyzing 20 individual aortas, only on composited aorta as was analyzed.

The circuits described herein and shown functionally in Fig. 1 are to be added to the HRT control circuit to provide control and protection for the revised HRT cooling water system. The circuitry will provide protection against excess pressure in the demineralized cooling water loop and cooling water activity, will initiate action to insure containment of activity in event of an explosion and will provide emergency cooling water from the tower basin when required.

The measured pressure drop across the reactor core, with the sample holder in place, is 6.9 psi, more than twice the estimated value. Better estimates, based on more rigorous mathematical analysis, should be possible for future problems of this type. The 2% density difference which produced the relatively high velocity of approximately 1 fps, in this experiment, will result from a temperature difference of about 8 C. It is concluded that the bulk fluid temperature near the sample holder will be less than 8 C above the average temperature at the same elevation in the core.

The failure of the valve bellows would appear to be due to a combination of stress corrosion and crevice corrosion. Stress corrosion occurred as evidenced by the transgranular branched cracking found in the bellows and in the base which which was joined to the bellows. It seems probable that chlorides were present, which, along with the residual stresses present in the bellows assembly, created the necessary conditions for stress corrosion to occur. Crevice corrosion occurred probably due to heavy deposits of solids at the base of the bellows, which created a condition of oxygen impoverishment. While the crater in the base may have been related to a galvanic effect created by the gold gasket, the contour of the crater would suggest that the cause of the crater was due more to crevice corrosion.

One of the considerations in designing a flat plate fuel element is the resistance of the fuel plates, especially the outer plates in the fuel plate array, to deflection and permanent deformation as a result of pressure differentials. An investigation was recently initiated wit the objective of obtaining preliminary information on the APPR-type fuel element to determine the effect of pressure differentials on the outer plates in the fuel assembly. The APPR-1 fuel element consists of 18 flat composite stainless steel fuel plates, joined to grooved 50 mil thick type 304L stainless steel side plates by brazing with Coast Metals N. P. alloy.

The fuel requirement of a heavy-water moderated, homogenous, power reactor were estimated for a variety of initial loadings, for both bath and continuous methods of fuel removal. This study considered a 12-ft spherical reactor, temperature 250 C, 500 Mw thermal power, 125 Mw electrical power capability, 0.8 load factor, and 4%/year inventory charges for U and D2O. The fuel shipping-and-processing charges were assumed to be $1/gm of fissionable fuel for the "batch" processed reactors, and $0.37/gm for the "continuous" processed reactors, Under these conditions, the minimum fuel costs associated with a 10-year 'batch" operating period were about 1.8 or 3.1 mills/kw-hr, if highly enriched U cost $15/gm or $20/gm, respectively. the analogous costs for the "continuous" processed reactor were about 1.6 and 2.6 mills/kw-hr, respectively.

Blake and his co-workers have shown that uranium and other elements can be extracted from acid solutions by various type of organo-phosphorous compounds. Early investigations in the laboratory have demonstrated the applicability of tri-n-alkyl phosphine oxides to the extraction of metal ions from acidic solutions for analytical purposes. This paper is concerned with a similar qualitative investigation of the extraction of metal ions with a di-alkyl phosphoric acid, di-2-ethylhexyl phosphoric acid (D2EPHA).

Lewis in 1929 determined the thermal conductivity of beryllium metal for the first time. The next measurement reported in the open literature on the thermal conductivity of beryllium was by Powell in 1953. Measurements were also made by Grenell, Linebrink, and Johnson at Battelle Memorial Institute in 1947. Their values are in good agreement with those of Powell.

Small Heat Exchanger ORNL #1, type SHE #2, was removed form test stand B after 2071 hours of operation. Thirty-five samples were removed form the entire heat exchanger. The corrosion found on the outside of the tubes exposed to the fluoride mixture ranged to a maximum depth of .004 inches; however, the frequency of occurrence along the tube wall was heavier at the NaK outlet header, which was the hottest area in the heat exchanger. The depth of attack observed on the fluoride side of this heat exchanger was uniform from header to header and did not exceed .004 inches.

An order of magnitude estimate was made to determine a minimum cost for 10% pure oxygen-17 when produced by a chemical exchange process. the calculations were based on separations factors of 1.03, 1.01, and 1.003. the cost of product was found to vary from $23 per gram for the large factor to $165 per gram for the smaller.

A simplified method was developed for calculating wall-scatter gamma radiation escaping through a shield opening. The method was applied to the HRT and the results showed that next to the line of sight contribution, scattering of the wall of the shield opening was the main contribution to the dose at the rear edge of the shield. Design charts were prepared that give the dose as a function of the gamma source location with the reactor cell.

In the HRP Welding program, a major part of the welding and fabrication, and some of the testing, is performed using an inert atmosphere. the use of the inert gas consumable arc welding process and dry box welding or other work in a dry box comes within this category. Since much of the work in the project make use of, or requires, inert atmospheres, a general discussion follows of the methods used, description of equipment, processes, quality of atmospheres, purity requirements, kinetics of metal-gas reactions, and the proper application of the equipment and methods use din obtaining the desired results.

Thorium nitrate, recovered from irradiated thorium metal processed in the ORNL Thorex Pilot Plant, was converted to thorium oxide and then to the fluoride in one pilot-plant-scale and two laboratory-scale runs. Activity distributions, decontamination factors, and safety of the process are treated. (D.L.C.)