While used extensively, the term intermediate-level waste is not a clearly defined waste category. Assuming the ILW includes all radioactive wastes requiring shielding but not ordinarily included in a high-level waste canister, its major sources include power plant operations, spent fuel storage, and spent fuel reprocessing. While the volume is approx. 10/sup 2/ greater than that of high-level waste, ILW contains only approx. 1% of the radioactivity. Power plant waste, constituting approx. 87% of the waste volume, is generally nontransuranic waste. The other approximately 13% from fuel reprocessing is generally transuranic. Intermediate-level wastes fall into the general categories of highly radioactive hardware, failed equipment, HEPA filters, wet wastes, and noncombustible solids. Within each category, however, the waste characteristics can vary widely, necessitating different treatments. The wet wastes, primarily power plant resins and sludges, contribute the largest volume; fuel hulls and core hardware represent the greatest activity. Numerous treatments for intermediate-level wastes are available and have been used successfully. Packaging and transportation systems are also available. Intermediate-level wastes from power plants are disposed of by shallow-land burial. However, the alpha-bearing wastes are being stored pending eventual disposal to a geologic repository or by other means, e.g., intermediate-depth burial, sea disposal. Problem …

Two wells, GT-2 and EE-1, were originally drilled to a depth of 9600 ft (2.93 km) and 10,000 ft (3.05 km), respectively, and, after some difficulties, including redrilling of the bottom portion of GT-2, a good fracture connection was made between EE-1 and GT-2B, as the modified GT-2 was called. The circulation system was studied extensively for the purpose of establishing a number of fracture properties. Techniques were developed to determine orientation, geometry, heat exchange area, volume, flow impedance and impedance distribution. A much larger fracture system was then created from a depth of 9620 ft (2.93 km) in EE-1. The techniques used and results obtained in the study of the new and old fracture systems are discussed. (MHR)

We study nonintegrable Hamiltonian dynamics: H(I,{theta}}) = H{sub 0}(I)+kH{sub 1}(I,{theta}) for large k; that is, far from integrability. An integral representation is given for the conditional probability P(I,{theta},t|I{sub 0},{theta}{sub 0},t{sub 0}) that the system is at I,{theta} at t, given it was at I{sub 0},{theta}{sub 0} at t{sub 0}. By discretizing time into steps of size {epsilon}, we show how to evatuate physical observables for large k, fixed {epsilon}. An explicit calculation of a diffusion coefficient in a two degree of freedom problem is reported. Passage to {epsilon} = 0, the original Hamiltonian flow, discussed.

We present a method for studying nonintegrable Hamiltonian systems H(I,{theta})=H{sub 0}(I)+kH{sub 1}(I,{theta}) (I,{theta} are action-angle variables) in the reg1me of large k. Our central tool is the conditional probability P(I,{theta},t|I{sub 0},{theta}{sub 0},t{sub 0}) that the system is at I,{theta} at time t given that it resided at I{sub 0},{theta}{sub 0} at t{sub 0}. An integral representation is given for this conditional probability. By discretizing the Hamiltonian equations of motion in small time steps, {epsilon}, we arrive at a phase volume preserving mapping which replaces the actual flow. When the motion on the energy surface E=H(I,{theta}) is bounded we are able to evaluate physical quantities of interest for large k and fixed {epsilon}. We also discuss the representation of P(I,{theta},t|I{sub 0},{theta}{sub 0},t{sub 0}) when an external random forcing is added in order to smooth the singular functions associated with the deterministic flow. Explicit calculations of a "diffusion" coefficient are given for a non-integrable system with two degrees of freedom. The limit {epsilon}{approaches}0 , which returns us to the actual flow, is subtle and is discussed.

The STARFIRE design study focused on solid tritium breeder blankets in order to minimize the stored chemical energy. The most suitable coolant candidates with solid tritium breeders are water and helium. This paper presents the results of a comparative study of the two coolants. The study shows clear advantages for the choice of pressurized water for the conditions of the STARFIRE tokamak power plant design. The study also identifies those areas where development is required in order to utilize the potential advantages of helium.

STARFIRE is a conceptual design for a commercial tokamak power plant based on the deuterium/tritium/lithium fuel cycle. The emphasis of the study is on the simplicity of the engineering design, maintainability, lower electricity cost, and improved safety and environmental features. The reactor has a 7-m major radius and produces 1200 MW of electric power. STARFIRE operates in a steady-state mode with the plasma current driven by a lower hybrid rf system. The plasma purity control and exhaust system is based on the limiter/vacuum concept, which offers unique advantages for commercial power reactors. The blanket utilizes a solid lithium compound for tritium breeding and pressurized water as the coolant.

A detailed design of a limiter/vacuum system for plasma impurity control and exhaust has been developed for the STARFIRE tokamak power plant. It is shown that the limiter/vacuum concept is a very attractive option for power reactors. It is relatively simple and inexpensive and deserves serious experimental verification.

An engineering model of a multiwire position-sensitive proportional-counter (PSPC) was developed, tested, and installed at the US National Small-Angle Neutron Scattering Facility at ORNL. The PSPC is based on the RC-encoding and time-difference decoding method to measure the spatial coordinates of the interaction loci of individual scattered neutrons. The active area of the PSPC is 65 cm x 65 cm, and the active depth is 3.6 cm. The spatial uncertainty in both coordinates is approx. 1.0 cm (fwhm) for thermal neutrons; thus, a matrix of 64 x 64 picture elements is resolved. The count rate capability for randomly detected neutrons is 10/sup 4/ counts per second, with < 3% coincidence loss. The PSPC gas composition is 63% /sup 3/He, 32% Xe, and 5% CO/sub 2/ at an absolute pressure of approx. 3 x 10/sup 5/ Pa (3 atm). The detection efficiency is approx. 90% for the 0.475-nm (4.75-A) neutrons used in the scattering experiments.

Lanthanide orthophosphates formed from elements in the first half of the 4f transition series are analogs of the monoclinic mineral monazite. The known geologic properties of this mineral make the general class of lanthanide orthophosphate compounds attractive substances for long-term containment and disposal of ..cap alpha..-active actinide nuclear wastes. EPR spectroscopy has been used to investigate the structural properties and solid state chemical properties of impurities in these materials and to compare the characteristics of single crystals and polycrystalline bodies.

The shower counter system of the SLAC-LBL Mark II detector is a large lead/liquid argon system of the type pioneered by Willis and Radekal; however, it differs in most details and is much larger than other such detectors currently in operation, It contains, for example, 8000 liters of liquid argon and 3000 channels of low noise electronics, which is about eight times the size of the system of Willis et al. in the CERN ISR. This paper reports, with little reference to design, on the operation and performance of the Mark II system during approximately a year and a half of operation at the Stanford Linear Accelerator Center&#x27;s e{sup +}-e{sup -} facility, SPEAR. The design and construction of the system have previously been described and a detailed discussion of all aspects -- design, construction, operation, and performance -- is in preparation.

We present a study of the decay properties of charmed D mesons produced near the peak of the {psi}&quot; (3770) resonance in e{sup +}e{sup -} annihilation. Branching fractions for nine Cabibbo-favored and three Cabibbo-suppressed decay modes are presented along with upper limits on one additional Cabibbo-favored and four additional Cabibbo-suppressed decay modes. A study of K{pi}{pi} decay mode Dalitz plots reveals a large quasi-two-body pseudoscalar-vector component for the D{sup 0} decays and an apparent nonuniform population an the Dalitz plot for the D{sup +} decay into K{sup -}{pi}{sup +}{pi}{sup +}. Using tagged events, we measure the charged particle multiplicity and strange particle content of D decays. A measurement of the D{sup +} and D{sup 0} semileptonic decay fractions indicates that the D{sup +} has a significantly longer lifetime than the D{sup 0}.

An experiment was performed at the Multiparticle Spectrometer using 17 GeV/c ..pi../sup -/ from the BNL AGS, triggering upon inclusive e/sup +/e/sup -/ production. Electron identification was based on two transition radiator detectors and lead-scintillator shower detectors. Good acceptance for the e/sup +/e/sup -/ pair covered the region x/sub F/ &gt; 0.3 for all p/sub T/ and pair masses. Charged particles and photons associated with the e/sup +/e/sup -/ pair are detected over a large solid angle. e/sup +/e/sup -/ pairs of mass up to 1.2 GeV/c/sup 2/ were produced. A clear peak due to rho, ..omega.. ..--&gt;.. e/sup +/e/sup -/ is observed. For e/sup +/e/sup -/ masses below the rho, ..omega.., an excess of events is found over those expected from known sources such as eta ..--&gt;.. e/sup +/e/sup -/..gamma.. and ..omega.. ..--&gt;.. e/sup +/e/sup -/..pi../sup 0/. This anomalous excess is more strongly produced at small x/sub F/. The structure of events containing anomalous e/sup +/e/sup -/ pairs is reported in an attempt to elucidate their origin. In particular, effective mass distributions of e/sup +/e/sup -/..gamma.., e/sup +/e/sup -/..pi../sup 0/, e/sup +/e/sup -/ charged hadrons are presented.

The oxidation behavior of type 316 stainless steel and candidate advanced alloys for the gas-cooled fast reactor (GCFR) is being investigated at General Atomic Company. The test program consists of oxidation tests in prototypic GCFR environments. Two tests have been completed to date and a third test is under way. The first test was performed in an environment containing a hydrogen/water ratio of 10. The oxidation behavior of all the alloys was good to excellent in this environment. Preferential oxidation of chromium was responsible for this behavior. The second test was performed in an environment containing a hydrogen/water ratio of 0.25, where both chromium and iron oxides are thermodynamically stable. Some of the alloys and some of the ribbed type 316 stainless steel test specimens showed unacceptable oxidation resistance in this environment. In the third test, presently under way, two different pretreatment procedures are being used to control the poor oxidation behavior observed in the second test. Early results show some degree of success.

As digital integrated circuit size and complexity increases, the need for accurate and efficient computer simulation increases. Logic simulators such as SALOGS (SAndia LOGic Simulator), which utilize transition states in addition to the normal stable states, provide more accurate analysis than is possible with traditional logic simulators. Furthermore, the computational complexity of this analysis is far lower than that of circuit simulation such as SPICE. An eight-value logic simulation environment allows the use of accurate delay models that incorporate both element response and transition times. Thus, timing simulation with an accuracy approaching that of circuit simulation can be accomplished with an efficiency comparable to that of logic simulation. 4 figures.

The new standard Fortran 77 has not been available long, but the American National Standards Institute (ANSI) committee X3J3 responsible for Fortran standardization is already working on the next revision. Since the result of this work will be a candidate for an international (ISO) standard, it is important that work being done now become known to all persons interested in Fortran. A new set of problems related to the accommodation of related standards in data base management and real-time process control, as well as the ever-increasing size of the language, have caused the standardization committee to consider some new approaches to the development of the next standard. These new approaches and many of the new features that probably will be in the next Fortran standard are described. It is hoped that this presentation will stimulate comments and suggestions in time to include them before work on the next standard is completed. 2 figures.

Analysis of self-powered neutron detector (SPND) data taken during loss-of-coolant experiment (LOCE) L2-3 conducted in the Loss-of-Fluid Test (LOFT) facility has shown that the detectors are sensitive, in a qualitative sense, to coolant density variations in the reactor core. These results provide a strong impetus for determining the applicability of self-powered nuclear flux detectors for liquid level measurement in the reactor vessel of a pressurized water reactor (PWR) during transient conditions. This summary discusses the response of the LOFT SPNDs to core coolant density variations in the L2-3 large break experiment and shows the potential application to reactor vessel liquid level measurements during small break accidents in PWRs.

The results are presented of three small break loss-of-coolant experiments performed in the LOFT Pressurized Water Reactor (PWR) system. Experiment L3-0, performed without reactor power, represented a loss of coolant from the power operated relief valve on the top of the pressurizer. Experiments L3-1 and L3-2 were initiated with the reactor at full power (maximum linear heat generation rate approximately 52 kW/m) and represented 4-in and 1-in diameter breaks, respectively, in the reactor inlet piping of a commercial PWR. Comparisons of data to analytical model calculations with a number of different models indicate that most major phenomena were correctly calculated, but that improvements in modeling small break behavior are necessary.

Results are reported for experiments on the behavior of sodium oxide aerosols, uranium oxide aerosols, and various mixtures of these in the Nuclear Safety Pilot Plant (NSPP) vessel at Oak Ridge National Laboratory. The NSPP program, sponsored by the Nuclear Regulatory Commission, is to provide experimental qualification of the HAARM-3 aerosol behavioral code being independently developed at Battelle-Columbus Laboratories. The results of fourteen single-component and six mixed-component aerosol experiments are presented along with selected comparisons with code predictions.

In situ recovery methods for many of our hydrocarbon and mineral resources depend on the ability to create or enhance permeability in the resource bed to allow uniform and predictable flow. To meet this need, a new branch of geomechanics devoted to computer prediction of explosive rock breakage and permeability enhancement has developed. The computer is used to solve the nonlinear equations of compressible flow, with the explosive behavior and constitutive properties of the medium providing the initial/boundary conditions and material response. Once the resulting computational tool has been verified and calibrated with appropriate large-scale field tests, it can be used to develop and optimize commercially useful explosive techniques for in situ resource recovery.

The steps in assembling the computational tools needed to simulate the explosive fracture of oil shale have been described. The resulting code, with its input data, was then used to simulate three explosive field experiments. The results of the calculations are in good agreement with what actually occurred in the field. Further detailed comparisons are in progress for these experiments and the others that have been conducted. As this is done, improvements will be made in the input data and in the code physics. The development of computer codes as tools to predict rock breakage makes a variety of interesting studies possible. The properties of the explosive can be changed to see how the extent of rubbling is affected. Studies of spacing and delays for decked charges are also possible. Finally, the codes can be applied in situations, such as confined-volume blasting, at the frontiers of blasting technology. These areas are vital to the effective utilization of our oil shale resources, especially with in situ techniques. Computer simulation will play a central role in the development of new technology for energy and mineral resource recovery.

The results of a controlled study of the effiacy of administration of calcitonin in arresting the progression of osteoporosis in male patients is presented. (ACR)

The progress of the LLL Laser Fusion Program to achieve high gain thermonuclear micro-explosions is discussed. Many experiments have been successfully performed and diagnosed using the large complex, 10-beam, 30 TW Shiva laser system. A 400 kJ design of the 20-beam Nova laser has been completed. The construction of the first phase of this facility has begun. New diagnostic instruments are described which provide one with new and improved resolution, information on laser absorption and scattering, thermal energy flow, suprathermal electrons and their effects, and final fuel conditions. Measurements were made on the absorption and Brillouin scattering for target irradiations at both 1.064 ..mu..m and 532 nm. These measurements confirm the expected increased absorption and reduced scattering at the shorter wavelength. Implosion experiments have been performed which have produced final fuel densities over the range of 10x to 100x liquid DT density.

The limitations of the integral boundary layer methods and the potential of the differential boundary layer method in analyzing MHD channel flows are assessed. The sensitivity of results from the integral method to the parametrization of boundary layer profiles and calculation of wall heat transfer is established. A mixing-length type turbulence model for flow on rough walls is developed and validated by comparison with experimental data. The turbulence model is used in a quasi-three-dimensional boundary layer model to evaluate the influence of wall roughness and pressure gradients on the flow characteristics and performance of MHD channels. The behaviors of skin friction and Stanton number calculated from the analytical model are found to differ considerably from the empirical correlations valid for non-MHD flows without pressure gradients.

Heat transfer in subsonic MHD diffusers, by convection and by gas and slag particle radiation, is analyzed by simultaneously solving the radiation transport equation and the quasi-three-dimensional gasdynamic equations. The efficiency factors for extinction and scattering by particles are calculated from the Mie theory. For a reference diffuser geometry, the heat transfer by convection is found to be 25 MW, and the radiative heat transfer varies from 44 MW to 79 MW, depending on the rate of ash carryover into the channel. Results reveal that the heat transfer is sensitive to the ash carryover into the channel, slag particles spectrum, electrical conductivity of ash, gas composition, and wall emissivity. It is observed that, because of multiple scattering, the particles shield the short wavelength radiation emitted by potassium atoms. The impacts of heat transfer enhancement by gas radiation in the channel and by gas-plus-particles radiation in the diffuser on MHD system design are assessed. It is suggested that, from the system design point of view, the diffuser be regarded as a part of the radiant boiler. No significant effect of radiation enhancement on the ability to decompose NO/sub x/ is anticipated.