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Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 1, Discussion and glass durability data

Description: The Materials Characterization Center (MCC) at Pacific Northwest Laboratory is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. Data collected throughout the world are included in the data base; current emphasis is on waste glasses and their properties. The goal is to provide a data base of properties and compositions and an analysis of dominant property trends as a function of composition. This data base is a resource that nuclear waste producers, disposers, and regulators can use to compare properties of a particular high-level nuclear waste glass product with the properties of other glasses of similar compositions. Researchers may use the data base to guide experimental tests to fill gaps in the available knowledge or to refine empirical models. The data are incorporated into a computerized data base that will allow the data to be extracted based on, for example, glass composition or test duration. 3 figs.
Date: December 1, 1987
Creator: Kindle, C. H. & Kreiter, M. R.
Partner: UNT Libraries Government Documents Department
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Comprehensive data base of high-level nuclear waste glasses: September 1987 status report: Volume 2, Additional appendices

Description: The Materials Characterization Center (MCC) is assembling a comprehensive data base (CDB) of experimental data collected for high-level nuclear waste package components. The status of the CDB is summarized in Volume I of this report. Volume II contains appendices that present data from the data base and an evaluation of glass durability models applied to the data base.
Date: December 1, 1987
Creator: Kindle, C. H. & Kreiter, M. R.
Partner: UNT Libraries Government Documents Department
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Examples of the changes in temperature, pressure and flow following a reactor power surge

Description: While it is obvious that a power surge in a Hanford reactor will result in temperature increases of the fuel and water, it is also true that there will be changes in the coolant flow rate and the pressures in each process tube. In addition, if the power increase is of sufficient magnitude, a self-induced flow reduction will occur as a result of pressure changes due to boiling within the process tube. The relations between power, temperatures, flow and pressure following a power surge are difficult to calculate with precise accuracy. However, a knowledge of these relations is important for incorporation with physics considerations to allow prediction of maximum fuel temperatures following a power surge and to specify detection instrumentation for proper reactor control. Fairly precise information of temperatures, flow, and pressure in a single process tube following a power surge can be obtained using an electrically heated model in the 189-D heat transfer laboratory. Obtaining such information for all possible conditions and for all the Hanford reactors would be a fairly large job. However, two cases were run with an existing full scale electrically heated model of a single column of fuel elements in a C reactor overbore process tube. These demonstrate the general relationship between temperature, pressure and flow following a power surge and provide examples of the type of experiments that are possible.
Date: June 7, 1962
Creator: Kreiter, M. R. & Batch, J. M.
Partner: UNT Libraries Government Documents Department
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Heat transfer experiments simulating a failure of the inlet piping to a BDF reactor process tube

Description: Laboratory heat transfer experiments were conduced to investigate fuel element temperatures which could result from coolant flow loss following a failure of the inlet piping to a process tube at a B, D, F, DR, or H reactor. The results are reported herein. Failure of the inlet coolant piping between the front header and the process tube on a reactor would stop the normal flow of cooling water to the fuel elements. Such a failure should immediately initiate a reactor shutdown, but the only means of removing the heat released during the post-shutdown period would be by reverse flow of hot water from the rear cross header. The subject experiments were conducted to determine what rear header pressure would be required to achieve adequate cooling of a BDF type reactor fuel assembly following such a piping rupture. Experimental studies were previously reported concerning failure of inlet piping to a K reactor geometry. The analytical techniques and experimental procedures used previously were also used in the present experiments.
Date: August 16, 1961
Creator: Waters, E. D. & Kreiter, M. R.
Partner: UNT Libraries Government Documents Department
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Results of Laboratory Heat Transfer Experiments for C-Reactor Overbore Fuel Channels

Description: The purpose of this report is to present experimental data concerning the heat transfer and fluid flow conditions within a C-overbore geometry process channel for the cases of steady state operation, flow plugging incidents, and inlet piping failure incidents.
Date: November 10, 1961
Creator: Waters, E. D. & Kreiter, M. R.
Partner: UNT Libraries Government Documents Department
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Report of the AD HOC Study Group on integrated versus dispersed fuel cycle facilities

Description: To provide isolation of strategic materials and confinement of nuclear wastes, the basic facilities considered in assessing the DFCF and IFCF were mixed plutonium and uranium oxide and HTGR fuel fabrication, fuel reprocessing, high- enrichment isotopic separation and interim waste storage. Reactors, low- enrichment isotopic separation, and low-enrichment uranium facilities were excluded. It is expected that the IFCF would attract uranium fuel fabrication and possibly reactors. An assumption was made for the study that the choice of either IFCF or DFCF would not alter the nuclear power generation pattern postulated to exist up to the year 2000. The advantages of IFCF are seen to outweigh disadvantages. (auth)
Date: April 1, 1975
Creator: Kreiter, M. R. & Platt, A. M.
Partner: UNT Libraries Government Documents Department
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Influence of plutonium recycle on the radioactive waste management system

Description: The effect that variations in the back-end of the fuel cycle have upon radioactive waste characteristics is of increasing interest. An overview treatment of the characteristics for the once-through cycle, recycle of uranium only, and recycle of uranium and plutonium modes of operation is presented. Characteristics which are considered include waste volumes, heat generation, neutron source strength, specific activities, and radioactive airborne releases.
Date: February 1, 1977
Creator: Kreiter, M R; Fleischman, R M & Muckerheide, W A
Partner: UNT Libraries Government Documents Department
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