Quarterly report discussing fuel cell research and development work at Argonne National Laboratory (ANL). This report describes activities directed toward understanding and improvement of molten-carbonate-electrolyte fuel cells operating at temperatures near 923 Kelvin.

Quarterly report discussing fuel cell research and development work at Argonne National Laboratory (ANL). This report describes the development of electrolyte structures which have good electrolyte retention and mechanical properties as well as long term stability, and on developing methods of synthesis amenable to mass production.

Quarterly report discussing fuel cell research and development work at Argonne National Laboratory (ANL).

The fuel utilization characteristics for LWR, SSCR, CANDU and LMFBR reactor concepts are quantified for various fuel cycle options, including once-through cycles, thorium cycles, and denatured cycles. The implications of various alternative reactor deployment strategies on the long-term nuclear power growth potential are then quantified in terms of the maximum nuclear capacity that can be achieved and the growth pattern over time, subject to the constraint of a fixed uranium-resource base. The overall objective of this study is to shed light on any large differences in the long-term potential that exist between various alternative reactor/fuel cycle deployment strategies.

This report compares calculated reaction rates based on neutron-transport calculations in RZ and XY geometries with measured values from a low-power dosimetry test in EBR-II. Axial distributions of Uranium-235 and uranium-238 fission rates and uranium-238 capture rate are given for various radial locations along the length of the core and the axial reflectors, and along the length of the radial steel reflectors. Reaction rates, primarily at the reactor midplane, are given for a number of fission and capture reactions. The analytical RZ- and XY-geometry models used for the neutronics calculations are described.

Particulate agglomeration and separation at high temperatures and pressures are examined, with particular emphasis on the unique features of the direct-cycle application of fluidized-bed combustion. The basic long-range mechanisms of aerosol separation are examined, and the effects of high temperature and high pressure on usable collection techniques are assessed. Primary emphasis is placed on those avenues that are not currently attracting widespread research. The high-temperature, particulate-collection problem is surveyed, together with the peculiar requirements associated with operation of turbines with particulate-bearing gas streams.

This report describes the pulsed-neutron-activation (PNA) flow-measuring technique as applied to in situ fluid-flow measurement at EBR-II. Analytic relationships are derived for modeling the process and estimating the uncertainty in measurement. Results from measurements of both water flow and secondary-sodium flow are presented. Results from PNA measurements of water side of the EBR-II steam system have led better definition of plant parameters. Results from sodium-flow measurements are used to provide a correlation for in situ calibration of the electromagnetic sodium flowmeter in the secondary system.

A nondestructive autoradiographic technique is described which can provide a verification of the piece count and the plutonium content of plutonium-containing fuel elements. This technique uses the spontaneously emitted gamma rays from plutonium to form images of fuel elements on photographic film. Autoradiography has the advantage of providing an inventory verification without the opening of containers or the handling of fuel elements. Missing fuel elements, substitution of nonradioactive material, and substitution of elements of different size are detectable. Results are presented for fuel elements in various storage configurations and for fuel elements contained in a fast critical assembly.

Biaxial creep-rupture tests were conducted on 12 prototypic GCFR fuel-cladding specimens at 650 deg C and a nominal hoop stress of 241.3 MPa. All test specimens were fabricated from 20% cold-worked Type 316 stainless steel tubes that were ribbed on the outer surface by mechanical grinding or electro-chemical etching. Test variables included specimen length and the presence or absence of weld-reinforcing end collars.

This report describes the bilinear stress-strain analysis under cyclic loading conditions for the alloy Incoloy 800. Although the method for determining the bilinear stress-strain parameters is based on a procedure proposed in the RDT Standard F9-1 for inelastic analysis of Fast Flux Test Facility (FFTF) components, the accuracy and consistency of results have been improved by an analytical technique, which also resulted in certain simplifications. The bilinear stress-strain parameters of solution-annealed Incoloy 800 (Heat HH7058A) under cyclic loading conditions at a strain rate of 8.6 x 10⁻⁵ s⁻¹, total strain range of 0.2 to 0.8 percent, and temperatures of room temperature to 593 degrees C (1100 degrees F) have been determined. The dependence of bilinear parameters on temperature and strain is discussed. The cyclic-hardening characteristics based on correlation of yield parameter k with accumulated plastic strain are also presented.

This report presents an extensive literature survey on bubble dynamics. Growth of a single spherical bubble moving in a uniformly superheated liquid is considered. Equations of motion and energy are presented in the forms that take into consideration the interaction between the motion and the growth. The fourth-order Runge-Kutta method is used to obtain a simultaneous solution of equations of motion and growth rate, and the solution is compared with available experimental results. Results for liquid sodium are presented for a range of pressures and Jakob numbers.

Report on fuel-cycle studies including pyrochemical separation of plutonium and americium oxides from contaminated materials of construction such as steel.

The Sodium Technology program currently comprises three parts. The first part is aimed at developing a model for accurately describing the behavior of tritium in LMFBRs from its formation in the core to its ultimate retention in the cold traps or release to the environment. Two important parts of this model are the behavior of the sodium cold traps and permeation of tritium through the steam-generator heat-transfer surfaces. A tritium monitor has been developed and installed on EBR-II to measure tritium specific activities and to test the model of an operating LMFBR. The second part of the program is focused in two areas: 91) on-reactor-site conversion of commercial-grade sodium and (2) requalifying sodium from decommissioned reactors for reuse in future LMFBRs.

Cladding failure is one of the major considerations in analysis of fuel-pin behavior during hypothetical accident transients since time, location, and nature of failure govern the early post-failure material motion and reactivity feedback. Out-of-pile thermal transient tests of both irradiated and unirradiated fast-reactor cladding show that local plastic instability, or bulging, often precedes rupture and that the extent of local instability limits the initial rip length. To investigate the details of bulge formation and growth, a perturbation analysis of the equations governing large deformation of a cylindrical shell has been developed, resulting in a set of linear differential equations for the bulge geometry. These equations have been solved along with appropriate constitutive equations and various constraints on the ends of the cladding. Sources for bulge formation that have been considered include initial geometric imperfections and thermal perturbations due to either eccentric fuel pellets or non-symmetric cooling. Of these, only the first is relevant to out-of-pile burst tests. Here it has been found that the most likely imperfection that will grow unstably to failure leads to a bulge around half the circumference with an axial length 1.1 times the deformed diameter. This is in general agreement with burst-test results. For the …

A development program is being carried out to obtain information applicable to the SYNTHOIL process for converting coal to liquid fuel of low sulfur content. This report presents information on (1) the design of a calorimeter to measure heat of reaction of hydrogen with coal slurries, (2) the design of apparatus and calculations for measuring coefficients of heat transfer from SYNTHOIL process feed and effluent products to process vessel walls, (3) tests on the use of additives to facilitate the removal of solids from oil produced in coal liquefaction processes, and (4) the design and construction of a test unit for evaluating new catalysts for coal liquefaction processes.

A development program is being carried out to obtain information applicable to the SYNTHOIL process for converting coal to liquid fuel of low sulfur content. This report presents information on (1) the design of a calorimeter to measure heat of reaction of hydrogen with coal slurries, (2) the design of apparatus and calculations for measuring coefficients of heat transfer from SYNTHOIL process feed and effluent products to process vessel walls, (3) tests on the use of additives to facilitate the removal of solids from oil produced in coal liquefaction processes, and (4) the design and construction of a test unit for evaluating new catalysts for coal liquefaction processes.

This is a sequel to the paper ''A comparison of two ODE codes: GEAR and EPISODE,'' and is concerned with the testing of two superficially similar ODE packages, GEAR and EPISODE. Fourteen basic test problems, some with several cases, are the basis for the testing. These problems represent several types-nonlinear systems with real and complex eigenvalues, linear systems with varied diagonal dominance, linear scalar problems, stiff and non-stiff problems, chemical kinetics with and without diurnal effect, and systems arising from the use of the numerical method of lines. Some problems are included in order to examine the options and error returns. The test results are presented in two forms: raw output and a comparative display of operation counts and of timings for the best method in the GEAR package and the best method in the EPISODE package. This approach allows a comparison of the consequences of the fixed-step interpolate strategy (GEAR) for changing step size against the truly variable step size strategy (EPISODE). It is concluded that EPISODE is generally faster than GEAR for problems involving wave fronts or transients on the interior of the interval of integration. For linear or simply decaying problems, these roles are usually reversed.

As part of the studies preliminary to the acquisition of additional computing capability at Argonne National Laboratory, six groups of jobs were run on the IBM 370/195 at the Applied Mathematics Division of Argonne National Laboratory, on an Amdahl 470V/6 at the Amdahl manufacturing facilities in Sunnyvale, California, and on an IBM 370/168 MODEL 3 at the IBM Field Support Center in Gaithersburg, Maryland. This report compares the performance of the IBM 370/168 MOD 3 with that of the other two machines. Differences in machine configurations were minimized. The memory size of each machine was identical, the I/O configurations were as similar as possible, and the same versions of OS/MVT 21.7 and ASP 3.1 were used on all three machines. This allowed the comparison to be based on the relative performance of the three CPUs.

The Liquid Metal Fast Breeder Reactor (LMFBR) core structure consists of a matrix of hexagonal subassembly ducts. Evaluation of the safety aspects of the core structure requires that reliable computational procedures be available to predict the deformation response of the subassembly configuration to postulated local energy releases. Finite-element computer codes have been developed to calculate deflections and strains of a hexcan subassembly wrapper subjected to internal and external dynamic pressure loadings over a wide range of material-property conditions. An experimental and analytical program has been undertaken to validate and extend the codes for describing the core structural mechanics under reactor operating conditions, including, in particular, descriptions of possible subassembly-to-subassembly damage propagation. This report describes results of the first phase of the experimental program in which single hexcan sections were internally and externally hydrostatically pressurized out-of-pile at room temperature. The experimental data are compared with calculations from a two-dimensional finite-element structural-dynamics code, STRAW. Some additional comparisons were also made with calculations from a three-dimensional code, SADCAT. The correlations obtained between the computations and the hydrostatic experimental results were sufficiently good to validate the STRAW code and proceed to the next phase of the program involving the dynamic structural response.

A simple representation for the crack-face displacement is employed to compute a weight function solely from stress intensity factors for a reference loading configuration. Crack face displacements given by the representation are shown to be in good agreement with analytical results for cracked tensile strips, and stress intensity factors computed from the weight function agree well with those for edge cracks in half planes, radial cracks from circular holes, and radially cracked rings. The technique involves only simple quadrature and its efficacy is demonstrated by the example computations. The weight function for a corner crack in an LMFBR hexagonal sub-assembly duct is constructed from stress-intensity-factor results for the uniformly over-pressurized case, and it is shown how this may be used to determine the stress intensity factors.

This conference was organized to identify, encourage, and promote greater understanding through basic research of the problems encountered in open-cycle, coal-fired MHD generators. The development of this system presents many challenging areas of research in materials sciences, thermodynamics, kinetics, solid state and ion-molecule chemistry and physics, all focused on phenomena occurring at high temperature. The scope of the conference has been designed to improve interdisciplinary communication by involving (1) persons in MHD science and engineering; (2) persons in industry interested in materials research and development; and (3) persons in universities and national laboratories engaged in related basic research. The presentations in the introductory session describe the nature of the MHD system and identify the near-term problems. Sessions following in Gas-Plasma Chemistry; Electronic, Ionic and Molecular processes; Materials; Slag/Seed Properties and Slag/Seed Interactions.

Present concepts of disposal of low-level aqueous wastes (LLAW) that contain much of the fission-product tritium from light water reactors involve dispersal to the atmosphere or to surface streams at fuel reprocessing plants. These concepts have been challenged in recent years. Deep-well injection of low-level aqueous wastes, an alternative to biospheric dispersal, is the subject of this presentation. Many factors must be considered in assessing its feasibility, including technology, costs, environmental impact, legal and regulatory constraints, and siting. Examination of these factors indicates that the technology of deep-well injection, extensively developed for other industrial wastes, would require little innovation before application to low-level aqueous wastes. Costs would be low, of the order of magnitude of 10⁻⁴ mill/kWh. The environmental impact of normal deep-well disposal would be small, compared with dispersal to the atmosphere or to surface streams; abnormal operation would not be expected to produce catastrophic results. Geologically suitable sites are abundant in the U.S., but a well would best be co-located with the fuel-reprocessing plant where the LLAW is produced. Legal and regulatory constraints now being developed will be the most important determinants of the feasibility of applying the method.

Portions of the cesium-uranium-oxygen system have been investigated between 873 and 1273°K and a phase diagram has been constructed using our data and the data of other workers in the field. Thermodynamic and kinetic data have been used to examine the reactions that occur in fast-reactor fuel pins between fission-product cesium and the uranium oxide blanket. It was concluded that at the low oxygen potentials existing at the interface between the uranium-plutonium mixed-oxide and the uranium oxide blanket, Cs₂UO₄ is the only Cs-U-O compound expected to be formed in the uranium oxide blanket.

This report presents results from an investigation on the creep-fatigue and cyclic-deformation of Type 16-8-2 (16% Cr-8% Ni-2% Mo) stainless steel weld metal. Tests were conducted in air at 593 degrees C and a strain rate of 0.004 s⁻¹. Comparisons with data for Type 316 stainless steel base metal indicated that the weld metal has significantly longer fatigue lives for several tension hold-time tests. This is attributed to the fine duplex microstructure of the weld metal that inhibits the growth rate of cracks. Additional studies on the cyclic deformation behavior of the weld metal indicate that the material is strain-history dependent; therefore a unique stress-strain curve does not exist. Monotonic tension tests after cyclic straining result in a different stress-strain curve than obtained from companion fatigue tests at various completely reversed constant strain ranges. A comparison of the fracture morphology and creep-rupture specimens indicates that differences resulting from these tests can be attributed to different failure mechanisms.