During this six-month period, Hydrogeochemical and Stream Sediment Reconnaissance samples were collected by the Los Alamos Scientific Laboratory (LASL) from 623 locations in Colorado for a special study. Additional special studies are planned for nine other areas in the Rocky Mountain states in fiscal year 1979, plus some cleanup reconnaissance sampling in parts of four quadrangles that overlap state borders. Preparations for reconnaissance sampling of the entire North Slope region of Alaska have been made but sampling is contingent upon receiving supplemental funds. Water samples from 6725 locations were analyzed for uranium by fluorometry or delayed-neutron counting (DNC). Water samples from 9390 locations were analyzed by emission spectroscopy for 12 other elements. Sediment samples from 14,414 locations were analyzed for uranium and 31 other elements by combined DNC and neutron activation analysis. Sediment samples from 9876 locations were analyzed for two additional elements by x-ray fluorescence. Nine LASL reconnaissance reports, one pilot study data release, one supplemental multielement data release, a multielement data release for three areas in western Alaska, and one quarterly report were open filed by the Department of Energy, Grand Junction Office during the six-month period. Numerous other reports are in various stages of completion and several …

A method for the solidification and fixation of commercial and defense high-level radioactive wastes, namely a cermet process, is currently under development. Chemical and physical processing techniques, as developed up to this time, and the subsequent properties of the resulting cermet bodies are described; unique and advantageous features of the cermet waste form and the conversion process are discussed.

Cermets designed to isolate high-level wastes in a solid form are a composite consisting of various ceramic phase particles uniformly dispersed in and microencapsulated by an iron-nickel base alloy matrix. The metal matrix provides this waste form with many advantageous features including excellent thermal conductivity and mechanical strength. These cermets are formed by first dissolving the waste in molten urea, precipitating and calcining all the constituents, compacting the calcine, and sintering and reduction to form the final product. The exact formulation of cermets through additions to the waste is designed to fix most of the fission products in stable, leach resistant ceramic phases which are subsequently microencapsulated by an alloy matrix. The alloy matrix, which is derived primarily from the waste itself and includes the reducible fission and activation products from the waste, can be compositionally adjusted through additions to optimize its corrosion resistance under conditions existing in various disposal environments. The processes by which cermets are formed include several new and unique materials preparation options that are being developed to permit engineering scale-up and to be compatible with remote operations. Cermets formed by alternate processing methods are being characterized. Initially, cermet samples were prepared using a laboratory scale, batch …

The Navajo have substantial resources of coal and uranium, and water use is certain to accompany development of these resources. A variety of supplies, however, are available--water in storage in Navajo Reservoir, water in existing uses which may be transferred, and groundwater. Furthermore, the quantity of water use varies over a wide range depending on the use of water conservation technologies such as dry coolers and wastewater treatment units. Joint management of energy and water resources requires a basic understanding of the water supply and demand alternatives available to the energy industry. Thus, the uses of water for key energy activities--coal and uranium mining, coal transportation (slurry pipelines), and coal conversion (electricity and synthetic gas production) are reviewed. For those activities for which water conservation is feasible, the technologies and estimate costs ($/af saved) are described. The range of water requirements are then compared to energy and water resource estimates. Finally, alternative (not necessarily exclusive) criteria for energy and water resource management are discussed: a) promote energy activities with the lowest minimum water requirements; b) require industry to use low-quality water resources and the most effective water conservation technology; and c) maximize the economic return on Navajo water resources ($/af …

The status of QCD with respect to recent results from deep-inelastic neutrino scattering is described emphasizing the theoretical uncertainties coming from effects of nonleading order in 1/Q/sup 2/ and in ..cap alpha../sub s/. 31 references.

A fuel cycle that employs /sup 233/U denatured with /sup 238/U and mixed with thorium fertile material is examined with respect to its proliferation-resistance characteristics and its technical and economic feasibility. The rationale for considering the denatured /sup 233/U fuel cycle is presented, and the impact of the denatured fuel on the performance of Light-Water Reactors, Spectral-Shift-Controlled Reactors, Gas-Cooled Reactors, Heavy-Water Reactors, and Fast Breeder Reactors is discussed. The scope of the R, D and D programs to commercialize these reactors and their associated fuel cycles is also summarized and the resource requirements and economics of denatured /sup 233/U cycles are compared to those of the conventional Pu/U cycle. In addition, several nuclear power systems that employ denatured /sup 233/U fuel and are based on the energy center concept are evaluated.

A 500 KeV one-ampere Cs/sup +1/ ion beam has been generated by contact ionization with a 30 cm dia. iridium hot plate. Reproducibility of space charge limited ion current wave forms at repetition rates up to 1 Hz has been verified. The beam is characterized to be very bright and suitable as an ion source for the induction linac based heavy ion fusion scheme. The hot anode plate was found to be reliable and self-cleaning during the operation.

QIKMIX is a quick-turnaround computer code developed to compute the radiative Rosseland mean opacity of specified mixtures at specified temperature and density points. The QIKLIB data base, which QIKMIX uses, has been derived from the OPLIB library. For most mixtures, QIKMIX can compute opacities over a temperature range of 50 to 25,000 eV in less than 1 min of CDC 7600 computer time. The purpose of this report is to discuss the QIKLIB data base and the operation of the QIKMIX code.

The heat and mass transfer processes taking place in molten core debris/concrete systems have been experimentally investigated. Two types of experiments have been conducted. The first experiment simulates the growth of a molten debris pool in a composite sacrificial bed. This experiment models debris pool growth in an inner, low-melting point, sacrificial material zone followed by a melting attack on the concrete bed. The purpose of the inner zone is to quickly melt and dilute the debris pool so that its subsequent downward growth in the concrete may be slowed. In the second experiment a two-layer immiscible liquid system is volumetrically heated and allowed to melt into a low-density gas releasing solid bed which is miscible in the initially-higher-density bottom liquid. The solid melts, mixes with, and dilutes the bottom liquid pool until its density is lower than that of the top liquid.

The heat and mass transfer processes taking place in molten core debris/concrete systems have been experimentally investigated. Two types of experiments have been conducted. The first experiment simulates the growth of a molten debris pool in a composite sacrificial bed. This experiment models debris pool growth in an inner, low-melting point, sacrificial material zone followed by a melting attack on the concrete bed. The purpose of the inner zone is to quickly melt and dilute the debris pool so that its subsequent downward growth in the concrete may be slowed. In the second experiment a two-layer immiscible liquid system is volumetrically heated and allowed to melt into a low-density gas-releasing solid bed which is miscible in the initially-higher-density bottom liquid. The solid melts, mixes with, and dilutes the bottom liquid pool until its density is lower than that of the top liquid. At this time pool inversion occurs and the immiscible liquid sinks to the bottom of the pool displacing the now lighter fuel-concrete simulant.

Extensive system and tradeoff studies were performed to support the selection process for the major parameters and design features of the STARFIRE commercial reactor. With a thermal power of 3800 MW, a neutron wall load of 3.5 MW/m/sup 2/ results in a relatively small-size reactor without imposing excessive requirements on the first-wall cooling capability, maximum toroidal-magnetic field, and frequency of structural material requirements. This moderately high-wall load requires that the first-wall coolant be liquid (water or lithium) and the lifetime of the structural material is > 15 MW-y/m/sup 2/. With moderate plasma elongation and beta the required maximum toroidal-field is approx. 11 T. STARFIRE is operated steady-state with no OH coil. The absence of an OH coil makes it possible to design the reactor with a low-aspect ratio (approx. 2.5) and small major radius. However, higher aspect ratios (approx. 3.5-4) are favored when the plasma current is driven with rf because the power required for the current drive, P/sub rf/, is much larger at lower aspect ratio. Since P/sub rf/ increases at lower plasma temperature, the optimum design for STARFIRE requires operation with plasma temperatures higher than those normally selected for designs with OH-driven current.

This paper presents the results of an attempt to develop an understanding of the various factors involved. This work was performed as a part of the task assigned to one of the expert groups on the International Tokamak Reactor (INTOR). However, the results of this investigation are believed to be generally applicable to the broad class of the next generation of experimental tokamak facilities such as ETF. The shielding penalties for requiring personnel access are quantified. This is followed by a quantitative estimate of the benefits associated with personnel access. The penalties are compared to the benefits and conclusions and recommendations are developed on resolving the issue.

Computing soil moisture content with a neutron probe requires use of a calibration curve that considers the thermal neutron capture cross section of the hole liner, as well as the hole diameter. The influence of steel, polyvinyl chloride, and aluminum casings that fit 0.051- to 0.102-m hole diameters was determined by comparison with neutron probe readings in uncased holes of corresponding diameters. Eccentricity of probe location was considered a potentially significant variable. The experiment was run in disturbed Bandelier tuff with an average dry density of 1.35g . cm/sup -3/ and moisture content of 3.8 to 26.7% by volume. The casing material and hole diameter influenced the probe readings significantly, whereas eccentric location of the probe did not. Regression analyses showed an almost perfect inverse linear correlation between hole diameter and count rate.

A use for GNLS, a general nonlinear least-squares curve fitting program, is presented. In this use GNLS determines a set of best parameters for relationships that cannot be written in closed form. This program can be useful for experimenters who need to determine experimental parameters by fitting nonclosed-form relationships to experimental data. Two examples are given to indicate how GNLS can be used for this type of problem. 11 figures.

This paper presents the radioactive waste production and management at a Liquid Metal Fast Breeder Reactor-II (EBR-II), which is operated for the US Department of Energy by the Argonne National Laboratory at the Idaho National Engineering Laboratory (INEL). Since this facility, in addition to supplying power has been used to demonstrate the breeder, fuel cycling, and recently operations with defective fuel elements, various categories of waste have been handled safely over some 14 years of operation. Liquid wastes are processed such that the resulting effluent can be discharged to an uncontrolled area. Solid wastes up to 10,000 R/hr are packaged and shipped contamination-free to a disposal site or interim storage with exposures to personnel approximately 10 mrem. Gaseous waste discharges are low such as 143 Ci of noble gases in 1978 and do not have a significant effect on the environment even with operations with breached fuel.

First results from the SLAC-LBL Mark II magnetic detector at SPEAR are presented. The performance of the detector is discussed and preliminary results are given on inclusive baryon production R/sub p + anti p/, R/sub ..lambda.. + anti ..lambda../, on decay modes of the D mesons and on two-photon production of eta' mesons.

Some recent results are presented from the Mark II detector at SPEAR: observation of some new D meson decay modes, including the Cabibbo suppressed K/sup -/K/sup +/ and ..pi../sup -/..pi../sup +/ modes, measurements of the rho ..nu.. and ..pi nu.. decays of the tau, and some new data on psi decays. 17 references.

Preliminary results on two-photon interactions from the SLAC-LBL Mark II magnetic detector at SPEAR are presented. The cross section for eta' production by the reaction e/sup +/e/sup -/ ..-->.. e/sup +/e/sup -/ eta' has been measured over the beam energy range from 2 to 4 GeV. The radiative width GAMMA/sub ..gamma gamma../(eta') has been determined to be 5.8 +- 1.1 keV (+- 20% systematic uncertainty). Upper limits on the radiative widths of the f(1270), and A/sub 2/(1310) and f'(1515) mesons have been determined.

Preliminary results are given from the Mark II experiment at SPEAR on radiative decays of the psi' (3684) and on inclusive baryon production from 3.67 to 7.4 GeV center-of-mass energy. A 90% confidence level upper limit of 0.12% is given for BR(psi' ..-->.. ..gamma..eta'/sub c/(3455)) x BR(eta'/sub c/(3455 ..-->.. ..gamma..psi)). 10 references.

The effects of long-term exposure to elevated temperatures on the physical properties of concrete mixes used in Hanford radioactive waste storage tanks were determined. Temperature had a significant effect on the elastic modulus of concretes. Poisson's ratio determined by the sonic method remained relatively constant. The splitting tensile strength increased rapidly up to 190 days of age. Then strength decreased to about 350 days and either leveled off or increased from that point on. Compressive strength data were erratic. (FS)

The reliability of the thermodynamic data for the vaporization of UO/sub 2/ was improved by presenting precise mass spectrometric measurements of the enthalpy of sublimation ever a wide temperature range; calculation of thermodynamic functions for UO/sub 2/ from the critical assessment of available data, and analyzing the departure from molecular flow and the failure of the Knudsen effusion method at vapor pressures of UO/sub 2/ in excess of 10/sup -4/ atm.

Thermodynamic properties of the Th-O system at high temperatures were investigated by means of mass effusion, mass spectrometric, and transpiration techniques over compositions ranging from the congruently vaporizing composition to the lower phase boundary composition. A very sharp decrease in O potential occurs when stoichiometric thoria is reduced but slightly from the stoichiometric composition. The small extent of reduction over a wide range of O potentials measured in the temperature range 2400 to 2655/sup 0/K is a clear illustration of the higher stability of the ThO/sub 2-x/ phase compared with actinide oxides such as UO/sub 2-x/, PuO/sub 2-x/, as well as other metal oxide systems. Estimates of ..delta..H-bar/sub O/sub 2//, ..delta..S-bar/sub O/sub 2//, and the standard free energy of formation of bivariant ThO/sub 2-x/ compositions were made. A substantial increase in the total pressure of Th-bearing species occurs when stoichiometric thoria is reduced toward the lower phase boundary. The increase in pressure can be attributed primarily to the increasing contribution of the ThO gaseous species. Total pressures of Th-bearing species above ThO/sub 2-x/ compositions calculated by means of experimental O potentials and the known or estimated free energy of formation values of the gaseous species and condensed phase are in …

In 1971, the Isotope Research Materials Laboratory (IRML) at the Oak Ridge National Laboratory (ORNL), began a neutron dosimeter development program. The main goal of the program is to produce neutron dosimeter materials for use in determining the neutron energy spectra, flux, fluence and burnup in fast reactor environments. Since 1971, many developments have been made which have reduced the fabrication cost and improved the characterization of the dosimeter materials. The IRML program, which now includes preparation of dosimeters that can be used in light water, controlled thermonuclear as well as fast breeder reactor environments, will be summarized with particular emphasis on materials development and characterization.

A physical model study is ongoing to investigate the sensitivity of recirculation and near field plume dynamics to variation in OTEC plant design and ambient ocean conditions. A thermally-stratified 18 x 12 x 0.6 m basin at a nominal scale of 1/300 allows the upper 180 m of the ocean to be studied for plant sizes up to 600 MW(e). Tests have been conducted for stagnant conditions and for conditions with a current, using both the mixed discharge (combined evaporator and condenser) and non-mixed discharge concepts. Separate tests were made to investigate interactions between evaporator and condenser discharges in a non-mixed concept. Measurements include temperature, dye concentration, mean velocity and visual observations obtained from still and motion pictures. Results for the stagnant water tests showed no significant recirculation except for those tests where the discharge ports were oriented (slightly) upward or where the largest plant size (600 MW(e)) was tested. No significant difference in recirculation could be discerned between the mixed and the non-mixed discharge designs although differences in the equilibrium positions of the discharge plumes were noted. Tests in a current are still in progress but some preliminary results are presented.