The magnitude of the energy gap in nuclear matter associated with a highly correlated ground state of the type believed to be important in the theory of superconductivity has been evaluated theoretically. The integral equation of Cooper, Mills, and Sessler is linearized and transformed into a form suitable for numerical solution. The energy gap, calculated by using an appropriate single-particle potential and the Gammel-Thaler two-body potential, is found to be a very strong function of the density of nuclear matter, and of the effective mass at the Fermi surface. It is concluded that the magnitude of the energy gap for nuclear matter should not be compared directly with experimental values for finite nuclei, although the results suggest that if the theory is extended to apply to finite nuclei it probably would be in agreement with experiment.

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

A possible phase transition in liquid He{sup 3} has been investigated theoretically by generalizing the Bardeen, Cooper, and Schrieffer equations for the transition temperature in the manner suggested by Cooper, Mills, and Sessler. The equations are transformed into a form suitable for numerical solution and an expression is given for the transition temperature at which liquid He{sup 3} will change to highly correlated phase. Following a suggestion of Hottelson, it is shown that the phase transition is a consequence of the interaction of particles in relative D-states. The predicted value of the transition temperature depends on the assumed form of the effective single-particle potential and the interaction between He{sup 3} atoms. The most important aspects of the single-particle potential are related to the thermodynamic properties of the liquid just above the transition temperature. Two choices of the two-particle interaction, oonstituent with experiments, yield a second-order transition at a temperature between approximately 0.01 K and 0.1 K. The highly correlated phase should exhibit enhanced fluidity.

Tables present excess reactivity vs pile rising periods, excess reactivity vs pile falling periods, mint-nominal reactivity vs charge length, Pb-Cd nominal reactivity vs charge length, and nominal reactivity of poison splines.

This report details the proposed E-N, tritium-plutonium producing reactor loading is made up of striped columns of lithium-aluminum (mint) target slugs and enriched uranium (.947 w/o U{sup 235}) slugs. Both target and uranium slugs are of the I & E geometry. The ratio of mint (N) to uranium (E) is determined by the requirement of sufficient k-excess to sustain an operable reactor. The designer of the lattice loading has a choice between natural lithium (7-5 w/o Li{sup 6}) or lithium enriched to {approximately} 38.5 w/o Li{sup 6} for the mint slugs, assuming enriched or ``black`` mint is available. It is possible to show at least 0.8% increase in total conversion ratio for an E-N load for enriched mint (38.5 w/o Li{sup 6}) over natural mint. The basis of the calculations rests on measured E-N length ratios for equivalent neutron multiplicative properties for both types of mint (in a dry pile) plus analytical calculations. It is shown that both increased Pu and H{sup 3} production are obtained by using blacker mint. The basic phenomena are (1) increased resonance capture in U{sup 238} due to more uranium volume in the black mint E-N lattice and (2) more efficient utilization of neutrons in …

Boron-impregnated polyethylene tape was irradiated in the Engineering Test Reactor Critical Facility to study the effect of boron as a burnable poison in reactor fuel. Isotopic analysis of the boron was performed with a conventional CEC Model 21-103 mass spectrometer. The tape was distilled off and the residual boron was converted to trimethyl borate. The reaction mixture was analyzed without separation. Good precision was obtained with samples containing less than 0.5 mg. boron. Features of the mass spectrum of trimethyl borate are discussed. Other applications of the method are suggested. (auth)

From Tamalpais debris crushed to <53 mu , 5.5, 19.6, and 12.5 f the alpha, beta, and gamma activities, respectively, were leached in 72 hr at room temperature by a 100 to 1 weight excess of ground water from the Nevada test site. The extracted alpha-activity material was> 97% Pu/sup 239/ and < 3% Am/sup 241/, and the gamma was 92% Ru/sup106/, 4.7% ZrNb/sup 95/, and 3.1% Cs/137/. The beta activities could not be identified. The ground water leached 10 to 20 times as high a percentage of activity from Tamalpais debris as from Rainier debris in a previous study, but on a count rate basis the total activity released to the ground water was a factor of 2 greater for Rainier than for Tamalpais. Increasing the leaching temperature from room temperature to boiling doubled the amount of active material extracted. increasing the particle size clsssification from <53 to 5901190 mu decreased the extraction efficiently approximately 3- fold. (auth)

During the report period, the nature of the ORNL supporting activities gradually changed, reflecting the progress in the conetruction of the N.S. Savannah. Design reviews continued to require considerable attention. The Inspection Engineering Dept. of the Jab. continued to act as an inspection agency for the ALC, witnessing inspections and tests during fabrication of components of the nuclear reactor system. An enviromnental analysis was made of safeguard aspects of operation of the N.S. Savannah at the NYSC site in Cannden. A study of the safety response of the reactor on the ORNL Analog Computer further defined the important role of the Doppler coefficient in controlling reactivity excursions. Health physics aspects of the operation were studied. Limited waste disposal studies indicated that sea disposal of exhausted demineralizer resins may be facilitated by casting the radioactive resins into concrete. Installation of a pressurized-water in-pile test loop in the ORR neared completion. The neutron flux distribution in the loop was determined by experimental measurements in a nuclear mockup of the in-pile section. Metallurgical activities included nil-ductility testing of steel from the reactor vessel, chemical analyses of primary system components, and investigations of the properties of electroless - nickel brazed joints. Limited fabrication studies …

The formation of heavy negative ions by the attachment of low-energy electrons to oxygen molecules was studied for small amounts of oxygen mixed with methane or carbon dioxide. The rate of attachment in both cases was found to depend on the electron energy, the pressure of the oxygen and the non-attaching gas, and on the kind of non-attaching gas. In general, the attachment increases as electron enprgy decreases or as either oxygen or total pressure increases. The value of the attachment coefficient in oxygencarbon dioxide mixtures is about 100 times its value in oxygen-methane mixtures. This large difference is probably due in part to differences in electron energy and partly to differences in the stabilizing qualities of the two molecules. Dissociative attachment, which should be pressure independent, does not occur at the low energies that were used in this work. Both methane and carbon dioxide are to differences in the stabilizing qualities of the two molecules. Dissociative attachment, which should be pressure independent, does not occur at the low energies that were used in this work. Both methane and carbon dioxide are sometimes used as filling gases for Geiger and proportional counters. The high sensitivity of carbon dioxide to oxygen …

This document details activities of the irradiation processing department during the month of December, 1959. A general summary is included at the start of the report, after which the report is divided into the following sections: Research and Engineering Operations; Production and Reactor Operations; Facilities Engineering Operation; Employee Relations Operation; and Financial Operation.

This study proposes improvements in the process water piping adjacent to the front and rear faces of these reactors. This report covers the external piping of the reactors from the incoming valve pit to the inlets of the outgoing retention basins.

The objectives of this production test detailed in this report is to evaluate the behavior during irradiation of tubular fuel elements with hot-headed end closures. With natural uranium twenty-inch tube- and-tube elements will be irradiated to a goal of 2500 MWD/T in the KER loops. The inner tubes will be closed by the hot-heading technique and the outer tubes will have normal welded closures.

Current information relative to the pyrophoricity of uranium is reviewed and aspects of this troublesome phenomenon which can support further investigation are considered. Information on metallic uranium fuel element fires, up to the present, provide an inadequate base for extrapolation to future operations because of existing uncertainty with respect to causative mechanisms and fire control techniques. Experience at Hanford reactors indicates conditions which tend to reduce the fuel element fire hazard. These include minimizing of the in-core residence time of a fuel element with a cladding defect which permits uranium corrosion, employment of removal and discharge techniques which do not contribute to the deterioration of the fuel element, and minimizing of the time interval between discharge and entry into the retention basin. (J.R.D.)

On the 0000-0800 shift, October 15, 1959, the delayed neutron monitor on KER Loop 2 gave a high coolant activity signal indicating a possible fuel element failure in this loop. KE reactor was shutdown immediately thereafter. This report is being written to summarize the events pertinent to this KE reactor scram and to discuss the results and significance of data from analyses on coolant and coupon samples taken from KER Loop 2.

Production of Pu nitrate from separations plants exceeded commitments. Two Np recovery runs were made, a routine one at Purex and a demonstration run at Redox. Total Np recovered in CY 1959 was 5.7 kg. The UO{sub 3} plant production and shipments also met commitments. Production of unfabricated Pu metal continued. Process studies are being conducted in Purex to resolve the problems in the Np recovery effort. A new solvent (hexone) condensate decanter was operated in Redox. Reprocessing of nonproduction reactor fuels is being reviewed. (DLC)

The objective of this document is to authorize large-scale irradiations of alloyed, low hydrogen, dingot uranium fuel elements in order to evaluate their suitability as a reactor fuel. Although purer than standard ingot uranium, the dingot uranium produced in the past by Mallinckrodt Chemical Plant has demonstrated inferior dimensional stability during irradiation. This instability has been manifest as warp and as swelling at the center of the fuel elements. To reduce this instability 150 ppM iron and 100 ppM silicon haze been added to refine the grain structure of the uranium. Destructive examination of grain structure and nondestructive examination of apparent core quality have indicated that this material is equivalent or superior to ingot material. In light of forecasted metal requirements, the Mallinckrodt Chemical Plant must be brought on stream as rapidly as possible. This PITA schedule will meet the programmed metal delivery requirements with a minimum of risk potential while accelerating the evaluation of dingot uranium as to its suitability as a reactor fuel. Completed irradiation performance data from small scale testing will not be available before dingot material is charged under this PITA. Therefore, the possibility exists that unforeseen dingot uranium properties will result in poor performance. Low …

This report discusses the use of liquid ammonia as a tracer for the detection of cooling water leaks into the piles. It is both safe and cheap and can be detected by methods adaptable to direct-reading instrumentation on a continuous-flow sample. Moderate capital costs and materials costs of less than $50 per pile test are anticipated.

A method of correlating irradiation effects in dispersion fuels was proposed in which the effects of irradiation conditions are considered independently of material variables. Two simple failure models were devised (failure by creep and by short-term stress yield). Criteria which permit estimates of the relative severity of tests made under different test conditions but on identical specimens were developed. Numerical application of the procedures for 18-8 stainless steel with 25 and 30 wt.% UO/sub 2/ specimens was attempted. No positive verification of the short-term stress model was obtained, but the creep model yielded the approximate failure limits for both specimen compositions. (auth)

Neutron multiplication measurements were made on tamped and untamped cylindrical assemblies. The assemblies consisted of Pu metal sheet moderated with Plexiglas, Experiments were performed to evaluate the effects of inhomogeneity. (auth)

The optical models which may be used to describe the scattering and the possible forms of spin-orbit coupling are discussed. The scattering amplitudes are defined and described, including the effects of an isolated resonance, and used to give expressions for the differential cross section and polarization. The latter results are extended to include the change of polarization induced on scattering a polarized beam. Double scattering and the change in polarization produced by deflection in a magnetic field are described. A detailed discussion of possible tensor spinorbit couplings is given. (W.D.M.)

The solubllity of zirconium dibutyl phosphate in aqueous uranyl nitrate and nitric acid solutions was found to vary from 0.35 to 159 rng of zirconium per liter, increasing with increasing uranium concentration. The solubility in 1.126 M TBP in Amsco 125-82 equilibrated with uranyl nitrate-nitric acid aqueous solutions was found to vary from 0.33 to 3.55 g of zirconium per liter. Particularly in the exbactant solutions, the solubillty of zirconium as zirconium dibutyl phosphate is well above the zirconium content of extensively burned natural uranium fuels under Purex process conditions. The zirconium dibutyl phosphate studied was prepared by direct synthesis in aqueous solution and found to have a variable composition. Precipitation from 2 M HNO/sub 3/ at 40 deg C yielded a product which had the approximate composition Zr (OH)(NO/sub 3/)(DBP)/ sub 2/. (auth)

The major effort of the Livermore Mirror Machine group is directed toward study of plasma stabllity and confinement in mirror geometries. The status and radial density distribution and diffusion of confined plasma are briefly summarized. The ALICE Experiment (Adiabatic Lowenergy Injection and Capture Experiment) is discussed in some detnil. (W.D.M.)

The development of rapid, selective, and sensitive methods for the determination of osmium in liquid samples is reported. The specific application of primary interest was the estimation of microgram or milligram quantities of osmium in homogeneous reactcr fuel solutions which contain uranyl sulfate as the major component and corrosion products of stainless steel as minor components. (W.L.H.)

Corrosion data for types 304L, 309SCb, and 347 stainless steel were obtained in a number of process vessels in the ORNL Thorex pilot plant during the development and the production-development periods of operation occurring between December 1954 and September 1956. Stressed corrosion-test specimens were exposed in the batch dissolver tank, the feed adjustment tank, the BT vapor separator, the A-column feed tuink, and the BTC catch tank Generally, types 304L and 3O9SCb stainless steel exhibited comparable corrosion resistance in all environments examined. Most of the studies were conducted with these two alloys. Severe corrosion damage was encountered in the vapor phase of both the batch dissolver twin, operated at a maximum temperature of 115 deg C, and the feed adjustment tank, operated at a maximum temperature of 155 deg C. Corrosion rates for types 304L and 309SCb stainless steel varied from approximately 30 to 55 mpy in the batch dissolver tank during the development and the production-development periods. Vapor-phase corrosion rates in the feed adjustment tank during the lant haif of the development period ranged from 85 to 100 mpy. Severe corrosion attack was experienced also in the BT vapor separator, which operated at a maximum temperature of 115 deg …