Analog studies of Icelandic geothermal fields have shown that the design of nuclear waste repositories in basalt can benefit by comparison to the data base already available from the development of these geothermal fields. A high degree of similarity exists between these two systems: their petrology, groundwater geochemistry, mineral solubilities, hydrologic parameters, temperature ranges, water-rock redox equilibria, hydrothermal pH values, and secondary mineralogies all show considerable overlap in the range of values. The experimentally-simulated hydrothermal studies of the basaltic nuclear waste repository rocks have, at this time, produced a data base that receives a strong confirmation from the Icelandic analog. Furthermore, the Icelandic analog should eventually be employed to extrapolate into higher and lower temperatures, into longer time-base chemical comparisons, and into more realistic mineral deposition studies, than have been possible in the laboratory evaluations of the nuclear waste repository designs. This eventual use of the Icelandic analog will require cooperative work with the Icelandic Geological Survey. 46 refs., 4 figs., 2 tabs.

The development of stable, efficient, photoelectrochemical cells based on silicon and gallium arsenide in non-aqueous electrolyte systems is being investigated. Redox reactions of ferrocene, anthracene and anthraquinone have been studied on platinum and n-silicon electrodes. The latter have been further characterized by differential capacitance measurements. Cells and equipment have been designed and set up for long-term stability studies.

Twenty five glass compositions were selected for a Phase II study to assess the potential for reducing the conservatism in the nepheline discriminator. The glass compositions were restricted to regions that fell within the validation ranges of the DWPF PCCS models. In addition, the liquidus temperature model was used to restrict the glass compositions so that they could all be melted at the same temperature. The nepheline discriminator was used to force the glass compositions into regions where nepheline formation was predicted to occur. The glasses were fabricated in the laboratory and characterized for crystallization and chemical durability after both quenching and slow cooling. Chemical analysis showed that the fabricated glasses met the target compositions. Nepheline was identified in one of the quenched glasses and several of the CCC glasses. There was no clear relationship between the types of crystallization that occurred in a particular glass and its location on the Al{sub 2}O{sub 3}-Na{sub 2}O-SiO{sub 2} ternary diagram. A partitioning algorithm was used to identify trends in crystallization behavior based on glass composition. Generally, for the CCC glasses MnO influenced the crystallization of spinels and B{sub 2}O{sub 3} and SiO{sub 2} influenced the crystallization of nepheline. Measured durability responses varied …

The main purpose of this project was to advance the basic scientific understanding of colloid and colloid-facilitated Cs transport of radionuclides in the vadose zone. We focused our research on the hydrological and geochemical conditions beneath the leaking waste tanks at the USDOE Hanford reservation. Specific objectives were (1) to determine the lability and thermodynamic stability of colloidal materials, which form after reacting Hanford sediments with simulated Hanford Tank Waste, (2) to characterize the interactions between colloidal particles and contaminants, i.e., Cs and Eu, (3) to determine the potential of Hanford sediments for \textit{in situ} mobilization of colloids, (4) to evaluate colloid-facilitated radionuclide transport through sediments under unsaturated flow, (5) to implement colloid-facilitated contaminant transport mechanisms into a transport model, and (6) to improve conceptual characterization of colloid-contaminant-soil interactions and colloid-facili\-tated transport for clean-up procedures and long-term risk assessment. We have previously shown that upon contact with simulated waste tank solutions, Hanford sediments change their mineralogical composition. Certain minerals, i.e., quartz, smectite, and kaolinite, are partially dissolved, and new mineral phases, i.e., the feldspathoids cancrinite and sodalite, are formed. We have characterized these mineral transformations and clarified the mineral transformation pathways. The new minerals were mainly in the colloidal size …

The Hanford Waste Management Project Master Documented Safety Analysis (MDSA) (HNF-14741, 2003) identifies derived safety controls to prevent or mitigate the risks of a single-container deflagration during operations requiring moving, venting or opening transuranic (TRU)-waste containers. The issue is whether these safety controls are necessary for operations involving TRU-waste boxes that are being retrieved from burial at the Hanford Site. This paper investigates the potential for a deflagration hazard within these boxes and whether safety controls identified for drum deflagration hazards should be applied to operations involving these boxes. The study evaluates the accumulation of hydrogen and VOCs within the waste box and the transport of these gases and vapors out of the waste box. To perform the analysis, there were numerous and major assumptions made regarding the generation rate and the transport pathway dimensions and their number. Since there is little actual data with regards to these assumptions, analyses of three potential configurations were performed to obtain some indication of the bounds of the issue (the concentration of hydrogen or flammable VOCs within a waste box). A brief description of each of the three cases along with the results of the analysis is summarized.

The removal of CO/sub 2/ from air in a calcium hydroxide slurry-agitated reactor was investigated to aid the design of such vessels. Gas-liquid interfacial areas were calculated using theoretical rate expression and experimental data at specific operating conditions. A correlation for interfacial areas was then determined as a function of impeller speed, impeller diameter, gas flow rate, and concentration of the slurry. Decontamination factors were also determined.

This study was conducted to identify processes that are most applicable for control of H/sub 2/S from geothermal sources. Both vapor-dominated and liquid-dominated sources were considered within the electric power generation category. The source characteristics, H/sub 2/S control requirements, and applicable technologies are discussed for the two geothermal sources. An evaluation of the applicable control technology indicates that there are three major approaches for H/sub 2/S removal. These are (a) upstream cleaning (ahead of the power plant), (b) removal of H/sub 2/S from condenser vent emissions, and (c) H/sub 2/S removal from cooling water, including condensate. The most promising processes for these emission points, based on current information, are as follows: the EIC process for upstream cleaning of liquid-dominated sources. For condenser vent emissions, the Stretford process appears to be most applicable; for cooling tower emissions, the iron catalyst process, followed by the H/sub 2/O/sub 2/ process, seems most appropriate.

Electric utilities are on the brink of a new era in waste disposal problems. This research project addresses the issue of how to effectively dispose of flyash, bottom ash, desulfurization sludge through the generation of chemically-hardened material that could potentially be used as a cement or as a synthetic aggregate. The specific goals of this study were: (1) to study the hardness of mixtures of flyash, bottom ash, and DSG treated with lime and other hardening agents; (2) to determine the optimum solids content, setting time, moisture content, and post setting treatments that will yield the greatest strength and hardness out of these mixtures; and (3) to determine the leachability of the synthetic material as a measure of its ability to retain absorbed and/or entrained toxic metals. 50 refs., 15 figs., 8 tabs.

An efficient convective waste heat recovery heat exchanger was designed and tested. The prototype appliance was designed for use in laundromats and other small commercial operations which use large amounts of hot water. Information on general characteristics of the coin-op laundry business, energy use in laundromats, energy saving resources already in use, and the potential market for energy saving devices in laundromats was collected through a literature search and interviews with local laundromat operators in Fort Collins, Colorado. A brief survey of time-use patterns in two local laundromats was conducted. The results were used, with additional information from interviews with owners, as the basis for the statistical model developed. Mathematical models for the advanced and conventional types were developed and the resulting computer program listed. Computer simulations were made using a variety of parameters; for example, different load profiles, hold-up volumes, wall resistances, and wall areas. The computer simulation results are discussed with regard to the overall conclusions. Various materials were explored for use in fabricating the appliance. Resistance to corrosion, workability, and overall suitability for laundromat installations were considered for each material.

It is generally expected that the depolarizing effects of the linear accelerator RF fields will be small. Recently Bill Atwood raised the question whether this conclusion is still correct in view of the fact that the particles in the SLC spend a larger fraction of their time at phase angles off crest'' due to BNS damping; since radial fields are in quadrature with the accelerating field this might imply that depolarizing effects are larger. On the other hand, because of the smaller emittance of the SLC relative to the earlier linac radial excursions would be smaller. The anticipation is therefore that the depolarizing effect will again be negligible but it might be worthwhile to update the early calculations of SLAC TN-63-97 revised in this paper.

In 1985, we reviewed the Department's procedures and practices for managing and controlling its major acquisition program, both for major systems and major projects. The 1985 review resulted in the identification of significant deficiencies. We found, for example, deficiencies relating to documentation and reporting requirements for major acquisitions. The purpose of this review was to determine if this condition had been corrected. The review included an examination of applicable laws, Executive Orders, Office of Management and Budget (OMB) Circulars, and Department policies. We examined key documents prepared for major acquisitions and reviewed reports used by senior Departmental officials to monitor these projects. Our audit was based primarily on a limited review of documentation available at Department of Energy (DOE) Headquarters. We did not extend our review of the issues raised in this report because we concluded that the management of major acquisitions was of such importance to the Department at this time that expedited reporting was needed.

Two prototype Thermal Cycling Absorption Process (TCAP) columns have been metallurgically examined after retirement, to determine the causes of failure and to evaluate the performance of the column container materials in this application. Leaking of the fluid heating and cooling subsystems caused retirement of both TCAP columns, not leaking of the main hydrogen-containing column. The aluminum block design TCAP column (ABL block TCAP) used in the Advanced Hydride Laboratory, Building 773-A, failed in one nitrogen inlet tube that was crimped during fabrication, which lead to fatigue crack growth in the tube and subsequent leaking of nitrogen from this tube. The Third Generation stainless steel design TCAP column (Third generation TCAP), operated in 773-A room C-061, failed in a braze joint between the freon heating and cooling tubes (made of copper) and the main stainless steel column. In both cases, stresses from thermal cycling and local constraint likely caused the nucleation and growth of fatigue cracks. No materials compatibility problems between palladium coated kieselguhr (the material contained in the TCAP column) and either aluminum or stainless steel column materials were observed. The aluminum-stainless steel transition junction appeared to be unaffected by service in the AHL block TCAP. Also, no evidence of …

This report, from the Chemical Processing Department at HAPO for October 1960, discusses the following: Production operation; Purex and Redox operation; Finished products operation; maintenance; Financial operations, facilities engineering; research; employee relations; and special separation processing and auxiliaries operation.

This project specific quality assurance program plan establishes the responsibility for the implementation of QA requirements, defines and documents the QA requirements associated with design, procurement, and construction, and defines and documents the degree of QA reviews and verifications on the design and construction necessary to assure compliance to project and DOE requirements. Revision 2 updates the QAPP to provide concurrence with approved work scope deletion. In addition, the Quality Assurance Program Index is being updated to reflect the current Quality Assurance Program requirements per DOE Order 5700.6C.

This document discusses the test program of evaluating the temperature effect of crud film build-up on fuel element heat generating surfaces in a carbon steel system. This program has three phases: Measurement of the effect of film build-up during normal high pH equilibrium operation; measurement of the temperature effect of film build-up subsequent to a loop decontamination; and measurement of the effect of film build-up in the event of loss of pH control.

In response to specific requests of the AEC and as part of a Combined Operations over-all review, a number of engineering studies have been made of alternative methods for increasing availabiltiy of neptunium at Hanford. The report updates an earlier study in which recycling of both natural and enriched uranium was considered for Hanford. The earlier study showed that recycled natural uranium would provide appreciable gains in neptunium availability but at an excessive cost. Recycle of the slightly enriched uranium streams proved a more economical means of realizing smaller but still significant gains in neptunium production. Subsequent to the earlier report, a feasible and immediately applicable scheme for UFI blending has been conceived demonstrating further advantages for recycling the enriched uranium. Approximately 34 kilograins of supplementary neptunium could be produced at Hanford during the next seven to eight years by upgrading irradiated E-metal and NPR uranium through a blending operation at Oak Ridge (rather than in the diffusion cascades) and then recycling the material through the Hanford reactors. (cf Table 1) Such a scheme would conserve uranium-236 for use as a source of neptunium-237 in the reactors without incurring major capital costs. Oak Ridge Operations personnel have estimated that capital …

This document presents a summary of work and progress at the Hanford Engineer Works for October 1951. The report is divided into sections by department. A plant wide general summary is included at the beginning of the report, after which the departmental summaries begin. The Manufacturing Department reports plant statistics, and summaries for the Metal Preparation, Reactor and Separation sections. The Engineering Department`s section summarizes work for the Technical Design, and Project Sections. Costs for the various departments are presented in the Financial Department`s summary. The Medical, Radiological Sciences, Utilities and General Services, Employee and Public Relations, and Community Real Estate and Services departments have sections presenting their monthly statistics, work, progress, and summaries.

Production efficiency studies of the E-N loading require that conversion ratio values for both plutonium and tritium be defined. Changing of the core-loading charge makeup with each core load to obtain a more efficient loading has theoretically resulted in increasing the tritium conversion ratio. Tritium recovery data from the third H Reactor E-N loading has recently become available for analysis. The buildup of product is much slower in fringe blanket material then in core target pieces. The second group of fringe conversion ratio data has only recently been obtained for analysis, the material analyzed was from the DR Reactor, irradiated under conditions closely paralleling those in the H Reactor blanket load. This document reports the conversion efficiencies for tritium production in Hanford E-N core and blanket loadings `indicated by these most recent data.

This document is the third of a series of question and answer lists issued as a review of the material discussed in the Reactor Processing Fundamentals Course. Each document represents the material covered by the Reactor Specialists during a typical three-month training program. Each question is discussed individually and the entire list completed during the three-month session. Each three-month-training period is devoted to the complete discussion of a single chapter of the IPD Physics Primer series. The questions are compiled in a logical sequence with the material as presented in Chapter V of the Primer series control. As the course progresses, a certain amount of recall of earlier chapters is essential to a thorough knowledge of the specifics of reactivity and distribution control. Therefore the first half of this document consists of material previously presented in a slightly different manner. Basic points are stressed; the intent is, of course, that informative discussions lead to a better understanding of the material presented in the Primer.

October performance of the plant production facilities was outstanding (915.2 tons U processed by Purex; 625.4 kg Pu separated by Redox/Purex). Redox processed three types of feed. Operation of incinerator furnace was resumed in Pu finishing processes. Capital cost estimates were prepared for several schemes for power reactor fuel reprocessing in Redox. Redox encased waste lines and line support system were found to be in good condition. H concentration in Redox dissolver off-gases occasionally exceeded lower flammable limits while sodium nitrate from high level waste storage tanks was used to suppress hydrogen.

This is the monthly report for the Hanford Laboratories Operation. Metallurgy, reactor fuels, physics and instrumentation, reactor technology, chemistry, separation processes, biology, financial activities, employee relations, laboratories auxiliaries, radiation protection, operation research, inventions, visits, and personnel status are discussed. This report is for October 1956.

Output of plutonium from the separations plants Purex, Redox was slightly less than that scheduled; however, year-to-date production exceeds the corresponding commitment. Production of UO{sub 3} met the commitment, and UO{sub 3} shipments conformed to the established shipping schedule. Production of buttons and shapes was slightly less than the commitment as shown in the current Official Forecast. Shipments of buttons conformed to the current Forecast, while large shape shipments were slightly less than those forecasted. Feasibility of neptunium recovery was studied. Employee relations are reported.

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The objective of this production test is to compare the dimensional and structural behavior of fuel elements with various heat treatments during irradiation at conditions similar to those anticipated for the NPR. Zircaloy-2 jacketed unalloyed natural uranium fuel elements of the same diametral dimensions an those anticipated for the N Reactor inner fuel tube, incorporating various heat treatments, will be irradiated to about 2000 MWD/T in the KER Loops. The elements will be enclosed in a Zircaloy-2 sleeve to provide a more balanced flow distribution through and around the pieces. An optimum heat treatment for N Reactor fuel elements providing minimum fuel distortion through control of uranium grain size and orientation has not been determined. This production test provides for the irradiation of fuel elements nearly equivalent to N Reactor inner tubes in size, geometry, and temperature conditions during irradiation. These elements have been subjected to varying heat treatments providing a range of grain sizes and degrees of orientation. Post-irradiation examination of the elements will show the fuel element distortion associated with each heat treatment and provide information on which to base further irradiation tenting in the development of a heat treating procedure for N Reactor fuel elements.