EDIT allows the user to edit or create text, either interactively or in the batch mode. The text may be a source program, program data, or documents. Special provisions permit the creation or editing of APT source-language programs. EDIT processes ASCII files from any accessible Data Director library and outputs files to any such library as specified by the user.

The preliminary results of an unsteady three-dimensional potential flow analysis of a five-liter flask air injection experiment (small-scale model simulation of a nuclear reactor steam condensation system) are presented. The location and velocity of the free water surface in the flask as a function of time are determined during pipe venting and bubble expansion processes. The analyses were performed using an extended version of the NASA-Ames Three-Dimensional Potential Flow Analysis System (POTFAN), which uses the vortex lattice singularity method of potential flow analysis. The pressure boundary condition at the free water surface and the boundary condition along the free jet boundary near the pipe exit were ignored for the purposes of the present study. The results of the analysis indicate that large time steps can be taken without significantly reducing the accuracy of the solutions and that the assumption of inviscid flow should not have an appreciable effect on the geometry and velocity of the free water surface. In addition, the computation time required for the solutions was well within acceptable limits.

This report discusses the current status of liquid-metal fast breeder reactor (LMFBR) development and one of the principal safety issues, a hypothetical core-disruptive accident (HCDA). Bibliographic information on worldwide LMFBRs relative to the development of the breeder reactor as a safe source of nuclear power is presented for the period 1974 through 1975. The bibliography consists of approximately 1554 abstracts covering early research and development and operating experiences leading up to the present design practices that are necessary for the licensing of breeder reactors. Key-word, author, and permuted-title indexes are included for completeness.

A geometrical multigrid method for solving the linearized matrix equations arising from node-on-face three-dimensional finite element contact is described. The development of an efficient implementation of this combination that minimizes both the memory requirements and the computational cost requires careful construction and storage of the portion of the coarse mesh stiffness matrices that are associated with the contact stiffness on the fine mesh. The multigrid contact algorithm is parallelized in a manner suitable for distributed memory architectures: results are presented that demonstrates the scheme's scalability. The solution of a large contact problem derived from an analysis of the factory joints present in the Space Shuttle reusable solid rocket motor demonstrates the usefulness of the general approach.

Graphical curves and text tables are presented that map out time and space distortions for data obtained from film records of the Many Beam Fabry Perot Velocimeter. Effective distortion corrections extracted from these mappings can be applied to upcoming velocimetry experiments, but only with limited success over periods of a year or more into the future. A method of using three fiducials to provide fresh time and space distortion data on each film record is presented as a more reliable procedure to correct distortions to an acceptable level of accuracy.

Isotopically enriched xenon isotopes are ideal for tracking the flow of relatively large volumes of groundwater. Dissolved noble gas tracers behave conservatively in the saturated zone, pose no health risk to drinking water supplies, and can be used with a large dynamic range. Different Xe isotopes can be used simultaneously at multiple recharge sources in a single experiment. Results from a tracer experiment at a California water district suggests that a small fraction of tracer moved from the recharge ponds through the thick, unconfined, coarse-grained alluvial aquifer to high capacity production wells at a horizontal velocity of 6 m/day. In contrast, mean water residence times indicate that the average rate of transport is 0.5 to 1 m/day.

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During the heat treatment step of High Temperature Gas-Cooled Reactor (HTGR) fuel beads, the /sup 233/U-loaded fuel beads are exposed to temperatures as high as 1700/sup 0/C. Since /sup 233/U fuel beads unavoidably contain a small percentage of /sup 232/U, decay products of /sup 232/U might be released during the heat treatment step. To avoid shielding requirements, experiments were conducted using /sup 232/Th-loaded fuel beads rather than /sup 232/U-loaded fuel beads. Decay products of /sup 232/U and /sup 232/Th are the same after /sup 228/Th. Experimental data were obtained by monitoring the 583 and 2614 keV gamma emissions of /sup 208/Tl. Computer simulations of the decay process and experimental data were used to infer what relative amount of the daughter products was removed. The carbonization step (300 to 500/sup 0/C) has an effect on the loss of daughter products in the conversion step. Using uncarbonized fuel beads, approximately 50 percent of /sup 220/Rn was removed at 300/sup 0/C and approximately 20 percent of /sup 220/Rn was removed at 200/sup 0/C. At most, 5 percent of /sup 220/Rn was removed from uncarbonized beads at 25/sup 0/C. Marginal /sup 220/Rn loss was detected at 100/sup 0/C using pre-carbonized beads. At 1400/sup 0/C, …

The purpose of this memo is to lay out the criteria for the Be capsule strength necessary for room temperature transport. Ultimately we will test full thickness capsules by sealing high pressures inside, but currently we are limited to both thinner capsules and alternative measures of capsule material strength.

COGEMA Engineering Corporation (COGEMA), under a contract from CH2M Hill Hanford Group (CH2M Hill), has performed an ultrasonic nondestructive examination of selected portions of Double-Shell Tank 241-AP-102. The purpose of this examination was to provide information that could be used to evaluate the integrity of the wall of the primary tank. The requirements for the ultrasonic examination of Tank 241-AP-102 were to detect, characterize (identify, size, and locate), and record measurements made of any wall thinning, pitting, or cracks that might be present in the wall of the primary tank. Any measurements that exceed the requirements set forth in the Engineering Task Plan (ETP), RPP-22571 (Jensen 2004) and summarized on page 1 of this document, are reported to CH2M Hill and the Pacific Northwest National Laboratory (PNNL) for further evaluation. Under the contract with CH2M Hill, all data is to be recorded on disk and paper copies of all measurements are provided to PNNL for third-party evaluation. PNNL is responsible for preparing a report that describes the results of the COGEMA ultrasonic examinations.

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Production Pu nitrate and unfabricated Pu metal during Feb. was below forecast; however FY output is above forecast. Production of UO{sub 3} exceeded commitments; shipments met schedule. Decontamination performance of Purex solvent extraction system was subnormal. Pu nitrate solutions were concentrated. A fire occurred in Purex N Cell during conversion of ion exchange prototype to production facility.

Specific conductance in one of the downgradient well exceeded its background value during the fourth quarter of FY 1999, triggering an upgrade from detection monitoring to a groundwater quality assessment program. This document presents the investigatory approach to be used for the assessment program. The objective of the first phase of the assessment program is to make the first determination to determine whether the WMA is the source of dangerous waste or dangerous waste constituents in the groundwater.

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Data have recently been obtained on a number of C-Reactor process tubes. These data make is possible to reassess the current conditions of the tubes in the C-Reactor and hence to re-evaluate the probable data when retubing of the reactor will have to be started. The results of this reassessment are presented below.

Project W-113 will provide the equipment and facilities necessary to retrieve suspect transuranic (TRU) waste from Trench 04 of the 218W-4C burial ground. As part of the retrieval process, waste drums will be assayed, overpacked, vented, head-gas sampled, and x-rayed prior to shipment to the Phase V storage facility in preparation for receipt at the Waste Receiving and Processing Facility (WRAP). Advanced Conceptual Design (ACD) studies focused on project items warranting further definition prior to Title I design and areas where the potential for cost savings existed. This ACD Report documents the studies performed during FY93 to optimize the equipment and facilities provided in relation to other SWOC facilities and to provide additional design information for Definitive Design.

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The February, 1958 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)

This document details activities of the irradiation processing department during the month of February 1958. A general summary is included at the start of the report, after which the report is divided into the following sections: research and engineering operations; production and reactor operations; facilities engineering operations; employee relations operation; and, financial operation.

This memorandum records the minutes of the PRA Senior Review Panel meeting held at Savannah River Laboratory (SRL) on December 1--3, 1987, and the report on that meeting written subsequently by the panel members. The minutes are contained as Attachment 2 of this memorandum, and the report as Attachment 1. The Panel indicated two principal concerns in their report: (1) that insufficient emphasis is being placed on the reliability data development program, and (2) that excessive detail is being built into the fault trees. These concerns have been addressed in a subsequent meeting with the Panel, held March 2--4, 1988. In addition, the members have been provided with a program document (Reference 1) indicating the extent, the timing, and the limitations of the data analysis effort for the PRA.

This report, for February 1962 fro 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.

This report, for February 1963 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; employee relations; weapons manufacturing operation; and safety and security.

Laboratory data for Sixth Round Groundwater samples collected during the 300-FF-5 Operable Unit Remedial Investigation have been reviewed and validated to ensure that they are of sufficient quality to support decisions regarding further actions to be taken at the 300-FF-5 Operable Unit. Table 1-1 is a summary of the validated samples. This report summarizes the results previously presented to Westinghouse Hanford in a series of Preliminary Quality Assurance Reports (PQAR) for the 300-FF-5 Sixth Round Groundwater samples. In some instances, the data qualifiers originally presented in the PQARs have been changed based upon further review of the data; these changes are highlighted in the text.

Westinghouse Hanford Company (WHC) is conducting a two-phased demonstration testing and evaluation of candidate melter system technologies for vitrification of Hanford Site low-level tank wastes. The testing is to be performed by melter equipment and vitrification technology commercial suppliers. This Statement of Work is for Phases 1 and 2 of the demonstration testing program. The primary objective of the demonstration testing is to identify the best available melter system technology for the Hanford Site LLW vitrification facility. Data obtained also will support various WHC engineering studies and conceptual design of the LLW vitrification facility. Multiple technologies will be selected for demonstration and evaluation. Testing will be conducted using non-radioactive LLW simulants in Seller-specified pilot/testing facilities.