7 research outputs found
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Characterization of spent fuel approved testing material--ATM-104
The characterization data obtained to date are described for Approved Testing Material 104 (ATM-104), which is spent fuel from Assembly DO47 of the Calvert Cliffs Nuclear Power Plant (Unit 1), a pressurized-water reactor. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory (PNL) on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well-characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCRWM) Program. ATM-104 consists of 128 full-length irradiated fuel rods with rod-average burnups of about 42 MWd/kgM and expected fission gas release of about 1%. A variety of analyses were performed to investigate cladding characteristics, radionuclide inventory, and redistribution of fission products. Characterization data include (1) fabricated fuel design, irradiation history, and subsequent storage and handling history; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) special fuel studies involving analytical transmission electron microscopy (AEM) and electron probe microanalyses (EPMA); (6) calculated nuclide inventories and radioactivities in the fuel and cladding; and (7) radiochemical analyses of the fuel and cladding
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Identification of radionuclides of concern in Hanford Site environmental cleanup
The purpose of this document is to consider which radionuclides should be included in conducting environmental surveys relative to site remediation at Hanford. During the operation of the Hanford site, the fission product radionuclides and a large number of activation products including the transuranic radionuclides were formed. The reactor operations and subsequent chemical processing and metallurgical operations resulted in the environmental release of gaseous and liquid effluents containing some radionuclides; however, the majority of the radionuclides were stored in waste tanks or disposed to trenches and cribs. Since some contamination of both soils and subsurface waters occurred, one must decide which radionuclides still remain in sufficient amounts to be of concern at the time when site remediation is to be complete. Many of the radionuclides which have constituted the principal hazard during site operation have half-lives on the order of a year or less; therefore, they will have decayed to insignificant amounts by the year 2030, a possible date for completion of the remediation process
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Factors affecting criticality for spent fuel materials in a geologic setting
Following closure of a geologic repository for spent fuel, geologic process may change geometries and spacings, and water may enter the repository. In this study the conditions required for the criticality of spent fuel constituents are determined. Many factors affect criticality, and the effects of various possible post-closure changes are investigated. Factors having the greatest effect on criticality are identified to provide guidance for research programs and for design and evaluation studies. Section II describes the calculational methods and computer codes used to determine critical conditions. Section III of this document addresses effects of the fissile content of spent fuel on criticality. Calculations have been performed to determine the minimum critical mass of spent fuel actinides as a function of the duration of in-reactor fuel exposure for a variety of possible conditions. Section IV addresses the conditions required for criticality under a scenario believed to be highly unlikely but having a unique possibility. Pu quantities and concentrations required for criticality without water were determined for various conditions of Pu separation, rock moderation and reflection, rock impurities and isotopic content of the Pu. Section V addresses the possibility of geochemical processes separating Pu from other spent fuel constituents. Solubilities of U and Pu are calculated for groundwaters characteristic of basalt, tuff, granite, bedded and dome salt. Maximum concentrations which could be adsorbed on geologic media in contact with these groundwaters are then calculated. Comparison of these maximum adsorbed concentrations with the results presented in Section IV yields the conclusion that criticality cannot occur in sorbed deposits of Pu in geologic media due to the low Pu concentrations achievable. The possibility of selective Pu precipitation, however, is not ruled out by these arguments
LATTICES OF PLUTONIUM-ENRICHED RODS IN LIGHT WATER. PART II. THEORETICAL ANALYSIS OF PLUTONIUM-FUELED SYSTEMS.
Phoenix fuel evaluation in a maritime reactor design /
"October, 1968.""UC-80, Reactor Technology."Mode of access: Internet
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Characterization of spent fuel approved testing material: ATM-106
The characterization data obtained to date are described for Approved Testing Material (ATM)-106 spent fuel from Assembly BT03 of pressurized-water reactor Calvert Cliffs No. 1. This report is one in a series being prepared by the Materials Characterization Center at Pacific Northwest Laboratory on spent fuel ATMs. The ATMs are receiving extensive examinations to provide a source of well- characterized spent fuel for testing in the US Department of Energy Office of Civilian Radioactive Waste Management (OCWRM) program. ATM-106 consists of 20 full-length irradiated fuel rods with rod-average burnups of about 3700 GJ/kgM (43 MWd/kgM) and expected fission gas release of /approximately/10%. Characterization data include (1) as-fabricated fuel design, irradiation history, and subsequent storage and handling; (2) isotopic gamma scans; (3) fission gas analyses; (4) ceramography of the fuel and metallography of the cladding; (5) calculated nuclide inventories and radioactivities in the fuel and cladding; and (6) radiochemical analyses of the fuel and cladding. Additional analyses of the fuel rod are being conducted and will be included in planned revisions of this report. 12 refs., 110 figs., 81 tabs
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BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses
This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions