[en] At the beginning of FY 1991, the major components of the experiment apparatus had been constructed and transported to Argonne National Laboratory (ANL). The fluids to be used as fuel and coolant simulants had also been tentatively selected. Experimentation began in FY 1991, with eight experiments being completed as of 4/22/91. In each each experiment, a single jet of molten Wood's metal (a Bi-Pb-Sn-Cd alloy, melting point 70 1/2 C) was poured into a pool of room-temperature Freon-11 at atmospheric pressure. The interaction was photographed using high speed motion picture cameras. Eight additional experiments are planned for the remainder of FY 1991. X-ray still or motion picture photography will be used in several. As a result of the work completed during the first seven months of FY 1991, data showing the variation of the jet breakup length with its initial velocity and diameter have been obtained. These data allow an evaluation of the applicability of breakup length correlations in the literature to jet penetration of a volatile liquid with high vapor generation. They also indicate the locations at which an X-ray beam should be placed to best photograph the breakup of the jet within the opaque melt/coolant/vapor mixture. A brief scaling analysis is given. It provides the rationale for the current test matrix. Following this is a description of the status of the experiment apparatus and procedures. Then a brief discussion of completed and planned experiments is given. Finally, the implications of the test results are discussed

[en] The DOE/CECo Nuclear Power Engineering Education Program at the University of Illinois in its first year has significantly impacted the quality of the power education which our students receive. It has contributed to: the recently completed upgrade of the console of our Advanced TRIGA reactor which increases the reactor's utility for training, the procurement of new equipment to upgrade and refurbish several of the undergraduate laboratory set-ups, and the procurement of computational workstations in support of the instructional computing laboratory. In addition, smaller amounts of funds were used for the recruitment and retention of top quality graduate students, the support of faculty to visit other institutions to attract top students into the discipline, and to provide funds for faculty to participate in short courses to improve their skills and background in the power area. These items and activities have helped elevate in the student's perspective the role of nuclear power in the discipline. We feel this is having a favorable impact on student career selection and on ensuring the continued supply of well educated nuclear engineering graduates

[en] The nodal methods computer code utilizing hexagonal geometry, which is being developed as part of this DOE contract, is called THMZ. The computational objective of the code is to calculate the steady-state thermal-hydraulic conditions in a hexagonal geometry reactor core given the appropriate initial conditions and the axial neutron flux profile. The latter is given by a companion nodal neutronics code which was developed in an earlier part of the contact. The joining of these two codes to provide a coupled analysis tool for hexagonal lattice cores is the ultimate objective of the contract and its follow-on work. The remaining part of this report presents the current status of the development and the results which have been obtained to date. These will appear in the MS thesis of Mr. Terrill Laughton in the Department of Nuclear Engineering which is currently in preparation

[en] It is proposed to develop a multigroup, three-dimensional nodal neutron diffusion method for hexagonal geometry reactors. The proposed method differs from all existing hexagonal nodal diffusion methods by its use of net currents rather than partial currents and differs from all but one by its use of a local analytical solution rather than a polynomial approach. These two differences are expected to yield a faster, more accurate method that is easily vectorizable and concurrency capable. The method will treat feedback and depletion induced homogeneities explicitly via in-node variable cross sections. It will also incorporate discontinuity factors and thermal hydraulic feedback via coupling to a newly developed semi-analytical thermal-hydraulic code. The use of a tight coupling algorithm will allow the method to treat transients very efficiently. The research will also include the development of a hexagonal geometry, burnup-corrected, depletion method. The issue of compatibility of the variable cross section methods (and of kinetic feedback) with current formulations of the equivalence theory will be addressed and a resolution of the issue sought

[en] Over half the world's known uranium reserves occur in fluviatile rocks. These deposits include Archean quartz-pebble conglomerates of alluvial fan facies and arkosic braided and meandering fluviatile sandstone facies. Uranium-bearing quartz-pebble conglomerates are described. Approximately 40% of the world's uranium reserves have been found in epigenetic sandstone deposits. Deposits of uranium in braided or meandering fluviatile sandstones can be grouped into peneconcordant and roll-front types. Uranium deposits are widely distributed through central, northern and western Australia but only a very small proportion of the reserves occur in fluviatile sequences

[en] Department of Nuclear Engineering at the University of Illinois, Urbana-Champaign, has a broad-based series of courses and research studies in nuclear thermal hydraulics and related areas. In order to sustain a viable series of courses from introductory undergraduate through advanced graduate levels, a mixture of courses, some taught within the department and some offered by other engineering departments, is used. This conserves faculty teaching time and maintains sufficient class enrollments to provide a regular offering schedule

[en] An experimental study of the fluctuating velocity field, the fluctuating static wall pressure and the in-stream fluctuating static pressure in an annular turbulent air flow system with a radius ratio of 4.314 has been conducted. The study included direct measurements of the mean velocity profile, turbulent velocity field; fluctuating static wall pressure and in-stream fluctuating static pressure from which the statistical values of the turbulent intensity levels, power spectral densities of the turbulent quantities, the cross-correlation between the fluctuating static wall pressure and the fluctuating static pressure in the core region of the flow and the cross-correlation between the fluctuating static wall pressure and the fluctuating velocity field in the core region of the flow were obtained

No abstract available

[en] The primary coolant system in light water reactors is subjected to corrosion on surfaces and to the deposition of corrosion products on surfaces. These conditions occur both in-core and out-of-core and inhibit the thermal transfer across the layers. The deposition of such crud layers on reactor fuel element surfaces is a recognized problem that can affect performance and potentially limit the maximum power rating of the fuel. When boiling is present with such crud layers, wick boiling has been postulated to be the dominant mode of heat transfer, at least when it can be assumed that the structure of the deposit includes chimneys, i.e., large pore holes that penetrate deeply into the deposit and almost perpendicularly to the surface on which the deposit has formed. The wick boiling mechanism is explained. In this study, the wicking limit for porous deposits with chimneys is investigated. The major highlights gained from this study include: (1) the dryout heat flux for a heating surface may be caused by the Helmholtz instability due to high-velocity gas jets or by the wicking limit. (2) The dryout heat flux based on the wicking limit decreases very rapidly with decreasing porosity, with decreasing size of particles in the porous layer, and with increasing crud thickness. (3) Physically, it is expected that the chimney population would increase with increasing heat flux. Using a power law relation between heat flux and chimney population density, much lower wicking limits or thinner deposits are predicted

[en] Traditional Single-Sensor-Single Indicator (SSSI) displays are poorly matched to the cognitive abilities of operators, especially for large and complex systems. It is difficult for operators to monitor very large arrays of displays and controls, and to integrate the information displayed therein. In addition, standard operating procedures (SOPs) are bulky (running to many hundreds of pages) and difficult to use, and operators may become lost. For these reasons, and also because it is becoming increasingly difficult to find replacements for aging hardware components, there is a trend towards computerized graphical interfaces for nuclear power plants (NPPs). There is, however, little rational theory for display design in this domain. This report describes some recent theoretical developments and shows how to develop displays which will greatly reduce the cognitive load on the operator and allow the use of perceptual rather than cognitive mechanisms while using SON and to support state diagnosis and fault management. The report outlines the conceptual framework within which such a new approach could be developed, and provides an example of how the operating procedures for the start-up sequence of a NPP could be realized. A detailed description of a set of displays for a graphical interface for the SON of the feedwater system is provided as an example of how the proposed approach could be realized, and a general account of how it would fit into the overall start-up sequence is given. Examples of open-quotes direct perceptionclose quotes or open-quotes ecologicalclose quotes configural state space displays to support the use of the proposed direct manipulation SOP interface are provided, and also a critical discussion which identifies some difficulties which may be anticipated should the general approach herein advocated be adopted