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Potential of pyroprocessing for partitioning purex wastes

Description: The processes are extremely compact. The process reagents are highly resistant to radiation damage and, therefore, can be used to handle short-cooled, highly concentrated waste with negligible degradation. Most reagents can be recycled back through the process many times, thereby minimizing the generation of waste products, and also reducing the process cost. Fission-product wastes are discharged from the process as concentrated, solid wastes, typically in a metal matrix suitable for permanent … more
Date: July 23, 1980
Creator: Coops, M. S. & Sisson, D. H.
Partner: UNT Libraries Government Documents Department
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Analysis of Lime-Slurry Stirred Tank Carbonation Reactor

Description: Gas residence time distributions were determined for a stirred tank carbonation reactor. Empirical correlations for the first and second moments of the residence time distribution (RTD) curves as functions of flow rates and impeller speeds were obtained. Decontamination factors for /sup 85/Kr were measured.
Date: September 23, 1977
Creator: McAleese, J. P.; Belt, B. A.; Datesh, J. R. & Shaeffer, M. C.
Partner: UNT Libraries Government Documents Department
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CREEP STRAIN CORRELATION FOR IRRADIATED CLADDING

Description: In an attempt to predict the creep deformation of spent nuclear fuel cladding under the repository conditions, different correlations have been developed. One of them, which will be referred to as Murty's correlation in the following, and whose expression is given in Henningson (1998), was developed on the basis of experimental points related to unirradiated Zircaloy cladding (Henningson 1998, p. 56). The objective of this calculation is to adapt Murty's correlation to experimental points perta… more
Date: January 23, 2001
Creator: Macheret, P.
Partner: UNT Libraries Government Documents Department
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Production-scale Direct Oxide Reduction demonstration

Description: A detailed, statistically valid, examination of the direct oxide reduction parameters affecting process yield and purity was planned and executed. Guidelines for attaining yields approaching 100% are presented. Feed oxide, percent excess calcium, and stirrer design affected yield and product purity. Experiments were performed in production-scale equipment utilizing 800 grams of plutonium dioxide per charge. 1 ref., 9 figs., 3 tabs.
Date: January 23, 1989
Creator: Humiston, T.J.; Santi, D.J.; Long, J.L.; Thomas, R.L. (ed.) & Delaney, I.C. (comp.)
Partner: UNT Libraries Government Documents Department
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Scope design for conversion of Purex anion exchange prototype to a manufacturing facility

Description: This document is a HAPO report dated January 23, 1959, and describes the plutonium tail-end anion exchange system, installed in Purex as a prototype unit. Although some modifications, those considered most needed, were made to the unit, additional changes and refinements were still needed to convert the prototype to a fully acceptable manufacturing facility. This document covers the scope design of these modifications. The purpose of this document is to provide scope design criteria for a proje… more
Date: January 23, 1959
Creator: Gustafson, L. D.
Partner: UNT Libraries Government Documents Department
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Palm content of production metal

Description: Purpose of this study was to determine the {sup 237}Np input to the Purex process. The {sup 237}Np content of 639 g Pu/ton U irradiated fuel was found to be 1.78 {plus_minus} .09 g/ton of uranium at the 95% confidence level. Standard recovery for the chemical method was 96.7%, 98.0% for the sampling.
Date: November 23, 1959
Creator: Campbell, M. H. & Swift, W. H.
Partner: UNT Libraries Government Documents Department
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Transfer measurement statistics derived from Purex Processing Measurements` Quality Control Program

Description: The Purex Processing Measurements Quality Control Program was designed to provide for orderly accumulation of measurements data suitable for statistical evaluation and to identify those elements of a measurement which, if improved, would result in significant reduction in overall measurement variation. This report gives an analysis of the audit data obtained at four transfer stations: L9 plutonium product loadout, K6 uranium product to storage, F15 salt waste to F16 and D5 dissolved feed to pro… more
Date: November 23, 1965
Creator: Hough, C. G.
Partner: UNT Libraries Government Documents Department
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Cold vacuum drying residual free water test description

Description: Residual free water expected to remain in a Multi-Canister Overpack (MCO) after processing in the Cold Vacuum Drying (CVD) Facility is investigated based on three alternative models of fuel crevices. Tests and operating conditions for the CVD process are defined based on the analysis of these models. The models consider water pockets constrained by cladding defects, water constrained in a pore or crack by flow through a porous bed, and water constrained in pores by diffusion. An analysis of com… more
Date: December 23, 1997
Creator: Pajunen, A. L.
Partner: UNT Libraries Government Documents Department
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Chemical Processing Department monthly report, November 1957

Description: The November, 1957 monthly report for the Chemical Processing Department of the Hanford Atomic Products Operation includes information regarding research and engineering efforts with respect to the Purex and Redox process technology. Also discussed is the production operation, finished product operation, power and general maintenance, financial operation, engineering and research operations, and employee operation. (MB)
Date: December 23, 1957
Partner: UNT Libraries Government Documents Department
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Preliminary hazards analysis of thermal scrap stabilization system. Revision 1

Description: This preliminary analysis examined the HA-21I glovebox and its supporting systems for potential process hazards. Upon further analysis, the thermal stabilization system has been installed in gloveboxes HC-21A and HC-21C. The use of HC-21C and HC-21A simplified the initial safety analysis. In addition, these gloveboxes were cleaner and required less modification for operation than glovebox HA-21I. While this document refers to glovebox HA-21I for the hazards analysis performed, glovebox HC-21C i… more
Date: August 23, 1994
Creator: Lewis, W. S.
Partner: UNT Libraries Government Documents Department
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FB-Line resin testing final report

Description: The Dowex 50W-X8 and 50W-Xl2 resin samples are both strong acid cation materials in the hydrogen form. Each material has a water retention capacity characteristic of its respective marketed degree of cross-linking. Dowex 21K gives confirmatory responses to tests for a strong anion exchange resin in the nitrate form. All three resins have the manufacturer`s specified ionic type and form, and the Dowex 50W resins have characteristic water retention capacities. These tests conclude that the ion ex… more
Date: January 23, 1992
Creator: Bannochie, C. J.
Partner: UNT Libraries Government Documents Department
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FB-Line resin testing final report

Description: The Dowex 50W-X8 and 50W-Xl2 resin samples are both strong acid cation materials in the hydrogen form. Each material has a water retention capacity characteristic of its respective marketed degree of cross-linking. Dowex 21K gives confirmatory responses to tests for a strong anion exchange resin in the nitrate form. All three resins have the manufacturer's specified ionic type and form, and the Dowex 50W resins have characteristic water retention capacities. These tests conclude that the ion ex… more
Date: January 23, 1992
Creator: Bannochie, C. J.
Partner: UNT Libraries Government Documents Department
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Cold Vacuum Drying Facility hazard analysis report

Description: This report describes the methodology used in conducting the Cold Vacuum Drying Facility (CVDF) hazard analysis to support the CVDF phase 2 safety analysis report (SAR), and documents the results. The hazard analysis was performed in accordance with DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, and implements the requirements of US Department of Energy (DOE) Order 5480.23, Nuclear Safety Analysis Reports.
Date: February 23, 1998
Creator: Krahn, D. E.
Partner: UNT Libraries Government Documents Department
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Chemical Processing Department Monthly Report: September 1962

Description: This report, for September 1962 from the Chemical Processing Department at HAPO, discusses the following; Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations; facilities engineering; research; and employee relations.
Date: October 23, 1962
Creator: Hanford Atomic Products Operation. Chemical Processing Department.
Partner: UNT Libraries Government Documents Department
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Simulating the Effect on Criticality of Simultaneous Matrix Degradation and Assembly Collapse for the 21 PWR Waste Package

Description: The purpose of this calculation is to evaluate the effects of fission products loss on the reactivity of commercial pressurized water reactor (PWR) spent nuclear fuel (SNF) in 21 PWR waste packages (WPs) in the event of simultaneous fuel matrix degradation and assembly collapse.
Date: September 23, 1999
Creator: Alsaed, A. A.
Partner: UNT Libraries Government Documents Department
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Strategies for Application of Isotopic Uncertainties in Burnup Credit

Description: Uncertainties in the predicted isotopic concentrations in spent nuclear fuel represent one of the largest sources of overall uncertainty in criticality calculations that use burnup credit. The methods used to propagate the uncertainties in the calculated nuclide concentrations to the uncertainty in the predicted neutron multiplication factor (k{sub eff}) of the system can have a significant effect on the uncertainty in the safety margin in criticality calculations and ultimately affect the pote… more
Date: December 23, 2002
Creator: Gauld, I. C.
Partner: UNT Libraries Government Documents Department
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Fuel-cladding interaction layers in irradiated U-ZR and U-PU-ZR fuel elements.

Description: Argonne National Laboratory is developing an electrometallurgical treatment for spent nuclear fuels. The initial demonstration of this process is being conducted on U-Zr and U-Pu-Zr alloy fuel elements irradiated in the Experimental Breeder Reactor-II (EBR-II). The electrometallurgical treatment process extracts usable uranium from irradiated fuel elements and places residual fission products, actinides, process Zr, and cladding hulls (small segments of tubing) into two waste forms--a ceramic a… more
Date: January 23, 2006
Creator: Keiser, D. D.
Partner: UNT Libraries Government Documents Department
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TRIGA FUEL PHASE I AND II CRITICALITY CALCULATION

Description: The purpose of this calculation is to characterize the criticality aspect of the codisposal of TRIGA (Training, Research, Isotopes, General Atomic) reactor spent nuclear fuel (SNF) with Savannah River Site (SRS) high-level waste (HLW). The TRIGA SNF is loaded into a Department of Energy (DOE) standardized SNF canister which is centrally positioned inside a five-canister defense SRS HLW waste package (WP). The objective of the calculation is to investigate the criticality issues for the WP conta… more
Date: November 23, 1999
Creator: Angers, L.
Partner: UNT Libraries Government Documents Department
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Aqueous corrosion of aluminum-based nuclear fuel.

Description: As part of the U.S. National Spent Nuclear Fuel Program, aluminide fuels (UAl{sub x}) are being tested under conditions that might exist in the proposed repository at Yucca Mountain, Nevada. Intermittent drip tests at 90 C were completed for up to 183 days on partially declad, unirradiated, low-enriched UAl{sub x} samples. Through 183 days of exposure to modified water from the J-13 well at 90 C, the fuel coupon remained in good mechanical condition. Only a tarnishing of the surface was observe… more
Date: April 23, 2003
Creator: Kaminski, M.
Partner: UNT Libraries Government Documents Department
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NORMAL LOAD BEARING BY SITE SPECIFIC CANISTER

Description: The overall purpose of this calculation is to perform a preliminary analysis of the Site Specific Canister/Basket, subject to static gravity loads that include the self weight of the Canister Shell, the Basket, the Spent Nuclear Fuel, the Shield Plug and the related hardware, so that the loads are approximately known for sizing purposes. Based on these loads the stress levels in various components of the Site Specific Canister/Basket are evaluated.
Date: March 23, 2005
Partner: UNT Libraries Government Documents Department
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Memorandum of Understanding (MOU) Completion and Acceptance of the Spent Nuclear Fuel (SNF) Project

Description: This Memorandum of Understanding (MOU) is written to provide clear direction with respect to roles, responsibilities, obligations, and expectations of each organization identified. It functions as an agreement between the Operations, Construction Projects and Startup Organizations within the Spent Nuclear Fuels Project.
Date: November 23, 1999
Creator: NISHIKAWA, L.D.
Partner: UNT Libraries Government Documents Department
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