The management of high-level radioactive wastes

International Nuclear Information System (INIS)

Lennemann, Wm.L.

1979-01-01

The definition of high-level radioactive wastes is given. The following aspects of high-level radioactive wastes' management are discussed: fuel reprocessing and high-level waste; storage of high-level liquid waste; solidification of high-level waste; interim storage of solidified high-level waste; disposal of high-level waste; disposal of irradiated fuel elements as a waste

Testing of variables which affect stablity of cement solidified low-level waste

International Nuclear Information System (INIS)

Boris, G.F.

1989-01-01

This paper describes the test program undertaken to investigate variables which could affect the stability of cement solidified low-level waste and to evaluate the effect of these variables on certain tests prescribed in the Technical Position on Waste Form. The majority of the testing was performed on solidified undepleted bead resin, however, six additional waste types, suggested by the NRC, were tested. The tested variables included waste loading, immersion duration, depletion level, ambient cure duration, curing environment, immersion medium and waste type. Of these, lower waste loadings, longer ambient cures prior to testing and immersion in demineralized water versus simulated sea water and potable water resulted in higher compressive strengths for bead resin samples. Immersion times longer than 90 days did not affect the resin samples. Compressive strengths for other waste types varied depending upon the waste. The strengths of all waste types exceeded the minimum criterion by at least a factor of four, up to a factor of forty. The higher waste loadings exhibit strengths less than the lower waste loadings

Method of solidifying radioactive solid wastes

International Nuclear Information System (INIS)

Fukazawa, Tetsuo; Kawamura, Fumio; Kikuchi, Makoto.

1984-01-01

Purpose: To obtain solidification products of radioactive wastes satisfactorily and safely with no destruction even under a high pressure atmosphere by preventing the stress concentration by considering the relationships of the elastic module between the solidifying material and radioactive solid wastes. Method: Solidification products of radioactive wastes with safety and securing an aimed safety ratio are produced by conditioning the modules of elasticity of the solidifying material equal to or less than that of the radioactive wastes in a case where the elastic module of radioactive solid wastes to be solidified is smaller than that of the solidifying material (the elastic module of wastes having the minimum elastic module among various wastes). The method of decreasing the elastic module of the solidifying material usable herein includes the use of such a resin having a long distance between cross-linking points of a polymer in the case of plastic solidifying materials, and addition of rubber-like binders in the case of cement or like other inorganic solidifying materials. (Yoshihara, H.)

Radioactive liquid waste solidifying device

International Nuclear Information System (INIS)

Uchiyama, Yoshio.

1987-01-01

Purpose: To eliminate the requirement for discharge gas processing and avoid powder clogging in a facility suitable to the volume-reducing solidification of regenerated liquid wastes containing sodium sulfate. Constitution: Liquid wastes supplied to a liquid waste preheater are heated under a pressure higher than the atmospheric pressure at a level below the saturation temperature for that pressure. The heated liquid wastes are sprayed from a spray nozzle from the inside of an evaporator into the super-heated state and subjected to flash distillation. They are further heated to deposit and solidify the solidification components in the solidifying evaporation steams. The solidified powder is fallen downwardly and heated for removing water content. The recovered powder is vibrated so as not to be solidified and then reclaimed in a solidification storage vessel. Steams after flash distillation are separated into gas, liquid and solids by buffles. (Horiuchi, T.)

Microbially influenced degradation of cement-solidified low-level radioactive waste forms

International Nuclear Information System (INIS)

Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr.

1996-01-01

Because of its apparent structural integrity, cement has been widely used in the United States as a binder to solidify Class B and C low-level radioactive waste (LLW). However, the resulting cement preparations are susceptible to failure due to the actions of stress and environment. This paper contains information on three groups of microoganisms that are associated with the degradation of cement materials: sulfur-oxidizing bacteria (Thiobacillus), nitrifying bacteria (Nitrosomonas and Nitrobacter), and heterotrophic bacteria, which produce organic acids. Preliminary work using laboratory- and vendor-manufactured, simulated waste forms exposed to thiobacilli has shown that microbiologically influenced degradation has the potential to severely compromise the structural integrity of ion-exchange resin and evaporator-bottoms waste that is solidified with cement. In addition, it was found that a significant percentage of calcium was leached from the treated waste forms. Also, the surface pH of the treated specimens was decreased to below 2. These conditions apparently contributed to the physical deterioration of simulated waste forms after 30 to 60 days of exposure

Heat transfer in high-level waste management

International Nuclear Information System (INIS)

Dickey, B.R.; Hogg, G.W.

1979-01-01

Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed

NWTS program criteria for mined geologic disposal of nuclear waste: functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel

International Nuclear Information System (INIS)

1982-07-01

The Department of Energy (DOE) has primary federal responsibility for the development and implementation of safe and environmentally acceptable nuclear waste disposal methods. Currently, the principal emphasis in the program is on emplacement of nuclear wastes in mined geologic repositories well beneath the earth's surface. A brief description of the mined geologic disposal system is provided. The National Waste Terminal Storage (NWTS) program was established under DOE's predecessor, the Energy Research and Development Administration, to provide facilities for the mined geologic disposal of radioactive wastes. The NWTS program includes both the development and the implementation of the technology necessary for designing, constructing, licensing, and operating repositories. The program does not include the management of processing radioactive wastes or of transporting the wastes to repositories. The NWTS-33 series, of which this document is a part, provides guidance for the NWTS program in the development and implementation of licensed mined geologic disposal systems for solidified high-level and transuranic (TRU) wastes. This document presents the functional requirements and performance criteria for waste packages for solidified high-level waste and spent fuel. A separate document to be developed, NWTS-33(4b), will present the requirements and criteria for waste packages for TRU wastes. The hierarchy and application of these requirements and criteria are discussed in Section 2.2

Liquid wastes concentrating and solidifying device

International Nuclear Information System (INIS)

Kamiyoshi, Hideki; Ninokata, Yoshihide.

1985-01-01

Purpose: To provide a device for concentrating to solidify radioactive liquid wastes at large solidifying speed and with high decontaminating coefficient, without requirement for automatic control. Constitution: An asphalt solidifying device is disposed below a centrifugal thin film drier, and powder resulted from the drier is directly solidified with asphalt by utilizing the rotation of the drier for the mixing operation in the asphalt vessel. If abnormality should occur in the operation of the drier, resulting liquid wastes can be received and solidified in the asphalt vessel. The liquid wastes are heated to dry in a vessel main body having the heating surface at the circumferential surface. The vessel main body provided with a nozzle for supplying liquid to be treated disposed slantwise at the upper portion of the heating face, scrapers which rotate and slidingly contact the heating face and nozzles which jet out chemicals to the heating face behind the scrapers. Below the vessel main body, are disposed a funnel-like hopper for receiving falling scales, rotary vanes, and the likes by which the scales are introduced into the asphalt solidifying vessel. (Moriyama, K.)

Near-surface storage facilities for vitrified high-level wastes

International Nuclear Information System (INIS)

Kondrat'ev, A.N.; Kulichenko, V.V.; Kryukov, I.I.; Krylova, N.V.; Paramoshkin, V.I.; Strakhov, M.V.

1980-01-01

Concurrently with the development of methods for solidifying liquid radioactive wastes, reliable and safe methods for the storage and disposal of solidified wastes are being devised in the USSR and other countries. One of the main factors affecting the choice of storage conditions for solidified wastes originating from the vitrification of high-level liquid wastes from fuel reprocessing plants is the problem of removing the heat produced by radioactive decay. In order to prevent the temperature of solidified wastes from exceeding the maximum permissible level for the material concerned, it is necessary to limit either the capacity of waste containers or the specific heat release of the wastes themselves. In order that disposal of high-level wastes in geological formations should be reliable and economic, solidified wastes undergo interim storage in near-surface storage facilities with engineered cooling systems. The paper demonstrates the relative influences of specific heat release, of the maximum permissible storage temperature for vitrified wastes and of the methods chosen for cooling wastes in order for the dimensions of waste containers to be reduced to the extent required. The effect of concentrating wastes to a given level in the vitrification process on the cost of storage in different types of storage facility is also examined. Calculations were performed for the amount of vitrified wastes produced by a reprocessing plant with a capacity of five tonnes of uranium per 24 hours. Fuel elements from reactors of the water-cooled, water-moderated type are sent for reprocessing after having been held for about two years. The dimensions of the storage facility are calculated on the assumption that it will take five years to fill

Accelerated leach testing of radionuclides from solidified low-level waste

International Nuclear Information System (INIS)

Pietrzak, R.F.; Fuhrmann, M.; Franz, E.M.; Heiser, J. III; Colombo, P.

1989-01-01

This paper describes some of the work performed to develop an accelerated leach test designed to provide data that show long-term leaching behavior of solidified waste in a relatively short period of testing (1,2). The need for an accelerated leach test stems from the fact that the response of an effectively solidified waste form to the leaching process is so slow that a very long time is required to complete a test which shows the long-term leaching behavior of a waste form. Because of time limitations, as well as economic considerations, most studies have been limited to the early stages of the leaching process which is predominantly controlled by diffusion, although acknowledged to be due to also dissolution, corrosion or ion-exchange

Performance of high level waste forms and engineered barriers under repository conditions

International Nuclear Information System (INIS)

1991-02-01

The IAEA initiated in 1977 a co-ordinated research programme on the ''Evaluation of Solidified High-Level Waste Forms'' which was terminated in 1983. As there was a continuing need for international collaboration in research on solidified high-level waste form and spent fuel, the IAEA initiated a new programme in 1984. The new programme, besides including spent fuel and SYNROC, also placed greater emphasis on the effect of the engineered barriers of future repositories on the properties of the waste form. These engineered barriers included containers, overpacks, buffer and backfill materials etc. as components of the ''near-field'' of the repository. The Co-ordinated Research Programme on the Performance of High-Level Waste Forms and Engineered Barriers Under Repository Conditions had the objectives of promoting the exchange of information on the experience gained by different Member States in experimental performance data and technical model evaluation of solidified high level waste forms, components of the waste package and the complete waste management system under conditions relevant to final repository disposal. The programme includes studies on both irradiated spent fuel and glass and ceramic forms as the final solidified waste forms. The following topics were discussed: Leaching of vitrified high-level wastes, modelling of glass behaviour in clay, salt and granite repositories, environmental impacts of radionuclide release, synroc use for high--level waste solidification, leachate-rock interactions, spent fuel disposal in deep geologic repositories and radionuclide release mechanisms from various fuel types, radiolysis and selective leaching correlated with matrix alteration. Refs, figs and tabs

The evaluation of solidifying performance of heavy metal waste using cementitious materials (2)

International Nuclear Information System (INIS)

Fujita, Hideki; Harasawa, Shuichi

2005-02-01

Some of radioactive waste generated from JNC's facilities contain the poisonous substances such as lead, cadmium and mercury. In order to establish an appropriate method of the treatment of these heavy metals, solidification performance was evaluated using cementitious materials. In this report, the solidification performance of lead and mercury, which accounts for relatively high ratio in total wastes, was evaluated. The results are summarized below: 1. The test of stabilization process of mercury. The conversion process from mercury to the powdery mercury sulfide (red) was examined on the beaker scale. As a result, it was confirmed that the conversion was possible using the liquid phase reaction at 80deg C by the addition of sulfur powder with the NaOH solution. After the process, the mercury concentration in the filtrate was relatively high (0.6 mass%), so it was judged that the reuse of the recovered mercury waste fluid was indispensable. 2. The fabrication and evaluation of solidified wastes. The solidified waste were fabricated with cementitious material, and were evaluated by the measurement of one-axis compressive strength, the elution ratio of lead, mercury and so on. Powdery lead sulfide and the mercury sulfide of reagent were used as model waste. (1) solidification test of the lead waste. It was confirmed one-axis compressive strength for all solidified waste to pass the technical standards 15 kg/cm 2 (1.5 Mpa) for homogeneously solidified waste as the Low-level Radioactive Waste Disposal Center in Aomori Prefecture, and as for the elution ratio of lead, it had obtained the better result (0.06 mg/L) at the case of solidification of sulfide lead 30 mass% packed in the total solidified waste by using Highly Fly-ash contained Silica fume Cement (HFSC) than standard value (0.3 mg/L) at Regulations of Waste Management and Public Cleansing Law. Additionally, it was confirmed the using admixture of the inorganic reducing agent such as the Iron (II) chloride

Other-than-high-level waste

International Nuclear Information System (INIS)

Bray, G.R.

1976-01-01

The main emphasis of the work in the area of partitioning transuranic elements from waste has been in the area of high-level liquid waste. But there are ''other-than-high-level wastes'' generated by the back end of the nuclear fuel cycle that are both large in volume and contaminated with significant quantities of transuranic elements. The combined volume of these other wastes is approximately 50 times that of the solidified high-level waste. These other wastes also contain up to 75% of the transuranic elements associated with waste generated by the back end of the fuel cycle. Therefore, any detailed evaluation of partitioning as a viable waste management option must address both high-level wastes and ''other-than-high-level wastes.''

On confirmation of abandonment of imported waste (glass solidified bodies) outside business places

International Nuclear Information System (INIS)

1996-01-01

Electric power companies entrust the reprocessing of spent fuel generated from nuclear power stations to COGEMA in France, and in April, 1995, 28 high level radioactive wastes (glass solidified bodies) generated by the reprocessing were returned. When these glass solidified wastes are abandoned in the waste management facility of Japan Nuclear Fuel Service Co., it was decided to receive the confirmation of the prime minister on the measures based on the relevant law. Four electric power companies submitted the application and the explanation paper. As to the contents of the glass solidified wastes, the technical inspection was carried out by Bureau Veritas. Considering that this import of glass solidified wastes is the first in Japan, Science and Technology Agency carried out the measurement of all 28 wastes. The results are reported. It was confirmed that the measures for the abandonment taken by four electric power companies conform to the stipulation. The contents of the confirmation are reported in the order of the stipulation. These wastes were solidified with borosilicate glass in 5 mm thick stainless steel vessels, and the welding was done properly. (K.I.)

Detection of free liquid in containers of solidified radioactive waste

Science.gov (United States)

Greenhalgh, Wilbur O.

1985-01-01

A method of nondestructively detecting the presence of free liquid within a sealed enclosure containing solidified waste by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solidified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

Method of solidifying radioactive laundry wastes

International Nuclear Information System (INIS)

Yasumura, Keijiro

1984-01-01

Purpose: To enable to solidify radioactive laundry wastes containing non-ionic liquid detergents less solidifiable by plastic solidification process in liquid laundry wastes for cloths or the likes discharged from a nuclear power plant. Method: Radioactive laundry wastes are solidified by using plastic solidifying agent comprising, as a main ingredient, unsaturated polyester resins and methylmethacrylate monomers. The plastic solidifying agents usable herein include, for example, unsaturated polyester resins prepared by condensating maleic anhydride and phthalic anhydride with propylene glycol and incorporated with methylmethacrylate monomers. The mixing ratio of the methylmethacrylate monomers is preferably 30 % by weight based on the unsaturated polyester resins. (Aizawa, K.)

Development of methodology to evaluate microbially influenced degradation of cement-solidified low-level radioactive waste forms

International Nuclear Information System (INIS)

Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W.

1994-01-01

Because of its apparent structural integrity, cement has been widely used in the United States as a binder to solidify Class B and C low-level radioactive waste (LLW). However, the resulting cement preparations are susceptible to failure due to the actions of stress and environment. An environmentally mediated process that could affect cement stability is the action of naturally occurring microorganisms. The US Nuclear Regulatory Commission (NRC), recognizing this eventuality, stated that the effects of microbial action on waste form integrity must be addressed. This paper provides present results from an ongoing program that addresses the effects of microbially influenced degradation (MID) on cement-solidified LLW. Data are provided on the development of an evaluation method using acid-producing bacteria. Results are from work with one type of these bacteria, the sulfur-oxidizing Thiobacillus. This work involved the use of a system in which laboratory- and vendor-manufactured, simulated waste forms were exposed on an intermittent basis to media containing thiobacilli. Testing demonstrated that MID has the potential to severely compromise the structural integrity of ion-exchange resin and evaporator-bottoms waste that is solidified with cement. In addition, it was found that a significant percentage of calcium and other elements were leached from the treated waste forms. Also, the surface pH of the treated specimens decreased to below 2. These conditions apparently contributed to the physical deterioration of simulated waste forms after 60 days of exposure to the thiobacilli

Analysis of cement solidified product and ash samples and preparation of a reference material

International Nuclear Information System (INIS)

Ishimori, Ken-ichiro; Haraga, Tomoko; Shimada, Asako; Kameo, Yutaka; Takahashi, Kuniaki

2010-08-01

Simple and rapid analytical methods for radionuclides in low-level radioactive waste have been developed by the present authors. The methods were applied to simulated solidified products and actual metal wastes to confirm their usefulness. The results were summarized as analytical guide lines. In the present work, cement solidified product and ash waste were analyzed followed by the analytical guide lines and subjects were picked up and solved for the application of the analytical guide lines to these wastes. Pulverization and homogenization method for ash waste was improved to prevent a contamination since the radioactivity concentrations of the ash samples were relatively high. Pre-treatment method was altered for the cement solidified product and ash samples taking account for their high concentration of Ca. Newly, an analytical method was also developed to measure 129 I with a dynamic reaction cell inductively coupled plasma mass spectrometer. In the analytical test based on the improved guide lines, gamma-ray emitting nuclides, 60 Co and 137 Cs, were measured to estimate the radioactivity of the other alpha and beta-ray emitting nuclides. The radionuclides assumed detectable, 3 H, 14 C, 36 Cl, 63 Ni, 90 Sr, and alpha-ray emitting nuclides, were analyzed with the improved analytical guide lines and their applicability for cement solidified product and ash samples were confirmed. Additionally a cement solidified product sample was evaluated in terms of the homogeneity and the radioactivity concentrations in order to prepare a reference material for radiochemical analysis. (author)

Safety analysis of the transportation of high-level radioactive waste

International Nuclear Information System (INIS)

Murphy, E.S.; Winegardner, W.K.

1975-01-01

An analysis of the risk from transportation of solidified high-level waste is being performed at Battelle-Northwest as part of a comprehensive study of the management of high-level waste. The risk analysis study makes use of fault trees to identify failure events and to specify combinations of events which could result in breach of containment and a release of radioactive material to the environment. Contributions to risk analysis methodology which have been made in connection with this study include procedures for identification of dominant failure sequences, methods for quantifying the effects of probabilistic failure events, and computer code development. Preliminary analysis based on evaluation of the rail transportation fault tree indicates that the dominant failure sequences for transportation of solidified high-level waste will be those related to railroad accidents. Detailed evaluation of rail accident failure sequences is proceeding and is making use of the limited frequency-severity data which is available in the literature. (U.S.)

The evaluation of solidifying performance of heavy metal waste using cementitious materials

International Nuclear Information System (INIS)

Takei, Akihiko; Fujita, Hideki; Harasawa, Shuichi

2004-02-01

Some of radioactive waste generated form JNC's facilities contain the poisonous substances such as lead, cadmium and mercury. In order to establish an appropriate method of the treatment of these heavy metals, solidification performance was evaluated using cementitious materials. In this report, the solidification performance of lead, which accounts for relatively high ratio in total wastes, was evaluated. The results are summarized below: 1. The test of stabilization process of lead: The conversion process from block lead to the powdery lead sulfide was examined on the beaker scale. As a result, it was confirmed that the conversion was possible using the liquid phase reaction by the addition of thiourea after block lead had been dissolved by the acetic acid with bubbling air. After the process, the lead concentration in the filtrate was extremely low (0.02 mg/L), so it was judged that almost all of the lead was converted and recovered as lead sulfide. 2. The fabrication and evaluation of solidified wastes: Five types of solidified waste were fabricated with different binder, and were evaluated by the measurement of one-axis compressive strength, porosity, the elution ratio of lead, and so on. Powdery lead and sulfide lead reagent were used as model waste. As a result of the test, it was confirmed one-axis compressive strength for all solidified waste to pass the technical standards 15 kg/cm 2 (1.5 MPa) for homogeneously solidified waste as the Low-level Radioactive Waste Disposal Center in Aomori Prefecture, and as for the elution ratio of lead, it had obtained the better result (0.27 mg/L) at the case of solidification of sulfide lead 20 mass% packed in the total solidified waste by using low alkaline cement (including Hauyne mineral) than standard value (0.3 mg/L) at Regulations of Waste Management and Public Cleansing Law. Moreover, it was understood that the elution of lead had high relationship with not only the character of the binder but also the physical

Solidifying processing device for radioactive waste

International Nuclear Information System (INIS)

Sueto, Kumiko; Toyohara, Naomi; Tomita, Toshihide; Sato, Tatsuaki

1990-01-01

The present invention concerns a solidifying device for radioactive wastes. Solidifying materials and mixing water are mixed by a mixer and then charged as solidifying and filling materials to a wastes processing container containing wastes. Then, cleaning water is sent from a cleaning water hopper to a mixer to remove the solidifying and filling materials deposited in the mixer. The cleaning liquid wastes are sent to a separator to separate aggregate components from cleaning water components. Then, the cleaning water components are sent to the cleaning water hopper and then mixed with dispersing materials and water, to be used again as the mixing water upon next solidifying operation. On the other hand, the aggregate components are sent to a processing mechanism as radioactive wastes. With such procedures, since the discharged wastes are only composed of the aggregates components, and the amount of the wastes are reduced, facilities and labors for the processing of cleaning liquid wastes can be decreased. (I.N.)

Characterization of rapidly solidified powder of high-speed steel

Czech Academy of Sciences Publication Activity Database

Miglierini, M.; Lančok, Adriana; Kusý, M.

2009-01-01

Roč. 190, 1-3 (2009), s. 51-57 ISSN 0304-3843 R&D Projects: GA ČR GP203/07/P011 Grant - others:GA(SK) VEGA1/3190/06 Institutional research plan: CEZ:AV0Z40320502 Keywords : Rapidly solidified powder * Tool steel * Mössbauer spectroscopy Subject RIV: CA - Inorganic Chemistry Impact factor: 0.209, year: 2007

Overview of high-level waste management accomplishments

International Nuclear Information System (INIS)

Lawroski, H.; Berreth, J.R.; Freeby, W.A.

1980-01-01

Storage of power reactor spent fuel is necessary at present because of the lack of reprocessing operations particularly in the U.S. By considering the above solidification and storage scenario, there is more than reasonable assurance that acceptable, stable, low heat generation rate, solidified waste can be produced, and safely disposed. The public perception of no waste disposal solutions is being exploited by detractors of nuclear power application. The inability to even point to one overall system demonstration lends credibility to the negative assertions. By delaying the gathering of on-line information to qualify repository sites, and to implement a demonstration, the actions of the nuclear power detractors are self serving in that they can continue to point out there is no demonstration of satisfactory high-level waste disposal. By maintaining the liquid and solidified high-level waste in secure above ground storage until acceptable decay heat generation rates are achieved, by producing a compatible, high integrity, solid waste form, by providing a second or even third barrier as a compound container and by inserting the enclosed waste form in a qualified repository with spacing to assure moderately low temperature disposal conditions, there appears to be no technical reason for not progressing further with the disposal of high-level wastes and needed implementation of the complete nuclear power fuel cycle

Welding and Weldability of Directionally Solidified Single Crystal Nickel-Base Superalloys

Energy Technology Data Exchange (ETDEWEB)

Vitek, J M; David, S A; Reed, R W; Burke, M A; Fitzgerald, T J

1997-09-01

Nickel-base superalloys are used extensively in high-temperature service applications, and in particular, in components of turbine engines. To improve high-temperature creep properties, these alloys are often used in the directionally-solidified or single-crystal form. The objective of this CRADA project was to investigate the weldability of both experimental and commercial nickel-base superalloys in polycrystalline, directionally-solidified, and single-crystal forms.

Permitting plan for the high-level waste interim storage

International Nuclear Information System (INIS)

Deffenbaugh, M.L.

1997-01-01

This document addresses the environmental permitting requirements for the transportation and interim storage of solidified high-level waste (HLW) produced during Phase 1 of the Hanford Site privatization effort. Solidified HLW consists of canisters containing vitrified HLW (glass) and containers that hold cesium separated during low-level waste pretreatment. The glass canisters and cesium containers will be transported to the Canister Storage Building (CSB) in a U.S. Department of Energy (DOE)-provided transportation cask via diesel-powered tractor trailer. Tri-Party Agreement (TPA) Milestone M-90 establishes a new major milestone, and associated interim milestones and target dates, governing acquisition and/or modification of facilities necessary for: (1) interim storage of Tank Waste Remediation Systems (TWRS) immobilized HLW (IHLW) and other canistered high-level waste forms; and (2) interim storage and disposal of TWRS immobilized low-activity tank waste (ILAW). An environmental requirements checklist and narrative was developed to identify the permitting path forward for the HLW interim storage (HLWIS) project (See Appendix B). This permitting plan will follow the permitting logic developed in that checklist

Method for the conditioning of high level radioactive wastes for their safe storage and disposal

International Nuclear Information System (INIS)

Geel, J. van; Eschrich, H.; Detilleux, E.

1976-01-01

A method is described for the treatment of solidified high level radioactive wastes to enable them to be safely stored or disposed of in an approved manner. The solidified waste is embedded in a matrix of pure metals or metal alloys. The metals may be Pb, Pb/Sb alloys, Pb/Sn alloys, Pb/Bi alloys, Pb/Zn alloys, or mixtures of these, or Al, Al/Si alloys, Al/Mg alloys, Al/Cu alloys, or mixtures. The matrix is clad with non-corrosive material, selected from stainless steel, Ti, Pb, Pb alloys, Al, Al alloys, or mixtures of same. A non-corrosive container is filled with the solidified waste and is heated to above the melting temperature of the metallic matrix material used to embed the waste. The matrix material is then added and the container is cooled. The container may then be degassed. The solidified waste feed may be in the form of a vitreous material containing the high level waste; this vitreous material may consist of a lead borosilicate or a mixture of non-lead borosilicates and phosphate glasses, and the method of preparing it is described. (U.K.)

Microstructure of directionally solidified Ti-Fe eutectic alloy with low interstitial and high mechanical strength

Science.gov (United States)

Contieri, R. J.; Lopes, E. S. N.; Taquire de La Cruz, M.; Costa, A. M.; Afonso, C. R. M.; Caram, R.

2011-10-01

The performance of Ti alloys can be considerably enhanced by combining Ti and other elements, causing an eutectic transformation and thereby producing composites in situ from the liquid phase. This paper reports on the processing and characterization of a directionally solidified Ti-Fe eutectic alloy. Directional solidification at different growth rates was carried out in a setup that employs a water-cooled copper crucible combined with a voltaic electric arc moving through the sample. The results obtained show that a regular fiber-like eutectic structure was produced and the interphase spacing was found to be a function of the growth rate. Mechanical properties were measured using compression, microindentation and nanoindentation tests to determine the Vickers hardness, compressive strength and elastic modulus. Directionally solidified eutectic samples presented high values of compressive strength in the range of 1844-3000 MPa and ductility between 21.6 and 25.2%.

Outline of facility for studying high level radioactive materials (CPF) and study programmes

International Nuclear Information System (INIS)

Sakamoto, Motoi

1983-01-01

The Chemical Processing Facility for studying high level radioactive materials in Tokai Works of Power Reactor and Nuclear Fuel Development Corp. is a facility for fundamental studies centering around hot cells, necessary for the development of fuel recycle techniques for fast breeder reactors, an important point of nuclear fuel cycle, and of the techniques for processing and disposing high level radioactive liquid wastes. The operation of the facility was started in 1982, for both the system A (the test of fuel recycle for fast breeder reactors) and the system B (the test of vitrification of high level liquid wastes). In this report, the outline of the facility, the contents of testings and the reflection of the results are described. For the fuel recycle test, the hot test of the spent fuel pins of JOYO MK-1 core was started, and now the uranium and plutonium extraction test is underway. The scheduled tests are fuel solubility, the confirmation of residual properties in fuel melting, the confirmation of extracting conditions, the electrolytic reduction of plutonium, off-gas behaviour and the test of material reliability. For the test of vitrification of high level liquid wastes, the fundamental test on the solidifying techniques for the actual high level wastes eluted from the Tokai reprocessing plant has been started, and the following tests are programmed: Assessment of the properties of actual liquid wastes, denitration and concentration test, vitrification test, off-gas treatment test, the test of evaluating solidified wastes, and the test of storing solidified wastes. These test results are programmed to be reflected to the safety deliberation and the demonstration operation of a vitrification pilot plant. (Wakatsuki, Y.)

Method for solidifying powdery radioactive wastes

International Nuclear Information System (INIS)

Yasumura, Keijiro; Matsuura, Hiroyuki; Tomita, Toshihide.

1978-01-01

Purpose: To solidify powdery radioactive wastes through polymerization in a vessel at a high impregnation speed with no cloggings in pipes. Method: A drum can is lined with an inner liner layer of a predetermined thickness made of inflammable material such as glass fiber. A plurality of pipes for supplying liquid plastic monomer are provided in adjacent to the upper end face of the inflammable material or inserted between the vessel and the inflammable material. Then powdery radioactive wastes are filled in the vessel and the liquid plastic monomer dissolving therein a polymerization initiator is supplied through the pipes. The liquid plastic monomer impregnates through the inflammable material layer into the radioactive wastes and the plastic monomer is polymerized by the aid of the polymerization initiator after a predetermined of time to produce solidified plastic products of radioactive wastes. (Seki, T.)

Preconceptual design study for solidifying high-level waste: West Valley Demonstration Project

International Nuclear Information System (INIS)

Hill, O.F.

1981-04-01

This report presents a preconceptual design study for processing radioactive high-level liquid waste presently stored in underground tanks at Western New York Nuclear Service Center (WNYNSC) near West Valley, New York, and for incorporating the radionculides in that waste into a solid. The high-level liquid waste accumulated from the operation of a chemical reprocessing plant by the Nuclear Fuel Services, Inc. from 1966 to 1972. The high-level liquid waste consists of approximately 560,000 gallons of alkaline waste from Purex process operations and 12,000 gallons of acidic (nitric acid) waste from one campaign of processing thoria fuels by a modified Thorex process (during this campaign thorium was left in the waste). The alkaline waste contains approximately 30 million curies and the acidic waste contains approximately 2.5 million curies. The reference process described in this report is concerned only with chemically processing the high-level liquid waste to remove radionuclides from the alkaline supernate and converting the radionuclide-containing nonsalt components in the waste into a borosilicate glass

Conceptual process for conversion of high level waste to glass

International Nuclear Information System (INIS)

1975-01-01

During a ten-year period highly radioactive wastes amounting to 22 million gallons of salt cake and 5 million gallons of wet sludge are to be converted to 1.2 million gallons of glass and 24 million gallons of decontaminated salt cake and placed in the new storage facilities which will provide high assurance of containment with minimal reliance on maintenance and surveillance. The glass will contain nearly all of the radioactivity in a form that is highly resistant to leaching and dispersion. The salt cake will contain a small amount of residual radioactivity. The process is shown in Figure 1 and the facilities may be arranged in seven modules to accomplish seven tasks, (1) remove wastes from tanks, (2) separate sludge and salt, (3) decontaminate salt, (4) solidify and package sludge and 137 Cs, (5) solidify and package decontaminated salt, (6) store high level waste, and (7) store decontaminated salt cake

Nanoscale microstructure effects on hydrogen behavior in rapidly solidified aluminum alloys

Energy Technology Data Exchange (ETDEWEB)

Tashlykova-Bushkevich, Iya I. [Belarusian State University of Informatics and Radioelectronics, Minsk (Belarus)

2015-12-31

The present work summarizes recent progress in the investigation of nanoscale microstructure effects on hydrogen behavior in rapidly solidified aluminum alloys foils produced at exceptionally high cooling rates. We focus here on the potential of modification of hydrogen desorption kinetics in respect to weak and strong trapping sites that could serve as hydrogen sinks in Al materials. It is shown that it is important to elucidate the surface microstructure of the Al alloy foils at the submicrometer scale because rapidly solidified microstructural features affect hydrogen trapping at nanostructured defects. We discuss the profound influence of solute atoms on hydrogen−lattice defect interactions in the alloys. with emphasis on role of vacancies in hydrogen evolution; both rapidly solidified pure Al and conventionally processed aluminum samples are considered.

Leaching behavior of solidified plastics radioactive wastes

International Nuclear Information System (INIS)

Yook, Chong Chul; Lee, Byung Hun; Jae, Won Mok; Kim, Kyung Eung

1986-01-01

It is highly needed to develope the solidification process to dispose safely the radioactive wastes increasing with the growth of the nuclear industry. The leaching mechanisms of the solidified plastic wastes were investigated and the leaching rates of the plastic wastes were also measured among the many solidification processes. In addition, the transport equation based on the diffusion or the diffusion-dissolution was compared with the empirical equation derived from the experimental data by graphical method. Consequently, leaching process of the solidified plastic wastes is quite well agreed with the mass transport theory, but it may be difficult to simulate leaching process by diffusion dissolution mechanism. But the theoretical equation could be applicable to the cumulative amount of radionuclides leached form the plastic wastes disposed into the environment. (Author)

High-level radioactive waste glass and storage canister design

International Nuclear Information System (INIS)

Slate, S.C.; Ross, W.A.

1979-01-01

Management of high-level radioactive wastes is a primary concern in nuclear operations today. The main objective in managing these wastes is to convert them into a solid, durable form which is then isolated from man. A description is given of the design and evaluation of this waste form. The waste form has two main components: the solidified waste and the storage canister. The solid waste form discussed in this study is glass. Waste glasses have been designed to be inert to water attack, physically rugged, low in volatility, and stable over time. Two glass-making processes are under development at PNL. The storage canister is being designed to provide high-integrity containment for solidified wastes from processing to terminal storage. An outline is given of the steps in canister design: material selection, stress and thermal analyses, quality verification, and postfill processing. Examples are given of results obtained from actual nonradioactive demonstration tests. 14 refs

Solidifier effectiveness : variation due to oil composition, oil thickness and temperature

International Nuclear Information System (INIS)

Fieldhouse, B.; Fingas, M.

2009-01-01

This paper provided an overview of solidifier types and composition. Solidifiers are a class of spill treating agents that offer an effective means to convert a liquid oil into a solid material. They are used as a treatment option for oil spills on water. This paper also reported on recent laboratory studies that consist of 4 components: (1) a qualitative examination of the characteristics of the interaction of a broad range of solidifier products with a standard oil to evaluate reaction rate, states of solidification, and the impact of dosage, (2) a comparison of a smaller subset of solidifiers on the standard oil at lower temperatures, (3) solidifier treatment on a range of oils of varying physical properties and composition to assess the potential scope of application, and (4) the treatment of a series of small-scale oil layers of varying thickness to determine the significance of oil thickness on solidifier effectiveness and recovery. This paper also reviewed solidifier chemistry with particular reference to polymer sorbents; cross-linking agents; and cross-linking agents and polymeric sorbents combined. Toxicity is also an important issue regarding solidifiers. The aquatic toxicity of solidifiers is low and not measurable as the products are not water-soluble. There have not been any studies on the effects of the solidifier or the treated oil on surface feeders and shoreline wildlife that may come into contact with the products. It was concluded that oil composition may play a major role in solidifier effectiveness. The effectiveness of solidifiers is also inhibited at reduced temperatures, increased viscosity and density of the oil. 25 refs., 5 tabs., 2 figs., 1 appendix

Radiochemical analysis of homogeneously solidified low level radioactive waste from nuclear power plants

International Nuclear Information System (INIS)

Sato, Kaneaki; Ikeuchi, Yoshihiro; Higuchi, Hideo

1995-01-01

As mentioned above, we have reliable radioanalytical methods for all kinds of homogeneously solidified wastes. We are now under studying an analytical method for pellets which are made from evaporator concentrates or resin. And we are going to study to establish new analytical method for the rad-waste including metal, cloths and so on in near future. (J.P.N.)

Microstructure and orientation evolution in unidirectional solidified Al–Zn alloys

Energy Technology Data Exchange (ETDEWEB)

Chen, Zhongwei, E-mail: chzw@nwpu.edu.cn; Wang, Enyuan; Hao, Xiaolei

2016-06-14

Morphological instability and growth orientation evolution during unidirectional solidification of Al–Zn alloys with different pulling speeds were investigated by X-ray diffraction (XRD) and electron back-scatter diffraction (EBSD) in scanning electron microscope (SEM). The experimental results show that, as the pulling speed increases, the primary dendrite spacing becomes smaller gradually and dendrite trunks incline to the heat flow direction perfectly in unidirectional solidified Al–9.8 wt%Zn and Al–89 wt%Zn alloys. However, regardless of the pulling speed in unidirectional solidified Al–Zn alloys under fixed thermal gradient, the regular dendrites with <100> directions of primary trunks and secondary arms in 9.8 wt% Zn composition are replaced by <110> dendrites of primary trunks and secondary arms in 89 wt% Zn composition. In unidirectional solidified Al–32 wt% Zn alloy, cellular, fractal seaweed, and stabilized seaweed structures were observed at high pulling speeds. At a high pulling speed of 1000 µm/s, seaweed structures transform to the columnar dendrites with <110> trunks and <100> arms. The above orientation evolution can be attributed to low anisotropy of solid-liquid interface energy and the seaweed structure is responsible for isotropy of {111} planes.

Energy asymmetry in melting and solidifying processes of PCM

International Nuclear Information System (INIS)

Jin, Xing; Hu, Huoyan; Shi, Xing; Zhang, Xiaosong

2015-01-01

Highlights: • The melting process and the solidifying process of PCM were asymmetrical. • The enthalpy and state of PCM were affected by its previous state. • The main reason for energy asymmetry of PCM was supercooling. - Abstract: The solidifying process of phase change material (PCM) was usually recognized as the exact inverse process of its melting process, especially when building the heat transfer model of PCM. To figure out that whether the melting process and the solidifying process of PCM were symmetrical, several kinds of PCMs were tested by a differential scanning calorimeter (DSC) in this paper. The experimental results showed that no matter using the DSC dynamic measurement method or the DSC step measurement method, the melting process and the solidifying process of PCM were asymmetrical. Because of the energy asymmetry in the melting and solidifying processes of PCM, it was also found that the enthalpy and the state of PCM were not only dependent on its temperature, but also affected by its “previous state”.

Recovering method for high level radioactive material

International Nuclear Information System (INIS)

Fukui, Toshiki

1998-01-01

Offgas filters such as of nuclear fuel reprocessing facilities and waste control facilities are burnt, and the burnt ash is melted by heating, and then the molten ashes are brought into contact with a molten metal having a low boiling point to transfer the high level radioactive materials in the molten ash to the molten metal. Then, only the molten metal is evaporated and solidified by drying, and residual high level radioactive materials are recovered. According to this method, the high level radioactive materials in the molten ashes are transferred to the molten metal and separated by the difference of the distribution rate of the molten ash and the molten metal. Subsequently, the molten metal to which the high level radioactive materials are transferred is heated to a temperature higher than the boiling point so that only the molten metal is evaporated and dried to be removed, and residual high level radioactive materials are recovered easily. On the other hand, the molten ash from which the high level radioactive material is removed can be discarded as ordinary industrial wastes as they are. (T.M.)

Method of solidifying radioactive liquid wastes

International Nuclear Information System (INIS)

Uetake, Naoto; Kawamura, Fumio; Kikuchi, Makoto; Fukazawa, Tetsuo.

1983-01-01

Purpose: To enable to confine the volatiling ingredients such as cesium in liquid wastes safely in glass solidification products while suppressing the volatilization thereof. Method: Acid salt of tetravalent metal such as titanium phosphate has an intense selective adsorption property to cesium. So liquid wastes stored in a high level liquid wastes tank is mixed with titanium phosphate gels stored in an adsorbent tank, then supplied to a mixer and mixed with a sodium silicate solution stored in a sodium silicate storage tank and boric acid stored in an additive tank, into gel-like state. The gel-like material thus formed is supplied to a drier. After being dried at a temperature of 200sup(o)C - 300sup(o)C, the material is melted under heating at a temperature of 1000sup(o)C - 1100sup(o)C, and then cooled to solidify. (Horiuchi, T.)

Radioactive substance solidifying device

International Nuclear Information System (INIS)

Sakoda, Kotaro.

1979-01-01

Purpose: To easily solidify radioactive substances adhering to the surfaces of solid wastes without scattering in the circumference by paints, and further to reduce surface contamination concentrations. Constitution: Solid wastes are placed on a hanging plate, and dipped in paints within a paint dipping treatment tank installed at the lower part of a treatment tank by means of a monorail hoist, and the surfaces of said solid wastes are coated with paints, thereby to solidify the radioactivity on the surfaces of the solid wastes. After dipping, the solid wastes are suspended up to a paint spraying tank to dry the paints. After drying, non-contaminated paints are atomized to apply through an atomizing tube onto the solid wastes. After drying the atomized paints, the solid wastes are carried outside the treatment tank by means of the monorail hoist. (Yoshino, Y.)

Cold Heat Release Characteristics of Solidified Oil Droplet-Water Solution Latent Heat Emulsion by Air Bubbles

Science.gov (United States)

Inaba, Hideo; Morita, Shin-Ichi

The present work investigates the cold heat-release characteristics of the solidified oil droplets (tetradecane, C14H30, freezing point 278.9 K)/water solution emulsion as a latent heat-storage material having a low melting point. An air bubbles-emulsion direct-contact heat exchange method is selected for the cold heat-results from the solidified oil droplet-emulsion layer. This type of direct-contact method results in the high thermal efficiency. The diameter of air bubbles in the emulsion increases as compared with that in the pure water. The air bubbles blown from a nozzle show a strong mixing behavior during rising in the emulsion. The temperature effectiveness, the sensible heat release time and the latent heat release time have been measured as experimental parameters. The useful nondimensional emulsion level equations for these parameters have been derived in terms of the nondimensional emalsion level expressed the emulsion layer dimensions, Reynolds number for air flow, Stefan number and heat capacity ratio.

Parameters of Solidifying Mixtures Transporting at Underground Ore Mining

Directory of Open Access Journals (Sweden)

Golik Vladimir

2017-01-01

Full Text Available The article is devoted to the problem of providing mining enterprises with solidifying filling mixtures at underground mining. The results of analytical studies using the data of foreign and domestic practice of solidifying mixtures delivery to stopes are given. On the basis of experimental practice the parameters of transportation of solidifying filling mixtures are given with an increase in their quality due to the effect of vibration in the pipeline. The mechanism of the delivery process and the procedure for determining the parameters of the forced oscillations of the pipeline, the characteristics of the transporting processes, the rigidity of the elastic elements of pipeline section supports and the magnitude of vibrator’ driving force are detailed. It is determined that the quality of solidifying filling mixtures can be increased due to the rational use of technical resources during the transportation of mixtures, and as a result the mixtures are characterized by a more even distribution of the aggregate. The algorithm for calculating the parameters of the pipe vibro-transport of solidifying filling mixtures can be in demand in the design of mineral deposits underground mining technology.

Parameters of Solidifying Mixtures Transporting at Underground Ore Mining

Science.gov (United States)

Golik, Vladimir; Dmitrak, Yury

2017-11-01

The article is devoted to the problem of providing mining enterprises with solidifying filling mixtures at underground mining. The results of analytical studies using the data of foreign and domestic practice of solidifying mixtures delivery to stopes are given. On the basis of experimental practice the parameters of transportation of solidifying filling mixtures are given with an increase in their quality due to the effect of vibration in the pipeline. The mechanism of the delivery process and the procedure for determining the parameters of the forced oscillations of the pipeline, the characteristics of the transporting processes, the rigidity of the elastic elements of pipeline section supports and the magnitude of vibrator' driving force are detailed. It is determined that the quality of solidifying filling mixtures can be increased due to the rational use of technical resources during the transportation of mixtures, and as a result the mixtures are characterized by a more even distribution of the aggregate. The algorithm for calculating the parameters of the pipe vibro-transport of solidifying filling mixtures can be in demand in the design of mineral deposits underground mining technology.

Way of thinking and method of promotion of disposal of high level radioactive waste

International Nuclear Information System (INIS)

Toyota, Masatoshi

1993-01-01

It is decided that the high level waste separated from spent fuel is solidified with glass, stored for 30-50 years to cool it down, and the final disposal is done under the responsibility of the government. As to the final disposal of high level waste, the method of enclosing glass-solidified waste in robust containers and burying them in deep stable strata to isolate from human environment is considered to be the safest. The significance of fuel reprocessing is the proper and safe separation and control of high level waste besides the reuse of unburned uranium and newly formed plutonium in spent fuel. The features of the high level waste solids are that their amount to be generated is little, the radioactivity attenuates with the lapse of time, the heat generation decreases with the lapse of time, and they are hard to elute and move. In order to prevent radioactive substances from appearing in human environment by being dissolved in groundwater, those are isolated with the combination of natural and artificial barriers. The requirements for the barriers are discussed. The research and development are in progress on the establishment of stratum disposal technology, the evaluation of suitability of geological environment and the selection of expected disposal grounds. (K.I.)

Effects of leachate concentration on the integrity of solidified clay liners.

Science.gov (United States)

Xue, Qiang; Zhang, Qian

2014-03-01

This study aimed to evaluate the impact of landfill leachate concentration on the degradation behaviour of solidified clay liners and to propose a viable mechanism for the observed degradation. The results indicated that the unconfined compressive strength of the solidified clay decreased significantly, while the hydraulic conductivity increased with the leachate concentration. The large pore proportion in the solidified clay increased and the sum of medium and micro pore proportions decreased, demonstrating that the effect on the solidified clay was evident after the degradation caused by exposure to landfill leachate. The unconfined compressive strength of the solidified clay decreased with increasing leachate concentration as the leachate changed the compact structure of the solidified clay, which are prone to deformation and fracture. The hydraulic conductivity and the large pore proportion of the solidified clay increased with the increase in leachate concentration. In contrast, the sum of medium and micro pore proportions showed an opposite trend in relation to leachate concentration, because the leachate gradually caused the medium and micro pores to form larger pores. Notably, higher leachate concentrations resulted in a much more distinctive variation in pore proportions. The hydraulic conductivity of the solidified clay was closely related to the size, distribution, and connection of pores. The proportion of the large pores showed a positive correlation with the increase of hydraulic conductivity, while the sum of the proportions of medium and micro pores showed a negative correlation.

A Study on Factors Affecting Strength of Solidified Peat through XRD and FESEM Analysis

Science.gov (United States)

Rahman, J. A.; Napia, A. M. A.; Nazri, M. A. A.; Mohamed, R. M. S. R.; Al-Geethi, A. S.

2018-04-01

Peat is soft soil that often causes multiple problems to construction. Peat has low shear strength and high deformation characteristics. Thus, peat soil needs to be stabilized or treated. Study on peat stabilization has been conducted for decades with various admixtures and mixing formulations. This project intends to provide an overview of the solidification of peat soil and the factors that affecting the strength of solidified peat soil. Three types of peats which are fabric, hemic and sapric were used in this study to understand the differences on the effect. The understanding of the factors affecting strength of solidified peat in this study is limited to XRD and FESEM analysis only. Peat samples were collected at Pontian, Johor and Parit Raja, Johor. Peat soil was solidified using fly ash, bottom ash and Portland cement with two mixing formulation following literature review. The solidified peat were cured for 7 days, 14 days, 28 days and 56 days. All samples were tested using Unconfined Compressive Strength Test (UCS), X-ray diffraction (XRD) and Field Emission Scanning Electron Microscope (FESEM). The compressive strength test of solidified peat had shown consistently increase of sheer strength, qu for Mixing 1 while decrease of its compressive strength value for Mixing 2. All samples were tested and compared for each curing days. Through XRD, it is found that all solidified peat are dominated with pargasite and richterite. The highest qu is Fabric Mixing 1(FM1) with the value of 105.94 kPa. This sample were proven contain pargasite. Samples with high qu were observed to be having fly ash and bottom ash bound together with the help of pargasite. Sample with decreasing strength showed less amount of pargasite in it. In can be concluded that XRD and FESEM findings are in line with UCS values.

Procedure for conditioning high-level solidified wastes

Energy Technology Data Exchange (ETDEWEB)

Hild, W; Krause, H; Scheffler, K

1974-05-30

The molds of glass, ceramic or basalt-similar mass in which highly radioactive wastes are incorporated are used for the conditioning of waste waters and/or of sewage or precipitating sludge or of natural water to obtain drinking water, prior to the end storage. By means of the gamma-radiation they emit, the viruses and bacteria and worm eggs are killed off as well as the poisonous, and organic substances such as, e.g., chlorated aromatics are destroyed. Furthermore, the filtration power is increased by coagulation, and the sludge is drained. Natural water is degermed. In particular, fission product mixtures of light water reactors can be incorporated in the molds. The molds are immersed in the media.

Method of solidifying radioactive wastes with plastics

International Nuclear Information System (INIS)

Matsuura, Hiroyuki; Yasumura, Keijiro; Minami, Yuji; Tomita, Toshihide

1980-01-01

Purpose: To prevent solidification of solidifying agents in the mixer by conducting the mixing process for the solidifying agents and the radioactive wastes at a temperature below the initiation point for the solidification of the agents thereby separating the mixing process from the solidification-integration process. Method: Catalyst such as cobalt naphthenate is charged into an unsaturated polyester resin in a mixer previously cooled, for example, to -10 0 C. They are well mixed with radioactive wastes and the mixture in the mixer is charged in a radioactive waste storage container. The temperature of the mixture, although kept at a low temperature initially, gradually increases to an ambient temperature whereby curing reaction is promoted and the reaction is completed about one day after to provide firm plastic solidification products. This can prevent the solidification of the solidifying agents in the mixer to thereby improve the circumstance's safety. (Kawakami, Y.)

The use of Nb in rapid solidified Al alloys and composites

Energy Technology Data Exchange (ETDEWEB)

Audebert, F., E-mail: metal@fi.uba.ar [Advanced Materials Group, Facultad de Ingeniería, Universidad de Buenos Aires, Paseo Colón 850, Ciudad de Buenos Aires 1063 (Argentina); Department of Materials, University of Oxford, Parks Road, OX1 3PH Oxford (United Kingdom); Department of Mechanical Engineering and Mathematical Sciences, Oxford Brookes University, Wheatley Campus, OX33 1HX Oxford (United Kingdom); Galano, M. [Department of Materials, University of Oxford, Parks Road, OX1 3PH Oxford (United Kingdom); Saporiti, F. [Advanced Materials Group, Facultad de Ingeniería, Universidad de Buenos Aires, Paseo Colón 850, Ciudad de Buenos Aires 1063 (Argentina)

2014-12-05

Highlights: • The use of Nb in RS Al alloys and composites has been reviewed. • Nb was found to improve the GFA of rapid solidified Al–Fe and Al–Ni alloys. • Nb has higher effect in increasing the corrosion resistance than RE in Al–Fe alloys. • Nb improves the stability of the Al–Fe–Cr icosahedral phase. • Nb improves strength, ductility and toughness of nanoquasicrystalline Al matrix composites. - Abstract: The worldwide requirements for reducing the energy consumption and pollution have increased the demand of new and high performance lightweight materials. The development of nanostructured Al-based alloys and composites is a key direction towards solving this demand. High energy prices and decreased availability of some alloying elements open up the opportunity to use non-conventional elements in Al alloys and composites. In this work the application of Nb in rapid solidified Al-based alloys and Al alloys matrix composites is reviewed. New results that clarify the effect of Nb on rapid solidified Al alloys and composites are also presented. It is observed that Nb stabilises the icosahedral Al–Fe/Cr clusters, enhances the glass forming ability and shifts the icosahedral phase decomposition towards higher temperatures. Nb provides higher corrosion resistance with respect to the pure Al and Al–Fe–RE (RE: rare earth) alloys in the amorphous and crystalline states. The use of Nb as a reinforcement to produce new Al alloy matrix composites is explored. It is observed that Nb provides higher strength, ductility and toughness to the nanoquasicrystalline matrix composite. Nb appears as a new key element that can improve several properties in rapid solidified Al alloys and composites.

Detection of free liquid in containers of solidified radioactive waste

Science.gov (United States)

Greenhalgh, W.O.

Nondestructive detection of the presence of free liquid within a sealed enclosure containing solidified waste is accomplished by measuring the levels of waste at two diametrically opposite locations while slowly tilting the enclosure toward one of said locations. When the measured level remains constant at the other location, the measured level at said one location is noted and any measured difference of levels indicates the presence of liquid on the surface of the solifified waste. The absence of liquid in the enclosure is verified when the measured levels at both locations are equal.

Radioactive waste solidifying material

International Nuclear Information System (INIS)

Ono, Keiichi; Sakai, Etsuro.

1989-01-01

The solidifying material according to this invention comprises cement material, superfine powder, highly water reducing agent, Al-containing rapid curing material and coagulation controller. As the cement material, various kinds of quickly hardening, super quickly hardening and white portland cement, etc. are usually used. As the superfine powder, those having average grain size smaller by one order than that of the cement material are desirable and silica dusts, etc. by-produced upon preparing silicon, etc. are used. As the highly water reducing agent, surface active agents of high decomposing performance and comprising naphthalene sulfonate, etc. as the main ingredient are used. As the Al-containing rapidly curing material, calcium aluminate, etc. is used in an amount of less than 10 parts by weight based on 100 parts by weight of the powdery body. As the coagulation controller, boric acid etc. usually employed as a retarder is used. This can prevent dissolution or collaption of pellets and reduce the leaching of radioactive material. (T.M.)

Structure and transformation behaviour of a rapidly solidified Al-Y-Ni-Co-Pd alloy

International Nuclear Information System (INIS)

Louzguine-Luzgin, D.V.; Inoue, A.

2005-01-01

An as-solidified structure and transformation behaviour on heating of the rapidly solidified Al-Y-Ni-Co-Pd alloy was studied by X-ray diffractometry (XRD), transmission electron microscopy (TEM), differential scanning and isothermal calorimetries. The Al-Y-Ni-Co-Pd ribbon samples have been produced by the melt spinning technique and heat treated using a differential scanning calorimeter (DSC). The addition of Pd to Al-Y-Ni-Co alloys caused disappearance of the supercooled liquid region as well as the formation of the highly dispersed primary α-Al nanoparticles about 3-7 nm in size homogeneously embedded in the glassy matrix upon solidification. An extremely high density of precipitates of the order of 10 24 m -3 is obtained. These particles start growing at the temperature below a glass-transition temperature. The results presented in this paper indicate that some of so-called 'marginal' glass-formers in as-solidified state are actually not glassy alloys with pre-existed nuclei but crystal-glassy nanocomposites

Improvement in mechanical properties of hypereutectic Al-Si-Cu alloys through sono-solidified

Directory of Open Access Journals (Sweden)

Yoshiki Tsunekawa

2014-07-01

Full Text Available For the wider applications, it is necessary to improve the ductility as well as the strength and wear-resistance of hypereutectic Al-Si-Cu alloys, which are typical light-weight wear-resistant materials. An increase in the amounts of primary silicon particles causes the modified wear-resistance of hypereutectic Al-Si-Cu alloys, but leads to the poor strength and ductility. It is known that dual phase steels composed of hetero-structure have succeeded in bringing contradictory mechanical properties of high strength and ductility concurrently. In order to apply the idea of hetero-structure to hypereutectic Al-Si-Cu alloys for the achievement of high strength and ductility along with wear resistance, ultrasonic irradiation of the molten metal during the solidification, which is called sono-solidification, was carried out from its molten state to just above the eutectic temperature. The sono-solidified Al-17Si-4Cu alloy is composed of hetero-structure, which are, hard primary silicon particles, soft non-equilibrium a -Al phase and the eutectic region. Rheo-casting was performed at just above the eutectic temperature with sono-solidified slurry to shape a disk specimen. After the rheo-casting with modified sonosolidified slurry held for 45 s at 570 篊, the quantitative optical microscope observation exhibits that the microstructure is composed of 18area% of hard primary silicon particles and 57area% of soft a -Al phase. In contrast, there exist only 5 area% of primary silicon particles and no a -Al phase in rheo-cast specimen with normally solidified slurry. Hence the tensile tests of T6 treated rheo-cast specimens with modified sono-solidified slurry exhibit improved strength and 5% of elongation, regardless of having more than 3 times higher amounts of primary silicon particles compared to that of rheo-cast specimen with normally solidified slurry.

West Valley demonstration project: alternative processes for solidifying the high-level wastes

International Nuclear Information System (INIS)

Holton, L.K.; Larson, D.E.; Partain, W.L.; Treat, R.L.

1981-10-01

In 1980, the US Department of Energy (DOE) established the West Valley Solidification Project as the result of legislation passed by the US Congress. The purpose of this project was to carry out a high level nuclear waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The DOE authorized the Pacific Northwest Laboratory (PNL), which is operated by Battelle Memorial Institute, to assess alternative processes for treatment and solidification of the WNYNSC high-level wastes. The Process Alternatives Study is the suject of this report. Two pretreatment approaches and several waste form processes were selected for evaluation in this study. The two waste treatment approaches were the salt/sludge separation process and the combined waste process. Both terminal and interim waste form processes were studied. The terminal waste form processes considered were: borosilicate glass, low-alkali glass, marbles-in-lead matrix, and crystallinolecular potential and molecular dynamics calculations of the effect are yet to be completed. Cous oxide was also investigated. The reaction is first order in nitrite ion, second order in hydrogen ion, and between zero and first order in hydroxylamine monosulfonate, depending on the concentration

Management of high level radioactive waste

International Nuclear Information System (INIS)

Redon, A.; Mamelle, J.; Chambon, M.

1977-01-01

The world wide needs in reprocessing will reach the value of 10.000 t/y of irradiated fuels, in the mid of the 80's. Several countries will have planned, in their nuclear programme, the construction of reprocessing plants with a 1500 t/y capacity, corresponding to 50.000 MWe installed. At such a level, the solidification of the radioactive waste will become imperative. For this reason, all efforts, in France, have been directed towards the realization of industrial plants able of solidifying the fission products as a glassy material. The advantages of this decision, and the reasons for it are presented. The continuing development work, and the conditions and methods of storing the high-level wastes prior to solidification, and of the interim storage (for thermal decay) and the ultimate disposal after solidification are described [fr

Evaluation of Carbonation Effects on Cement-Solidified Contaminated Soil Used in Road Subgrade

Directory of Open Access Journals (Sweden)

Yundong Zhou

2018-01-01

Full Text Available Cement solidification/stabilization is widely used towards contaminated soil since it has a low price and significant improvement for the structural capacity of soil. To increase the usage of the solidified matrix, cement-solidified contaminated soil was used as road subgrade material. In this study, carbonation effect that reflected the durability on strength characteristics of cement-solidified contaminated soil and the settlement of pavement were evaluated through experimental and numerical analysis, respectively. According to results, compressive strengths of specimens with 1% Pb(II under carbonation and standard curing range from 0.44 MPa to 1.17 MPa and 0.14 MPa to 2.67 MPa, respectively. The relatively low strengths were attributed to immobilization of heavy metal, which consumed part of SiO2, Al2O3, and CaO components in the cement or kaolin and reduced the hydration and pozzolanic reaction materials. This phenomenon further decreased the strength of solidified soils. The carbonation depth of 1% Cu(II or Zn(II contaminated soils was 18 mm, which significantly increased with the increase of curing time and contamination concentration. Furthermore, the finite element calculation results showed that surface settlements decreased with the increase of modulus of subgrade and the distance away from the center. At the center, the pavement settlement was proportional to the level of traffic load.

Filling of recovered mining areas using solidifying backfill

Directory of Open Access Journals (Sweden)

Zeman Róbert

2001-12-01

Full Text Available The aim of this article is to explore the possibilities for filling recovered mining areas using solidifying backfill .The article describes the preparation of the backfill (backfill formulation with an eventual application using low quality sands, wastes from treatment plants and ash from power plants etc now to transport it as well as its application in practice. Advantageous and disadvantageous of this method are also mentioned.Several factors must be taken info consideration during the preparation process of the backfill mixture. Firstly, the quantities of each individual component must be constantly regulated. Secondly, the properties of each component must be respected. In addition, the needs of the pipeline transport system and the specific conditions of the recovered area to be filled must also be considered.Hydraulic transport and pneumo-hydraulic pipeline transport are used for handling the backfill. Pumps for transporting the solidifying backfill have to carry out demanding tasks.Due to the physical-mechanical properties of the backfill, only highly powerful pumps can be considered. Piston type pumps such as Abel Simplex and Duplex pumps with capacities of up to 100 m3.h-1 and operating pressures of up to 16 MPa would be suitable.This method has been applied abroad for different purposes. For example, solid backfill was used in the Hamr mine during exploitation of uranium using the room-and-pillar system mining method.In the Ostrava–Karvina Coal field, backfill was used in decontamination work, filling areas in a zone of dangerous deformations and for creating a dividing stratum during thick seam mining.Research info the use of solidifying backfill was also done in the Walsum mine in Germany. The aim of this research was:- to investigate the possibilities of filling a collapsing area in a working face using a solidifying mixture of power plant ash and water,- to verify whether towing pipelines proposed by the DMT corporation would be

Method of solidifying radioactive wastes

International Nuclear Information System (INIS)

Fukazawa, Tetsuo; Ootsuka, Masaharu; Uetake, Naoto; Ozawa, Yoshihiro.

1984-01-01

Purpose: To prepare radioactive solidified wastes excellent in strength, heat resistance, weather-proof, water resistance, dampproof and low-leaching property. Method: A hardening material reactive with alkali silicates to form less soluble salts is used as a hardener for alkali silicates which are solidification filler for the radioactive wastes, and mixed with cement as a water absorbent and water to solidify the radioactive wastes. The hardening agent includes, for example, CaCO 3 , Ca(ClO 4 ) 2 , CaSiF 6 and CaSiO 3 . Further, in order to reduce the water content in the wastes and reduce the gap ratio in the solidification products, the hardener adding rate, cement adding rate and water content are selected adequately. As the result, solidification products can be prepared with no deposition of easily soluble salts to the surface thereof, with extremely low leaching of radioactive nucleides. (Kamimura, M.)

Experimental study on intermediate level radioactive waste processing

International Nuclear Information System (INIS)

Nagakura, Tadashi; Abe, Hirotoshi; Okazawa, Takao; Hattori, Seiichi; Maki, Yasuro

1977-01-01

In the disposal of intermediate level radioactive wastes, multilayer package will be adopted. The multilayer package consists of cement-solidified waste and a container such as a drum - can with concrete liner or a concrete container. So, on the waste to be cement-solidified in such container, experimental study was carried out as follows. (1) Cement-solidification method. (2) Mechanical behaviour of cement-solidified waste. The mechanical behaviour of the containers was studied by the finite element method and experiment, and the function of pressure-balancing valves was also studied. The following data on processing intermediate level radioactive wastes were obtained. (1) In the case of cement-solidified waste, the data to select the suitable solidifying material and the standard mixing proportion were determined. (2) The basic data concerning the uniaxial compressive strength of cement-solidified waste, the mechanical behaviour of cement-solidified waste packed in a drum under high hydrostatic pressure, the shock response of cement-solidified waste at the time of falling and so on were obtained. (3) The pressure-balancing valves worked at about 0.5 Kg/cm 2 pressure difference inside and outside a container, and the deformation of a drum cover was 10 to 13 mm. In case of the pressure difference less than 0,5 Kg/cm 2 , the valves shut, and water flow did occur. (auth.)

Method of solidifying and disposing radioactive waste plastic

International Nuclear Information System (INIS)

Matsuura, Hiroyuki; Yasumura, Keijiro

1981-01-01

Purpose: To solidify radioactive waste as it is with plastic by forming a W/O (Water-in-Oil) emulsion with the radioactive waste and a plastic solidifier, and treating it with a polymerization starting agent, an accelerator, and the like. Method: A predetermined amount of alkaline substance such as sodium hydroxide, triethanol, or the like is added quantitatively to radioactive waste and it is mixed by an agitator. A predetermined amount of solidifier such as unsaturated polyester or the like is added to the mixture and it is further mixed by the agitator to form a stable W/O emulsion. Subsequently, predetermined amounts of polymerization starting agent such as methyl ethyl ketone peroxide and polymerization accelerator such as cobalt naphthenate or the like are added thereto, the mixture is mixed, and is then allowed to stand for at room temperature for the plastic solidification thereof. No reaction occurs after the solidification. (Sekiya, K.)

Safety of geologic disposal of high level radioactive waste

International Nuclear Information System (INIS)

Zaitsu, Tomohisa; Ishiguro, Katsuhiko; Masuda, Sumio

1992-01-01

This article introduces current concepts of geologic disposal of high level radioactive waste and its safety. High level radioactive waste is physically stabilized by solidifying it in a glass form. Characteristics of deep geologic layer are presented from the viewpoint of geologic disposal. Reconstruction of multi-barrier system receives much attention to secure the safety of geologic disposal. It is important to research performance assessment of multi-barrier system for preventing dissolution or transfer of radionuclides into the ground water. Physical and chemical modeling for the performance assessment is outlined in the following terms: (1) chemical property of deep ground water, (2) geochemical modeling of artificial barrier spatial water, (3) hydrology of deep ground water, (4) hydrology of the inside of artificial barrier, and (5) modeling of radionuclide transfer from artificial barrier. (N.K.)

Evaluation of the performance of solidified commercial low-level wastes in an arid climate

International Nuclear Information System (INIS)

Graham, M.J.; Walter, M.B.

1984-01-01

Shallow land burial is being used as a disposal method for commercial low-level waste at waste disposal sites in arid (Hanford, Washington) and humid (Barnwell, South Carolina) climatic regions. A field lysimeter facility has been established at Hanford in which to conduct waste-form leaching tests. The primary objective of this research is to determine typical source terms generated by commercial solidified low-level wastes. The field lysimeter facility consists of 10, 3 M deep by 1.8 M diameter, closed-bottomed lysimeters around a central 4 M deep by 4 M diameter instrument caisson. Commercial cement and dow polymer waste samples were removed from 210 L drums and placed in the 1.8 M diameter lysimeters. Two bitumen samples are planned to be emplaced in the facility this year. The central caisson provides access to the instrumentation in the individual lysimeters and allows selective sampling of the soil and waste forms. Suction candles (ceramic cups) placed around the waste will be used to periodically collect soil water samples for chemical analysis. Meteorological data, moisture content, and soil temperature are being automatically monitored at the facility. Characterization of the soils and waste forms have been partially completed. These data consist of moisture release characteristics, particle size distribution, concentrations and distributions of radionuclides in the waste streams, and concentrations of hydrophilic organic species in one of the waste streams

Small-scale integrated demonstration of high-level radioactive waste processing and vitrification using actual SRP waste

International Nuclear Information System (INIS)

Ferguson, R.B.; Woolsey, G.B.; Galloway, R.M.; Baumgarten, P.M.; Eibling, R.E.

1980-01-01

Experiments have been made to demonstrate the feasibility of immobilizing SRP high-level waste in borosilicate glass. Results to date are encouraging. Equipment performance and processing characteristics for solidifying small batches of actual SRP waste have agreed well with previous experience with small- and large-scale tests synthetic waste, and with theoretical predictions

Analysis of environmental effects from disposal of solidified ICPP high-level wastes

International Nuclear Information System (INIS)

Chipman, N.A.; Simpson, G.G.; Lawroski, H.; Rodger, W.A.; Frendberg, R.L.

1979-01-01

This work is part of a comprehensive study to assess possible environmental impacts from six different options for managing high-level defense wastes generated at the ICPP. Only radiological consequences are considered in this report; population doses to those within 80 km of ICPP were estimated for time periods up to 100 million years. The population dose to future generations from any option is insignificant compared with that from natural background radiation: less than 1 cancer death in 1,000 years compared with 20,000 cancer deaths from natural background radiation. 16 tables

High Pressure Soxhlet Type Leachability testing device and leaching test of simulated high-level waste glass at high temperature

International Nuclear Information System (INIS)

Senoo, Muneaki; Banba, Tsunetaka; Tashiro, Shingo; Shimooka, Kenji; Araki, Kunio

1979-11-01

A High Pressure Soxhlet Type Leachability Testing Device (HIPSOL) was developed to evaluate long-period stability of high-level waste (HLW) solids. For simulated HLW solids, temperature dependency of the leachability was investigated at higher temperatures from 100 0 C to 300 0 C at 80 atm. Leachabilities of cesium and sodium at 295 0 C were 20 and 7 times higher than at 100 0 C, respectively. In the repository, the temperatures around solidified products may be hundred 0 C. It is essential to test them at such elevated temperatures. HIPSOL is also usable for accelerated test to evaluate long-period leaching behavior of HLW products. (author)

Assessment of the radiological protection aspects of disposal of high level waste on the ocean floor

International Nuclear Information System (INIS)

Grimwood, P.D.; Webb, G.A.M.

1976-10-01

This study is a preliminary assessment of the potential radiological consequences of disposal of solidified high-level radioactive waste on the floor of the deep ocean. As an input to the modelling used in the assessment, an arbitrary choice is made to consider the total high-level waste which would be generated by a postulated world nuclear power programme to the year 2000. It is assumed that all this waste, in solidified form, is disposed of on to the floor of the North Atlantic. The body of this report is the modelling of the subsequent release of activity into the water, its dispersion in the ocean and eventual uptake in marine organisms and sediments. The consequent radiation exposure of man is assessed in terms of both individual and collective doses. It is intended that only broad conclusions should be drawn from this study. The objective of the assessment is to highlight those subject areas where more study of information is required before a decision can be reached regarding this method of disposal. No overriding reason connected with the radiological protection considerations has been identified which would preclude the disposal of suitably conditioned high-level waste on the ocean floor. Further evaluation of this disposal option is therefore justified. (author)

Glass-solidification method for high level radioactive liquid waste

International Nuclear Information System (INIS)

Kawamura, Kazuhiro; Kometani, Masayuki; Sasage, Ken-ichi.

1996-01-01

High level liquid wastes are removed with precipitates mainly comprising Mo and Zr, thereafter, the high level liquid wastes are mixed with a glass raw material comprising a composition having a B 2 O 3 /SiO 2 ratio of not less than 0.41, a ZnO/Li 2 O ratio of not less than 1.00, and an Al 2 O 3 /Li 2 O ratio of not less than 2.58, and they are melted and solidified into glass-solidification products. The liquid waste content in the glass-solidification products can be increased up to about 45% by using the glass raw material having such a predetermined composition. In addition, deposition of a yellow phase does not occur, and a leaching rate identical with that in a conventional case can be maintained. (T.M.)

Leaching studies of radionuclides from solidified wastes with thermosetting resin

International Nuclear Information System (INIS)

Suzuki, K.; Kuribayashi, H.; Morimitsu, W.; Ono, I.

1982-01-01

This paper reports on studies of the leachability of Co-60 and Cs-137 from simulated LWR radwastes solidified with thermosetting resin and evaluates the effects of chemical fixation on leachability. It is concluded that insolubilization by a nickel-ferrocyanide compound offers an effective chemical fixation of these radionuclides and is a recommended pretreating method for radwastes that are to be solidified. 2 figures

Evaluation of the performance of solidified commercial low-level wastes in an arid climate

International Nuclear Information System (INIS)

Graham, M.J.; Walter, M.B.

1984-09-01

Shallow land burial is being used as a disposal method for commercial low-level waste at waste disposal sites in arid (Hanford site near Richland, Washington) and humid (Barnwell, South Carolina) climatic regions. A field lysimeter facility has been established at the Hanford site in which to conduct waste-form leaching tests. The primary objective of this research is to determine typical source terms generated by commercial solidified low-level wastes. The field lysimeter facility consists of ten 3-m-deep by 1.8-m-diameter, closed-bottom lysimeters around a central instrument caisson, 4 m in diameter. Commercial cement and vinyl ester-styrene waste samples were removed from 210-L drums and placed in the 1.8-m-diameter lysimeters. Two bitumen samples are planned to be emplaced in the facility in 1984. The central caisson provides access to the instrumentation in the individual lysimeters and allows selective sampling of the soil and waste forms. Suction candles (ceramic cups) placed around the waste will be used to periodically collect soil water samples for chemical analysis. Meteorological data, moisture content, and soil temperature are automatically monitored at the facility. Characterization of the soils and waste forms have been partially completed. These data consist of moisture release characteristics, particle size distribution, concentrations and distributions of radionuclides in the waste forms, concentrations of radionuclides in the waste streams, and concentrations of hydrophilic organic species in one of the waste steams. 8 references, 3 figures, 5 tables

Characteristics of high-level radioactive waste forms for their disposal

International Nuclear Information System (INIS)

Kim, Seung Soo; Chun, Kwan Sik; Kang, Chul Hyung

2000-12-01

In order to develop a deep geological repository for a high-level radioactive waste coming from reprocessing of spent nuclear fuels discharged from our domestic nuclear power plants, the the required characteristics of waste form are dependent upon a solidifying medium and the amount of waste loading in the medium. And so, by the comparative analysis of the characteristics of various waste forms developed up to the present, a suitable medium is recommended.The overall characteristics of the latter is much better than those of the former, but the change of the properties due to an amorphysation by radiation exposure and its thermal expansion has not been clearly identified yet. And its process has not been commercialized. However, the overall properties of the borosilicate glass waste forms are acceptable for their disposal, their production cost is reasonable and their processes have already been commercialized. And plenty informations of their characteristics and operational experiences have been accumulated. Consequently, it is recommended that a suitable medium solidifying the HLW is a borosilicate glass and its composition for the identification of a reference waste form would be based on the glass frit of R7T7

The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants.

Science.gov (United States)

Park, S D; Kim, J S; Han, S H; Ha, Y K; Song, K S; Jee, K Y

2009-09-01

In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of (129)I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The (129)I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67+/-3% and 5.43+/-0.53 g, 70+/-7% and 10.40+/-1.60 g, respectively. And the minimum detectable activity (MDA) of (129)I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, (129)I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher (129)I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

The measurement of 129I for the cement and the paraffin solidified low and intermediate level wastes (LILWs), spent resin or evaporated bottom from the pressurized water reactor (PWR) nuclear power plants

International Nuclear Information System (INIS)

Park, S.D.; Kim, J.S.; Han, S.H.; Ha, Y.K.; Song, K.S.; Jee, K.Y.

2009-01-01

In this paper a relatively simple and low cost analysis procedure to apply to a routine analysis of 129 I in low and intermediate level radioactive wastes (LILWs), cement and paraffin solidified evaporated bottom and spent resin, which are produced from nuclear power plants (NPPs), pressurized water reactors (PWR), is presented. The 129 I is separated from other nuclides in LILWs using an anion exchange adsorption and solvent extraction by controlling the oxidation and reduction state and is then precipitated as silver iodide for counting the beta activity with a low background gas proportional counter (GPC). The counting efficiency of GPC was varied from 4% to 8% and it was reversely proportional to the weight of AgI by a self absorption of the beta activity. Compared to a higher pH, the chemical recovery of iodide as AgI was lowered at pH 4. It was found that the chemical recovery of iodide for the cement powder showed a lower trend by increasing the cement powder weight, but it was not affected for the paraffin sample. In this experiment, the overall chemical recovery yield of the cement and paraffin solidified LILW samples and the average weight of them were 67±3% and 5.43±0.53 g, 70±7% and 10.40±1.60 g, respectively. And the minimum detectable activity (MDA) of 129 I for the cement and paraffin solidified LILW samples was calculated as 0.070 and 0.036 Bq/g, respectively. Among the analyzed cement solidified LILW samples, 129 I activity concentration of four samples was slightly higher than the MDA and their ranges were 0.076-0.114 Bq/g. Also of the analyzed paraffin solidified LILW samples, five samples contained a little higher 129 I activity concentration than the MDA and their ranges were 0.036-0.107 Bq/g.

High-level-waste containment for a thousand years: unique technical and research problems

International Nuclear Information System (INIS)

Davis, M.S.

1982-01-01

In the United States the present policy for disposal of high level nuclear wastes is focused on isolation of solidified wastes in a mined geologic repository. Safe isolation is to be achieved by utilizing both natural and man-made barriers which will act in concert to assure the overall conservative performance of the disposal system. The incorporation of predictable man-made barriers into the waste disposal strategy has generated some new and unique problems for the scientific community

Micro and Macro Segregation in Alloys Solidifying with Equiaxed Morphology

Science.gov (United States)

Stefanescu, Doru M.; Curreri, Peter A.; Leon-Torres, Jose; Sen, Subhayu

1996-01-01

To understand macro segregation formation in Al-Cu alloys, experiments were run under terrestrial gravity (1g) and under low gravity during parabolic flights (10(exp -2) g). Alloys of two different compositions (2% and 5% Cu) were solidified at two different cooling rates. Systematic microscopic and SEM observations produced microstructural and segregation maps for all samples. These maps may be used as benchmark experiments for validation of microstructure evolution and segregation models. As expected, the macro segregation maps are very complex. When segregation was measured along the central axis of the sample, the highest macro segregation for samples solidified at 1g was obtained for the lowest cooling rate. This behavior is attributed to the longer time available for natural convection and shrinkage flow to affect solute redistribution. In samples solidified under low-g, the highest macro-segregation was obtained at the highest cooling rate. In general, low-gravity solidification resulted in less segregation. To explain the experimental findings, an analytical (Flemings-Nereo) and a numerical model were used. For the numerical model, the continuum formulation was employed to describe the macroscopic transports of mass, energy, and momentum, associated with the microscopic transport phenomena, for a two-phase system. The model proposed considers that liquid flow is driven by thermal and solutal buoyancy, and by solidification shrinkage. The Flemings-Nereo model explains well macro segregation in the initial stages of low-gravity segregation. The numerical model can describe the complex macro segregation pattern and the differences between low- and high-gravity solidification.

Microstructure of rapidly solidified materials

Science.gov (United States)

Jones, H.

1984-07-01

The basic features of rapidly solidified microstructures are described and differences arising from alternative processing strategies are discussed. The possibility of achieving substantial undercooling prior to solidification in processes such as quench atomization and chill block melt spinning can give rise to striking microstructural transitions even when external heat extraction is nominally Newtonian. The increased opportunity in laser and electron beam surface melting for epitaxial growth on the parent solid at an accelerating rate, however, does not exclude the formation of nonequilibrium phases since the required undercooling can be locally attained at the solidification front which is itself advancing at a sufficiently high velocity. The effects of fluid flow indicated particularly in melt spinning and surface melting are additional to the transformational and heat flow considerations that form the present basis for interpretation of such microstructural effects.

Study on dissolution behavior of molten solidified waste

International Nuclear Information System (INIS)

Mizuno, Tsuyoshi; Maeda, Toshikatsu

2005-01-01

Radioactive molten solidified waste (slag) has been generated by melting non-metallic low-level radioactive wastes (LLW). Slag is expected to immobilize radionuclides in the waste repository. The chemical durability of slag is an important factor for the safety assessment of the disposal in that the durability provides the source term in the assessment. Since a chemical characteristic of slag is similar to that of glass, the general information on the chemical durability of slag might be provided from previous studies on nuclear waste glass. We have investigated effects of chemical compositions of slag and alkaline environments of repository on the chemical durability of slag. (author)

Cation distributions on rapidly solidified cobalt ferrite

Science.gov (United States)

De Guire, Mark R.; Kalonji, Gretchen; O'Handley, Robert C.

1990-01-01

The cation distributions in two rapidly solidified cobalt ferrites have been determined using Moessbauer spectroscopy at 4.2 K in an 8-T magnetic field. The samples were obtained by gas atomization of a Co0-Fe2O3-P2O5 melt. The degree of cation disorder in both cases was greater than is obtainable by cooling unmelted cobalt ferrite. The more rapidly cooled sample exhibited a smaller departure from the equilibrium cation distribution than did the more slowly cooled sample. This result is explained on the basis of two competing effects of rapid solidification: high cooling rate of the solid, and large undercooling.

The principal radionuclides in high level radioactive waste management

International Nuclear Information System (INIS)

Mulyanto

1998-01-01

The principal radionuclides in high level radioactive waste management. The selection of the principal radionuclides in the high level waste (HLW) management was developed in order to improve the disposal scenario of HLW. In this study the unified criteria for selection of the principal radionuclides were proposed as; (1) the value of hazard index estimated by annual limit of intake (ALI) for long-term tendency,(2) the relative dose factor related to adsorbed migration rate transferred by ground water, and (3) heat generation in the repository. From this study it can be concluded that the principal radionuclides in the HLW management were minor actinide (MA=Np, Am, Cm, etc), Tc, I, Cs and Sr, based on the unified basic criteria introduced in this study. The remaining short-lived fission product (SLFPs), after the selected nuclides are removed, should be immobilized and solidified in a glass matrix. Potential risk due to the remaining SLFPs can be lower than that of uranium ore after about 300 year. (author)

R and D Activities on high-level nuclear waste management

International Nuclear Information System (INIS)

Watanabe, Shosuke

1985-01-01

High-level liquid waste (HLLW) at Tokai Reprocessing Plant has been generated from reprocessing of spent fuels from the light water reactors, and successfully managed since 1977. At the time of 1984, about 154m 3 of HLLW from 170 tons of spent fuels were stored in three high-integrity stainless steel tanks (90m 3 for each) as a nitric acid aqueous solution. The HLLW arises mainly from the first cycle solvent extraction phase. Alkaline solution to scrub the extraction solvent is another source of HLLW. The Advisory Committee on Radioactive Waste Management reported the concept on disposal of high-level waste (HLW) in Japan in 1980 report, that the waste be solidified into borosilicate glass and then be disposed in deep geologic formation so as to minimize the influence of the waste on human environment, with the aid of multibarrier system which is the combination of natural barrier and engineered barrier

Rapidly solidified aluminium for optical applications

NARCIS (Netherlands)

Gubbels, G.P.H.; Venrooy, B.W.H. van; Bosch, A.J.; Senden, R.

2008-01-01

This paper present the results of a diamond turning study of a rapidly solidified aluminium 6061 alloy grade, known as RSA6061. It is shown that this small grain material can be diamond turned to smaller roughness values than standard AA6061 aluminium grades. Also, the results are nearly as good as

Microstructural development in a rapidly solidified Al-Fe-V-Si alloy

International Nuclear Information System (INIS)

Park, W.J.; Baek, E.R.; Lee, Sunghak; Kim, N.J.

1991-01-01

TEM is used to investigate microstructural development in a rapidly solidified Al-Fe-V-Si alloy. The as-cast microstructure of a rapidly solidified Al-Fe-V-Si alloy was found to vary depending on casting conditions and also through the thickness of ribbon. For completely Zone A ribbon, intercellular phase consists of a microquasi-crystalline phase, while for the Zone A and Zone B mixed ribbon, it consists of a silicide phase. In either case, formation of globular particles of a cluster microquasi-crystalline phase is observed near the air side of the ribbon. Annealing study shows significant differences in the final microstructure depending on the initial status of the ribbon. Completely Zone A ribbon, whose microstructure is composed of a microquasi-crystalline phase, results in a very coarse microstructure after annealing as compared to the Zone A and Zone B mixed ribbon. This result has important implications for the development of high-performance elevated-temperature Al alloys. 12 refs

Hydration products and mechanical properties of hydroceramics solidified waste for simulated Non-alpha low and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Wang Jin; Hong Ming; Wang Junxia; Li Yuxiang; Teng Yuancheng; Wu Xiuling

2011-01-01

In this paper, simulated non-alpha low and intermediate level radioactive wastes was handled as curing object and that of 'alkali-slag-coal fly ash-metakaolin' hydroceramics waste forms were prepared by hydrothermal synthesis method. The hydration products were analyzed by X ray diffraction. The composition of hydrates and the compressive strength of waste forms were determined and measured. The results indicate that the main crystalline phase of hydration products were analcite when the temperature was 150 to 180 degree C and the salt content ratio was 0.10 to 0.30. Analcite diffraction peaks in hydration products is increasing when the temperature was raised and the reaction time prolonged. Strength test results show that the solidified waste forms have superior compressive strength. The compressive strength gradually decreased with the increase in salt content ratio in waste forms. (authors)

Survey of matrix materials for solidified radioactive high-level waste

Energy Technology Data Exchange (ETDEWEB)

Gurwell, W.E.

1981-09-01

Pacific Northwest Laboratory (PNL) has been investigating advanced waste forms, including matrix waste forms, that may provide a very high degree of stability under the most severe repository conditions. The purpose of this study was to recommend practical matrix materials for future development that most enhance the stability of the matrix waste forms. The functions of the matrix were reviewed. Desirable matrix material properties were discussed and listed relative to the matrix functions. Potential matrix materials were discussed and recommendations were made for future matrix development. The matrix mechanically contains waste cores, reduces waste form temperatures, and is capable of providing a high-quality barrier to leach waters. High-quality barrier matrices that separate and individually encapsulate the waste cores are fabricated by powder fabrication methods, such as sintering, hot pressing, and hot isostatic pressing. Viable barrier materials are impermeable, extremely corrosion resistant, and mechanically strong. Three material classes potentially satisfy the requirements for a barrier matrix and are recommended for development: titanium, glass, and graphite. Polymers appear to be marginally adequate, and a more thorough engineering assessment of their potential should be made.

Survey of matrix materials for solidified radioactive high-level waste

International Nuclear Information System (INIS)

Gurwell, W.E.

1981-09-01

Pacific Northwest Laboratory (PNL) has been investigating advanced waste forms, including matrix waste forms, that may provide a very high degree of stability under the most severe repository conditions. The purpose of this study was to recommend practical matrix materials for future development that most enhance the stability of the matrix waste forms. The functions of the matrix were reviewed. Desirable matrix material properties were discussed and listed relative to the matrix functions. Potential matrix materials were discussed and recommendations were made for future matrix development. The matrix mechanically contains waste cores, reduces waste form temperatures, and is capable of providing a high-quality barrier to leach waters. High-quality barrier matrices that separate and individually encapsulate the waste cores are fabricated by powder fabrication methods, such as sintering, hot pressing, and hot isostatic pressing. Viable barrier materials are impermeable, extremely corrosion resistant, and mechanically strong. Three material classes potentially satisfy the requirements for a barrier matrix and are recommended for development: titanium, glass, and graphite. Polymers appear to be marginally adequate, and a more thorough engineering assessment of their potential should be made

Analysis of Light Gathering Abilities of Dynamically Solidified Micro-lenses, and Their Implementation to Improve Sensitivity of Fluorescent PCR Micro-detectors.

Science.gov (United States)

Wu, Jian; Guo, Wei; Wang, Chunyan; Yu, Kuanxin; Chen, Tao; Li, Yinghui

2015-06-01

Fluorescent polymerase chain reaction (PCR) is becoming the preferred method of quantitative analysis due to its high specificity and sensitivity. We propose to use a new kind of micro-lens, dynamically solidified with optic glue, to improve the sensitivity of fluorescent PCR micro-detector. We developed light ray track equations for these lenses and used them to calculate relative light intensity distribution curve for stimulation lenses and illumination point probability distribution curve for detection lenses. We manufactured dynamically solidified micro-lenses using optic glue NOA61, and measured their light gathering ability. Lenses with radius/thickness (R/H) ratio of 4 reached light focusing ratio of 3.85 (stimulation lens) and photon collection efficiency of 0.86 (detection lens). We then used dynamically solidified lenses in PCR fluorescence micro-detector and analyzed their effect on the detector sensitivity. We showed that the use of dynamically solidified micro-lenses with R/H = 4 resulted in over 4.4-fold increased sensitivity of the detector.

Vessel for solidifying water-impermeable radioactive waste

International Nuclear Information System (INIS)

Kiuchi, Yoshimasa; Tamada, Shin; Suzuki, Yasushi.

1993-01-01

A blend prepared by admixing silica sand, alumina powder or glass fiber, as aggregates, to epoxy resin elastic adhesives is coated on an inner surface of a steel drum can or an inner surface of a concrete vessel at a thickness of greater than 1mm followed by hardening. The addition amount of the silica sand, alumina powder or glass fiber is determined as 20 to 40% by weight, 30 to 60% by weight or 5 to 15% by weight respectively. A lid having a hole for injecting fillers is previously bonded to a container for use in solidifying radioactive materials. The strength of the coating layer is increased and a coating performance and an adhesion force are improved by admixing the aggregates, to provide a satisfactory water-impermeability. The container for use in solidifying radioactive wastes having a coating layer with an advantage of the elastic resin adhesives, strong strength and adhesion and being excellent in the water-impermeability can be obtained relatively economically. (N.H.)

Engineering-scale vitrification of commercial high-level waste

International Nuclear Information System (INIS)

Bonner, W.F.; Bjorklund, W.J.; Hanson, M.S.; Knowlton, D.E.

1980-04-01

To date, technology for immobilizing commercial high-level waste (HLW) has been extensively developed, and two major demonstration projects have been completed, the Waste Solidification Engineering Prototypes (WSEP) Program and the Nuclear Waste Vitrification Project (NWVP). The feasibility of radioactive waste solidification was demonstrated in the WSEP program between 1966 and 1970 (McElroy et al. 1972) using simulated power-reactor waste composed of nonradioactive chemicals and HLW from spent, Hanford reactor fuel. Thirty-three engineering-scale canisters of solidified HLW were produced during the operations. In early 79, the NWVP demonstrated the vitrification of HLW from the processing of actual commercial nuclear fuel. This program consisted of two parts, (1) waste preparation and (2) vitrification by spray calcination and in-can melting. This report presents results from the NWVP

High damping Al-Fe-Mo-Si/Zn-Al composites produced by rapidly solidified powder metallurgy process

International Nuclear Information System (INIS)

Li, P.Y.; Dai, S.L.; Chai, S.C.; Li, Y.R.

2000-01-01

The metallic materials commonly used in aircraft and aerospace fields, such as aluminum and titanium alloys, steels, etc., show extremely low damping capacity (usually of the order of or less than 10 -3 ). Thus, some problems related to vibration may emerge and influence the reliability, safety and life of airplanes, satellites, etc. It has been reported that almost two thirds of errors for rockets and satellites are related to vibration and noise. One effective way to solve these vibration-related problems is to adopt high damping metallic materials. Conventional high damping alloys exhibit damping capacity above 10 -2 , however, their densities are usually great than 5 x 10 3 kg m -3 , or their strengths are less than 200 MPa (for alloys based on dislocation damping), making them impossible to be applied to aircraft and aerospace areas. Recently, some low-density high-damping metal/metal composites based on aluminum and high damping alloys have been developed in Beijing Institute of Aeronautical Materials (BIAM) by the rapidly solidified power metallurgy process. This paper aims to report the properties of the composites based on a high temperature Al-Fe-Mo-Si alloy and a high damping Zn-Al alloy, and compare them with that of 2618-T61 alloy produced by the ingot metallurgy process

Evaluation of forms for the immobilization of high-level and transuranic wastes

International Nuclear Information System (INIS)

Schuman, R.P.; Cox, N.D.; Gibson, G.W.; Kelsey, P.V. Jr.

1982-08-01

A figure-of-merit (FOM) analysis has been made of a number of waste forms for solidifying both defense and commercial high-level reprocessing waste (HLW) and transuranic (TRU) wastes. The evaluation includes iron-enriched basalt (IEB), a fusion-produced glass-ceramic, which has not been included in other assessments. For HLW, concrete receives the highest FOM, but may not meet regulatory requirements; IEB and glass are the best choices of the materials that should easily meet regulatory requirements. Concrete waste forms are the best choice for TRU wastes, with IEB a close contender. 116 references, 3 figures, 112 tables

Precipitation in as-solidified undercooled Ni-Si hypoeutectic alloy: Effect of non-equilibrium solidification

Energy Technology Data Exchange (ETDEWEB)

Fan Kai [State Key Laboratory of Solidification Processing, Northwestern Polytechnical University, Xi' an, Shaanxi 710072 (China); Liu Feng, E-mail: liufeng@nwpu.edu.cn [State Key Laboratory of Solidification Processing, Northwestern Polytechnical University, Xi' an, Shaanxi 710072 (China); Yang Gencang; Zhou Yaohe [State Key Laboratory of Solidification Processing, Northwestern Polytechnical University, Xi' an, Shaanxi 710072 (China)

2011-08-25

Highlights: {yields} The solid solubility of Si atom in {alpha}-Ni matrix increased with undercooling in the as-solidified sample. {yields} The effect of non-equilibrium solidification on precipitation has been theoretically described. {yields} The nucleation density, the real-time particle size and the precipitation rate are all increased upon annealing. {yields} The precipitate process can be artificially controlled by modifying the initial melt undercooling and the annealing time. - Abstract: Applying glass fluxing and cyclic superheating, high undercooling up to {approx}350 K was achieved for Ni-Si hypoeutectic alloy melt. By isothermally annealing the as-solidified alloy subjected to different undercoolings, precipitation behavior of Ni{sub 3}Si particle, at 973 K, was systematically studied. It was found that, the nucleation density and the real-time particle size, as well as the precipitation rate, were all increased, provided the sample was solidified subjected to higher undercooling. This was ascribed mainly to the increased solid solubility of Si atom in {alpha}-Ni matrix upon non-equilibrium solidification. On this basis, the non-equilibrium dendrite growth upon solidification and the soft impingement prevailing upon solid-state precipitation have been quantitatively connected. As such, the effect of liquid/solid transformation on subsequent precipitation was described.

Precipitation in as-solidified undercooled Ni-Si hypoeutectic alloy: Effect of non-equilibrium solidification

International Nuclear Information System (INIS)

Fan Kai; Liu Feng; Yang Gencang; Zhou Yaohe

2011-01-01

Highlights: → The solid solubility of Si atom in α-Ni matrix increased with undercooling in the as-solidified sample. → The effect of non-equilibrium solidification on precipitation has been theoretically described. → The nucleation density, the real-time particle size and the precipitation rate are all increased upon annealing. → The precipitate process can be artificially controlled by modifying the initial melt undercooling and the annealing time. - Abstract: Applying glass fluxing and cyclic superheating, high undercooling up to ∼350 K was achieved for Ni-Si hypoeutectic alloy melt. By isothermally annealing the as-solidified alloy subjected to different undercoolings, precipitation behavior of Ni 3 Si particle, at 973 K, was systematically studied. It was found that, the nucleation density and the real-time particle size, as well as the precipitation rate, were all increased, provided the sample was solidified subjected to higher undercooling. This was ascribed mainly to the increased solid solubility of Si atom in α-Ni matrix upon non-equilibrium solidification. On this basis, the non-equilibrium dendrite growth upon solidification and the soft impingement prevailing upon solid-state precipitation have been quantitatively connected. As such, the effect of liquid/solid transformation on subsequent precipitation was described.

Characterization of aluminium alloys rapidly solidified

International Nuclear Information System (INIS)

Monteiro, W.A.

1988-01-01

This paper discussed the investigation of the microstructural and mechanical properties of the aluminium alloys (3003; 7050; Al-9% Mg) rapidly solidified by melt spinning process (cooling rate 10 4 - 10 6 K/s). The rapidly solidification process of the studied aluminium alloys brought a microcrystallinity, a minimum presence of coarse precipitation and, also, better mechanical properties of them comparing to the same alloys using ingot process. (author) [pt

Site Simulation of Solidified Peat: Lab Monitoring

Science.gov (United States)

Durahim, N. H. Ab; Rahman, J. Abd; Tajuddin, S. F. Mohd; Mohamed, R. M. S. R.; Al-Gheethi, A. A.; Kassim, A. H. Mohd

2018-04-01

In the present research, the solidified peat on site simulation is conducted to obtain soil leaching from soil column study. Few raw materials used in testing such as Ordinary Portland Cement (OPC), Fly ash (FA) and bottom ash (BA) which containing in solidified peat (SP), fertilizer (F), and rainwater (RW) are also admixed in soil column in order to assess their effects. This research was conducted in two conditions which dry and wet condition. Distilled water used to represent rainfall during flushing process while rainwater used to gain leaching during dry and wet condition. The first testing made after leaching process done was Moisture Content (MC). Secondly, Unconfined Compressive Strength (UCS) will be conducted on SP to know the ability of SP strength. These MC and UCS were made before and after SP were applied in soil column. Hence, the both results were compared to see the reliability occur on SP. All leachate samples were tested using Absorption Atomic Spectroscopy (AAS), Ion Chromatography (IC) and Inductively-Coupled Plasma Spectrophotometry (ICP-MS) testing to know the anion and cation present in it.

Microstructure formation and in situ phase identification from undercooled Co-61.8 at.% Si melts solidified on an electromagnetic levitator and an electrostatic levitator

Energy Technology Data Exchange (ETDEWEB)

Li Mingjun [National Institute of Advanced Industrial Science and Technology (AIST), Materials Research Institute for Sustainable Development, 2266-98 Shimo-Shidami, Moriyama, Nagoya, Aichi 463-8560 (Japan); Japan Aerospace Exploration Agency (JAXA), Institute of Space and Astronautical Science (ISAS), Tsukuba Space Centre, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan)], E-mail: li.mingjun@aist.go.jp; Nagashio, Kosuke [Japan Aerospace Exploration Agency (JAXA), Institute of Space and Astronautical Science (ISAS), Sagamihara Campus, 3-1-1 Yoshinodai, Sagamihara, Kanagawa 229-8510 (Japan); Ishikawa, Takehiko [Japan Aerospace Exploration Agency (JAXA), Institute of Space and Astronautical Science (ISAS), Tsukuba Space Centre, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan); Mizuno, Akitoshi; Adachi, Masayoshi; Watanabe, Masahito [Department of Physics, Gakushuin University, 1-5-1 Mejiro, Toshima, Tokyo 171-8588 (Japan); Yoda, Shinichi [Japan Aerospace Exploration Agency (JAXA), Institute of Space and Astronautical Science (ISAS), Tsukuba Space Centre, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan); Kuribayashi, Kazuhiko [Japan Aerospace Exploration Agency (JAXA), Institute of Space and Astronautical Science (ISAS), Sagamihara Campus, 3-1-1 Yoshinodai, Sagamihara, Kanagawa 229-8510 (Japan); Katayama, Yoshinori [Japan Atomic Energy Agency (JAEA), 1-1-1 Kouto, Mikazuki, Sayo, Hyogo 679-5148 (Japan)

2008-06-15

Co-61.8 at.% Si (CoSi-CoSi{sub 2}) eutectic alloys were solidified on an electromagnetic levitator (EML) and an electrostatic levitator (ESL) at different undercooling levels. The results indicated that there is only a single recalescence event at low undercooling with the CoSi intermetallic compound as primary phase, which is independent of processing facilities, on either an EML or an ESL. The microstructure, however, is strongly dependent on the processing facility. The interior melt flow behavior in the sphere solidified at the EML differs substantially from that at the ESL, thus yielding different microstructures. On high undercooling, double recalescence takes place regardless of levitation condition. In situ X-ray diffraction of alloys solidified on the EML demonstrates that the CoSi{sub 2} compound becomes the primary phase upon the first recalescence, and the CoSi intermetallic phase crystallizes during the second recalescence. In addition to phase identification, real-time diffraction patterns can also provide additional evidence of the fragmentation of the primary phase and the ripening feature in the subsequent cooling process in the semisolid state. The phase competition between the CoSi and CoSi{sub 2} compounds is discussed when considering the nucleation barrier. The low interfacial energy of the CoSi{sub 2} phase favors a preferential nucleation event over the CoSi phase, which also plays a critical role in non-reciprocity nucleation and thus yields a double recalescence profile at high undercooling.

Hardness and microstructural characteristics of rapidly solidified Al-8-16 wt.%Si alloys

International Nuclear Information System (INIS)

Uzun, O.; Karaaslan, T.; Gogebakan, M.; Keskin, M.

2004-01-01

Al-Si alloys with nominal composition of Al-8 wt.%Si, Al-12 wt.%Si, and Al-16 wt.%Si were rapidly solidified by using melt-spinning technique to examine the influence of the cooling rate/conditions on microstructure and mechanical properties. The microstructures of the rapidly solidified ribbons and ingot samples were investigated by the optical microscopy, electron microscopy and X-ray diffraction (XRD) techniques. The results showed that the structures of all melt-spun ribbons were completely composed of finely dispersed α-Al and eutectic Si phase, and primary silicon was not observed. The XRD analysis indicated that the solubility of Si in the α-Al matrix was greatly increased with rapid solidification. Additionally, mechanical properties of both conventionally cast (ingot) and melt-spun ribbons were examined by using Vickers indenter for one applied load (0.098 N). The hardness values of the melt-spun ribbons were about three times higher than those of ingot counterparts. The high hardness of the rapidly solidified state can be attributed to the supersaturated solid solutions. Besides, hardness values with different applied loads were measured for melt-spun ribbons. The results indicated that Vickers hardness values (H v ) of the ribbons depended on the applied load. Applying the concept of Hays-Kendall, the load independent hardness values were calculated as 694.0, 982.8 and 1186.8 MN/m 2 for Al-8 wt.%Si, Al-12 wt.%Si and Al-16 wt.%Si, respectively

Method of solidifying powderous wastes

International Nuclear Information System (INIS)

Kakimoto, Akira; Miyake, Takashi; Sato, Shuichi; Inagaki, Yuzo.

1985-01-01

Purpose: To improve the properties of solidification products, in the case of solidifying powderous wastes with thermosetting resins. Method. A solvent for the solution of the thermosetting resin is admixed with the powderous wastes into a paste-like form prior to adding the resin to the wastes, which are then mixed with the resin solution. As the result, those solidification products having the specific gravity and the compression strength more excellent than those of the conventional ones, and much higher than the reference values can be obtained. (Kamimura, M.)

Propertis of solidified radioactive wastes from commercial LWRs

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1978-01-01

A study has been performed to characterize the properties of solidified radioactive wastes generated in the liquid radwaste treatment systems at LWRs. The properties which have been studied are those which are pertinent in defining the relative potential for the release of radionuclides to the environment as well as others relating to the evaluation of various solidification agents on an economic and feasibility basis. The use of standard testing procedures in measuring these properties allows an intercomparison of respective properties between various types of solidified waste forms. The leachability, mechanical properties, thermal stability, radiation stability, and thermal properties of hydraulic cement, ureaformaldehyde, bitumen, and addition type polymer waste forms have been measured. In addition, the chemical sensitivity, volumetric efficiency and radiation shielding characteristics of these waste forms have been studied. Emphasis in this paper is placed on the results of studies concerning chemical compatibility of solidification agents with specific waste streams, volumetric efficiency, free standing water, and leachability

Research and development on the melting test of low-level radioactive miscellaneous solid waste

International Nuclear Information System (INIS)

Nakashio, Nobuyuki; Hoshi, Akiko; Kameo, Yutaka; Nakashima, Mikio

2007-02-01

The Nuclear Science Research Institute of the Japan Atomic Energy Agency constructed the Advanced Volume Reduction Facilities (AVRF) in February 2003 for treatment of low-level radioactive miscellaneous solid waste (LLW). The waste volume reduction is carried out by a high-compaction process or melting processes in the AVRF. In advance of operating the melting process in the AVRF, melting tests of simulated LLW with RI tracers ( 60 Co, 137 Cs and 152 Eu) have been conducted by using the plasma melter in pilot scale. Viscosity of molten waste, chemical composition and physical properties of solidified products and distribution of the tracers in each product were investigated in various melting conditions. It was confirmed that the viscosity of molten waste was able to be controlled by adjusting chemical composition of molten waste. The RI tracer were almost uniformly distributed in the solidified products. The retention of 137 Cs depended on the basicity (CaO/SiO 2 ) of the solidified products. The solidified product possessed satisfactory compressive strength. In the case of basicity less than 0.8, the leachability of RI tracers from the solidified products was less than or equal to that of a high-level vitrified waste. In this review, experimental results of the melting tests were discussed in order to contribute to actual treatment of LLW in the AVRF. (author)

Porous glass matrix method for encapsulating high-level nuclear wastes

International Nuclear Information System (INIS)

Macedo, P.B.; Tran, D.C.; Simmons, J.H.; Saleh, M.; Barkatt, A.; Simmons, C.J.; Lagakos, N.; DeWitt, E.

1979-01-01

A novel process which uses solidified porous high-silica glass powder to fixate radioactive high-level wastes is described. The process yields cylinders consisting of a core of high-silica glass containing the waste elements in its structure and a protective layer also of high-silica glass completely free of waste elements. The process can be applied to waste streams containing 0 to 100% solids. The core region exhibits a higher coefficient of thermal expansion and a lower glass transition temperature than the outer protective layer. This leads to mechanical strengthening of the glass and good resistance to stress corrosion by the development of a high residual compressive stress on the surface of the sample. Both the core and the protective layer exhibit extremely high chemical durability and offer an effective fixation of the radioactive waste elements, including 239 Pu and 99 Tc which have long half-lives, for calculated periods of more than 1 million years, when temperatures are not allowed to rise above 100 0 C

Method and apparatus for solidifying radioactive waste

International Nuclear Information System (INIS)

Kadota, Hiroko; Kikuchi, Makoto; Tsuchiya, Hiroyuki; Tamada, Shin.

1989-01-01

The present invention concerns a method of solidifying radioactive wastes that generate heat with water curing solidifying material and the object there of is suppress the effect of heat generation of the wastes given on the solidification material. That is, it is a feature of the invention to inject water content contained in the water curable solidification material in the form of ice into the wastes. Thus, since the water content in the water curable solidification material is ice, the solidification products can be obtained by way of the following three steps: (1) ice is dissolved into water, (2) solid content of the solidification material is dissolved into water, and(3) curing reaction of the solidification material is started. Acccordingly, since the heat generated from the wastes contributes as heat of reaction when ice is dissolved into water till the solidification material has been completely filled, promotion for the curing reaction causing problems so far can be suppressed to enable easy filling. Then, after the completion of the filling of the solidification material, the heat of the wastes has an effect of promoting the second and the third steps described above to accelerate the curing reaction. (K.M.)

Method and device for solidifying radioactive waste

International Nuclear Information System (INIS)

Hayashi, Tadamasa.

1981-01-01

Purpose: To solidify radioactive waste without producing radioactive dusts by always heating and evaporating the water from liquid radioactive waste in a mixture of liquid plastic and exhausting the molten mixture of the waste residue and the plastic material. Constitution: Liquid plastic material in a tank cooled to prevent polymerization or changes of its properties is continuously supplied to the top of a heating and mixing evaporator by a constant supply pump. After the heat transfer surface of the evaporator is covered with the plastic material, radioactive waste in the tank is supplied to the evaporator via the constant supply pump. The waste is abruptly mixed with the plastic material by an agitating rotor, heated by a heater, and the evaporated water is fed to a condenser. An anhydrous molten mixture is continuously exhausted from the bottom of the evaporator into a mixture cooler, a polymerizing agent and catalyst are introduced thereinto from a polymerizing agent tank and a catalyst tank, inhibitor is introduced thereinto from a polymerization inhibitor tank as required, and is filled with the mixture a solidifying container while it is cooled for its polymerization and solidification. (Yoshino, Y.)

Effect of drying-wetting cycles on leaching behavior of cement solidified lead-contaminated soil.

Science.gov (United States)

Li, Jiang-Shan; Xue, Qiang; Wang, Ping; Li, Zhen-Ze; Liu, Lei

2014-12-01

Lead contaminated soil was treated by different concentration of ordinary Portland cement (OPC). Solidified cylindrical samples were dried at 40°C in oven for 48 h subsequent to 24h of immersing in different solution for one drying-wetting. 10 cycles were conducted on specimens. The changes in mass loss of specimens, as well as leaching concentration and pH of filtered leachates were studied after each cycle. Results indicated that drying-wetting cycles could accelerate the leaching and deterioration of solidified specimens. The cumulative leached lead with acetic acid (pH=2.88) in this study was 109, 83 and 71 mg respectively for solidified specimens of cement-to-dry soil (C/Sd) ratios 0.2, 0.3 and 0.4, compared to 37, 30, and 25mg for a semi-dynamic leaching test. With the increase of cycle times, the cumulative mass loss of specimens increased linearly, but pH of filtered leachates decreased. The leachability and deterioration of solidified specimens increased with acidity of solution. Increases of C/Sd clearly reduced the leachability and deterioration behavior. Copyright © 2014 Elsevier Ltd. All rights reserved.

Formation of metastable phases and nanocomposite structures in rapidly solidified Al-Fe alloys

International Nuclear Information System (INIS)

Nayak, S.S.; Chang, H.J.; Kim, D.H.; Pabi, S.K.; Murty, B.S.

2011-01-01

Highlights: → Structures of nanocomposites in rapidly solidified Al-Fe alloys were investigated. → Nanoquasicrystalline, amorphous and intermetallics phases coexist with α-Al. → Nanoquasicrystalline phase was observed for the first time in the dilute Al alloys. → Thermodynamic driving force plays dominant role in precipitation of Fe-rich phases. → High hardness (3.57 GPa) was observed for nanocomposite of Al-10Fe alloy. - Abstract: In the present work the structure and morphology of the phases of nanocomposites formed in rapidly solidified Al-Fe alloys were investigated in details using analytical transmission electron microscopy and X-ray diffraction. Nanoquasicrystalline phases, amorphous phase and intermetallics like Al 5 Fe 2 , Al 13 F 4 coexisted with α-Al in nanocomposites of the melt spun alloys. It was seen that the Fe supersaturation in α-Al diminished with the increase in Fe content and wheel speed indicating the dominant role of the thermodynamic driving force in the precipitation of Fe-rich phases. Nanoquasicrystalline phases were observed for the first time in the dilute Al alloys like Al-2.5Fe and Al-5Fe as confirmed by high resolution TEM. High hardness (3.57 GPa) was measured in nanocomposite of Al-10Fe alloy, which was attributed to synergistic effect of solid solution strengthening due to high solute content (9.17 at.% Fe), dispersion strengthening by high volume fraction of nanoquasicrystalline phase; and Hall-Petch strengthening from finer cell size (20-30 nm) of α-Al matrix.

Microstructural Quantification of Rapidly Solidified Undercooled D2 Tool Steel

Science.gov (United States)

Valloton, J.; Herlach, D. M.; Henein, H.; Sediako, D.

2017-10-01

Rapid solidification of D2 tool steel is investigated experimentally using electromagnetic levitation (EML) under terrestrial and reduced gravity conditions and impulse atomization (IA), a drop tube type of apparatus. IA produces powders 300 to 1400 μm in size. This allows the investigation of a large range of cooling rates ( 100 to 10,000 K/s) with a single experiment. On the other hand, EML allows direct measurements of the thermal history, including primary and eutectic nucleation undercoolings, for samples 6 to 7 mm in diameter. The final microstructures at room temperature consist of retained supersaturated austenite surrounded by eutectic of austenite and M7C3 carbides. Rapid solidification effectively suppresses the formation of ferrite in IA, while a small amount of ferrite is detected in EML samples. High primary phase undercoolings and high cooling rates tend to refine the microstructure, which results in a better dispersion of the eutectic carbides. Evaluation of the cell spacing in EML and IA samples shows that the scale of the final microstructure is mainly governed by coarsening. Electron backscattered diffraction (EBSD) analysis of IA samples reveals that IA powders are polycrystalline, regardless of the solidification conditions. EBSD on EML samples reveals strong differences between the microstructure of droplets solidified on the ground and in microgravity conditions. While the former ones are polycrystalline with many different grains, the EML sample solidified in microgravity shows a strong texture with few much larger grains having twinning relationships. This indicates that fluid flow has a strong influence on grain refinement in this system.

Evaluating Primary Dendrite Trunk Diameters in Directionally Solidified Al-Si Alloys

Science.gov (United States)

Grugel, R. N.; Tewari, S. N.; Poirier, D. R.

2014-01-01

The primary dendrite trunk diameters of Al-Si alloys that were directionally solidified over a range of processing conditions have been measured. These data are analyzed with a model based primarily on an assessment of secondary dendrite arm dissolution in the mushy zone. Good fit with the experimental data is seen and it is suggested that the primary dendrite trunk diameter is a useful metric that correlates well with the actual solidification processing parameters. These results are placed in context with the limited results from the aluminium - 7 wt. % silicon samples directionally solidified aboard the International Space Station as part of the MICAST project.

Durability, mechanical, and thermal properties of experimental glass-ceramic forms for immobilizing ICPP high level waste

International Nuclear Information System (INIS)

Vinjamuri, K.

1990-01-01

The high-level liquid waste generated at the Idaho Chemical Processing Plant (ICPP) is routinely solidified into granular calcined high-level waste (HLW) and stored onsite. Research is being conducted at the ICPP on methods of immobilizing the HLW, including developing a durable glass-ceramic form which has the potential to significantly reduce the final waste volume by up to 60% compared to a glass form. Simulated, pilot plant, non-radioactive, calcines similar to the composition of the calcined HLW and glass forming additives are used to produce experimental glass-ceramic forms. The objective of the research reported in this paper is to study the impact of ground calcine particle size on durability and mechanical and thermal properties of experimental glass-ceramic forms

Influence of Short-time Oxidation on Corrosion Properties of Directionally Solidified Superalloys with Different Orientations

Directory of Open Access Journals (Sweden)

MA Luo-ning

2016-07-01

Full Text Available In order to investigate the corrosion performance on intersecting and longitudinal surfaces of unoxidized and oxidized directionally solidified superalloys, Ni-base directionally solidified superalloy DZ125 and Co-base directionally solidified superalloy DZ40M were selected. Oxidation behavior on both alloys with different orientations was investigated at 1050℃ at different times, simulating the oxidation process of vanes or blades in service; subsequent electrochemical performance in 3.5%NaCl aqueous solution was studied on two orientations of unoxidized and oxidized alloys, simulating the corrosion process of superalloy during downtime. The results show that grain boundaries and sub-boundaries of directionally solidified superalloys are susceptible to corrosion and thus longitudinal surface with lower area fraction of grain boundaries has higher corrosion resistance. Compared to intersecting surface of alloys, the structure of grain boundaries of longitudinal surface is less conducive to diffusion and thus the oxidation rate on longitudinal surface is lower. Formation of oxide layers on alloys after short-time oxidation provides protective effect and enhances the corrosion resistance.

Elution behavior of heavy metals from cement solidified products of incinerated ash waste - 59102

International Nuclear Information System (INIS)

Meguro, Yoshihiro; Kawato, Yoshimi; Nakayama, Takuya; Tomioka, Osamu; Mitsuda, Motoyuki

2012-01-01

A method, in which incinerated ash is solidified with a cement material, has been developed to dispose radioactive incinerated ash waste. In order to bury the solidified product, it is required that elution of hazardous heavy metals included in the ash from the solidified products is inhibited. In this study, the elution behavior of the heavy metals from the synthetic solidified products, which included Pb(II), Cd(II), and Cr(VI) and were prepared using ordinary portland cement (OPC), blast furnace slag cement (BFS), or a cement material that showed low alkalinity (LA-Cement), was investigated. Several chemicals and materials were added as additive agents to prevent the elution of the heavy metals. When OPC was used, Cd elution was inhibited, but Pb and Cr were not enough even using the additive agent examined. FeSO 4 and Na 2 S additive agents worked effective to inhibit elution of Cr. When BFS was used, the elution of Pb, Cd and Cr was inhibited for the all products prepared. In the case of LA-Cement, the elution of Pb and Cd was inhibited for the all products, but only the product that was added FeSO 4 showed good result of the elution of Cr. (authors)

Evolution of the microstructure and nanohardness of Ti-48 at.%Al alloy solidified under high pressure

International Nuclear Information System (INIS)

Wang, Hongwei; Zhu, Dongdong; Zou, Chunming; Wei, Zunjie

2012-01-01

Highlights: → The microstructure of Ti-48Al alloy changes under high pressure. → With increasing pressure, the amount of γ s phase decreases. → High pressure leads to the decreasing of lamellar spacing. → The nanohardness of lamellar structure increases with pressure. -- Abstract: In this work the microstructure and nanohardness of Ti-48 at.%Al alloy solidified under different pressures (normal pressure, 2 GPa, 4 GPa) were experimental investigated by using a tungsten-carbide six-anvil apparatus. The results indicate that high pressure does not change the phase constitution of Ti-48 at.%Al alloy. However, the microstructure changes under high pressure. With increasing pressure, the volume fraction of interdendritic γ (γ s ) phase decreases and Al concentration in lamellae increases. When the pressure is 4 GPa, there is only a little γ s embedded in lamellar structure. The volume fraction of γ s phase is approximately 17.0% for normal pressure, 8.73% for 2 GPa, 0.69% for 4 GPa. The lamellar spacings also decrease with pressure, which are 495 nm, 345 nm, 227 nm under normal pressure, 2 GPa, 4 GPa, respectively. The change in nanohardness was discussed based on the microstructural observations. It shows a certain increase of the nanohardness as the pressure increases from normal pressure to 4 GPa. When the pressure is 4 GPa, the nanohardness increases by 50.2% compared with that of normal pressure.

Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

Science.gov (United States)

Knox, C. A.; Farnsworth, R. K.

1981-08-01

The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In facility, installed in a radiochemical cell, is described in which installed in a radiochemical cell is described in which small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. During the spray calcination of commercial high-level liquid waste spiked with Tc-99 and I-131 and 31 wt% loss of I-131 past the sintered-metal filters. These filters and venturi scrubber were very efficient in removing particulates and Tc-99 from the the off-gas stream. Liquid scrubbers were not efficient in removing I-131 as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents are needed to remove iodine. For all future operations where iodine is present, a silver zeolite adsorber is to be used.

Vitrification of high level nuclear waste inside ambient temperature disposal containers using inductive heating: The SMILE system

International Nuclear Information System (INIS)

Powell, J.; Reich, M.; Barletta, R.

1996-01-01

A new approach, termed SMILE (Small Module Inductively Loaded Energy), for the vitrification of high level nuclear wastes (HLW) is described. Present vitrification systems liquefy the HLW solids and associated frit material in large high temperature melters. The molten mix is then poured into small (∼1 m 3 ) disposal canisters, where it solidifies and cools. SMILE eliminates the separate, large high temperature melter. Instead, the BLW solids and frit melt inside the final disposal containers, using inductive heating. The contents then solidify and cool in place. The SMILE modules and the inductive heating process are designed so that the outer stainless can of the module remains at near ambient temperature during the process cycle. Module dimensions are similar to those of present disposal containers. The can is thermally insulated from the high temperature inner container by a thin layer of refractory alumina firebricks. The inner container is a graphite crucible lined with a dense alumina refractory that holds the HLW and fiit materials. After the SMILE module is loaded with a slurry of HLW and frit solids, an external multi-turn coil is energized with 30-cycle AC current. The enclosing external coil is the primary of a power transformer, with the graphite crucible acting as a single turn ''secondary.'' The induced current in the ''secondary'' heats the graphite, which in turn heats the HLW and frit materials. The first stage of the heating process is carried out at an intermediate temperature to drive off remnant liquid water and water of hydration, which takes about 1 day. The small fill/vent tube to the module is then sealed off and the interior temperature raised to the vitrification range, i.e., ∼1200C. Liquefaction is complete after approximately 1 day. The inductive heating then ceases and the module slowly loses heat to the environment, allowing the molten material to solidify and cool down to ambient temperature

Nickel speciation in cement-stabilized/solidified metal treatment filtercakes

Energy Technology Data Exchange (ETDEWEB)

Roy, Amitava, E-mail: reroy@lsu.edu [J. Bennett Johnston, Sr., Center for Advanced Microstructures and Devices, Louisiana State University, Baton Rouge, LA 70806, USA (United States); Stegemann, Julia A., E-mail: j.stegemann@ucl.ac.uk [Centre for Resource Efficiency & the Environment, Department of Civil, Environmental & Geomatic Engineering, University College London, Chadwick Building, Gower Street, London WC1E 6BT, UK (United Kingdom)

2017-01-05

Highlights: • XAS shows the same Ni speciation in untreated and stabilized/solidified filtercake. • Ni solubility is the same for untreated and stabilized/solidified filtercake. • Leaching is controlled by pH and physical encapsulation for all binders. - Abstract: Cement-based stabilization/solidification (S/S) is used to decrease environmental leaching of contaminants from industrial wastes. In this study, two industrial metal treatment filtercakes were characterized by X-ray diffractometry (XRD), thermogravimetric and differential thermogravimetric analysis (TG/DTG) and Fourier transform infrared (FTIR); speciation of nickel was examined by X-ray absorption (XAS) spectroscopy. Although the degree of carbonation and crystallinity of the two untreated filtercakes differed, α-nickel hydroxide was identified as the primary nickel-containing phase by XRD and nickel K edge XAS. XAS showed that the speciation of nickel in the filtercake was unaltered by treatment with any of five different S/S binder systems. Nickel leaching from the untreated filtercakes and all their stabilized/solidified products, as a function of pH in the acid neutralization capacity test, was essentially complete below pH ∼5, but was 3–4 orders of magnitude lower at pH 8–12. S/S does not respeciate nickel from metal treatment filtercakes and any reduction of nickel leaching by S/S is attributable to pH control and physical mechanisms only. pH-dependent leaching of Cr, Cu and Ni is similar for the wastes and s/s products, except that availability of Cr, Cu and Zn at decreased pH is reduced in matrices containing ground granulated blast furnace slag.

Solidified ceramics of radioactive wastes and method of producing it

International Nuclear Information System (INIS)

Oota, Takao; Matake, Shigeru; Ooka, Kazuo.

1980-01-01

Purpose: To provide solidified ceramics which have low leaching properties to water of radioactive substance, excellent heat dissipating and resistive properties and high mechanical strength by mixing and sintering limited amounts of titanium and aluminum compounds with calcined radioactive wastes containing special compound. Method: More than 20% by weight of titanium compound (as TiO 2 ) and more than 5% by weight of aluminum compound (as Al 2 O 3 ) are mixed with the calcined radioactive wasted containing, as converted by oxide, 5 to 40% by weight of Na 2 O, 5 to 20% by weight of Fe 2 O 3 , 5 to 15% by weight of MoO 3 , 5 to 15% by weight of ZrO 2 , 2 to 10% by weight of CeO 2 , 2 to 10% by weight of Cs 2 O, 1 to 5% by weight of BaO, 1 to 5% by weight of SrO, 0.2 to 2% by weight of Rb 2 O, 0.2% by weight of Y 2 O 3 , 0.2 to 2% by weight of NiO, 5 to 20% by weight of rare earth metal oxide, and 0.2 to 2% by weight of Cr 2 O 3 . The mixture is molded, sintered, and solidified to ceramics which contains no Mo phase, Na 2 O, MoO 3 , K 2 O, MoO 3 and Cs 2 O, MoO 3 phases and the like. (Yoshino, Y.)

Rapidly solidified prealloyed powders by laser spin atomization

Science.gov (United States)

Konitzer, D. G.; Walters, K. W.; Heiser, E. L.; Fraser, H. L.

1984-01-01

A new technique, termed laser spin atomization, for the production of rapidly solidified prealloyed powders is described. The results of experiments involving the production of powders of two alloys, one based on Ni, the other on Ti, are presented. The powders have been characterized using light optical metallography, scanning electron microscopy, energy dispersive X-ray spectroscopy, and Auger elec-tron spectroscopy, and these various observations are described.

Ultrasensitive determination of mercury in human saliva by atomic fluorescence spectrometry based on solidified floating organic drop microextraction

International Nuclear Information System (INIS)

Yuan, C.-G.; Wang, J.; Jin, Y.

2012-01-01

We report on a new, rapid and simple method for the determination of ultra-trace quantities of mercury ion in human saliva. It is based on solidified floating organic drop microextraction and detection by cold vapor atomic fluorescence spectrometry (CV-AFS). Mercury ion was complexed with diethyldithiocarbamate, and the hydrophobic complex was then extracted into fine droplets of 1-undecanol. By cooling in an ice bath after extraction, the droplets in solution solidify to form a single ball floating on the surface of solution. The solidified micro drop containing the mercury complex was then transferred for determination by CV-AFS. The effects of pH value, concentration of chelating reagent, quantity of 1-undecanol, sample volume, equilibration temperature and time were investigated. Under the optimum conditions, the preconcentration of a 25-mL sample is accomplished with an enrichment factor of 182. The limit of detection is 2.5 ng L -1 . The relative standard deviation for seven replicate determinations at 0.1 ng mL -1 level is 4.1%. The method was applied to the determination of mercury in saliva samples collected from four volunteers. Two volunteers having dental amalgam fillings had 0.4 ng mL -1 mercury in their saliva, whereas mercury was not detectable in the saliva of two volunteers who had no dental fillings. (author)

Annual report on the development and characterization of solidified forms for nuclear wastes, 1979

International Nuclear Information System (INIS)

Chick, L.A.; McVay, G.L.; Mellinger, G.B.; Roberts, F.P.

1980-12-01

Development and characterization of solidified nuclear waste forms is a major continuing effort at Pacific Northwest Laboratory. Contributions from seven programs directed at understanding chemical composition, process conditions, and long-term behaviors of various nuclear waste forms are included in this report. The major findings of the report are included in extended figure captions that can be read as brief technical summaries of the research, with additional information included in a traditional narrative format. Waste form development proceeded on crystalline and glass materials for high-level and transuranic (TRU) wastes. Leaching studies emphasized new areas of research aimed at more basic understanding of waste form/aqueous solution interactions. Phase behavior and thermal effects research included studies on crystal phases in defense and TRU waste glasses and on liquid-liquid phase separation in borosilicate waste glasses. Radiation damage effects in crystals and glasses from alpha decay and from transmutation are reported

Importance of microscopy in durability studies of solidified and stabilized contaminated soils

Science.gov (United States)

Klich, I.; Wilding, L.P.; Drees, L.R.; Landa, E.R.

1999-01-01

Solidification/stabilization (S/S) is recognized by the U.S. EPA as a best demonstrated available technology for the containment of contaminated soils and other hazardous wastes that cannot be destroyed by chemical, thermal, or biological means. Despite the increased use of S/S technologies, little research has been conducted on the weathering and degradation of solidified and stabilized wastes once the treated materials have been buried. Published data to verify the performance and durability of landfilled treated wastes over time are rare. In this preliminary study, optical and electron microscopy (scanning electron microscopy [SEM], transmission electron microscopy [TEM] and electron probe microanalyses [EPMA]) were used to evaluate weathering features associated with metal-bearing contaminated soil that had been solidified and stabilized with Portland cement and subsequently buried on site, stored outdoors aboveground, or achieved in a laboratory warehouse for up to 6 yr. Physical and chemical alteration processes identified include: freeze-thaw cracking, cracking caused by the formation of expansive minerals such as ettringite, carbonation, and the movement of metals from waste aggregates into the cement micromass. Although the extent of degradation after 6 yr is considered slight to moderate, results of this study show that the same environmental concerns that affect the durability of concrete must be considered when evaluating the durability and permanence of the solidification and stabilization of contaminated soils with cement. In addition, such evaluations cannot be based on leaching and chemical analyses alone. The use of all levels of microscopic analyses must be incorporated into studies of the long-term performance of S/S technologies.Solidification/stabilization (S/S) is recognized by the U.S. EPA as a best demonstrated available technology for the containment of contaminated soils and other hazardous wastes that cannot be destroyed by chemical

A comparative EBSP study of microstructure and microtexture formation from undercooled Ni99B1 melts solidified on an electrostatic levitator and an electromagnetic levitator

International Nuclear Information System (INIS)

Li Mingjun; Ishikawa, Takehiko; Nagashio, Kosuke; Kuribayashi, Kazuhiko; Yoda, Shinichi

2006-01-01

Ni 99 B 1 alloys were solidified by containerless processing at various melt undercoolings on an electrostatic levitator (ESL) and an electromagnetic levitator (EML). A scanning electron microscope in combination with an electron backscatter diffraction pattern mapping technique was employed to reveal microstructures and microtextures formed on these two facilities. The microstructure consists of well-developed primary dendrites with coarse secondary arms in the alloys solidified on the ESL at low and medium undercooling levels, whereas equiaxed grains are yielded in alloys solidified on the EML at almost the same undercoolings. Further analysis indicates that the melt flow induced by the electromagnetic field in the EML may play a significant role in promoting fragmentation of primary dendrites in the mushy zone and thus resulting in equiaxed grains. In contrast, the primary dendrites in the alloy processed on the ESL can fully develop in the absence of melt flow. The fluid flow in the sample on the EML can rotate, move, and displace surviving fragments, yielding a random distribution of grain orientation and thus leading to a random microtexture at low and medium undercoolings. At high undercoolings, refined equiaxed grains can be obtained on both the ESL and the EML and the influence of melt flow on refinement seems negligible due to the enhanced driving force in capillarity and solute effects. A great number of coherent annealing twins are formed, making the pole figures more complex and random

Solidified self-nanoemulsifying formulation for oral delivery of combinatorial therapeutic regimen

DEFF Research Database (Denmark)

Jain, Amit K; Thanki, Kaushik; Jain, Sanyog

2014-01-01

PURPOSE: The present work reports rationalized development and characterization of solidified self-nanoemulsifying drug delivery system for oral delivery of combinatorial (tamoxifen and quercetin) therapeutic regimen. METHODS: Suitable oil for the preparation of liquid SNEDDS was selected based...

Performance demonstration program plan for RCRA constituent analysis of solidified wastes

International Nuclear Information System (INIS)

1995-06-01

Performance Demonstration Programs (PDPS) are designed to help ensure compliance with the Quality Assurance Objectives (QAOs) for the Waste Isolation Pilot Plant (WIPP). The PDPs are intended for use by the Department of Energy (DOE) Carlsbad Area Office (CAO) to assess and approve the laboratories and other measurement facilities supplying services for the characterization of WIPP TRU waste. The PDPs may also be used by CAO in qualifying laboratories proposing to supply additional analytical services that are required for other than waste characterization, such as WIPP site operations. The purpose of this PDP is to test laboratory performance for the analysis of solidified waste samples for TRU waste characterization. This performance will be demonstrated by the successful analysis of blind audit samples of simulated, solidified TRU waste according to the criteria established in this plan. Blind audit samples (hereinafter referred to as PDP samples) will be used as an independent means to assess laboratory performance regarding compliance with the QAOs. The concentration of analytes in the PDP samples will address levels of regulatory concern and will encompass the range of concentrations anticipated in actual waste characterization samples. Analyses that are required by the WIPP to demonstrate compliance with various regulatory requirements and which are included in the PDP must be performed by laboratories that demonstrate acceptable performance in the PDP. These analyses are referred to as WIPP analyses and the samples on which they are performed are referred to as WIPP samples for the balance of this document

Vitrification of high level wastes in France

International Nuclear Information System (INIS)

Sombret, C.

1984-02-01

A brief historical background of the research and development work conducted in France over 25 years is first presented. Then, the papers deals with the vitrification at (1) the UP1 reprocessing plant (Marcoule) and (2) the UP2 and UP3 reprocessing plants (La Hague). 1) The properties of glass required for high-level radioactive waste vitrification are recalled. The vitrification process and facility of Marcoule are presented. (2) The average characteristics (chemical composition, activity) of LWR fission product solution are given. The glass formulations developed to solidify LWR waste solution must meet the same requirements as those used in the UP1 facility at Marcoule. Three important aspects must be considered with respect to the glass fabrication process: corrosiveness of the molten glass with regard to metals, viscosity of the molten glass, and, volatization during glass fabrication. The glass properties required in view of interim storage and long-term disposal are then largely developed. Two identical vitrification facilities are planned for the site: T7, to process the UP2 throughput, and T7 for the UP3 plant. A prototype unit was built and operated at Marcoule

Effect of solidification parameters on mechanical properties of directionally solidified Al-Rich Al-Cu alloys

Science.gov (United States)

Çadırlı, Emin

2013-05-01

Al(100-x)-Cux alloys (x=3 wt%, 6 wt%, 15 wt%, 24 wt% and 33 wt%) were prepared using metals of 99.99% high purity in vacuum atmosphere. These alloys were directionally solidified under steady-state conditions by using a Bridgman-type directional solidification furnace. Solidification parameters (G, V and ), microstructure parameters (λ1, λ2 and λE) and mechanical properties (HV, σ) of the Al-Cu alloys were measured. Microstructure parameters were expressed as functions of solidification parameters by using a linear regression analysis. The dependency of HV, σ on the cooling rate, microstructure parameters and composition were determined. According to experimental results, the microhardness and ultimate tensile strength of the solidified samples was increased by increasing the cooling rate and Cu content, but decreased with increasing microstructure parameters. The microscopic fracture surfaces of the different samples were observed using scanning electron microscopy. Fractographic analysis of the tensile fracture surfaces showed that the type of fracture significantly changed from ductile to brittle depending on the composition.

Nial and Nial-Based Composites Directionally Solidified by a Containerless Zone Process. Ph.D. Thesis

Science.gov (United States)

Joslin, Steven M.

1995-01-01

A containerless electromagnetically levitated zone (CELZ) process has been used to directionally solidify NiAl and NiAl-based composites. The CELZ processing results in single crystal NiAl (HP-NiAl) having higher purity than commercially pure NiAl grown by a modified Bridgman process (CP-NiAl). The mechanical properties, specifically fracture toughness and creep strength, of the HP-NiAl are superior to binary CP-NiAl and are used as a base-line for comparison with the composite materials subsequently studied. Two-phase composite materials (NiAl-based eutectic alloys) show improvement in room temperature fracture toughness and 1200 to 1400 K creep strength over that of binary HP-NiAl. Metallic phase reinforcements produce the greatest improvement in fracture toughness, while intermetallic reinforcement produces the largest improvement in high temperature strength. Three-phase eutectic alloys and composite materials were identified and directionally solidified with the intent to combine the improvements observed in the two-phase alloys into one alloy. The room temperature fracture toughness and high temperature strength (in air) serve as the basis for comparison between all of the alloys. Finally, the composite materials are discussed in terms of dominant fracture mechanism observed by fractography.

Detection of free liquid in cement-solidified radioactive waste drums using computed tomography

International Nuclear Information System (INIS)

Steude, J.S.; Tonner, P.D.

1991-01-01

Acceptance criteria for disposal of radioactive waste drums require that the cement-solidified material in the drum contain minimal free liquid after the cement has hardened. Free liquid is to be avoided because it may corrode the drum, escape and cause environmental contamination. The DOE has requested that a nondestructive evaluation method be developed to detect free liquid in quantities in excess of 0.5% by volume. This corresponds to about 1 liter in a standard 208 liter (55 gallon) drum. In this study, the detection of volumes of free liquid in a 57 cm (2 ft.) diameter cement-solidified drum is demonstrated using high-energy X-ray computed tomography (CT0. In this paper it is shown that liquid concentrations of simulated radioactive waste inside glass tubes imbedded in cement can easily be detected, even for tubes with inner diameters less than 2 mm (0.08 in.). Furthermore, it is demonstrated that tubes containing water and liquid concentrations of simulated radioactive waste can be distinguished from tubes of the same size containing air. The CT images were obtained at a rate of about 6 minutes per slice on a commercially available CT system using a 9 MeV linear accelerator source

Modeling Macrosegregation in Directionally Solidified Aluminum Alloys under Gravitational and Microgravitational Conditions.

Energy Technology Data Exchange (ETDEWEB)

Lauer, Mark A.; Poirier, David R.; Erdmann, Robert G.; Tewari, Surendra N.; Madison, Jonathan D

2014-09-01

This report covers the modeling of seven directionally solidified samples, five under normal gravitational conditions and two in microgravity. A model is presented to predict macrosegregation during the melting phases of samples solidified under microgravitational conditions. The results of this model are compared against two samples processed in microgravity and good agreement is found. A second model is presented that captures thermosolutal convection during directional solidification. Results for this model are compared across several experiments and quantitative comparisons are made between the model and the experimentally obtained radial macrosegregation profiles with good agreement being found. Changes in cross section were present in some samples and micrographs of these are qualitatively compared with the results of the simulations. It is found that macrosegregation patterns can be affected by changing the mold material.

Immobilization of low and intermediate level radioactive liquid wastes using some industrial by-product materials

International Nuclear Information System (INIS)

Sami, N.M.; EI-Dessouky, M.I.; Abou EI-Nour, F.H.; Abdel-Khalik, M.

2006-01-01

Immobilization of low and intermediate level.radioactive liquid wastes in different matrices: ordinary Portland cement and cement mixed with some industrial byproduct: by-pass kiln cement dust, blast furnace slag and ceramic sludge was studied. The effect of these industrial by-product materials on the compressive strength, water immersion, radiation effect and teachability were investigated. The obtained results showed that, these industrial by-product improve the cement pastes where they increase the compressive strength, decrease the leaching rate for radioactive cesium-137 and cobalt-60 ions through the solidified waste forms and increase resistance for y-radiation. It is found that, solidified waste forms of intermediate level liquid waste (ILLW) had high compressive strength values more than those obtained from low level liquid waste (LLLW). The compressive strength increased after immersion in different leachant for one and three months for samples with LLLW higher than those obtained for ILLW. The cumulative fractions released of cesium-137 and cobalt-60 of solidified waste forms of LLLW was lower than those obtained for ILLW

A new technology for concentrating and solidifying liquid LLRW

Energy Technology Data Exchange (ETDEWEB)

Newell, N. [TMC, Inc., Portland, OR (United States); Osborn, M.W.; Carey, C.C. [Oregon Health Sciences Univ., Portland, OR (United States)] [and others

1995-12-31

One of the unsolved problem areas of low level radioactive waste management is the radiolabeled material generated by life sciences research and clinical diagnostics. In hundreds of academic, biotechnology, and pharmaceutical institutions, there exists large amounts of both aqueous and organic solutions containing radioactively labeled nucleic acids, proteins, peptides, and their monomeric components. We have invented a generic slurry capable of binding all these compounds, thus making it possible to concentrate and solidify the radioactive molecules into a very small and lightweight material. The slurry can be contained in both large and small disposal plastic devices designed for the size of any particular operation. The savings in disposal costs and convenience of this procedure is a very attractive alternative to the present methods of long and short term storage. Additionally, the slurry can remove radiolabeled biological compounds from organic solvents, thus solving the major problem of {open_quotes}mixed{close_quotes} waste. We are now proceeding with the field application stage for the testing of these devices and anticipate widespread use of the process. We also are exploring the use of the slurry on other types of liquid low level radioactive waste.

Primary Dendrite Arm Spacings in Al-7Si Alloy Directionally Solidified on the International Space Station

Science.gov (United States)

Angart, Samuel; Lauer, Mark; Poirier, David; Tewari, Surendra; Rajamure, Ravi; Grugel, Richard

2015-01-01

Samples from directionally solidified Al- 7 wt. % Si have been analyzed for primary dendrite arm spacing (lambda) and radial macrosegregation. The alloy was directionally solidified (DS) aboard the ISS to determine the effect of mitigating convection on lambda and macrosegregation. Samples from terrestrial DS-experiments thermal histories are discussed for comparison. In some experiments, lambda was measured in microstructures that developed during the transition from one speed to another. To represent DS in the presence of no convection, the Hunt-Lu model was used to represent diffusion controlled growth under steady-state conditions. By sectioning cross-sections throughout the entire length of a solidified sample, lambda was measured and calculated using the model. During steady-state, there was reasonable agreement between the measured and calculated lambda's in the space-grown samples. In terrestrial samples, the differences between measured and calculated lambda's indicated that the dendritic growth was influenced by convection.

A comparative EBSP study of microstructure and microtexture formation from undercooled Ni{sub 99}B{sub 1} melts solidified on an electrostatic levitator and an electromagnetic levitator

Energy Technology Data Exchange (ETDEWEB)

Li Mingjun [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Tsukuba Space Center, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan)]. E-mail: li.mingjun@aist.go.jp; Ishikawa, Takehiko [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Tsukuba Space Center, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan); Nagashio, Kosuke [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Sagamihara Campus, 3-1-1 Yoshinodai, Sagamihara, Kanagawa 229-8510 (Japan); Kuribayashi, Kazuhiko [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Sagamihara Campus, 3-1-1 Yoshinodai, Sagamihara, Kanagawa 229-8510 (Japan); Yoda, Shinichi [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Tsukuba Space Center, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan)

2006-08-15

Ni{sub 99}B{sub 1} alloys were solidified by containerless processing at various melt undercoolings on an electrostatic levitator (ESL) and an electromagnetic levitator (EML). A scanning electron microscope in combination with an electron backscatter diffraction pattern mapping technique was employed to reveal microstructures and microtextures formed on these two facilities. The microstructure consists of well-developed primary dendrites with coarse secondary arms in the alloys solidified on the ESL at low and medium undercooling levels, whereas equiaxed grains are yielded in alloys solidified on the EML at almost the same undercoolings. Further analysis indicates that the melt flow induced by the electromagnetic field in the EML may play a significant role in promoting fragmentation of primary dendrites in the mushy zone and thus resulting in equiaxed grains. In contrast, the primary dendrites in the alloy processed on the ESL can fully develop in the absence of melt flow. The fluid flow in the sample on the EML can rotate, move, and displace surviving fragments, yielding a random distribution of grain orientation and thus leading to a random microtexture at low and medium undercoolings. At high undercoolings, refined equiaxed grains can be obtained on both the ESL and the EML and the influence of melt flow on refinement seems negligible due to the enhanced driving force in capillarity and solute effects. A great number of coherent annealing twins are formed, making the pole figures more complex and random.

Microstructure and mechanical properties of a novel rapidly solidified, high-temperature Al-alloy

Energy Technology Data Exchange (ETDEWEB)

Overman, N.R., E-mail: Nicole.Overman@pnnl.gov [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99352 (United States); Mathaudhu, S.N. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99352 (United States); University of California, Riverside, 3401 Watkins Dr., Riverside, CA 92521 (United States); Choi, J.P.; Roosendaal, T.J.; Pitman, S. [Pacific Northwest National Laboratory, P.O. Box 999, Richland, WA 99352 (United States)

2016-02-15

Rapid solidification (RS) processing, as a production method, offers a variety of unique properties based on far-from-equilibrium microstructures obtained through rapid cooling rates. In this study, we seek to investigate the microstructures and properties of a novel Al-alloy specifically designed for high temperature mechanical stability. Synthesis of, AlFe{sub 11.4}Si{sub 1.8}V{sub 1.6}Mn{sub 0.9} (wt.%), was performed by two approaches: rotating cup atomization (“shot”) and melt spinning (“flake”). These methods were chosen because of their ability to produce alloys with tailored microstructures due to their inherent differences in cooling rate. The as-solidified precursor materials were microstructurally characterized with electron microscopy. The results show that the higher cooling rate flake material exhibited the formation of nanocrystalline regions as well additional phase morphologies not seen in the shot material. Secondary dendritic branching in the flake material was on the order of 0.1–0.25 μm whereas branching in the shot material was 0.5–1.0 μm. Consolidated and extruded material from both precursor materials was mechanically evaluated at both ambient and high (300 °C) temperature. The consolidated RS flake material is shown to exhibit higher strengths than the shot material. The ultimate tensile strength of the melt spun flake was reported as 544.2 MPa at room temperature and 298.0 MPa at 300 °C. These results forecast the ability to design alloys and processing approaches with unique non-equilibrium microstructures with robust mechanical properties at elevated temperatures. - Highlights: • A novel alloy, AlFe{sub 11.4}Si{sub 1.8}V{sub 1.6}Mn{sub 0.9} was fabricated by rapid solidification. • Room temperature yield strength exceeded 500 MPa. • Elevated temperature (300 °C) yield strength exceeded 275 MPa. • Forging, after extrusion of the alloy resulted in microstructural coarsening. • Decreased strength and ductility was

Leaching experiment of cement solidified waste form under unsaturated condition

International Nuclear Information System (INIS)

Wang Zhiming; Yao Laigen; Li Shushen; Zhao Yingjie; Cai Yun; Li Dan; Han Xinsheng; An Yongfeng

2003-01-01

A device for unsaturated leaching experiments was designed and built up. 8 different sizes, ranging from 40.2 cm 3 to 16945.5 cm 3 , of solidified waste form were tested in the experiment. 5 different water contents, from 0.15 to 0.40, were used for the experiment. The results show that the cumulative leaching fraction increases with water content when the sizes of the forms are equal to and less than 4586.7 cm 3 , for example, the ratios of the cumulative leaching fractions are between 1.24-1.41 under water content of 0.35 and 0.15 on 360 day of Teaching. It can also be seen that the cumulative leaching fraction under higher water content is close to that under saturated condition. The cumulative leaching fraction decreases with size of the form. Maximum leached depth of the solidified waste forms is about 2.25 cm after one year Teaching. Moreover, it has no clear effect on cumulative leaching fraction that sampling or non-sampling during the experiment

High-temperature deformation behavior and mechanical properties of rapidly solidified Al-Li-Co and Al-Li-Zr alloys

International Nuclear Information System (INIS)

Sastry, S.M.L.; Oneal, J.E.

1984-01-01

The deformation behavior at 25-300 C of rapidly solidified Al-3Li-0.6Co and Al-3Li-0.3Zr alloys was studied by tensile property measurements and transmission electron microscopic examination of dislocation substructures. In binary Al-3Li and Al-3Li-Co alloys, the modulus normalized yield stress increases with an increase in temperature up to 150 C and then decreases. The yield stress at 25 C of Al-3Li-0.3Zr alloys is 180-200 MPa higher than that of Al-3Li alloys. However, the yield stress of the Zr-containing alloy decreases drastically with increasing temperatures above 75 C. The short-term yield stresses at 100-200 C of the Al-3Li-based alloys are higher than that of the conventional high-temperature Al alloys. The temperature dependences of the flow stresses of the alloys were analyzed in terms of the magnitudes and temperature dependences of the various strengthening contributions in the two alloys. The dislocation substructures at 25-300 C were correlated with mechanical properties. 19 references

Validation of the solidifying soil process using laser-induced breakdown spectroscopy

Science.gov (United States)

Lin, Zhao-Xiang; Liu, Lin-Mei; Liu, Lu-Wen

2016-09-01

Although an Ionic Soil Stabilizer (ISS) has been widely used in landslide control, it is desirable to effectively monitor the stabilization process. With the application of laser-induced breakdown spectroscopy (LIBS), the ion contents of K, Ca, Na, Mg, Al, and Si in the permeable fluid are detected after the solidified soil samples have been permeated. The processes of the Ca ion exchange are analyzed at pressures of 2 and 3 atm, and it was determined that the cation exchanged faster as the pressure increased. The Ca ion exchanges were monitored for different stabilizer mixtures, and it was found that a ratio of 1:200 of ISS to soil is most effective. The investigated plasticity and liquidity indexes also showed that the 1:200 ratio delivers the best performance. The research work indicates that it is possible to evaluate the engineering performances of soil solidified by ISS in real time and online by LIBS.

Method of solidifying radioactive wastes

International Nuclear Information System (INIS)

Maeda, Masahiko; Kira, Satoshi; Watanabe, Naotoshi; Nagaoka, Takeshi; Akane, Junta.

1982-01-01

Purpose: To obtain solidification products of radioactive wastes having sufficient monoaxial compression strength and excellent in water durability upon ocean disposal of the wastes. Method: Solidification products having sufficient strength and filled with a great amount of radioactive wastes are obtained by filling and solidifying 100 parts by weight of chlorinated polyethylene resin and 100 - 500 parts by weight of particular or powderous spent ion exchange resin as radioactive wastes. The chlorinated polyethylene resin preferably used herein is prepared by chlorinating powderous or particulate polyethylene resin in an aqueous suspending medium or by chlorinating polyethylene resin dissolved in an organic solvent capable of dissolving the polyethylene resin, and it is crystalline or non-crystalline chlorinated polyethylene resin comprising 20 - 50% by weight of chlorine, non-crystalline resin with 25 - 40% by weight of chlorine being particularly preferred. (Horiuchi, T.)

High-level waste solidification - why we chose glass

International Nuclear Information System (INIS)

Grover, J.R.

1979-05-01

This paper considers the desirable properties and factors to be assessed in the selection of a solidified waste product, surveys the possible product options and then analyses in detail their suitability in meeting the criteria. (author)

Experimental Investigation of Closed Porosity of Inorganic Solidified Foam Designed to Prevent Coal Fires

Directory of Open Access Journals (Sweden)

Yi Lu

2015-01-01

Full Text Available In order to overcome the deficiency of the existing fire control technology and control coal spontaneous combustion by sealing air leakages in coal mines, inorganic solidified foam (ISF with high closed porosity was developed. The effect of sodium dodecyl sulfate (SDS concentration on the porosity of the foams was investigated. The results showed that the optimized closed porosity of the solidified foam was 38.65 wt.% for an SDS concentration of approximately 7.4×10-3 mol/L. Based on observations of the microstructure of the pore walls after solidification, it was inferred that an equilibrium between the hydration process and the drainage process existed. Therefore, the ISF was improved using three different systems. Gelatin can increase the viscosity of the continuous phase to form a viscoelastic film around the air cells, and the SDS + gelatin system can create a mixed surfactant layer at gas/liquid interfaces. The accelerator (AC accelerates the hydration process and coagulation of the pore walls before the end of drainage. The mixed SDS + gelatin + AC systems produced an ISF with a total porosity of 79.89% and a closed porosity of 66.89%, which verified the proposed stabilization mechanism.

Evaluating the freeze-thaw durability of portland cement-stabilized-solidified heavy metal waste using acoustic measurements

International Nuclear Information System (INIS)

El-Korchi, T.; Gress, D.; Baldwin, K.; Bishop, P.

1989-01-01

The use of stress wave propagation to assess freeze-thaw resistance of portland cement solidified/stabilized waste is presented. The stress wave technique is sensitive to the internal structure of the specimens and would detect structural deterioration independent of weight loss or visual observations. The freeze-thaw resistance of a cement-solidified cadmium waste and a control was evaluated. The control and cadmium wastes both showed poor freeze-thaw resistance. However, the addition of cadmium and seawater curing increased the resistance to more cycles of freezing and thawing. This is attributed to microstructural changes

Experimental study on the properties of drum-packed, cement solidified waste package of pre and after sea dumping test of sea depth 30m and 100m

International Nuclear Information System (INIS)

Maki, Yasuro; Abe, Hirotoshi; Hattori, Seiichi

1976-01-01

Japan Marine Science and Technology Center has been tackling with the development of the monitoring system to confirm the soundness of drum-packed, cement-solidified low level radioactive waste (the package) during falling and after reaching at sea bed when it is dumped into sea. The test was carried out at Sagami Bay of 30 m and 100 m sea depth using non-radioactive packages. The observation of the falling behaviour of packages in sea was carried out by taking photographs of the motion of packages with an underwater 16 mm movie camera and an underwater industrial TV (ITV), and the observation of the soundness and the area of packages scattered on sea bed was carried out with an underwater ITV and an underwater 70 mm snap camera which were set up on the frame. The proportion of cement-solidified waste was decided so that the uni-axial compressive strength of the cement-solidified waste satisfies the condition of ''The tentative guideline''. Prior to tests at sea, hydrostatic pressure test of packages are carried out on land. After that, core specimens were sampled to obtain the uniaxial compressive strength from packages and were tested. After sea dumping tests, 6 packages were recovered from sea bed, and the soundness were tested. As the results, the deformation and damage of drums and cement solidified waste packages did not occur at all. (Kako, I.)

Cryogenic EBSD reveals structure of directionally solidified ice–polymer composite

Energy Technology Data Exchange (ETDEWEB)

Donius, Amalie E., E-mail: amalie.donius@gmail.com [Thayer School of Engineering, Dartmouth College, 14 Engineering Drive, Hanover, NH 03755 (United States); Department of Materials Science and Engineering, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Obbard, Rachel W., E-mail: Rachel.W.Obbard@dartmouth.edu [Thayer School of Engineering, Dartmouth College, 14 Engineering Drive, Hanover, NH 03755 (United States); Burger, Joan N., E-mail: ridge.of.the.ancients@gmail.com [Thayer School of Engineering, Dartmouth College, 14 Engineering Drive, Hanover, NH 03755 (United States); Department of Materials Science and Engineering, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Hunger, Philipp M., E-mail: philipp.m.hunger@gmail.com [Thayer School of Engineering, Dartmouth College, 14 Engineering Drive, Hanover, NH 03755 (United States); Department of Materials Science and Engineering, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Baker, Ian, E-mail: Ian.Baker@dartmouth.edu [Thayer School of Engineering, Dartmouth College, 14 Engineering Drive, Hanover, NH 03755 (United States); Doherty, Roger D., E-mail: dohertrd@drexel.edu [Department of Materials Science and Engineering, Drexel University, 3141 Chestnut Street, Philadelphia, PA 19104 (United States); Wegst, Ulrike G.K., E-mail: ulrike.wegst@dartmouth.edu [Thayer School of Engineering, Dartmouth College, 14 Engineering Drive, Hanover, NH 03755 (United States)

2014-07-01

Despite considerable research efforts on directionally solidified or freeze-cast materials in recent years, little fundamental knowledge has been gained that links model with experiment. In this contribution, the cryogenic characterization of directionally solidified polymer solutions illustrates, how powerful cryo-scanning electron microscopy combined with electron backscatter diffraction is for the structural characterization of ice–polymer composite materials. Under controlled sublimation, the freeze-cast polymer scaffold structure is revealed and imaged with secondary electrons. Electron backscatter diffraction fabric analysis shows that the ice crystals, which template the polymer scaffold and create the lamellar structure, have a-axes oriented parallel to the direction of solidification and the c-axes perpendicular to it. These results indicate the great potential of both cryo-scanning electron microscopy and cryo-electron backscatter diffraction in gaining fundamental knowledge of structure–property–processing correlations. - Highlights: • Cryo-SEM of freeze-cast polymer solution reveals an ice-templated structure. • Cryo-EBSD reveals the ice crystal a-axis to parallel the solidification direction. • The honeycomb-like polymer phase favors columnar ridges only on one side. • Combining cryo-SEM with EBSD links solidification theory with experiment.

Cryogenic EBSD reveals structure of directionally solidified ice–polymer composite

International Nuclear Information System (INIS)

Donius, Amalie E.; Obbard, Rachel W.; Burger, Joan N.; Hunger, Philipp M.; Baker, Ian; Doherty, Roger D.; Wegst, Ulrike G.K.

2014-01-01

Despite considerable research efforts on directionally solidified or freeze-cast materials in recent years, little fundamental knowledge has been gained that links model with experiment. In this contribution, the cryogenic characterization of directionally solidified polymer solutions illustrates, how powerful cryo-scanning electron microscopy combined with electron backscatter diffraction is for the structural characterization of ice–polymer composite materials. Under controlled sublimation, the freeze-cast polymer scaffold structure is revealed and imaged with secondary electrons. Electron backscatter diffraction fabric analysis shows that the ice crystals, which template the polymer scaffold and create the lamellar structure, have a-axes oriented parallel to the direction of solidification and the c-axes perpendicular to it. These results indicate the great potential of both cryo-scanning electron microscopy and cryo-electron backscatter diffraction in gaining fundamental knowledge of structure–property–processing correlations. - Highlights: • Cryo-SEM of freeze-cast polymer solution reveals an ice-templated structure. • Cryo-EBSD reveals the ice crystal a-axis to parallel the solidification direction. • The honeycomb-like polymer phase favors columnar ridges only on one side. • Combining cryo-SEM with EBSD links solidification theory with experiment

Chemical leaching of rapidly solidified Al-Si binary alloys

International Nuclear Information System (INIS)

Yamauchi, I.; Takahara, K.; Tanaka, T.; Matsubara, K.

2005-01-01

Various particulate precursors of Al 100-x Si x (x = 5-12) alloys were prepared by a rapid solidification process. The rapidly solidified structures of the precursors were examined by XRD, DSC and SEM. Most of Si atoms were dissolved into the α-Al(fcc) phase by rapid solidification though the solubility of Si in the α-Al phase is negligibly small in conventional solidification. In the case of 5 at.% Si alloy, a single α-Al phase was only formed. The amount of the primary Si phase increased with increase of Si content for the alloys beyond 8 at.% Si. Rapid solidification was effective to form super-saturated α-Al precursors. These precursors were chemically leached by using a basic solution (NaOH) or a hydrochloric acid (HCl) solution. All Al atoms were removed by a HCl solution as well as a NaOH solution. Granules of the Si phase were newly formed during leaching. The specific surface area was about 50-70 m 2 /g independent of Si content. The leaching behavior in both solutions was slightly different. In the case of a NaOH solution, the shape of the precursor often degenerated after leaching. On the other hand, it was retained after leaching by a HCl solution. Fine Si particles precipitated in the α-Al phase by annealing of as-rapidly solidified precursors at 773 K for 7.2 x 10 3 s. In this case, it was difficult to obtain any products by NaOH leaching, but a few of Si particles were obtained by HCl leaching. Precipitated Si particles were dissolved by the NaOH solution. The X-ray diffraction patterns of leached specimens showed broad lines of the Si phase and its lattice constant was slightly larger than that of the pure Si phase. The microstructures of the leached specimens were examined by transmission electron microscopy. It showed that the leached specimens had a skeletal structure composed of slightly elongated particles of the Si phase and quite fine pores. The particle size was about 30-50 nm. It was of comparable order with that evaluated by Scherer

Study on Magnesium in Rainwater and Fertilizer Infiltration to Solidified Peat

Science.gov (United States)

Tajuddin, S. A. M.; Rahman, J. A.; Mohamed, R. M. S. R.

2018-04-01

Magnesium is a component of several primary and secondary minerals in the soil which are essentially insoluble for agricultural purpose. The presence of water infiltrate in the soil allows magnesium to dissolve together into the groundwater. In fertilizers, magnesium is categorized as secondary macronutrient which supplies food and encouraging for plants growth. The main objective of this study was to determine the concentration of magnesium in fibric peat when applied the solidification under different conditions. Physical model was used as a mechanism for the analysis of the experimental data using a soil column as an equipment to produce water leaching. In this investigation, there were four outlets in the soil column which were prepared from the top of the column to the bottom with the purpose of identifying the concentration of magnesium for each soil level. The water leaching of each outlet was tested using atomic absorption spectroscopy (AAS). The results obtained showed that the highest concentrations of magnesium for flush and control condition at outlet 4 was 12.50 ppm and 1.29 ppm respectively. Similarly, fibric with solidified peat under rainwater recorded the highest value of 3.16 at outlet 1 for wet condition while for dry condition at outlet 4 of 1.33 ppm. However, the difference in fibric with solidified peat under rainwater and fertilizer condition showed that the highest value for the wet condition was achieved at outlet 1 with 5.43 ppm while highest value of 1.26 ppm was obtained for the dry condition at the outlet 4. It was concluded that the outlets in the soil column gave a detailed analysis of the concentration of magnesium in the soil which was influenced by the environmental conditions.

Long-term leach testing of solidified radioactive waste forms (International Standard Publication ISO 6961:1982)

International Nuclear Information System (INIS)

Stefanik, J.

2001-01-01

Processes are developed for the immobilization of radionuclides by solidification of radioactive wastes. The resulting solidification products are characterized by strong resistance to leaching aimed at low release rates of the radionuclides to the environment. To measure this resistance to leaching of the solidified materials: glass, glass-ceramics, bitumen, cement, concrete, plastics, a long-term leach test is presented. The long-term leach test is aimed at: a) the comparison of different kinds or compositions of solidified waste forms; b) the intercomparison between leach test results from different laboratories on one product; c) the intercomparison between leach test results on products from different processes

The hot bench scale plant Ester for the vitrification of high level wastes

International Nuclear Information System (INIS)

Nannicini, R.; Strazzer, A.; Cantale, C; Donato, A.; Grossi, G.

1985-01-01

In this paper the hot bench-scale plant ESTER for the vitrification of the high-level radioactive wastes is described, and the main results of the first radioactive campaign are reported. The ESTER plant, which is placed in the ADECO-ESSOR hot cells of the C.C.R.-EURATOM-ISPRA, has been built and is operated by the ENEA, Departement of Fuel Cycle. It began operating with real radioactive wastes about 1 year ago, solidifying a total of 12 Ci of fission products into 2,02 Kg of borosilicate glass, corresponding to 757 ml of glass. During the vitrification many samples of liquid and gaseous streams have been taken and analyzed. A radioactivity balance in the plant has been calculated, as well as a mass balance of nitrates and of the 137 Cs and 106 Ru volatized in the process

Microstructure and property of directionally solidified Ni-Si hypereutectic alloy

Science.gov (United States)

Cui, Chunjuan; Tian, Lulu; Zhang, Jun; Yu, Shengnan; Liu, Lin; Fu, Hengzhi

2016-03-01

This paper investigates the influence of the solidification rate on the microstructure, solid/liquid interface, and micro-hardness of the directionally solidified Ni-Si hypereutectic alloy. Microstructure of the Ni-Si hypereutectic alloy is refined with the increase of the solidification rate. The Ni-Si hypereutectic composite is mainly composed of α-Ni matrix, Ni-Ni3Si eutectic phase, and metastable Ni31Si12 phase. The solid/liquid interface always keeps planar interface no matter how high the solidification rate is increased. This is proved by the calculation in terms of M-S interface stability criterion. Moreover, the Ni-Si hypereutectic composites present higher micro-hardness as compared with that of the pure Ni3Si compound. This is caused by the formation of the metastable Ni31Si12 phase and NiSi phase during the directional solidification process.

Research on the compressive strength of basic magnesium salts and cyanide slag solidified body

Science.gov (United States)

Tu, Yubo; Han, Peiwei; Ye, Shufeng; Wei, Lianqi; Zhang, Xiaomeng; Fu, Guoyan; Yu, Bo

2018-02-01

The solidification of cyanide slag by using basic magnesium salts could reduce pollution and protect the environment. Experiments were carried out to investigate the effects of age, mixing amount of cyanide slag, water cement ratio and molar ratio of MgO to MgSO4 on the compressive strength of basic magnesium salts and cyanide slag solidified body in the present paper. It was found that compressive strength of solidified body increased with the increase of age, and decreased with the increase of mixing amount of cyanide slag and water cement ratio. The molar ratio of MgO to MgSO4 should be controlled in the range from 9 to 11 when the mixing amount of cyanide slag was larger than 80 mass%.

Study on metal material corrosion behavior of packaging of cement solidified form

International Nuclear Information System (INIS)

He Zhouguo; Lin Meiqiong; Fan Xianhua

1997-01-01

The corrosion behavior of A3 carbon steel is studied by the specimens that are exposed to atmosphere, embedded in cement solidified form or immersed in corrosion liquid. The corrosion rate is determined by mass change of the specimens. In order to compare the corrosion resistant performance of various coatings, the specimens painted with various material such as epoxide resin, propionic acid resin, propane ether resin and Ti-white paint are tested. The results of the tests show that corrosion rate of A3 carbon steel is less than 10 -3 mm·a -1 in the atmosphere and the cement solidified from, less than 0.1 mm·a -1 in the corrosion liquids, and pH value in the corrosion liquids also affect the corrosion rate of A3 carbon steel. The corrosion resistant performance of Ti-white paint is better than that of other paints. So, A3 carbon steel as packaging material can meet the requirements during storage

Determination of the leaching rate of radionuclide 134Cs from the solidified radioactive wastes in Syrian Portland cement and cement-microsilica matrixes

International Nuclear Information System (INIS)

Ismail Shaaban; Nasim Assi

2010-01-01

The suitability of Syrian Portland cement for disposal of solidified low-level radioactive waste was assessed by measuring the leaching rate of 134 Cs. In ordinary cement concrete, a leaching rate of 1.309 x 10 -3 g/cm 2 per day was measured. Mixing this concrete with microsilica reduced significantly the leaching rate to 3.106 x 10 -4 g/cm 2 per day for 1% mixing, and to 9.645 x 10 -5 g/cm 2 per day for 3% mixing. It was also found that the application of a latex paint reduced these leaching rates by about 10%. These results, along with mechanical strength tests (under radiation exposure, high temperature, long water immersion and freeze-thaw cycling) indicate that Syrian Portland cement is suited for the disposal of low-level radioactive waste. (author)

Method of solidifying radioactive waste

International Nuclear Information System (INIS)

Hasegawa, Akira; Mihara, Shigeru; Yamashita, Koji; Sauda, Kenzo.

1988-01-01

Purpose: To obtain satisfactory plastic solidification products rapidly and more conveniently from radioactive wastes. Method: liquid wastes contain, in addition to sodium sulfate as the main ingredient, nitrates hindering the polymerizing curing reactions and various other unknown ingredients, while spent resins contain residual cationic exchange groups hindering the polymerizing reaction. Generally, as the acid value of unsaturated liquid polyester resins is lower, the number of terminal alkyd resins is small, formation of nitrates is reduced and the polymerizing curing reaction is taken place more smoothly. In view of the above, radioactive wastes obtained by dry powderization or dehydration of radioactive liquid wastes or spent resins are polymerized with unsaturated liquid polyester resins with the acid value of less than 13 to obtain plastic solidification. Thus, if the radioactive wastes contain a great amount of polymerization hindering material such as NaNO 2 , they can be solidified rapidly and conveniently with no requirement for pre-treatment. (Kamimura, Y.)

Simple and rapid determination methods for low-level radioactive wastes generated from nuclear research facilities. Guidelines for determination of radioactive waste samples

International Nuclear Information System (INIS)

Kameo, Yutaka; Shimada, Asako; Ishimori, Ken-ichiro; Haraga, Tomoko; Katayama, Atsushi; Nakashima, Mikio; Hoshi, Akiko

2009-10-01

Analytical methods were developed for simple and rapid determination of U, Th, and several nuclides, which are selected as important nuclides for safety assessment of disposal of wastes generated from research facilities at Nuclear Science Research Institute and Oarai Research and Development Center. The present analytical methods were assumed to apply to solidified products made from miscellaneous wastes by plasma melting in the Advanced Volume Reduction Facilities. In order to establish a system to analyze the important nuclides in the solidified products at low cost and routinely, we have advanced the development of a high-efficiency non-destructive measurement technique for γ-ray emitting nuclides, simple and rapid methods for pretreatment of solidified product samples and subsequent radiochemical separations, and rapid determination methods for long-lived nuclides. In the present paper, we summarized the methods developed as guidelines for determination of radionuclides in the low-level solidified products. (author)

High temperature low cycle fatigue behavior of a directionally solidified Ni-base superalloy DZ951

International Nuclear Information System (INIS)

Chu Zhaokuang; Yu Jinjiang; Sun Xiaofeng; Guan Hengrong; Hu Zhuangqi

2008-01-01

Total strain-controlled low cycle fatigue (LCF) tests were performed at a temperature range from 700 to 900 deg. C in ambient air condition on a directionally solidified Ni-base superalloy DZ951. The fatigue life of DZ951 alloy does not monotonously decrease with increasing temperature, but exhibits a strong dependence on the total strain range. The dislocation characteristics and failed surface observation were evaluated through transmission electron microscopy and scanning electron microscopy. The alloy exhibits cyclic hardening, softening or cyclic stability as a whole, which is dependent on the testing temperature and total strain range. At 700 deg. C, the cyclic plastic deformation process is the main cause of fatigue failure. At 900 deg. C, the failure mostly results from combined fatigue and creep damage under total strain range from 0.6 to 1.2% and the reduction in fatigue life can be taken as the cause of oxidation, creep and cyclic plastic deformation under total strain range of 0.5%

Testing and evaluation of solidified high-level waste forms

International Nuclear Information System (INIS)

Engelmann, C.

1984-01-01

The report describes research by several laboratories on the behaviour, in aqueous and salt environments, of borosilicate glass ceramics proposed for the solidification of nuclear wastes by the European Community. Results were obtained on inactive simulates, doped materials, and on borosilicate glass containing real radioactive waste. The influence of many important parameters were studied: leaching mode, nature of the leachant, pH, pressure, temperature, duration of the treatment, etc. The results of tests lasting for as little as a few hours or for as long as several hundred days, at temperatures up to 200 0 C or under pressures up to 200 bars, are presented. Numerous analytical techniques (ESCA, EMP, IRR, SEM, etc.) were used to determine the structure and the chemical composition of the altered layer developed by hydration at the glass surface. Information is also given on physical properties of the borosilicate glass: crystallization phase separation, alpha-irradiation stability, mechanical and thermal stability, etc. Finally, preliminary results on the structure and composition of hollandite ceramics are given

Testing and evaluation of solidified high-level waste forms

International Nuclear Information System (INIS)

De Batist Al, R.

1983-01-01

In addition to the preceding programme of the European Atomic Energy Community two new borosilicate glass compositions have been introduced. The chemical stability of these waste forms, in particular with respect to geological disposal conditions, is examined as well as effects of alpha-radiation and of devitrification. Leaching studies include theoretical and experimental investigations of the basic leaching mechanisms, the measurement of the leach rates of a number of critical radioisotopes and the influence on the leach rate of various parameters such as temperature, pressure pH and duration. Of particular interest is the simulation of repository conditions. Prelimimary results are described related to various mineral waters, granite and salt solutions. The surface layers generated on the waste forms during corrosion are investigated in detail using various experimental techniques such as scanning electron microscopy, X-ray analysis and alpha particle energy loss spectra measurements. The radiation stability was further tested by continuing investigations of the samples doped with 238 Pu in the course of the previous programme; density and leach rate variations were measured. Effects on the leach rate of devitrification resulting from various heat treatments of active glass samples were also investigated

Microtexture formation of Ni99B1 alloys solidified on an ESL and an EML-a study based on the EBSP technique

International Nuclear Information System (INIS)

Li Mingjun; Ishikawa, Takehiko; Nagashio, Kosuke; Kuribayashi, Kazuhiko; Yoda, Shinichi

2007-01-01

We employed an electrostatic levitator (ESL) and an electromagnetic levitator (EML) to solidify Ni 99 B 1 (at.%) alloys at various undercoolings. The microstructures and microtextures were revealed by using the electron backscatter diffraction pattern (EBSP) technique in a scanning electron microscope. It is found that that no significant refinement can be identified at the low and medium undercooling regimes for the primary trunk in the sample solidified on the ESL, while the fragmentation of the secondary and even tertiary branches may take place to generate equiaxed grains. Further investigation by the EBSP reveals that neighboring grains have small misorientation angles, which may be ascribed to the absence of mechanical stirring from electromagnetic eddy current. A sharp contrast is that the samples solidified on the EML at low and medium undercoolings have refined equiaxed microstructures. The EBSP mapping reveals that the equiaxed grains yielded on the EML have a random distribution in crystallographic orientations among neighboring grains, indicating that electromagnetic stirring (EMS) induced by the electromagnetic field in the EML plays a vital role in promoting fragmentation and thus generating refined grains and random distribution in orientation. Regarding to the refined microstructure at high undercoolings, no significant difference arises in the samples processed between the EML and ESL

Effect of a high magnetic field on the microstructures in directionally solidified Zn–Cu peritectic alloys

International Nuclear Information System (INIS)

Li, Xi; Gagnoud, Annie; Wang, Jiang; Li, Xiaolong; Fautrelle, Yves; Ren, Zhongming; Lu, Xionggang; Reinhart, Guillaume; Nguyen-Thi, Henri

2014-01-01

The effect of an axial high magnetic field on the microstructures in directionally solidified Zn–Cu peritectic alloys was investigated. The experimental results indicated that the magnetic field induced the destabilization of the liquid–solid interface and the formation of a band-like structure. The magnetic field also caused the disruption of the columnar η-Zn and ε-Zn 5 Cu dendrites. As the applied magnetic field increased, the columnar-to-equiaxed transition occurred, and the size of the equiaxed grains gradually decreased. The magnetic effects, the magnetic moment and the thermoelectric magnetic effects during the directional solidification of Zn–Cu peritectic alloys under an axial magnetic field were studied. Regular ε-Zn 5 Cu hexagons appeared on the transverse section of the sample fabricated with a high magnetic field (i.e. 16 T). In addition, electron backscatter diffraction analysis revealed that the 〈0 0 0 1〉-crystal direction of the Zn 5 Cu crystal is not only its easy magnetization direction but also its preferred growth direction. The thermoelectric magnetic effects were numerically simulated. The results indicated that a thermoelectric magnetic force acts on the solid near the liquid–solid interface and increases linearly with an increase in the magnetic field. As the effect of the magnetic moment arising from the magnetic crystalline anisotropy is eliminated, the thermoelectric magnetic effect has a substantial effect on the solidification structure. Therefore, the destabilization of the liquid–solid interface and the disruption of the dendrites during directional solidification under the magnetic field are primarily due to the thermoelectric magnetic force acting on the solid

Characterization of solidified radioactive waste and container due to the incorporation of high density polyethylene granules and powder in mortar matrices

International Nuclear Information System (INIS)

Peric, A.D.

1999-01-01

Powder and granules of the high density polyethylene (PEHD) were used to prepare mortar based matrices for immobilization of radioactive waste materials containing 137 Cs, as well as containers for solidified radioactive waste form. Seven types of matrices, differ due to the percentage of granules and filler material added, were investigated. PEHD powder and granules were added to mortar matrix preparations with the objective of improving physico-chemical characteristics of the radwaste-mortar matrix mixtures, in particular the leach-rate of the immobilized radionuclide, as well as mechanical characteristics either of mortar matrix and container. In this paper, only mechanical strength aspect of the investigated mortar and concrete container formulations, is presented. The equivalent diameter of the PEHD granules used was 2.0 mm. PEHD granules were used to replace 100 volume percent of stone granules, sifted size of 2.0 mm, normally used in the matrix preparation, in order to decrease the porosity and density of the mortar matrix and to avoid segregation of the stone particles at the bottom of the immobilized radioactive waste cylindrical form. PEHD powder, particle size of 250 micrometer, was added as filler to the mortar formulation, replacing 5, 8 and 10 wt% of the total cement weight in matrix formulation and 15 and 18 wt% of the total cement weight in container formulation. Cured samples were investigated on mechanical strength, using 150 MPa hydraulic press, in order to determine influence of added polyethylene granules and powder on samples resistance to mechanical forces that solidified waste materials and concrete containers may experience at the disposal site. Results of performed investigations have shown that samples prepared with polyethylene granules, replacing 100 wt% of the stone granules, have almost twice as much mechanical strength than samples prepared with stone aggregate. Samples prepared with PEHD granules and powder have mechanical strength

3D observation of the solidified structures by x-ray micro computerized tomography

International Nuclear Information System (INIS)

Yasuda, Hideyuki; Ohnaka, Itsuo; Tsuchiyama, Akira; Nakano, Tsukasa; Uesugi, Kentaro

2003-01-01

The high flux density of the monochromatized and well-collimated X-ray and the high-resolution detector provide a new 3D observation tool for microstructures of metallic alloys and ceramics. The X-ray micro computerized tomography in BL47XU of SPring-8 (SP-μCT) was applied to observe microstructures produced through the eutectic reaction for Sn-based alloys and an Al 2 O 3 -Y 2 O 3 oxide system. The constituent phases in the eutectic structures were three-dimensionally identified, in which the lamellar spacing ranged from several to 10 μm. Since the 3D structure of the unidirectionally solidified specimens contains history of the eutectic structure formation, the 3D structure obtained by SP-μCT gives useful information to consider the microstructure evolution. (author)

Options for the disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Mitchell, N.T.; Laughton, A.S.; Webb, G.A.M.

1977-01-01

The management of radioactive waste within the fuel cycle, especially the high-level wastes from reprocessing of nuclear fuel, is currently a matter of particular concern. In the short term (meaning a timescale of tens of years) management by engineered storage is considered to provide a satisfactory solution. Beyond this, however, the two main alternative options which are considered in the paper are: (a) disposal by burial into geologic formations on land; and (b) disposal by emplacement into or onto the seabed. Status of our present knowledge on the land and seabed disposal options is reviewed together with an assessment of the extent to which their reliability and safety can be judged on presently available information. Further information is needed on the environmental behaviour of radioactivity in the form of solidified waste in both situations in order to provide a more complete, scientific assessment. Work done so far has clarified the areas where further research is most needed - for instance modelling of the environmental transfer processes associated with the seabed option. This is discussed together with an indication of the research programmes which are now being pursued

Experiment of solidifying photo sensitive polymer by using UV LED

Science.gov (United States)

Kang, Byoung Hun; Shin, Sung Yeol

2008-11-01

The development of Nano/Micro manufacturing technologies is growing rapidly and in the same manner, the investments in these areas are increasing. The applications of Nano/Micro technologies are spreading out to semiconductor production technology, biotechnology, environmental engineering, chemical engineering and aerospace. Especially, SLA is one of the most popular applications which is to manufacture 3D shaped microstructure by using UV laser and photo sensitive polymer. To make a high accuracy and precision shape of microstructures that are required from the diverse industrial fields, the information of interaction relationship between the photo resin and the light source is necessary for further research. Experiment of solidifying photo sensitive polymer by using UV LED is the topic of this paper and the purpose of this study is to find out what relationships do the reaction of the resin have in various wavelength, power of the light and time.

Leach studies on cement-solidified ion exchange resins from decontamination processes at operating nuclear power stations

International Nuclear Information System (INIS)

McIsaac, C.V.; Akers, D.W.; McConnell, J.W.; Morcos, N.

1992-01-01

The effects of varying pH and leachant compositions on the physical stability and leachability of radionuclides and chelating agents were determined for cement-solidified decontamination ion-exchange resin wastes collected from two operating commercial light water reactors. Small scale waste-form specimens were collected during waste solidifications performed at the Brunswick Steam Electric Plant Unit 1 and at the James A. FitzPatrick Nuclear Power Station. The collected specimens were leach tested, and their compressive strength was measured in accordance with the Nuclear Regulatory Commission's ''Technical Position on Waste Form'' (Revision 1), from the Low-Level Waste Management Branch. Leachates from these studies were analyzed for radionuclides, selected transition metals, and chelating agents to assess the leachability of these waste form constituents. Leachants used for the study were deionized water, simulated seawater, and groundwater compositions similar to those found at Barnwell, South Carolina and Hanford, Washington. Results of this study indicate that initial leachant pH does not affect leachate pH or releases from cement-solidified decontamination ion-exchange resin waste forms. However, differences in leachant composition and the presence of chelating agents may affect the releases of radionuclides and chelating agents. In addition, results from this study indicate that the cumulative releases of radionuclides and chelating agents observed for forms that disintegrated were similar to those for forms that maintained their general physical integrity

A study on the microstructural characteristics of rapidly solidified Al-Fe alloys(I)

International Nuclear Information System (INIS)

Kim, D.H.; Lee, H.I.

1991-01-01

Solidification microstructures and phases in rapidly solidified Al-5, 10wt% Fe alloys have been investigated by TEM bright field and dark field imaging techniques and electron and x-ray diffraction techniques. Rapid solidification of Al-5, 10wt%Fe alloys produces various metastable and stable phases, such as Al m Fe, Al 6 Fe and Al 13 Fe 4 . In addition to these phases, clusters of randomly oriented few nm scale particles exist in the form of fine cellular network with α-Al or primary spherical particles. Solidification microstructures of the rapidly solidified Al-5, 10wt%Fe alloys consist of various combination of primary phases such as Al 13 Fe 4 , Al m Fe and cluster of nm scale particles, and cellular/dendritic structures such as fine cellular network structure of nm scale particle clusters and α-Al and cellular structure of Al m Fe and α-Al, depending upon alloy compositions and local cooling rates. (Author)

Magnetic domain structure, crystal orientation, and magnetostriction of Tb{sub 0.27}Dy{sub 0.73}Fe{sub 1.95} solidified in various high magnetic fields

Energy Technology Data Exchange (ETDEWEB)

Gao, Pengfei [Key Laboratory of Electromagnetic Processing of Materials (Ministry of Education), Northeastern University, Shenyang 110819 (China); Liu, Tie, E-mail: liutie@epm.neu.edu.cn [Key Laboratory of Electromagnetic Processing of Materials (Ministry of Education), Northeastern University, Shenyang 110819 (China); Dong, Meng [Key Laboratory of Electromagnetic Processing of Materials (Ministry of Education), Northeastern University, Shenyang 110819 (China); Yuan, Yi [School of Materials and Metallurgy, Northeastern University, Shenyang 110819 (China); Wang, Qiang [Key Laboratory of Electromagnetic Processing of Materials (Ministry of Education), Northeastern University, Shenyang 110819 (China)

2016-03-01

In this paper, we studied how applying a high magnetic field during solidification of Tb{sub 0.27}Dy{sub 0.73}Fe{sub 1.95} alloys affected their magnetic domain structure, crystal orientation, and magnetostriction. We observed the morphology of the magnetic domain during solidification, finding it change with the applied field: from fiber like (0 T) to dot like and closure mixed (4.4 T) to fiber like (8.8 T) to fishbone like (11.5 T). The alloy solidified at 4.4 T showed the best contrast of light and dark in its domain image, widest magnetic domain, fastest magnetization, and highest magnetostriction; this alloy is followed in descending order by the alloys solidified at 11.5 T, 8.8 T, and 0 T. The orientation of the (Tb, Dy)Fe{sub 2} phase changed with magnetic field from random (0 T) to 〈111〉 (4.4 T) to 〈113〉 (8.8 T) to 〈110〉 (11.5 T). The improvement in magnetostriction was likely caused by modification of both the magnetization process and the alloy microstructure. - Highlights: • We present how magnetic field affects magnetic domain structure of Tb{sub 0.27}Dy{sub 0.73}Fe{sub 1.95}. • Morphology and width of magnetic domain change with increasing magnetic field. • Magnetization and magnetostriction of alloy change with increasing magnetic field. • A transformation of random–〈111〉–〈113〉–〈110〉 for (Tb, Dy)Fe{sub 2} orientation forms.

Leach testing of simulated ion-exchange resin waste solidified in cement

International Nuclear Information System (INIS)

Muurinen, A.K.; Uotila, P.I.; Ovaskainen, R.M.

Leach tests were carried out on ion-exchange resins solidified in cement. Three product mixtures, two isotopes and four leachants at two temperatures, were tested. The increase of resin content increased the leaching of Cs-137; the effect of silix admixture was negligible. The type of the leachant has a stronger influence on Co-60 than on Cs-137. The increase of temperature usually also increased leaching. (author)

Microtexture formation of Ni{sub 99}B{sub 1} alloys solidified on an ESL and an EML-a study based on the EBSP technique

Energy Technology Data Exchange (ETDEWEB)

Li Mingjun [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Tsukuba Space Center, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan)], E-Mail: li.mingjun@aist.go.jp; Ishikawa, Takehiko [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Tsukuba Space Center, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan); Nagashio, Kosuke [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Sagamihara Campus, 3-1-1 Yoshinodai, Sagamihara, Kanagawa 229-8510 (Japan); Kuribayashi, Kazuhiko [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Sagamihara Campus, 3-1-1 Yoshinodai, Sagamihara, Kanagawa 229-8510 (Japan); Yoda, Shinichi [Japan Aerospace Exploration Agency, Institute of Space and Astronautical Science, Tsukuba Space Center, ISS Science Project Office, 2-1-1 Sengen, Tsukuba, Ibaraki 305-8505 (Japan)

2007-03-25

We employed an electrostatic levitator (ESL) and an electromagnetic levitator (EML) to solidify Ni{sub 99}B{sub 1} (at.%) alloys at various undercoolings. The microstructures and microtextures were revealed by using the electron backscatter diffraction pattern (EBSP) technique in a scanning electron microscope. It is found that that no significant refinement can be identified at the low and medium undercooling regimes for the primary trunk in the sample solidified on the ESL, while the fragmentation of the secondary and even tertiary branches may take place to generate equiaxed grains. Further investigation by the EBSP reveals that neighboring grains have small misorientation angles, which may be ascribed to the absence of mechanical stirring from electromagnetic eddy current. A sharp contrast is that the samples solidified on the EML at low and medium undercoolings have refined equiaxed microstructures. The EBSP mapping reveals that the equiaxed grains yielded on the EML have a random distribution in crystallographic orientations among neighboring grains, indicating that electromagnetic stirring (EMS) induced by the electromagnetic field in the EML plays a vital role in promoting fragmentation and thus generating refined grains and random distribution in orientation. Regarding to the refined microstructure at high undercoolings, no significant difference arises in the samples processed between the EML and ESL.

Decontamination of high-level waste canisters

International Nuclear Information System (INIS)

Nesbitt, J.F.; Slate, S.C.; Fetrow, L.K.

1980-12-01

This report presents evaluations of several methods for the in-process decontamination of metallic canisters containing any one of a number of solidified high-level waste (HLW) forms. The use of steam-water, steam, abrasive blasting, electropolishing, liquid honing, vibratory finishing and soaking have been tested or evaluated as potential techniques to decontaminate the outer surfaces of HLW canisters. Either these techniques have been tested or available literature has been examined to assess their applicability to the decontamination of HLW canisters. Electropolishing has been found to be the most thorough method to remove radionuclides and other foreign material that may be deposited on or in the outer surface of a canister during any of the HLW processes. Steam or steam-water spraying techniques may be adequate for some applications but fail to remove all contaminated forms that could be present in some of the HLW processes. Liquid honing and abrasive blasting remove contamination and foreign material very quickly and effectively from small areas and components although these blasting techniques tend to disperse the material removed from the cleaned surfaces. Vibratory finishing is very capable of removing the bulk of contamination and foreign matter from a variety of materials. However, special vibratory finishing equipment would have to be designed and adapted for a remote process. Soaking techniques take long periods of time and may not remove all of the smearable contamination. If soaking involves pickling baths that use corrosive agents, these agents may cause erosion of grain boundaries that results in rough surfaces

Behavior of radioactive iodine and technetium in the spray calcination of high-level waste

International Nuclear Information System (INIS)

Knox, C.A.; Farnsworth, R.K.

1981-08-01

The Remote Laboratory-Scale Waste Treatment Facility (RLSWTF) was designed and built as a part of the High-Level Waste Immobilization Program (now the High-Level Waste Process Development Program) at the Pacific Northwest Laboratory. In this facility, which is installed in a radiochemical cell, small volumes of radioactive liquid wastes can be solidified, the process off gas can be analyzed, and the methods for decontaminating this off gas can be tested. Initial operations were completed with nonradioactive, simulated waste solutions (Knox, Siemens and Berger 1981). The first radioactive operations in this facility were performed with a simulated, commercial waste composition containing tracer levels of 99 Tc and 131 I. This report describes the facility and test operations and presents the results of the behavior of 131 I and 99 Tc during solidification of radioactive liquid wastes. During the spray calcination of commercial high-level liquid waste spiked with 99 Tc and 131 I, there was a 0.3 wt% loss of particulates, a 0.15 wt% loss of 99 Tc and a 31 wt% loss of 131 I past the sintered-metal filters. These filters and a venturi scrubber were very efficient in removing particulates and 99 Tc from the off-gas stream. Liquid scrubbers were not efficient in removing 131 I, as 25% of the total lost went to the building off-gas system. Therefore, solid adsorbents will be needed to remove iodine. For all future RLSWTF operations where iodine is present, a silver zeolite adsorber will be used

High energy x-radiographic assessment of conditioned intermediate level waste blocks

International Nuclear Information System (INIS)

Lewcock, A.I.; Burch, S.F.; Reynolds, W.N.; Pullen, D.A.W.; Smith, D.

1985-07-01

This report describes an effective technique for examining the quality of the solidification matrix material in a 500 litre waste drum, testing for homogeneity and major cracks and the confirmation of set. A high energy x-ray source, (an 8 MeV Linac) and a special x-ray TV system, were used to examine several different types of solidified waste form, with and without background radiation, simulated by the use of an uncollimated radiographic isotope. The system as tested showed no discernable image degradation when the isotope was positioned to give a representative background dose as experienced with active ILW monoliths. (author)

Leaching behavior of cement solidified materials

International Nuclear Information System (INIS)

2002-03-01

An immersion test of mortar was carried out in order to solidify waste with uranium. The sample consists of 2000g cement, 950g ion exchange water, 1600g sound and 1g water reducing agent. The solid sample and ion exchange water (100 of immersion liquid/original sample) was put into polystyrene closed vessel in globe box and kept four weeks, and then it was separated to the immersion liquid and the solid phase. New ion exchange water was added to the solid and kept four weeks and then separated. Its ratio showed 200. The analysis was done at 100, 200 and 300 ratio of immersion liquid/sample. The solid phase was studied by the powder X-ray diffraction analysis, thermo gravimetric analysis and chemical analysis. The liquid phase was determined by pH values and composition analysis. The results showed Ca(OH) 2 , cement hydrate, was flowed out and it was not found in the solid phase at 200 ratio. (S.Y.)

Glass as a matrix for SRP high-level defense waste

International Nuclear Information System (INIS)

Wiley, J.R.; Bibler, N.E.; Dukes, M.D.; Plodinec, M.J.

1980-01-01

Work done at Savannah River Laboratory and elsewhere that has led to development of glass as a candidate for solidifying Savannah River Plant waste is summarized. Areas of development described are glass formulation and fabrication, and leaching and radiation effects

New evolution on the high level radioactive waste disposal in Japan

International Nuclear Information System (INIS)

Koumoto, Harumi

2001-01-01

On nuclear power generation, spent fuel is formed and reaches to about 30 ton from a 1 million kW class large power plant. As some nations deal with the spent fuel itself to waste, Japan adopts a reprocessing and recycling route to recover uranium and plutonium reusable for nuclear fuels by reprocessing of the spent fuels. As waste liquid containing about one ton of cinder (fission product) formed by nuclear fission after its recovery, a glass solid solidifying this to a stable glassy state is called the high level radioactive wastes (HLW). As it has extremely high radioactivity which continues for long term in spite of its decay with elapsing time, safety security must be paid enough attention to its countermeasure. Therefore, as a result of long-term research and development in Japan as well as in many other nations, it is admitted to be the most preferable countermeasure to bury HLW into deep stratum to safely isolate from human life environment for its scientific and technical method. Here was introduced on a framework of its disposal business in Japan of which preparation rapidly advanced as a turning point of 2000 at a center of its technical and regulative advancement. (G.K.)

Production of solidified high level wastes: a cost comparison of solidification processes

International Nuclear Information System (INIS)

1977-06-01

Differential cost estimates of the annual operating and maintenance costs and the capital costs for five HLW Waste Solidification Alternates were developed. The annual operating and maintenance cost estimates included the cost of labor, consumables, utilities, shipping casks, shipping and disposal at a federal repository. The capital cost included the cost of the component, installation and building. The differential cost estimates do not include equipment and facilities which are either shared with the reprocessing facility or are common between all of the alternates. Total annual cost differential between the five waste form alternates is summarized in tabular form. The Borosilicate Glass Alternate has the lowest total annual cost. The other alternates have higher costs which range from $6.6 M to $7.4 M per year higher than the Glass alternate with the Supercalcine being the highest cost at $7.4 M per year differential. The major items in the cost estimates are then disposal costs in the operating cost estimates and the HLW Storage Tanks in the capital cost estimates. The Supercalcine Multibarrier Alternate ships 180 canisters per year more than the other alternates and consequently has a significantly higher operating cost. However, off-setting this the Supercalcine Multibarrier Alternate does not require HLW Storage Tanks for decay because of the high heat conductivity of this product and correspondingly the capital cost for this alternate is significantly lower than the other alternates. The radiological risk values are correlated with the cost evaluation normalized to cost ($)/MWe-yr

Formation of an 18R long-period stacking ordered structure in rapidly solidified Mg88Y8Zn4 alloy

International Nuclear Information System (INIS)

Garcés, Gerardo; Requena, Guillermo; Tolnai, Domonkos; Pérez, Pablo; Medina, Judit; Stark, Andreas; Schell, Norbert; Adeva, Paloma

2016-01-01

The formation of the long-period stacking ordered structure (LPSO) in a Mg 88 Y 8 Zn 4 (at%) ribbon produced by melt spinning was studied using high energy X-ray synchrotron radiation diffraction during in-situ isochronal heating and transmission electron microscopy. The microstructure of the rapidly solidified ribbons is characterised by fine magnesium grains with yttrium and zinc in solid solution and primary 18R LPSO-phase segregated at grain boundaries. Using differential scanning calorimetry, a strong exothermal peak was observed around 300 °C which was associated with the development of the 18R-type LPSO-phase in the magnesium grains. The apparent activation energy calculated using the Kissinger model was 125 KJmol −1 and it is related to simultaneous diffusion of Y and Zn through magnesium basal plane. - Highlights: •The formation of the LPSO phase in rapidly solidified ribbons was studied. •The formation of the 18R LPSO starts at around 300 °C. •LPSO formation have to steps: Stacking faults along basal plane and then growth of 18R structure along the c direction.

Solidification of low-level radioactive liquid waste using a cement-silicate process

International Nuclear Information System (INIS)

Grandlund, R.W.; Hayes, J.F.

1979-01-01

Extensive use has been made of silicate and Portland cement for the solidification of industrial waste and recently this method has been successfully used to solidify a variety of low level radioactive wastes. The types of wastes processed to date include fuel fabrication sludges, power reactor waste, decontamination solution, and university laboratory waste. The cement-silicate process produces a stable solid with a minimal increase in volume and the chemicals are relatively inexpensive and readily available. The method is adaptable to either batch or continuous processing and the equipment is simple. The solid has leaching characteristics similar to or better than plain Portland cement mixtures and the leaching can be further reduced by the use of ion-exchange additives. The cement-silicate process has been used to solidify waste containing high levels of boric acid, oils, and organic solvents. The experience of handling the various types of liquid waste with a cement-silicate system is described

Properties of rapidly solidified Fe-Cr-Al ribbons for the use as automotive exhaust gas catalyst substrates

International Nuclear Information System (INIS)

Emmerich, K.

1993-01-01

Metallic honeycomb structures are used as catalyst substrates in automotive exhaust gas systems. This application requires an outstanding corrosion resistance at elevated temperatures of the substrate material. Technical improvements can be achieved by the use of rapid solidification technology for the production of the Fe-Cr-Al ribbons since the Al content can be substantially increased from about 5% Al in the conventionally rolled material to about 12% Al in the rapid solidified ribbon. As a result the lifetime of the ribbon in a higher-temperature corrosion environment is drastically increased. In addition the scale/metal adherance is improved. The impediment of recrystallization in the rapidly solidified ribbons prevents an embrittlement even in carbonizing atmospheres. (orig.)

A process for solidifying radioactive liquid waste

International Nuclear Information System (INIS)

Mergan, L.M.; Cordier, J.-P.

1981-01-01

In a process for solidifying radioactive liquid waste, its pH is adjusted, solids precipitated and then it is concentrated to about 50% solids content using a thin film evaporator, the concentrate then being dried to powder in a heated mixer. The mixer has a heated wall and working means, e.g. a rotor and helical screw, to shear the dried concentrate from the internal walls, subdivide it into a dry particulate powder, and advance the powder to the mixer outlet. The dried particles are then encapsulated in a suitable matrix. Vapour from the mixer and evaporator is condensed and recycled after any particles have been removed from it. The mixer may both dry the concentrate and mix the dry particles with the encapsulating matrix, and possibly, part of the mixer may be used for pH adjustment and precipitation. (author)

Sensitive determination of cadmium using solidified floating organic drop microextraction-slotted quartz tube-flame atomic absorption spectroscopy.

Science.gov (United States)

Akkaya, Erhan; Chormey, Dotse Selali; Bakırdere, Sezgin

2017-09-20

In this study, solidified floating organic drop microextraction (SFODME) by 1-undecanol was combined with slotted quartz tube flame atomic absorption spectrometry (SQT-FAAS) for the determination of cadmium at trace levels. Formation of a complex with 4,4'-dimethyl-2,2'-bipyridine facilitated the extraction of cadmium from aqueous solutions. Several chemical variables were optimized in order to obtain high extraction outputs. Parameters such as concentration of the ligand, pH, and amount of buffer solution were optimized to enhance the formation of cadmium complex. The SFODME method was assisted by dispersion of extractor solvent into aqueous solutions using 2-propanol. Under the optimum extraction and instrumental conditions, the limit of detection and limit of quantitation values obtained for cadmium using the combined methods (SFODME-SQT-FAAS) were found to be 0.4 and 1.3 μg L -1 , respectively. Matrix effects on the method were also examined for tap water and wastewater, and spiked recovery results were found to be very satisfactory. Graphical Abstract SFODME-SQT-FAAS system for sensitive determination of cadmium.

Strength, leachability and microstructure characteristics of cement-based solidified plating sludge

International Nuclear Information System (INIS)

Asavapisit, Suwimol; Naksrichum, Siripat; Harnwajanawong, Naraporn

2005-01-01

The solidification of the stabilized zinc-cyanide plating sludge was carried out using ordinary Portland cement (OPC) and pulverized fuel ash (PFA) as solidification binders. The plating sludge were used at the level of 0%, 10%, 20% and 30% dry weight, and PFA was used to replace OPC at 0%, 10%, 20% and 30% dry weight, respectively. Experimental results showed that a significant reduction in strength was observed when the plating sludge was added to both the OPC and OPC/PFA binders, but the negative effect was minimized when PFA was used as part substitute for OPC. SEM observation reveals that the deposition of the plating sludge on the surface of the clinkers and PFA could be the cause for hydration retardation. In addition, calcium zinc hydroxide hydrate complex and the unreacted di- and tricalcium silicates were the major phases in X-ray diffraction (XRD) patterns of the solidified plating waste hydrated for 28 days, although the retardation effect on hydration reactions but Cr concentration in toxicity characteristic leaching procedure (TCLP) leachates was lower than the U.S. EPA regulatory limit

Undercooling and demixing in rapidly solidified Cu-Co alloys

DEFF Research Database (Denmark)

Battezzati, L.; Curiotto, S.; Johnson, Erik

2007-01-01

The Cu–Co system displays a metastable miscibility gap in the liquid state. A considerable amount of work has been performed to study phase separation and related microstructures showing that demixing of the liquid is followed by coagulation before dendritic solidification. Due to kinetic...... competition of transformation phenomena, the mechanisms have not been fully disclosed. This contribution reviews such findings with the help of a computer calculation of the phase diagram and extends the present knowledge by presenting new results obtained by rapidly solidifying various Cu–Co compositions...... using a wide range of cooling rates achieved by forcing the liquid into cylindric and conic moulds and by melt spinning....

Freckle Defect Formation near the Casting Interfaces of Directionally Solidified Superalloys.

Science.gov (United States)

Hong, Jianping; Ma, Dexin; Wang, Jun; Wang, Fu; Sun, Baode; Dong, Anping; Li, Fei; Bührig-Polaczek, Andreas

2016-11-16

Freckle defects usually appear on the surface of castings and industrial ingots during the directional solidification process and most of them are located near the interface between the shell mold and superalloys. Ceramic cores create more interfaces in the directionally solidified (DS) and single crystal (SX) hollow turbine blades. In order to investigate the location of freckle occurrence in superalloys, superalloy CM247 LC was directionally solidified in an industrial-sized Bridgman furnace. Instead of ceramic cores, Alumina tubes were used inside of the casting specimens. It was found that freckles occur not only on the casting external surfaces, but also appear near the internal interfaces between the ceramic core and superalloys. Meanwhile, the size, initial position, and area of freckle were investigated in various diameters of the specimens. The initial position of the freckle chain reduces when the diameter of the rods increase. Freckle area follows a linear relationship in various diameters and the average freckle fraction is 1.1% of cross sectional area of casting specimens. The flow of liquid metal near the interfaces was stronger than that in the interdendritic region in the mushy zone, and explained why freckle tends to occur on the outer or inner surfaces of castings. This new phenomenon suggests that freckles are more likely to occur on the outer or inner surfaces of the hollow turbine blades.

Freckle Defect Formation near the Casting Interfaces of Directionally Solidified Superalloys

Directory of Open Access Journals (Sweden)

Jianping Hong

2016-11-01

Full Text Available Freckle defects usually appear on the surface of castings and industrial ingots during the directional solidification process and most of them are located near the interface between the shell mold and superalloys. Ceramic cores create more interfaces in the directionally solidified (DS and single crystal (SX hollow turbine blades. In order to investigate the location of freckle occurrence in superalloys, superalloy CM247 LC was directionally solidified in an industrial-sized Bridgman furnace. Instead of ceramic cores, Alumina tubes were used inside of the casting specimens. It was found that freckles occur not only on the casting external surfaces, but also appear near the internal interfaces between the ceramic core and superalloys. Meanwhile, the size, initial position, and area of freckle were investigated in various diameters of the specimens. The initial position of the freckle chain reduces when the diameter of the rods increase. Freckle area follows a linear relationship in various diameters and the average freckle fraction is 1.1% of cross sectional area of casting specimens. The flow of liquid metal near the interfaces was stronger than that in the interdendritic region in the mushy zone, and explained why freckle tends to occur on the outer or inner surfaces of castings. This new phenomenon suggests that freckles are more likely to occur on the outer or inner surfaces of the hollow turbine blades.

Effect of tensile mean stress on fatigue behavior of single-crystal and directionally solidified superalloys

Science.gov (United States)

Kalluri, Sreeramesh; Mcgaw, Michael A.

1990-01-01

Two nickel base superalloys, single crystal PWA 1480 and directionally solidified MAR-M 246 + Hf, were studied in view of the potential usage of the former and usage of the latter as blade materials for the turbomachinery of the space shuttle main engine. The baseline zero mean stress (ZMS) fatigue life (FL) behavior of these superalloys was established, and then the effect of tensile mean stress (TMS) on their FL behavior was characterized. At room temperature these superalloys have lower ductilities and higher strengths than most polycrystalline engineering alloys. The cycle stress-strain response was thus nominally elastic in most of the fatigue tests. Therefore, a stress range based FL prediction approach was used to characterize both the ZMS and TMS fatigue data. In the past, several researchers have developed methods to account for the detrimental effect of tensile mean stress on the FL for polycrystalline engineering alloys. However, the applicability of these methods to single crystal and directionally solidified superalloys has not been established. In this study, these methods were applied to characterize the TMS fatigue data of single crystal PWA 1480 and directionally solidified MAR-M 246 + Hf and were found to be unsatisfactory. Therefore, a method of accounting for the TMS effect on FL, that is based on a technique proposed by Heidmann and Manson was developed to characterize the TMS fatigue data of these superalloys. Details of this method and its relationship to the conventionally used mean stress methods in FL prediction are discussed.

Simultaneous extraction and quantification of lamotrigine, phenobarbital, and phenytoin in human plasma and urine samples using solidified floating organic drop microextraction and high-performance liquid chromatography.

Science.gov (United States)

Asadi, Mohammad; Dadfarnia, Shayessteh; Haji Shabani, Ali Mohammad; Abbasi, Bijan

2015-07-01

A novel and simple method based on solidified floating organic drop microextraction followed by high-performance liquid chromatography with ultraviolet detection has been developed for simultaneous preconcentration and determination of phenobarbital, lamotrigine, and phenytoin in human plasma and urine samples. Factors affecting microextraction efficiency such as the type and volume of the extraction solvent, sample pH, extraction time, stirring rate, extraction temperature, ionic strength, and sample volume were optimized. Under the optimum conditions (i.e. extraction solvent, 1-undecanol (40 μL); sample pH, 8.0; temperature, 25°C; stirring rate, 500 rpm; sample volume, 7 mL; potassium chloride concentration, 5% and extraction time, 50 min), the limits of detection for phenobarbital, lamotrigine, and phenytoin were 1.0, 0.1, and 0.3 μg/L, respectively. Also, the calibration curves for phenobarbital, lamotrigine, and phenytoin were linear in the concentration range of 2.0-300.0, 0.3-200.0, and 1.0-200.0 μg/L, respectively. The relative standard deviations for six replicate extractions and determinations of phenobarbital, lamotrigine, and phenytoin at 50 μg/L level were less than 4.6%. The method was successfully applied to determine phenobarbital, lamotrigine, and phenytoin in plasma and urine samples. © 2015 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Encapsulating of high-level radioactive waste with use of pyrocarbon and silicon carbide coatings

International Nuclear Information System (INIS)

Chernikov, A.

2007-01-01

It is known that high-level radioactive waste (HLW) constitute a real danger to biosphere, especially that their part, which contains transuranium and long-lived radionuclides resulting during reprocessing of nuclear fuel industrial and power reactors. Such waste contains approximately 99 % of long-lived fission products and transplutonium elements. At present, the concept of multi barrier protection of biosphere from radioactive waste is generally acknowledged. The main barriers are the physicochemical form of waste and enclosing strata of geological formation at places of waste-disposal. Applied methods of solidification of HLW with preparation of phosphatic and borosilicate glasses do not guarantee in full measure safety of places of waste-disposal of solidified waste in geological formations during thousand years. One promising way to improve HLW handling safety is placing of radionuclides in mineral-like matrixes similar to natural materials. The other possible way to increase safety of HLW disposal places is suggested for research by experts of Russian research institutes, for example, in the proposal for the Project of ISTC and considered in the present report, is to introduce an additional barrier on a radionuclides migration path by coating of HLW particles. Unique protective properties of pyrocarbon and silicon carbide such as low coefficients of diffusion of gaseous and solid fission products and high chemical and radiation stability [1] attract attention to these materials for coating of solidified HLW. The objective of the Project is the development of method of HLW encapsulating with use of pyrocarbon and silicon carbide coatings. To gain this end main direction of researches, including analysis of various encapsulation processes of fractionated HLW, and expected results are presented. Realization of the Project will allow to prove experimentally the efficiency of the proposed approach in the solution of the problem of HLW conditioning and ecological

Microbial degradation of low-level radioactive waste

International Nuclear Information System (INIS)

Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr.

1994-04-01

The Nuclear Regulatory Commission stipulates that disposed low-level radioactive waste (LLW) be stabilized. Because of apparent ease of use and normal structural integrity, cement has been widely used as a binder to solidify LLW. However, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. This report reviews laboratory efforts that are being developed to address the effects of microbiologically influenced chemical attack on cement-solidified LLW. Groups of microorganisms are being employed that are capable of metabolically converting organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with cement and can ultimately lead to structural failure. Results on the application of mechanisms inherent in microbially influenced degradation of cement-based material are the focus of this report. Sufficient data-validated evidence of the potential for microbially influenced deterioration of cement-solidified LLW has been developed during the course of this study. These data support the continued development of appropriate tests necessary to determine the resistance of cement-solidified LLW to microbially induced degradation that could impact the stability of the waste form. They also justify the continued effort of enumeration of the conditions necessary to support the microbiological growth and population expansion

Microstructure of rapidly solidified Al2O3-dispersion-strengthened Type 316 stainless steel

International Nuclear Information System (INIS)

Megusar, J.; Arnberg, L.; Vander Sande, J.B.; Grant, N.J.

1981-01-01

An aluminum oxide dispersion strengthened 316 stainless steel was developed by surface oxidation. Surface oxidation was chosen as a preferred method in order to minimize formation of less stable chromium oxides. Ultra low C+N 316 stainless steel was alloyed with 1 wt % Al, rapidly solidified to produce fine powders and attrited to approximately 0.5 μm thick flakes to provide for surface oxidation. Oxide particles in the extruded material were identified mostly as Al oxides. In the preirradiated condition, oxide dispersion retarded crystallization and grain growth and had an effect on room temperature tensile properties. These structural modifications are expected to have an effect on the swelling resistance, structure stability and high temperature strength of austenitic stainless steels

Structure fields in the solidifying cast iron roll

Directory of Open Access Journals (Sweden)

W.S. Wołczyński

2010-01-01

Full Text Available Some properties of the rolls depend on the ratio of columnar structure area to equiaxed structure area created during roll solidification. The transition is fundamental phenomenon that can be apply to characterize massive cast iron rolls produced by the casting house. As the first step of simulation, a temperature field for solidifying cast iron roll was created. The convection in the liquid is not comprised since in the first approximation, the convection does not influence the studied occurrence of the (columnar to equiaxed grains transition in the roll. The obtained temperature field allows to study the dynamics of its behavior observed in the middle of the mould thickness. This midpoint of the mould thickness was treated as an operating point for the transition. A full accumulation of the heat in the mould was postulated for the transition. Thus, a plateau at the curve was observed at the midpoint. The range of the plateau existence corresponded to the incubation period , that appeared before fully equiaxed grains formation. At the second step of simulation, behavior of the thermal gradients field was studied. Three ranges within the filed were visible: EC→EC→EC→EC→(tTECtt↔RERCtt↔a/ for the formation of columnar structure (the C – zone: ( and 0>>T&0>>=−>−=REREttGttG.The columnar structure formation was significantly slowed down during incubation period. It resulted from a competition between columnar growth and equiaxed growth expected at that period of time. The 0≈=−=RERCttGttG relationship was postulated to correspond well with the critical thermal gradient, known in the Hunt’s theory. A simulation was performed for the cast iron rolls solidifying as if in industrial condition. Since the incubation divides the roll into two zones: C and E; (the first with columnar structure and the second with fully equiaxed structure some experiments dealing with solidification were made on semi-industrial scale.

Centralized cement solidification technique for low-level radioactive wastes

International Nuclear Information System (INIS)

Matsuda, Masami; Nishi, Takashi; Izumida, Tatsuo; Tsuchiya, Hiroyuki.

1996-01-01

A centralized cement solidification system has been developed to enable a single facility to solidify such low-level radioactive wastes as liquid waste, spent ion exchange resin, incineration ash, and miscellaneous solid wastes. Since the system uses newly developed high-performance cement, waste loading is raised and deterioration of waste forms after land burial prevented. This paper describes the centralized cement solidification system and the features of the high-performance cement. Results of full-scale pilot plant tests are also shown from the viewpoint of industrial applicability. (author)

Experimental Study and Application of Inorganic Solidified Foam Filling Material for Coal Mines

Directory of Open Access Journals (Sweden)

Hu Wen

2017-01-01

Full Text Available Spontaneous combustion of residual coal in a gob due to air leakage poses a major risk to mining safety. Building an airtight wall is an effective measure for controlling air leakage. A new type of inorganic solidified foam-filled material was developed and its physical and chemical properties were analyzed experimentally. The compressive strength of this material increased with the amount of sulphoaluminate cement. With an increasing water–cement ratio, the initial setting time was gradually extended while the final setting time firstly shortened and then extended. The change in compressive strength had the opposite tendency. Additionally, as the foam expansion ratio increased, the solidification time tended to decrease but the compressive strength remained approximately constant. With an increase in foam production, the solidification time increased and the compressive strength decreased exponentially. The results can be used to determine the optimal material ratios of inorganic solidified foam-filled material for coal mines, and filling technology for an airtight wall was designed. A field application of the new material demonstrated that it seals crossheadings tightly, leaves no fissures, suppresses air leakage to the gob, and narrows the width of the spontaneous combustion and heat accumulation zone.

Leachability of radionuclides from cement solidified waste forms produced at operating nuclear power plants

International Nuclear Information System (INIS)

Croney, S.T.

1985-03-01

This study determined the leachability indexes of radionuclides contained in solidified liquid wastes from operating nuclear power plants. Different sizes of samples of cement-solidified liquid wastes were collected from two nuclear power plants - a pressurized water reactor and a boiling water reactor - to correlate radionuclide leaching from small- and full-sized (55-gallon) waste forms. Diffusion-based model analysis (ANS 16.1) of measured radionuclide leach data from both small- and full-sized samples was performed and indicate that leach data from small samples can be used to determine leachability indexes for full-sizes waste forms. The leachability indexes for cesium, strontium, and cobalt isotopes were determined for waste samples from both plants according to the models used for ANS 16.1. The leachability indexes for the pressurized water reactor samples were 6.4 for cesium, 7.1 for strontium, and 10.4 for cobalt. Leachability indexes for the boiling water reactor samples were 6.5 for cesium, 8.6 for strontium, and 11.1 for cobalt

Effective hydrogen diffusion coefficient for solidifying aluminium alloys

International Nuclear Information System (INIS)

Felberbaum, M.; Landry-Desy, E.; Weber, L.; Rappaz, M.

2011-01-01

An effective hydrogen diffusion coefficient has been calculated for two solidifying Al - 4.5 wt.% Cu and Al - 10 wt.% Cu alloys as a function of the volume fraction of solid. For this purpose, in situ X-ray tomography was performed on these alloys. For each volume fraction of solid between 0.6 and 0.9, a representative volume element of the microstructure was extracted. Solid and liquid voxels were assimilated to solid and liquid nodes in order to solve the hydrogen diffusion equation based on the chemical potential and using a finite volume formulation. An effective hydrogen diffusion coefficient based on the volume fraction of solid only could be deduced from the results of the numerical model at steady state. The results are compared with various effective medium theories.

Leaching test of bituminized waste and waste solidified by epoxy resin

International Nuclear Information System (INIS)

Yoshinaka, Kazuyuki; Sugaya, Atsushi; Onizawa, Toshikazu; Takano, Yugo; Kimura, Yukihiko

2008-10-01

About 30,000 bituminized waste drums and about 1800 drums of waste solidified by epoxy resin, generated from Tokai Reprocessing Plant, were stored in storage facilities. And study for disposal of these waste is performed. It was considered that radioactive nuclides and chemical components were released from these waste by contact of underground water, when disposed there waste. This paper is reported that result of leaching tests for these waste, done from 2003 to 2006. We've get precious knowledge and data, as follows. (1) In leaching tests for bituminized waste, it has detected iodine-129 peak, considered difficult too low energy gamma to detect. We've get data and knowledge of iodine-129 behavior first. Leached radioactivity for 50 days calculated by peak area was equal for about 40% and 100% of including radioactivity in bituminized waste sample. And we've get data of behavior of nitric acid ion and so on, important to study for disposal, in various condition of sample shape or leaching liquid temperature. (2) In leaching test for waste solidified by epoxy resin, we've get data of behavior of TBP, radionuclides and so on, important to study for disposal. Leached TBP was equal about 1% of including of sample. And we've get data of iodine-129 behavior, too. It was confirmed that leached iodine-129 was equal for about 60% and 100% of including sample, for 90 days. (author)

Studies of the Influence of Beam Profile and Cooling Conditions on the Laser Deposition of a Directionally-Solidified Superalloy

Directory of Open Access Journals (Sweden)

Shuo Yang

2018-02-01

Full Text Available In the laser deposition of single crystal and directionally-solidified superalloys, it is desired to form laser deposits with high volume fractions of columnar grains by suppressing the columnar-to-equiaxed transition efficiently. In this paper, the influence of beam profile (circular and square shapes and cooling conditions (natural cooling and forced cooling on the geometric morphology and microstructure of deposits were experimentally studied in the laser deposition of a directionally-solidified superalloy, IC10, and the mechanisms of influence were revealed through a numerical simulation of the thermal processes during laser deposition. The results show that wider and thinner deposits were obtained with the square laser beam than those with the circular laser beam, regardless of whether natural or forced cooling conditions was used. The heights and contact angles of deposits were notably increased due to the reduced substrate temperatures by the application of forced cooling for both laser beam profiles. Under natural cooling conditions, columnar grains formed epitaxially at both the center and the edges of the deposits with the square laser beam, but only at the center of the deposits with the circular laser beam; under forced cooling conditions, columnar grains formed at both the center and the edges of deposits regardless of the laser beam profile. The high ratios of thermal gradient and solidification velocity in the height direction of the deposits were favorable to forming deposits with higher volume fractions of columnar grains.

Method of solidifying radioactive waste by plastics

International Nuclear Information System (INIS)

Yasumura, Keijiro; Tomita, Toshihide.

1976-01-01

Purpose: To prevent leakage of radioactivity by providing corrosion-resistant layer on the inner surface of a waste container for radioactive waste. Constitution: The inner periphery and bottom of a drum can is lined with an non-flammable cloth of such material as asbestos. This drum is filled with a radioactive waste in the form of powder or pellets. Then, a mixture of a liquid plastic monomer and a polymerization starting agent is poured at a normal temperature, and the surface is covered with a non-flammable cloth. The plastic monomer and radioactive waste are permitted to impregnate the non-flammable cloth and are solidified there. Thus, even if the drum can is corroded at the sea bottom after disposal it in the ocean, it is possible to prevent the waste from permeating into the outer sea water because of the presence of the plastic layer on the inside. Styrene is used as the monomer. (Aizawa, K.)

Method for accelerated leaching of solidified waste

International Nuclear Information System (INIS)

Fuhrmann, M.; Heiser, J.H.; Pietrzak, R.F.; Franz, E.M.; Colombo, P.

1990-11-01

An accelerated leach test method has been developed to determine the maximum leachability of solidified waste. The approach we have taken is to use a semi-dynamic leach test; that is, the leachant is sampled and replaced periodically. Parameters such as temperature, leachant volume, and specimen size are used to obtain releases that are accelerated relative to other standard leach tests and to the leaching of full-scale waste forms. The data obtained with this test can be used to model releases from waste forms, or to extrapolate from laboratory-scale to full-scale waste forms if diffusion is the dominant leaching mechanism. Diffusion can be confirmed as the leaching mechanism by using a computerized mathematical model for diffusion from a finite cylinder. We have written a computer program containing several models including diffusion to accompany this test. The program and a Users' Guide that gives screen-by-screen instructions on the use of the program are available from the authors. 14 refs., 4 figs., 1 tab

FY 1999 report on the results of R and D projects by local consortiums for immediate effects. R and D regarding high quality/high performance of lithium tantalate single crystal's solidifying growth and SAW wafer; 1999 nendo sankabutsu tankessho no ikusei to wafer no kohinshitsu konoritsuka ni kansuru kenkyu kaihatsu seika hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-05-01

The R and D project has been implemented for establishing, e.g., methods for growing oxide single crystals (e.g., lithium tantalate, LiTaO{sub 3}) to have the large and/or long products, technologies for polishing/cleaing the wafer products, and technologies for evaluating device performance. For solidifying growth and production technologies for lithium tantalate single crystals, pulling-up of the single crystal, 154 mm in body length and 12.9 kg, is succeeded by reducing temperature gradient at the crystal solid-liquid interface, increasing oxygen concentration, and improving the seed-sustaining system. Bright prospects have been obtained for the automated crystal pulling-up system, and high-precision control of crystal weight. For technologies for polishing/cleaning the wafers, the investigated cleaning methods include ELID polishing, mechanochemical polishing, and supersonic cleaning which uses two frequency bands of multi-supersonic and megasonic waves. For development of the technologies for evaluation/examination of the highly functional devices, the non-contact type method has been developed, which can measure the absolute level of SAW speed at a high speed and precision. (NEDO)

Electron microscopy investigations of rapidly solidified Fe-Zr-B-Cu alloys

International Nuclear Information System (INIS)

Majumdar, B.; Arvindha Babu, D.; Akhtar, D.

2010-01-01

Rapidly solidified Fe-based nanocrystalline soft magnetic materials possess a unique combination of properties i,e high permeability, saturation and Curie temperature and very low coercivity which are otherwise not attainable in conventional soft magnetic materials. The alloys are processed by producing amorphous phase through melt spinning route followed by a partial devitrification for incorporation of nanocrystalline phase in the amorphous matrix. In this paper, detailed electron microscopic investigations of melt spun Fe-Zr-B-Cu alloys are presented. Melt spun ribbons of Fe 99-x-y Zr x BCu 1 alloys with x+y = 11 and x+y = 13 were prepared under different wheel speed conditions and then vacuum annealed for 1 h at different temperatures. The microstructure changes from completely amorphous to a cellular/dendritic bcc solid solution coexisting with the amorphous phase at intercellular/dendritic regions when Zr/B ratio or the process parameters are varied. Annealing leads to the precipitation of nanocrystalline bcc-Fe phase from both amorphous phase and already existing bcc solid solution. (author)

Solidified structure and leaching properties of metallurgical wastewater treatment sludge after solidification/stabilization process.

Science.gov (United States)

Radovanović, Dragana Đ; Kamberović, Željko J; Korać, Marija S; Rogan, Jelena R

2016-01-01

The presented study investigates solidification/stabilization process of hazardous heavy metals/arsenic sludge, generated after the treatment of the wastewater from a primary copper smelter. Fly ash and fly ash with addition of hydrated lime and Portland composite cement were studied as potential binders. The effectiveness of the process was evaluated by unconfined compressive strength (UCS) testing, leaching tests (EN 12457-4 and TCLP) and acid neutralization capacity (ANC) test. It was found that introduction of cement into the systems increased the UCS, led to reduced leaching of Cu, Ni and Zn, but had a negative effect on the ANC. Gradual addition of lime resulted in decreased UCS, significant reduction of metals leaching and high ANC, due to the excess of lime that remained unreacted in pozzolanic reaction. Stabilization of more than 99% of heavy metals and 90% of arsenic has been achieved. All the samples had UCS above required value for safe disposal. In addition to standard leaching tests, solidificates were exposed to atmospheric conditions during one year in order to determine the actual leaching level of metals in real environment. It can be concluded that the EN 12457-4 test is more similar to the real environmental conditions, while the TCLP test highly exaggerates the leaching of metals. The paper also presents results of differential acid neutralization (d-AN) analysis compared with mineralogical study done by scanning electron microscopy and X-ray diffraction analysis. The d-AN coupled with Eh-pH (Pourbaix) diagrams were proven to be a new effective method for analysis of amorphous solidified structure.

Setting of cesium residual ratio of molten solidified waste produced in Japan Atomic Power Company Tokai and Tokai No.2 Power Stations

International Nuclear Information System (INIS)

2013-02-01

JNES investigated the appropriateness of a view of the Japan Nuclear Fuel Co. on cesium residual content and the radioactivity measurement precision regarding the molten solidified (with lowered inorganic salt used) radioactive wastes which were produced from Japan Atomic Power Company Tokai and Tokai No. 2 Power Stations. Based on the written performance report from the request and past disposal confirmation experience, a view of the JNFC is confirmed as appropriate that setting of 15% cesium residual ratio for molten solidified with volume ratio larger than 4% and less than 10% cases. (S. Ohno)

A coupled analysis of fluid flow, heat transfer and deformation behavior of solidifying shell in continuously cast beam blank

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Eui; Yeo, Tae Jung; Oh, Kyu Hwan; Yoon, Jong Kyu [School of Materials Science and Engineering, Seoul Nat` l Univ., Seoul (Korea, Republic of); Han, Heung Nam [Oxford Center for Advanced Materials and Composites, Department of Materials, Univ. of Oxford (United Kingdom)

1998-12-31

A mathematical model for a coupled analysis of fluid flow, heat transfer and deformation behavior in the continuously cast beam blank has been developed. The fluid flow, heat transfer and solidification in the mold region were analyzed with 3-dimensional finite difference method (FDM) based on control volume method. A body fitted coordinate system was introduced for the complex geometry of the beam blank. The effects of turbulence and natural convection of molten steel were taken into account in determining the fluid flow in the strand. The thermo-elasto-plastic deformation behavior in the cast strand and the formation of air gap between the solidifying shell and the mold were analyzed by the finite element method (FEM) using the 2-dimensional slice temperature profile calculated by the FDM. The heat flow between the strand and the mold was evaluated by the coupled analysis between the fluid flow-heat transfer analysis and the thermo-elasto-plastic stress analysis. In order to determine the solid fraction in the mushy zone, the microsegregation of solute element was assessed. The effects of fluid flow on the heat transfer, the solidification of steel and the distribution of shell thickness during the casting of the beam blank were simulated. The deformation behavior of the solidifying shell and the possibility of cracking of the strand were also investigated. The recirculating flows were developed in the regions of the web and the flange tip. The impinging of the inlet flow from the nozzle retarded the growing of solidifying shell in the regions of the fillet and the flange. The air gap between the strand and the mold was formed near the region of the corner of the flange tip. At the initial stage of casting, the probability of the surface cracking was high in the regions of the fillet and the flange tip. After the middle stage of casting, the internal cracking was predicted in the regions of the flange tip, and between the fillet and the flange tip. (author) 38

A coupled analysis of fluid flow, heat transfer and deformation behavior of solidifying shell in continuously cast beam blank

Energy Technology Data Exchange (ETDEWEB)

Lee, Jung Eui; Yeo, Tae Jung; Oh, Kyu Hwan; Yoon, Jong Kyu [School of Materials Science and Engineering, Seoul Nat`l Univ., Seoul (Korea, Republic of); Han, Heung Nam [Oxford Center for Advanced Materials and Composites, Department of Materials, Univ. of Oxford (United Kingdom)

1997-12-31

A mathematical model for a coupled analysis of fluid flow, heat transfer and deformation behavior in the continuously cast beam blank has been developed. The fluid flow, heat transfer and solidification in the mold region were analyzed with 3-dimensional finite difference method (FDM) based on control volume method. A body fitted coordinate system was introduced for the complex geometry of the beam blank. The effects of turbulence and natural convection of molten steel were taken into account in determining the fluid flow in the strand. The thermo-elasto-plastic deformation behavior in the cast strand and the formation of air gap between the solidifying shell and the mold were analyzed by the finite element method (FEM) using the 2-dimensional slice temperature profile calculated by the FDM. The heat flow between the strand and the mold was evaluated by the coupled analysis between the fluid flow-heat transfer analysis and the thermo-elasto-plastic stress analysis. In order to determine the solid fraction in the mushy zone, the microsegregation of solute element was assessed. The effects of fluid flow on the heat transfer, the solidification of steel and the distribution of shell thickness during the casting of the beam blank were simulated. The deformation behavior of the solidifying shell and the possibility of cracking of the strand were also investigated. The recirculating flows were developed in the regions of the web and the flange tip. The impinging of the inlet flow from the nozzle retarded the growing of solidifying shell in the regions of the fillet and the flange. The air gap between the strand and the mold was formed near the region of the corner of the flange tip. At the initial stage of casting, the probability of the surface cracking was high in the regions of the fillet and the flange tip. After the middle stage of casting, the internal cracking was predicted in the regions of the flange tip, and between the fillet and the flange tip. (author) 38

Solidification of high-level radioactive wastes. Final report

International Nuclear Information System (INIS)

1979-06-01

A panel on waste solidification was formed at the request of the Nuclear Regulatory Commission to study the scientific and technological problems associated with the conversion of liquid and semiliquid high-level radioactive wastes into a stable form suitable for transportation and disposition. Conclusions reached and recommendations made are as follows. Many solid forms described in this report could meet standards as stringent as those currently applied to the handling, storage, and transportation of spent fuel assemblies. Solid waste forms should be selected only in the context of the total radioactive waste management system. Many solid forms are likely to be satisfactory for use in an appropriately designed system, The current United States policy of deferring the reprocessing of commercial reactor fuel provides additional time for R and D solidification technology for this class of wastes. Defense wastes which are relatively low in radioactivity and thermal power density can best be solidified by low-temperature processes. For solidification of fresh commercial wastes that are high in specific activity and thermal power density, the Panel recommends that, in addition to glass, the use of fully-crystalline ceramics and metal-matrix forms be actively considered. Preliminary analysis of the characteristics of spent fuel pins indicates that they may be eligible for consideration as a waste form. Because the differences in potential health hazards to the public resulting from the use of various solid form and disposal options are likely to be small, the Panel concludes that cost, reliability, and health hazards to operating personnel will be major considerations in choosing among the options that can meet safety requiremens. The Panel recommends that responsibility for all radioactive waste management operations (including solidification R and D) should be centralized

Simultaneous extraction and determination of albendazole and triclabendazole by a novel syringe to syringe dispersive liquid phase microextraction-solidified floating organic drop combined with high performance liquid chromatography.

Science.gov (United States)

Asadi, Mohammad; Dadfarnia, Shayessteh; Haji Shabani, Ali Mohammad

2016-08-17

A syringe to syringe dispersive liquid phase microextraction-solidified floating organic drop was introduced and used for the simultaneous extraction of trace amounts of albendazole and triclabendazole from different matrices. The extracted analytes were determined by high performance liquid chromatography along with fluorescence detection. The analytical parameters affecting the microextraction efficiency including the nature and volume of the extraction solvent, sample volume, sample pH, ionic strength and the cycles of extraction were optimized. The calibration curves were linear in the range of 0.1-30.0 μg L(-1) and 0.2-30.0 μg L(-1) with determination coefficients of 0.9999 and 0.9998 for albendazole and triclabendazole respectively. The detection limits defined as three folds of the signal to noise ratio were found to be 0.02 μg L(-1) for albendazole and 0.06 μg L(-1) for triclabendazole. The inter-day and intra-day precision (RSD%) for both analytes at three concentration levels (0.5, 2.0 and 10.0 μg L(-1)) were in the range of 6.3-10.1% and 5.0-7.5% respectively. The developed method was successfully applied to determine albendazole and triclabendazole in water, cow milk, honey, and urine samples. Copyright © 2016. Published by Elsevier B.V.

Measurements of Mercury Released From Solidified/Stabilized Waste Forms-FY2002

International Nuclear Information System (INIS)

Mattus, C.H.

2003-01-01

This report covers work performed during FY 2002 in support of treatment demonstrations conducted for the U.S. Department of Energy (DOE) Mixed Waste Focus Area (MWFA) Mercury Working Group. To comply with the requirements of the Resource Conservation and Recovery Act, as implemented by the U.S. Environmental Protection Agency (EPA), DOE must use one of the following procedures for mixed low-level radioactive wastes containing mercury at levels above 260 ppm: a retorting/roasting treatment or (if the wastes also contain organics) an incineration treatment. The recovered radioactively contaminated mercury must then be treated by an amalgamation process prior to disposal. The DOE MWFA Mercury Working Group is working with EPA to determine whether some alternative processes could be used to treat these types of waste directly, thereby avoiding a costly recovery step for DOE. In previous years, demonstrations were performed in which commercial vendors applied their technologies for the treatment of radiologically contaminated elemental mercury as well as radiologically contaminated and mercury-contaminated waste soils from Brookhaven National Laboratory. The test results for mercury release in the headspace were reported in two reports, ''Measurements of Mercury Released from Amalgams and Sulfide Compounds'' (ORNL/TM-13728) and ''Measurements of Mercury Released from Solidified/Stabilized Waste Forms'' (ORNL/TM-2001/17). The current work did not use a real waste; a surrogate sludge had been prepared and used in the testing in an effort to understand the consequences of mercury speciation on mercury release

Examination of solidified and stabilized matrices as a result of solidification and stabilization process of arseniccontaining sludge with portland cement and lime

Directory of Open Access Journals (Sweden)

Tanapon Phenrat

2004-02-01

Full Text Available By solidification and stabilization (S/S with Portland cement and lime, it is possible to reduce arsenic concentration in leachate of the arsenic-containing sludge from arsenic removal process by coagulation with ferric chloride. From the initial arsenic concentration in leachate of unsolidified /unstabilized sludge which was around 20.75 mg/L, the arsenic concentrations in leachate of solidified/stabilized waste were reduced to 0.3, 0.58, 1.09, and 1.85 mg/L for the waste-to-binder ratios of 0.15, 0.25, 0.5, and 1, respectively, due tothe formation of insoluble calcium-arsenic compounds. To be more cost effective for the future, alternative uses of these S/S products were also assessed by measurement of compressive strength of the mortar specimens. It was found that the compressive strengths of these matrices were from 28 ksc to 461 ksc. In conclusion, considering compressive strength and leachability of the solidified matrices, some of these solidified/ stabilized products have potential to serve as an interlocking concrete paving block.

FY 1999 report on the results of R and D projects by local consortiums for immediate effects. R and D regarding high quality/high performance of lithium tantalate single crystal's solidifying growth and SAW wafer; 1999 nendo sankabutsu tankessho no ikusei to wafer no kohinshitsu konoritsuka ni kansuru kenkyu kaihatsu seika hokokusho

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-05-01

The R and D project has been implemented for establishing, e.g., methods for growing oxide single crystals (e.g., lithium tantalate, LiTaO{sub 3}) to have the large and/or long products, technologies for polishing/cleaing the wafer products, and technologies for evaluating device performance. For solidifying growth and production technologies for lithium tantalate single crystals, pulling-up of the single crystal, 154 mm in body length and 12.9 kg, is succeeded by reducing temperature gradient at the crystal solid-liquid interface, increasing oxygen concentration, and improving the seed-sustaining system. Bright prospects have been obtained for the automated crystal pulling-up system, and high-precision control of crystal weight. For technologies for polishing/cleaning the wafers, the investigated cleaning methods include ELID polishing, mechanochemical polishing, and supersonic cleaning which uses two frequency bands of multi-supersonic and megasonic waves. For development of the technologies for evaluation/examination of the highly functional devices, the non-contact type method has been developed, which can measure the absolute level of SAW speed at a high speed and precision. (NEDO)

Solidified structure of Al-Pb-Cu alloys

International Nuclear Information System (INIS)

Ikeda, Tetsuyuki; Nishi, Seiki; Kumeuchi, Hiroyuki; Tatsuta, Yoshinori.

1986-01-01

Al-Pb-Cu alloys were cast into bars or plates in different two metal mold casting processes in order to suppress gravity segregation of Pb and to achieve homogeneous dispersion of Pb phase in the alloys. Solidified structures were analyzed by a video-pattern-analyzer. Plate castings 15 to 20 mm in thickness of Al-Pb-1 % Cu alloy containing Pb up to 5 % in which Pb phase particles up to 10 μm disperse are achieved through water cooled metal mold casting. The plates up to 5 mm in thickness containing Pb as much as 8 to 10 % cast in this process have dispersed Pb particles up to 5 μm in diameter in the surface layer. Al-8 % Pb-1 % Cu alloy bars 40 mm in diameter and 180 mm in height in which gravity segregation of Pb is prevented can be cast by movable and water sprayed metal mold casting at casting temperature 920 deg C and mold moving speed 1.0 mm/s. Pb phase particles 10 μm in mean size are dispersed in the bars. (author)

Phase composition of rapidly solidified Ag-Sn-Cu dental alloys

International Nuclear Information System (INIS)

Lecong Dzuong; Do Minh Nghiep; Nguyen van Dzan; Cao the Ha

1996-01-01

The phase composition of some rapidly solidified Ag-Sn-Cu dental alloys with different copper contents (6.22 wtpct) has been studied by XRD, EMPA and optical microscopy. The samples were prepared from melt-spun ribbons. The microstructure of the as-quenched ribbons was microcrystalline and consisted of the Ag sub 3 Sn, Ag sub 4 Sn, Cu sub 3 Sn and Cu sub 3 Sn sub 8 phases. Mixing with mercury (amalgamation) led to formation of the Ag sub 2 Hg sub 3, Sn sub 7 Hg and Cu sub 6 Sn sub 5 phases. The amount of copper atoms in the alloys played an important role in phase formation in the amalgams

Measurements of Mercury Released from Solidified/Stabilized Waste Forms

International Nuclear Information System (INIS)

Mattus, C.H.

2001-01-01

This report covers work performed during FY 1999-2000 in support of treatment demonstrations conducted for the Mercury Working Group of the U.S. Department of Energy (DOE) Mixed Waste Focus Area. In order to comply with the requirements of the Resource Conservation and Recovery Act, as implemented by the U.S. Environmental Protection Agency (EPA), DOE must use one of these procedures for wastes containing mercury at levels above 260 ppm: a retorting/roasting treatment or an incineration treatment (if the wastes also contain organics). The recovered radioactively contaminated mercury must then be treated by an amalgamation process prior to disposal. The DOE Mixed Waste Focus Area and Mercury Working Group are working with the EPA to determine if some alternative processes could treat these types of waste directly, thereby avoiding for DOE the costly recovery step. They sponsored a demonstration in which commercial vendors applied their technologies for the treatment of two contaminated waste soils from Brookhaven National Laboratory. Each soil was contaminated with ∼4500 ppm mercury; however, one soil had as a major radioelement americium-241, while the other contained mostly europium-152. The project described in this report addressed the need for data on the mercury vapor released by the solidified/stabilized mixed low-level mercury wastes generated during these demonstrations as well as the comparison between the untreated and treated soils. A related work began in FY 1998, with the measurement of the mercury released by amalgamated mercury, and the results were reported in ORNL/TM-13728. Four treatments were performed on these soils. The baseline was obtained by thermal treatment performed by SepraDyne Corp., and three forms of solidification/stabilization were employed: one using sulfur polymer cement (Brookhaven National Laboratory), one using portland cement [Allied Technology Group (ATG)], and a third using proprietary additives (Nuclear Fuel Services)

Evaluation of physical stability and leachability of Portland Pozzolona Cement (PPC) solidified chemical sludge generated from textile wastewater treatment plants

International Nuclear Information System (INIS)

Patel, Hema; Pandey, Suneel

2012-01-01

Highlights: ► Stabilization/solidification of chemical sludge from textile wastewater treatment plants using Portland Pozzolona Cement (PPC) containing fly ash. ► Physical engineering (compressive strength and block density) indicates that sludge has potential to be reused for construction purpose after stabilization/solidification. ► Leaching of heavy metals from stabilized/solidified materials were within stipulated limits. ► There is a modification of microstructural properties of PPC with sludge addition as indicated by XRD and SEM patterns. - Abstract: The chemical sludge generated from the treatment of textile dyeing wastewater is a hazardous waste as per Indian Hazardous Waste Management rules. In this paper, stabilization/solidification of chemical sludge was carried out to explore its reuse potential in the construction materials. Portland Pozzolona Cement (PPC) was selected as the binder system which is commercially available cement with 10–25% fly ash interground in it. The stabilized/solidified blocks were evaluated in terms of unconfined compressive strength, block density and leaching of heavy metals. The compressive strength (3.62–33.62 MPa) and block density (1222.17–1688.72 kg/m 3 ) values as well as the negligible leaching of heavy metals from the stabilized/solidified blocks indicate that there is a potential of its use for structural and non-structural applications.

Disposal of low-level radioactive waste using high-calcium fly ash. Final report

International Nuclear Information System (INIS)

Cogburn, C.O.; Hodgson, L.M.; Ragland, R.C.

1986-04-01

The feasibility of using calcium-rich fly ash from coal-fired power plants in the disposal of low-level radioactive waste was examined. The proposed areas of use were: (1) fly-ash cement as a trench lining material; (2) fly ash as a backfill material; and (3) fly ash as a liquid waste solidifier. The physical properties of fly-ash cement were determined to be adequate for trench liner construction, with compressive strengths attaining greater than 3000 psi. Hydraulic conductivities were determined to be less than that for clay mineral deposits, and were on the order of 10 -7 cm/sec, with some observed values as low as 10 -9 cm/sec. Removal of radioisotopes from acidified solutions by fly ash was good for all elements tested except cesium. The removal of cesium by fly ash was similar to that of montmorillonite clay. The corrosive effects on metals in fly ash environments was determined to be slight, if not non-existent. Coatings at the fly-ash/metal interfaces were observed which appeared to inhibit or diminish corrosion. The study has indicated that high-calcium fly ash appears to offer considerable potential for improved retention of low-level radioactive wastes in shallow land disposal sites. Further tests are needed to determine optimum methods of use. 8 refs., 4 figs., 7 tabs

Phase formation kinetics, hardness and magnetocaloric effect of sub-rapidly solidified LaFe11.6Si1.4 plates during isothermal annealing

Science.gov (United States)

Dai, Yuting; Xu, Zhishuai; Luo, Zhiping; Han, Ke; Zhai, Qijie; Zheng, Hongxing

2018-05-01

High-temperature phase transition behavior and intrinsic brittleness of NaZn13-type τ1 phase in La-Fe-Si magnetocaloric materials are two key problems from the viewpoint of materials production and practical applications. In the present work, the Johnson-Mehl-Avrami-Kolmogorov (JMAK) equation was introduced to quantitatively characterize the formation kinetics of τ1 phase in sub-rapidly solidified LaFe11.6Si1.4 plates during the isothermal annealing process. Avrami index was estimated to be 0.43 (∼0.5), which suggests that the formation of τ1 phase is in a diffusion-controlled one-dimensional growth mode. Meanwhile, it is found that the Vickers hardness as a function of annealing time for sub-rapidly solidified plates also agrees well with the JMAK equation. The Vickers hardness of τ1 phase was estimated to be about 754. Under a magnetic field change of 30 kOe, the maximum magnetic entropy change was about 22.31 J/(kg·K) for plates annealed at 1323 K for 48 h, and the effective magnetic refrigeration capacity reached 191 J/kg.

Influence of Thermal Parameters, Microstructure, and Morphology of Si on Machinability of an Al–7.0 wt.% Si Alloy Directionally Solidified

Directory of Open Access Journals (Sweden)

Cássio A. P. Silva

2018-01-01

Full Text Available This study aims to correlate the influence of thermal and microstructural parameters such as growth rate and cooling rate (VL and TR and secondary dendrite spacing (λ2, respectively, in the machining cutting temperature and tool wear on the necking process of the Al–7 wt.% Si alloy solidified in a horizontal directional device using a high-speed steel with a tungsten tool. The dependence of λ2 on VL and TR and dependence of the maximum cutting temperature and maximum flank wear on λ2 were determined by power experimental laws given by λ2 = constant (VL and TRn and TMAX, VBMAX = constant (λ2n, respectively. The maximum cutting temperature increased with increasing of λ2. The opposite occurred with the maximum flank wear. The role of Si alloying element on the aforementioned results has also been analyzed. A morphological change of Si along the solidified ingot length has been observed, that is, the morphology of Si in the eutectic matrix has indicated a transition from particles to fibers along the casting together with an increase of the particle diameters with the position from the metal/mold interface.

Characteristics of solidified products containing radioactive molten salt waste.

Science.gov (United States)

Park, Hwan-Seo; Kim, In-Tae; Cho, Yong-Zun; Eun, Hee-Chul; Kim, Joon-Hyung

2007-11-01

The molten salt waste from a pyroprocess to recover uranium and transuranic elements is one of the problematic radioactive wastes to be solidified into a durable wasteform for its final disposal. By using a novel method, named as the GRSS (gel-route stabilization/solidification) method, a molten salt waste was treated to produce a unique wasteform. A borosilicate glass as a chemical binder dissolves the silicate compounds in the gel products to produce one amorphous phase while most of the phosphates are encapsulated by the vitrified phase. Also, Cs in the gel product is preferentially situated in the silicate phase, and it is vitrified into a glassy phase after a heat treatment. The Sr-containing phase is mainly phosphate compounds and encapsulated by the glassy phase. These phenomena could be identified by the static and dynamic leaching test that revealed a high leach resistance of radionuclides. The leach rates were about 10(-3) - 10(-2) g/m2 x day for Cs and 10(-4) - 10(-3) g/m2 x day for Sr, and the leached fractions of them were predicted to be 0.89% and 0.39% at 900 days, respectively. This paper describes the characteristics of a unique wasteform containing a molten salt waste and provides important information on a newly developed immobilization technology for salt wastes, the GRSS method.

International program to study subseabed disposal of high-level radioactive wastes

International Nuclear Information System (INIS)

Carlin, E.M.; Hinga, K.R.; Knauss, J.A.

1984-01-01

This report provides an overview of the international program to study seabed disposal of nuclear wastes. Its purpose is to inform legislators, other policy makers, and the general public as to the history of the program, technological requirements necessary for feasibility assessment, legal questions involved, international coordination of research, national policies, and research and development activities. Each of these major aspects of the program is presented in a separate section. The objective of seabed burial, similar to its continental counterparts, is to contain and to isolate the wastes. The subseabed option should not be confuesed with past practices of ocean dumping which have introduced wastes into ocean waters. Seabed disposal refers to the emplacement of solidified high-level radioactive waste (with or without reprocessing) in certain geologically stable sediments of the deep ocean floor. Specially designed surface ships would transport waste canisters from a port facility to the disposal site. Canisters would be buried from a few tens to a few hundreds of meters below the surface of ocean bottom sediments, and hence would not be in contact with the overlying ocean water. The concept is a multi-barrier approach for disposal. Barriers, including waste form, canister, ad deep ocean sediments, will separate wastes from the ocean environment. High-level wastes (HLW) would be stabilized by conversion into a leach-resistant solid form such as glass. This solid would be placed inside a metallic canister or other type of package which represents a second barrier. The deep ocean sediments, a third barrier, are discussed in the Feasibility Assessment section. The waste form and canister would provide a barrier for several hundred years, and the sediments would be relied upon as a barrier for thousands of years. 62 references, 3 figures, 2 tables

Vitrification testing of simulated high-level radioactive waste at Hanford

International Nuclear Information System (INIS)

Perez, J.M. Jr.; Nakaoka, R.R.

1986-03-01

The Hanford Waste Vitrification Plant may apply vitrification technology, being developed at Pacific Northwest Laboratory, to solidify selected Hanford waste streams prior to disposal in a federal repository. Based on the first stage of flowsheet development and laboratory testing, a reference working glass and two candidate simulated feed slurries were recommended for vitrification testing. Over 500 hours of melter testing were performed in 1985 during prototype vitrification experiments. Testing demonstrated that the slurry compositions had acceptable processing characteristics in a ceramic melter. A pre-made glass-former frit was determined to be preferred as the method of glass-former addition. Due to a high chromium content in the waste, spinal crystal formation and settling occurred in the glass tank. The nature and extent of off-gas effluents were consistent with past experiments processing slurries containing formic acid

System design for retrieval of solidified high-level wastes at Hanford

International Nuclear Information System (INIS)

Wallskog, H.A.

1977-01-01

A Waste Retrieval System has been conceptually designed as a step in the process toward the demonstration of the capability to retrieve the projected 36,000,000 gallons of radioactive salt cake and sludge wastes from underground storage tanks at Hanford. This functionally complete, totally remotely operable system consists of a large mobile platform containing all of the tools and equipment necessary to recover, remove and package the wastes for transfer to an onsite processing facility

Analysis of factors influencing the reliability of retrievable storage canisters for containment of solid high-level radioactive waste

International Nuclear Information System (INIS)

Mecham, W.J.; Seefeldt, W.B.; Steindler, M.J.

1976-08-01

The reliability of stainless steel type 304L canisters for the containment of solidified high-level radioactive wastes in the glass and calcine forms was studied. A reference system, drawn largely from information furnished by Battelle Northwest Laboratories and Atlantic Richfield Hanford Company is described. Operations include filling the canister with the appropriate waste form, interim storage at a reprocessing plant, shipment in water to a Retrievable Surface Storage Facility (RSSF), interim storage at the RSSF, and shipment to a final disposal facility. The properties of stainless steel type 304L, fission product oxides, calcine, and glass were reviewed, and mechanisms of corrosion were identified and studied. The modes of corrosion important for reliability were stress-corrosion cracking, internal pressurization of the canister by residual impurities present, intergranular attack at the waste-canister interface, and potential local effects due to migration of fission products. The key role of temperature control throughout canister lifetime is considered together with interactive effects. Methods of ameliorating adverse effects and ensuring high reliability are identified and described. Conclusions and recommendations are presented

Directionally solidified Al2O3/GAP eutectic ceramics by micro-pulling-down method

Science.gov (United States)

Cao, Xue; Su, Haijun; Guo, Fengwei; Tan, Xi; Cao, Lamei

2016-11-01

We reported a novel route to prepare directionally solidified (DS) Al2O3/GAP eutectic ceramics by micro-pulling-down (μ-PD) method. The eutectic crystallizations, microstructure characters and evolutions, and their mechanical properties were investigated in detail. The results showed that the Al2O3/GAP eutectic composites can be successfully fabricated through μ-PD method, possessed smooth surface, full density and large crystal size (the maximal size: φ90 mm × 20 mm). At the process of Diameter, the as-solidified Al2O3/GAP eutectic presented a combination of "Chinese script" and elongated colony microstructure with complex regular structure. Inside the colonies, the rod-type or lamellar-type eutectic microstructures with ultra-fine GAP surrounded by the Al2O3 matrix were observed. At an appropriate solidificational rate, the binary eutectic exhibited a typical DS irregular eutectic structure of "chinese script" consisting of interpenetrating network of α-Al2O3 and GAP phases without any other phases. Therefore, the interphase spacing was refined to 1-2 µm and the irregular microstructure led to an outstanding vickers hardness of 17.04 GPa and fracture toughness of 6.3 MPa × m1/2 at room temperature.

Structural investigations of mechanical properties of Al based rapidly solidified alloys

International Nuclear Information System (INIS)

Karakoese, Ercan; Keskin, Mustafa

2011-01-01

Highlights: → Rapid solidification processing (RSP) involves exceptionally high cooling rates. → We correlate the microstructure of the intermetallic Al 3 Fe, Al 2 Cu and Al 3 Ni phases with the cooling rate. → The solidification rate is high enough to retain most of alloying elements in the Al matrix. → The rapid solidification has effect on the phase constitution. -- Abstract: In this study, Al based Al-3 wt.%Fe, Al-3 wt.%Cu and Al-3 wt.%Ni alloys were prepared by conventional casting. They were further processed using the melt-spinning technique and characterized by the X-ray diffraction (XRD), scanning electron microscopy (SEM) together with energy dispersive spectroscopy (EDS), transmission electron microscope (TEM), differential scanning calorimetry (DSC) and the Vickers microhardness tester. The rapidly solidified (RS) binary alloys were composed of supersaturated α-Al solid solution and finely dispersed intermetallic phases. Experimental results showed that the mechanical properties of RS alloys were enhanced, which can be attributed to significant changes in the microstructure. RS samples were measured using a microhardness test device. The dependence of microhardness H V on the solidification rate (V) was analysed. These results showed that with the increasing values of V, the values of H V increased. The enthalpies of fusion for the same alloys were determined by DSC.

Study on the barrier performance of molten solidified waste (I). Review of the performance assessment research

Energy Technology Data Exchange (ETDEWEB)

Maeda, Toshikatsu; Sakamoto, Yoshiaki; Nakayama, Shinichi; Yamaguchi, Tetsuji; Ogawa, Hiromichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-02-01

Application of melting technique is thought as one of the effective methods to treatment of the waste from the view point of its homogeneity and waste volume reduction. Solidified products by melting are expected as potential candidates of engineered barrier in a repository due to the good properties for their stabilization of radionuclides and hazardous elements. However, the methodology of performance evaluation has not been estimated so far. In this report, a literature survey on the properties of molten solidified waste was performed. It is clarified that the leachability of waste elements such as Co or Sr in molten waste form would be controlled by the corrosion behaviors of iron or silica which are the matrix elements of the waste form. While, no investigations into the durability of waste form have performed so far. Also noticed that the research items on performance evaluation such as the leachability for long-lived radionuclides and durability of waste form would be necessary for the long-term barrier assessment on the disposal. (author)

The effect of grain size and cement content on index properties of weakly solidified artificial sandstones

Science.gov (United States)

Atapour, Hadi; Mortazavi, Ali

2018-04-01

The effects of textural characteristics, especially grain size, on index properties of weakly solidified artificial sandstones are studied. For this purpose, a relatively large number of laboratory tests were carried out on artificial sandstones that were produced in the laboratory. The prepared samples represent fifteen sandstone types consisting of five different median grain sizes and three different cement contents. Indices rock properties including effective porosity, bulk density, point load strength index, and Schmidt hammer values (SHVs) were determined. Experimental results showed that the grain size has significant effects on index properties of weakly solidified sandstones. The porosity of samples is inversely related to the grain size and decreases linearly as grain size increases. While a direct relationship was observed between grain size and dry bulk density, as bulk density increased with increasing median grain size. Furthermore, it was observed that the point load strength index and SHV of samples increased as a result of grain size increase. These observations are indirectly related to the porosity decrease as a function of median grain size.

Interplay between temperature gradients field and C - E transformation in solidifying rolls

Directory of Open Access Journals (Sweden)

W. Wołczyński

2009-07-01

Full Text Available At first step of simulation a temperature field for solidifying cast steel and cast iron roll has been performed. The calculation does not take into account the convection in the liquid since convection has no influence on the proposed model for the localization of the C-E (columnar to equiaxed grains transformation. However, it allows to study the dynamics of temperature field temporal behavior in the middle of a mould. It is postulated that for the C-E transition a full accumulation of the heat in the mould has been observed (plateau at the T(t curve. The temporal range of plateau existence corresponds to the incubation time for the full equiaxed grains formation. At the second step of simulation temporal behavior of the temperature gradient field has been studied. Three ranges within temperature gradients field have been distinguished for the operating point situated at the middle of mould: a/ for the formation of columnar grains zone, ( and high temperature gradient 0>>T&0//>>∂∂−∂∂∂∂−∂∂>EttEtrTrT. T - temperature, r - roll radius. It is evident that the heat transfer across the mould decides on the temporal appearance of incubation during which the solidification is significantly arrested and competition between columnar and equiaxed growth occurs. Moreover solidification with positive temperature gradient transforms into solidification with negative temperature gradient (locally after the incubation. A simulation has been performed for the cast steel and cast iron rolls solidifying as in industry condition. Since the incubation divides the roll into to parts (first with columnar structure, second with equiaxed structure some experiments dealing with solidification have been made in laboratory scale. Finally, observations of the macrosegregation or microsegregation and phase or structure appearance in the cast iron ingot / roll (made in laboratory has also been done in order to confront them with theoretical predictions

Solidified structure of thin-walled titanium parts by vertical centrifugal casting

Directory of Open Access Journals (Sweden)

Wu Shiping

2011-05-01

Full Text Available The solidified structure of the thin-walled and complicated Ti-6Al-4V castings produced by the vertical centrifugal casting process was studied in the present work. The results show that the wall thickness of the section is featured with homogeneously distributed fine equiaxial grains, compared with the microstructure of the thick-walled section. The grain size of the castings has a tendency to decrease gradually with the increasing of the centrifugal radius. The inter-lamellar space in thick-walled casting parts is bigger than that of the thin-walled parts, and the profile of inter-lamellar space is not susceptible to the centrifugal radius.

Surface free energy of polypropylene and polycarbonate solidifying at different solid surfaces

International Nuclear Information System (INIS)

Chibowski, Emil; Terpilowski, Konrad

2009-01-01

Advancing and receding contact angles of water, formamide, glycerol and diiodomethane were measured on polypropylene (PP) and polycarbonate (PC) sample surfaces which solidified at Teflon, glass or stainless steel as matrix surfaces. Then from the contact angle hystereses (CAH) the apparent free energies γ s tot of the surfaces were evaluated. The original PP surface is practically nonpolar, possessing small electron donor interaction (γ s - =1.91mJ/m 2 ), as determined from the advancing contact angles of these liquids. It may result from impurities of the polymerization process. However, it increases up to 8-10 mJ/m 2 for PP surfaces contacted with the solids. The PC surfaces both original and modified show practically the same γ s - =6.56.7mJ/m 2 . No electron acceptor interaction is found on the surfaces. The γ s tot of modified PP and PC surfaces depend on the kind of probe liquid and contacted solid surface. The modified PP γ s tot values determined from CAH of polar liquids are greater than that of original surface and they increase in the sequence: Teflon, glass, stainless steel surface, at which they solidified. No clear dependence is observed between γ s tot and dielectric constant or dipole moment of the polar probe liquids. The changes in γ s tot of the polymer surfaces are due to the polymer nature and changes in its surface structure caused by the structure and force field of the contacting solid. It has been confirmed by AFM images.

Tensile behavior change depending on the microstructure of a Fe-Cu alloy produced from rapidly solidified powder

International Nuclear Information System (INIS)

Kakisawa, Hideki; Minagawa, Kazumi; Halada, Kohmei

2003-01-01

The relationship between consolidating temperature and the tensile behavior of iron alloy produced from Fe-Cu rapidly solidified powder is investigated. Fe-Cu powder fabricated by high-pressure water atomization was consolidated by heavy rolling at 873-1273 K. Microstructural changes were observed and tensile behavior was examined. Tensile behavior varies as the consolidating temperature changes, and these temperature-dependent differences depend on the morphology of the microstructure on the order of micrometers. The sample consolidated at 873 K shows a good strength/elongation balance because the powder microstructure and primary powder boundaries are maintained. The samples consolidated at the higher temperatures have a microstructure of recrystallized grains, and these recrystallized samples show the conventional relationship between tensile behavior and grain size in ordinal bulk materials

State-of-the-art review of quality assurance techniques for vitrified high level waste

International Nuclear Information System (INIS)

Miller, P.L.H.

1984-07-01

Quality assurance is required for certain chemical and physical properties of both the molten glass pour and the solidified glass within the stainless steel container. It is also required to monitor the physical condition of the container lid weld. A review is presented of techniques which are used or which might be adapted for use in the quality assurance of vitrified high level waste. For the most part only non-intrusive methods have been considered, however, some techniques which are not strictly non-intrusive have been reviewed where a non-intrusive technique has not been identified or where there are other advantages associated with the particular technique. In order to identify suitable candidate techniques reference has been made to an extensive literature survey and experts in the fields of nuclear waste technology, glass technology, non-destructive testing, chemical analysis and remote analysis have been contacted. The opinions of manufacturers and users of specific techniques have also been sought. A summary is also given of those techniques which can most readily be applied to the problem of quality assurance for vitrified waste as well as recommendations for further research into techniques which might be adapted to suit this application. (author)

Solidification structure and dispersoids in rapidly solidified Ti-Al-Sn-Zr-Er-B alloys

International Nuclear Information System (INIS)

Rowe, R.G.; Broderick, T.F.; Koch, E.F.; Froes, F.H.

1986-01-01

The microstructure of melt extracted and melt spun titanium alloys containing erbium and boron revealed a duplex solidification structure of columnar grains leading to equiaxed and dendritic structures near the free surface of melt extracted and melt spun alloys. The solidification structure was revealed by apparent boride segregation to cellular, interdendritic and grain boundaries. Precipitation of needle or lath-like TiB particles occurred adjacent to Er/sub 2/O/sub 3/ dispesoid particles in as-rapidly solidified ribbon

EPICOR-II: a field leaching test of solidified radioactively loaded ion exchange resin

International Nuclear Information System (INIS)

Davis, E.C.; Marshall, D.S.; Todd, R.A.; Craig, P.M.

1986-08-01

As part of an ongoing research program investigating the disposal of radioactive solid wastes in the environment' the Oak Ridge National Laboratory (ORNL) is participating with Argonne National Laboratory, the Idaho National Engineering Laboratory, and the Nuclear Regulatory Commission in a study of the leachability of solidified EPICOR-II ion-exchange resin under simulated disposal conditions. To simulate disposal, a group of five 2-m 3 soil lysimeters has been installed in Solid Waste Storage Area Six at ORNL, with each lysimeter containing a small sample of solidified resin at its center. Two solidification techniques are being investigated: a Portland cement and a vinyl ester-styrene treatment. During construction, soil moisture temperature cells were placed in each lysimeter, along with five porous ceramic tubes for sampling water near the waste source. A meteorological station was set up at the study site to monitor climatic conditions (primarily precipitation and air temperature), and a data acquisition system was installed to keep daily records of these meteorological parameters as well as lysimeter soil moisture and temperature conditions. This report documents the first year of the long-term field study and includes discussions of lysimeter installation, calibration of soil moisture probes, installation of the site meteorological station, and the results of the first-quarter sampling for radionuclides in lysimeter leachate. In addition, the data collection and processing system developed for this study is documented, and the results of the first three months of data collection are summarized in Appendix D

Site selection factors for repositories of solid high-level and alpha-bearing wastes in geological formations

International Nuclear Information System (INIS)

1977-01-01

The purpose of this report is to provide guidelines for the selection and evaluation of suitable areas and sites for the disposal of solid high-level and alpha-bearing wastes into geological formations. This report is also intended to provide summary information on many types of geological formations underlying the land masses that might be considered as well as guidance on the geological and hydrological factors that should be investigated to demonstrate the suitability of the formations. In addition, other factors that should be considered in selecting a site for a radioactive waste repository are discussed briefly. The information, as presented, was developed to the extent of current technology for application to the evaluation of deep (greater than about 300 metres below ground level) geological formations in the selection of suitable areas for the disposal of solid or solidified high-level and alpha-bearing wastes. The extreme complexity of many geological environments and of the rock features that govern the presence and circulation of groundwater does not make it feasible to derive strict criteria for the selection of a site for a radioactive waste repository in a geological formation. Each potential repository location must be evaluated according to its own unique geological and hydrological setting. Therefore, only general guidance is offered, and this is done through discussion of the many factors that need to be considered in order to obtain the necessary assurances that the radionuclides will be confined in the geological repository over the required period of time

Site selection factors for repositories of solid high-level and alpha-bearing wastes in geological formations

Energy Technology Data Exchange (ETDEWEB)

1977-01-01

The purpose of this report is to provide guidelines for the selection and evaluation of suitable areas and sites for the disposal of solid high-level and alpha-bearing wastes into geological formations. This report is also intended to provide summary information on many types of geological formations underlying the land masses that might be considered as well as guidance on the geological and hydrological factors that should be investigated to demonstrate the suitability of the formations. In addition, other factors that should be considered in selecting a site for a radioactive waste repository are discussed briefly. The information, as presented, was developed to the extent of current technology for application to the evaluation of deep (greater than about 300 meters below ground level) geological formations in the selection of suitable areas for the disposal of solid or solidified high-level and alpha-bearing wastes. The extreme complexity of many geological environments and of the rock features that govern the presence and circulation of groundwater does not make it feasible to derive strict criteria for the selection of a site for a radioactive waste repository in a geological formation. Each potential repository location must be evaluated according to its own unique geological and hydrological setting. Therefore, only general guidance is offered, and this is done through discussion of the many factors that need to be considered in order to obtain the necessary assurances that the radionuclides will be confined in the geological repository over the required period of time.

Structure and mechanical properties of Al-3Fe rapidly solidified alloy

International Nuclear Information System (INIS)

Karakoese, Ercan; Keskin, Mustafa

2011-01-01

The Al based Al-3 wt%Fe alloy was prepared by conventionally casting (ingot) and further processed the melt-spinning technique and characterized by X-ray diffraction (XRD), scanning electron microscopy (SEM) together with energy dispersive spectroscopy (EDS), differential scanning calorimetry (DSC) and the Vickers microhardness tester. The rapidly solidified (RS) binary alloys were composed of supersaturated α-Al solid solution and finely dispersed intermetallic phases. Experimental results showed that the mechanical properties of RS alloys were enhanced, which can be attributed to significant changes in the microstructure. The dependence of microhardness H V on the solidification rate (V) was analysed. These results showed that with the increasing values of V, the values of H V increased.

Formation and growth of crystal defects in directionally solidified multicrystalline silicon for solar cells

Energy Technology Data Exchange (ETDEWEB)

Ryningen, Birgit

2008-07-01

Included in this thesis are five publications and one report. The common theme is characterisation of directionally solidified multicrystalline silicon for solar cells. Material characterisation of solar cell silicon is naturally closely linked to both the casting process and to the solar cell processing: Many of the material properties are determined by the casting process, and the solar cell processing will to some extend determine which properties will influence the solar cell performance. Solar grade silicon (SoG-Si) made by metallurgical refining route and supplied by Elkem Solar was directionally solidified and subsequently characterised, and a simple solar cell process was applied. Except from some metallic co-precipitates in the top of the ingot, no abnormalities were found, and it is suggested that within the limits of the tests performed in this thesis, the casting and the solar cell processing, rather than the assumed higher impurity content, was the limiting factor. It is suggested in this thesis that the main quality problem in multicrystalline silicon wafers is the existence of dislocation clusters covering large wafer areas. The clusters will reduce the effect of gettering and even if gettering could be performed successfully, the clusters will still reduce the minority carrier mobility and hence the solar cell performance. It has further been pointed out that ingots solidified under seemingly equal conditions might have a pronounced difference in minority carrier lifetime. Ingots with low minority carrier lifetime have high dislocation densities. The ingots with the substantially higher lifetime seem all to be dominated by twins. It is also found a link between a higher undercooling and the ingots dominated by twins. It is suggested that the two types of ingots are subject to different nucleation and crystal growth mechanisms: For the ingots dominated by dislocations, which are over represented, the crystal growth is randomly nucleated at the

Optimization of dispersive liquid-phase microextraction based on solidified floating organic drop combined with high-performance liquid chromatography for the analysis of glucocorticoid residues in food.

Science.gov (United States)

Huang, Yuan; Zheng, Zhiqun; Huang, Liying; Yao, Hong; Wu, Xiao Shan; Li, Shaoguang; Lin, Dandan

2017-05-10

A rapid, simple, cost-effective dispersive liquid-phase microextraction based on solidified floating organic drop (SFOD-LPME) was developed in this study. Along with high-performance liquid chromatography, we used the developed approach to determine and enrich trace amounts of four glucocorticoids, namely, prednisone, betamethasone, dexamethasone, and cortisone acetate, in animal-derived food. We also investigated and optimized several important parameters that influenced the extraction efficiency of SFOD-LPME. These parameters include the extractant species, volumes of extraction and dispersant solvents, sodium chloride addition, sample pH, extraction time and temperature, and stirring rate. Under optimum experimental conditions, the calibration graph exhibited linearity over the range of 1.2-200.0ng/ml for the four analytes, with a reasonable linearity(r 2 : 0.9990-0.9999). The enrichment factor was 142-276, and the detection limits was 0.39-0.46ng/ml (0.078-0.23μg/kg). This method was successfully applied to analyze actual food samples, and good spiked recoveries of over 81.5%-114.3% were obtained. Copyright © 2017. Published by Elsevier B.V.

Decomposition for the analysis of radionuclides in solidified cement radioactive waste

International Nuclear Information System (INIS)

Lee, Jeong Jin; Pyo, Hyung Yeal; Jee, Kwang Yung; Jeon, Jong Seon

2004-01-01

Spent ion exchange resins make solid radioactive wastes when mixed with cement as solidifying material that was widely used in securing human environment from radionuclides for at least hundreds years. The cumulative increase of low and medium level radioactive wastes results in capacity problem of temporary storage in some NPPs (Nuclear Power Plants) of Korea around 2008. Radioactive wastes are scheduled to be disposed in a permanent disposal facility in accordance with the Korean Radioactive Wastes Management Program. It is mandatory to identify kinds and concentration of radionuclides immobilized for transporting them from temporary storage in NPPs to disposal facility. Accordingly, the effective sample decomposition prior to radiochemical separation is prerequisite to obtain the analytical data about radionuclides in cement waste forms. The closed-vessel microwave digestion technology among several sample preparation methods is taken into account to decompose cement waste forms. In this study, SRM 1880a (Portland cement) which is known for its certified values was used to optimize decomposition condition of cement waste forms containing nonradioactive ion exchange resins from NPP. With such variables as reagents, time, and power, the variation of the transparency and the color of the solution after closed-vessel microwave digestion can be examine. SRM 1880a is decomposed by suggested digestion procedure and the recoveries of constituents were investigated by ICP-AES and AAS

Organic semiconductor rubrene thin films deposited by pulsed laser evaporation of solidified solutions

Science.gov (United States)

Majewska, N.; Gazda, M.; Jendrzejewski, R.; Majumdar, S.; Sawczak, M.; Śliwiński, G.

2017-08-01

Organic semiconductor rubrene (C42H28) belongs to most preferred spintronic materials because of the high charge carrier mobility up to 40 cm2(V·s)-1. However, the fabrication of a defect-free, polycrystalline rubrene for spintronic applications represents a difficult task. We report preparation and properties of rubrene thin films deposited by pulsed laser evaporation of solidified solutions. Samples of rubrene dissolved in aromatic solvents toluene, xylene, dichloromethane and 1,1-dichloroethane (0.23-1% wt) were cooled to temperatures in the range of 16.5-163 K and served as targets. The target ablation was provided by a pulsed 1064 nm or 266 nm laser. For films of thickness up to 100 nm deposited on Si, glass and ITO glass substrates, the Raman and AFM data show presence of the mixed crystalline and amorphous rubrene phases. Agglomerates of rubrene crystals are revealed by SEM observation too, and presence of oxide/peroxide (C42H28O2) in the films is concluded from matrix-assisted laser desorption/ionization time-of-flight spectroscopic analysis.

Recycling stabilised/solidified drill cuttings for forage production in acidic soils.

Science.gov (United States)

Kogbara, Reginald B; Dumkhana, Bernard B; Ayotamuno, Josiah M; Okparanma, Reuben N

2017-10-01

Stabilisation/solidification (S/S), which involves fixation and immobilisation of contaminants using cementitious materials, is one method of treating drill cuttings before final fate. This work considers reuse of stabilised/solidified drill cuttings for forage production in acidic soils. It sought to improve the sustainability of S/S technique through supplementation with the phytoremediation potential of plants, eliminate the need for landfill disposal and reduce soil acidity for better plant growth. Drill cuttings with an initial total petroleum hydrocarbon (TPH) concentration of 17,125 mg kg -1 and low concentrations of metals were treated with 5%, 10%, and 20% cement dosages. The treated drill cuttings were reused in granular form for growing a forage, elephant grass (Pennisetum purpureum), after mixing with uncontaminated soil. The grasses were also grown in uncontaminated soil. The phytoremediation and growth potential of the plants was assessed over a 12-week period. A mix ratio of one part drill cuttings to three parts uncontaminated soil was required for active plant growth. The phytoremediation ability of elephant grass (alongside abiotic losses) reduced the TPH level (up to 8795 mg kg -1 ) in the soil-treated-drill cuttings mixtures below regulatory (1000 mg kg -1 ) levels. There were also decreased concentrations of metals. The grass showed better heights and leaf lengths in soil containing drill cuttings treated with 5% cement dosage than in uncontaminated soil. The results suggest that recycling S/S treated drill cuttings for forage production may be a potential end use of the treated waste. Copyright © 2017 Elsevier Ltd. All rights reserved.

Fabrication and tensile properties of rapidly solidified Cu-10wt. %Ni alloy. [Cu-10Ni

Energy Technology Data Exchange (ETDEWEB)

Baril, D; Angers, R; Baril, J [Dept. of Mining and Metallurgy, Laval Univ., Ste-Foy, Quebec (Canada)

1992-10-15

Cu-10wt.%Ni ribbons were produced by melt spinning and cut into small particles with a blade cutter mill. The powders were then hot consolidated to full density by hot pressing followed by hot extrusion. Tensile properties of the resulting pieces were measured. Cu-10wt.%Ni cast ingots were also hot extruded and mechanically tested to compare with the rapidly solidified alloy and to evaluate the possible benefits brought by the rapid solidification process.

Development of volume reduction treatment techniques for low level radioactive wastes

International Nuclear Information System (INIS)

Nabatame, Yasuzi

1984-01-01

The solid wastes packed in drums are preserved in the stores of nuclear establishments in Japan, and the quantity of preservation has already reached about 60 % of the capacity. It has become an important subject to reduce the quantity of generation of radioactive wastes and how to reduce the volume of generated wastes. As the result of the research aiming at the development of the solidified bodies which are excellent in the effect of volume reduction and physical properties, it was confirmed that the plastic solidified bodies using thermosetting resin were superior to conventional cement or asphalt solidification. The plastic solidifying system can treat various radioactive wastes. After radioactive wastes are dried and powdered, they are solidified with plastics, therefore, the effect of volume reduction is excellent. The specific gravity, strength and the resistance to water, fire and radiation were confirmed to be satisfacotory. The plastic solidifying system comprises three subsystems, that is, drying system, powder storing and supplying system and plastic solidifying system. Also the granulation technique after drying and powdering, acid decomposition technique, the microwave melting and solidifying technique for incineration ash, plasma melting process and electrolytic polishing decontamination are described. (Kako, I.)

The effects of low-molecular-weight emulsifiers in O/W-emulsions on microviscosity of non-solidified oil in fat globules and the mobility of emulsifiers at the globule surfaces

DEFF Research Database (Denmark)

Munk, Merete B.; Erichsen, Henriette Rifbjerg; Andersen, Mogens Larsen

2014-01-01

caseinate and different combinations of lactic acid ester of monoglyceride (LACTEM), unsaturated monoglycerides (GMU) or saturated monoglyceride (GMS) were studied. The non-solidified oil in emulsions made with LACTEM. +. GMU had a high microviscosity, whereas the emulsion made with GMS had a low...... of the spin probe on the droplet surfaces. Conversely, in presence of LACTEM and GMS, the protein surface loads decreased and high surface mobilities were observed. Based on these results it is argued that the high macroscopic viscosity and lipid agglomeration of emulsions containing GMU is due to a lipid...

A Laboratory Screening Study On The Use Of Solidifiers As A Response Tool To Remove Crude Oil Slicks On Seawater

Science.gov (United States)

The effectiveness of five solidifiers to remove Prudhoe Bay crude oil from artificial seawater in the laboratory was determined by ultraviolet-visible spectroscopy (UV-VIS) and gas chromatography/mass spectrometry (GC/MS). The performance of the solidifers was determined by US-V...

Microbial degradation of low-level radioactive waste. Final report

International Nuclear Information System (INIS)

Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr.

1996-06-01

The Nuclear Regulatory Commission stipulates in 10 CFR 61 that disposed low-level radioactive waste (LLW) be stabilized. To provide guidance to disposal vendors and nuclear station waste generators for implementing those requirements, the NRC developed the Technical Position on Waste Form, Revision 1. That document details a specified set of recommended testing procedures and criteria, including several tests for determining the biodegradation properties of waste forms. Information has been presented by a number of researchers, which indicated that those tests may be inappropriate for examining microbial degradation of cement-solidified LLW. Cement has been widely used to solidify LLW; however, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. The purpose of this research program was to develop modified microbial degradation test procedures that would be more appropriate than the existing procedures for evaluation of the effects of microbiologically influenced chemical attack on cement-solidified LLW. The procedures that have been developed in this work are presented and discussed. Groups of microorganisms indigenous to LLW disposal sites were employed that can metabolically convert organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with cement and can ultimately lead to structural failure. Results on the application of mechanisms inherent in microbially influenced degradation of cement-based material are the focus of this final report. Data-validated evidence of the potential for microbially influenced deterioration of cement-solidified LLW and subsequent release of radionuclides developed during this study are presented

Alternatives for conversion to solid interim waste forms of the radioactive liquid high-level wastes stored at the Western New York Nuclear Service Center

International Nuclear Information System (INIS)

Vogler, S.; Trevorrow, L.E.; Ziegler, A.A.; Steindler, M.J.

1981-08-01

Techniques for isolating and solidifying the nuclear wastes in the storage tanks at the Western New York Nuclear Service Center plant have been examined. One technique involves evaporating the water and forming a molten salt containing the precipitated sludge. The salt is allowed to solidify and is stored in canisters until processing into a final waste form is to be done. Other techniques involve calcining the waste material, then agglomerating the calcine with sodium silicate to reduce its dispersibility. This option can also involve a prior separation and decontamination of the supernatant salt. The sludge and all resins containing fission-product activity are then calcined together. The technique of removing the water and solidifying the salt may be the simplest method for removing the waste from the West Valley Plant

Conceptual design report for regional low-level waste interim storage site

International Nuclear Information System (INIS)

Bird, M.V.; Thompson, J.D.

1981-08-01

An interim storage site design concept was developed for receiving 100,000 ft 3 low-level waste per year, in the form of solidified wastes in 55-gallon drums with a dose rate of < 200 mrem per hour at contact

Phase Composition of a CrMo0.5NbTa0.5TiZr High Entropy Alloy: Comparison of Experimental and Simulated Data

OpenAIRE

Fan Zhang; Oleg N. Senkov; Jonathan D. Miller

2013-01-01

Microstructure and phase composition of a CrMo0.5NbTa0.5TiZr high entropy alloy were studied in the as-solidified and heat treated conditions. In the as-solidified condition, the alloy consisted of two disordered BCC phases and an ordered cubic Laves phase. The BCC1 phase solidified in the form of dendrites enriched with Mo, Ta and Nb, and its volume fraction was 42%. The BCC2 and Laves phases solidified by the eutectic-type reaction, and their volume fractions were 27% and 31%, respectively....

Comparative analysis of mechanical characteristics of solidified concentrates from BWR system using Yugoslav and Italian cements

International Nuclear Information System (INIS)

Plecas, I.; Peric, A.; Drljaca, J.; Kostadinovic, A.

1987-01-01

In this paper, properties of Italian and Yugoslav cement mixture with BWR evaporation concentrates were compared, research was held upon fifteen samples, according to the adequate formulations. Samples were made in standard cube form, side 10 cm. Functional relationship between decreasing the compressive strength and amount of incorporated BWR concentrate cement mixture was developed. The results of research showed nearly the same mechanical properties of solidified BWR concentrate with Italian and Yugoslav cements. (author)

Preconceptual design study for solidifying high-level waste: West Valley Demonstration Project

International Nuclear Information System (INIS)

1981-04-01

This Appendix contains the preconceptual design drawings prepared by Vitro Engineering Corporation for Pacific Northwest Laboratory. The following types of drawings are included in this Appendix: process flow diagrams; process and instrumentation diagrams; hydraulic diagrams; equipment arrangement drawings; service gallery drawings; electric power one-line diagram; equipment line lists; and outline specifications. The basic purpose of these drawings was to determine the feasibility of installing the reference solidification process in existing cells at the Western New York Nuclear Service Center. Most of the process and vitrification equipment will be installed in the former Chemical Processing Cell, while the salt solidification equipment will be housed in the former Scrap Removal Room. The design utilized a remote maintenance and operation concept

The research of Ti-rich zone on the interface between TiCx and aluminum melt and the formation of Ti3Al in rapid solidified Al-Ti-C master alloys

International Nuclear Information System (INIS)

Jiang Kun; Ma Xiaoguang; Liu Xiangfa

2009-01-01

In the present work, the thermodynamic tendency of formation of Ti-rich zone on the interface between TiC x and aluminum melt is calculated and a high titanium concentration can exist in the zone according to the thermodynamic calculation. Rapid solidified Al-5Ti-0.5C master alloy is analyzed by X-ray diffraction (XRD) and transmission electronic microscopy (TEM). The appearance of Ti 3 Al in the master alloy results from the existence of high-concentration Ti-rich zone.

Leaching studies of low-level radioactive waste forms

International Nuclear Information System (INIS)

Dayal, R.; Arora, H.; Milian, L.; Clinton, J.

1985-01-01

A research program has been underway at the Brookhaven National Laboratory to investigate the release of radionuclides from low-level waste forms under laboratory conditions. This paper describes the leaching behavior of Cs-137 from two major low-level waste streams, that is, ion exchange bead resin and boric acid concentrate, solidified in Portland cement. The resultant leach data are employed to evaluate and predict the release behavior of Cs-137 from low-level waste forms under field burial conditions

Interconnection of thermal parameters, microstructure and mechanical properties in directionally solidified Sn–Sb lead-free solder alloys

Energy Technology Data Exchange (ETDEWEB)

Dias, Marcelino; Costa, Thiago [Department of Manufacturing and Materials Engineering, University of Campinas — UNICAMP, 13083-860 Campinas, SP (Brazil); Rocha, Otávio [Federal Institute of Education, Science and Technology of Pará — IFPA, 66093-020 Belém, PA (Brazil); Spinelli, José E. [Department of Materials Engineering, Federal University of São Carlos — UFSCar, 13565-905 São Carlos, SP (Brazil); Cheung, Noé, E-mail: cheung@fem.unicamp.br [Department of Manufacturing and Materials Engineering, University of Campinas — UNICAMP, 13083-860 Campinas, SP (Brazil); Garcia, Amauri [Department of Manufacturing and Materials Engineering, University of Campinas — UNICAMP, 13083-860 Campinas, SP (Brazil)

2015-08-15

Considerable effort is being made to develop lead-free solders for assembling in environmental-conscious electronics, due to the inherent toxicity of Pb. The search for substitute alloys of Pb–Sn solders has increased in order to comply with different soldering purposes. The solder must not only meet the expected levels of electrical performance but may also have appropriate mechanical strength, with the absence of cracks in the solder joints. The Sn–Sb alloy system has a range of compositions that can be potentially included in the class of high temperature solders. This study aims to establish interrelations of solidification thermal parameters, microstructure and mechanical properties of Sn–Sb alloys (2 wt.%Sb and 5.5 wt.%Sb) samples, which were directionally solidified under cooling rates similar to those of reflow procedures in industrial practice. A complete high-cooling rate cellular growth is shown to be associated with the Sn–2.0 wt.%Sb alloy and a reverse dendrite-to-cell transition is observed for the Sn–5.5 wt.%Sb alloy. Strength and ductility of the Sn–2.0 wt.%Sb alloy are shown not to be affected by the cellular spacing. On the other hand, a considerable variation in these properties is associated with the cellular region of the Sn–5.5 wt.%Sb alloy casting. - Graphical abstract: Display Omitted - Highlights: • The microstructure of the Sn–2 wt.%Sb alloy is characterized by high-cooling rates cells. • Reverse dendrite > cell transition occurs for Sn–5.5 wt.%Sb alloy: cells prevail for cooling rates > 1.2 K/s. • Sn–5.5 wt.%Sb alloy: the dendritic region occurs for cooling rates < 0.9 K/s. • Sn–5.5 wt.%Sb alloy: tensile properties are improved with decreasing cellular spacing.

Experimental study on the leaching of radioactive materials from radioactive wastes solidified in cement into sea water. Part 2

International Nuclear Information System (INIS)

Hatta, H.; Ono, H.; Nagakura, T.; Machida, T.; Seki, T.; Maki, Y.

Results are presented from the study on leachability of 60 Co and 137 Cs from BWR concentrated wastes that had been solidified in cement. The leachability of 60 Co is very small compared to that of 137 Cs and varies greatly with the type of leaching medium. The effect of duration of immersion on leachability is comparatively large

Influence of carbonation on the acid neutralization capacity of cements and cement-solidified/stabilized electroplating sludge.

Science.gov (United States)

Chen, Quanyuan; Zhang, Lina; Ke, Yujuan; Hills, Colin; Kang, Yanming

2009-02-01

Portland cement (PC) and blended cements containing pulverized fuel ash (PFA) or granulated blast-furnace slag (GGBS) were used to solidify/stabilize an electroplating sludge in this work. The acid neutralization capacity (ANC) of the hydrated pastes increased in the order of PC > PC/GGBS > PC/PFA. The GGBS or PFA replacement (80 wt%) reduced the ANC of the hydrated pastes by 30-50%. The ANC of the blended cement-solidified electroplating sludge (cement/sludge 1:2) was 20-30% higher than that of the hydrated blended cement pastes. Upon carbonation, there was little difference in the ANC of the three cement pastes, but the presence of electroplating sludge (cement/sludge 1:2) increased the ANC by 20%. Blended cements were more effective binders for immobilization of Ni, Cr and Cu, compared with PC, whereas Zn was encapsulated more effectively in the latter. Accelerated carbonation improved the immobilization of Cr, Cu and Zn, but not Ni. The geochemical code PHREEQC, with the edited database from EQ3/6 and HATCHES, was used to calculate the saturation index and solubility of likely heavy metal precipitates in cement-based solidification/stabilization systems. The release of heavy metals could be related to the disruption of cement matrices and the remarkable variation of solubility of heavy metal precipitates at different pH values.

High-level verification

CERN Document Server

Lerner, Sorin; Kundu, Sudipta

2011-01-01

Given the growing size and heterogeneity of Systems on Chip (SOC), the design process from initial specification to chip fabrication has become increasingly complex. This growing complexity provides incentive for designers to use high-level languages such as C, SystemC, and SystemVerilog for system-level design. While a major goal of these high-level languages is to enable verification at a higher level of abstraction, allowing early exploration of system-level designs, the focus so far for validation purposes has been on traditional testing techniques such as random testing and scenario-based

Numerical Research on Magnetic Field, Temperature Field and Flow Field During Melting and Directionally Solidifying TiAl Alloys by Electromagnetic Cold Crucible

Science.gov (United States)

Chen, Ruirun; Yang, Yaohua; Gong, Xue; Guo, Jingjie; Su, Yanqing; Ding, Hongsheng; Fu, Hengzhi

2017-12-01

The electromagnetic cold crucible (EMCC) technique is an effective method to melt and directionally solidify reactive and high-temperature materials without contamination. The temperature field and fluid flow induced by the electromagnetic field are very important for melting and controlling the microstructure. In this article, a 3D EMCC model for calculating the magnetic field in the charges (TiAl alloys) using the T-Ω finite element method was established and verified. Magnetic fields in the charge under different electrical parameters, positions and dimensions of the charge were calculated and analyzed. The calculated results show that the magnetic field concentrates in the skin layer, and the magnetic flux density ( B) increases with increasing of the frequency, charge diameter and current. The maximum B in the charge is affected by the position of the charge in EMCC ( h 1) and the charge height ( h 2), which emerges at the middle of coils ( h c) when the relationship of h c < h 1 + h 2 < h c + δ is satisfied. Lower frequency and smaller charge diameter can improve the uniformity of the magnetic field in the charge. Consequently, the induced uniform electromagnetic stirring weakens the turbulence and improves temperature uniformity in the vicinity of the solid/liquid (S/L) interface, which is beneficial to forming a planar S/L interface during directional solidification. Based on the above conclusions, the TiAlNb alloy was successfully melted with lower power consumption and directionally solidified by the square EMCC.

Application of systems analysis to the disposal of high level waste in deep ocean sediments

International Nuclear Information System (INIS)

De Marsily, G.; Dorp, F. van

1982-01-01

Emplacement in deep ocean sediments is one of the disposal options being considered for solidified high level radioactive waste. Task groups set up within the framework of the NEA Seabed Working Group have been studying many aspects of this option since 1976. The methods of systems analysis have been applied to enable the various parts of the problem to be assessed within an integrated framework. This paper describes the progress made by the Systems Analysis Task Group towards the development of an overall system model. The Task Group began by separating the problem into elements and defining the interfaces between these elements. A simple overall system model was then developed and used in both a preliminary assessment and a sensitivity analysis to identify the most important parameters. These preliminary analyses used a very simple model of the overall system and therefore the results cannot be used to draw any conclusions as to the acceptability of the sub-seabed disposal option. However they served to show the utility of the systems analysis method. The work of the other task groups will focus on the important parameters so that improved results can be fed back into an improved system model. Subsequent iterations will eventually provide an input to an acceptability decision. (Auth.)

Examination on rational disposal concept, layout, and methods of molding and settling for high level radioactive waste

International Nuclear Information System (INIS)

Toyota, Masatoshi

1998-01-01

As for the concept of disposing high level radioactive waste in the place of disposal, the method of securing safety by isolating the waste from human environment with the combination of artificial barriers and natural barriers has been adopted. At present in Japan, Power Reactor and Nuclear Fuel Development Corporation has considered the concept of disposal, but it is considered to be necessary to review it from the viewpoints of the uncertainty in safety characteristics, the possibility of realizing construction and settlement, economical efficiency and others. Recently, the investigation of the rational disposal concept has been advanced jointly with Dr. McKinley. The conditions to be considered for artificial barriers at the time of reviewing the disposal concept are described on bentonite buffer and carbon steel overpack enclosing glass-solidified body. As the disposal concept, the private plan of Toyota and that of Toyota and McKinley are shown. Also the layout for settling two modules each in horizontal adits on both sides of the connecting tunnel is proposed. The methods of molding and settling the engineered barrier system are explained. This disposal concept can reduce uncertainty, heighten safety and reduce the cost. (K.I.)

French high level wastes management

International Nuclear Information System (INIS)

Gauvenet, A.J.; Sombret, C.G.

1980-06-01

The first French spent fuel reprocessing plant went on stream in 1956 at Marcoule. Since then, all French irradiated fuels and some foreign spent fuels have been reprocessed either in this plant or in a subsequent plant built at La Hague. Marcoule is primarily devoted to metallic fuels, and La Hague to oxide fuels. The fission products solutions generated by reprocessing are acid liquids. They are stored on site in double walled stainless steel tanks fitted with a cooling device to deal with thermal release due to radioactive decay. Although these liquids are retrievable and can be transfered from one tank to another, and in spite of the fact that no disturbance such as overheating or leakage has ever occurred, a decision was made to solidify these solutions in order to make interim storage and, later on, ultimate disposal, safer and easier to control. Glass was chosen, because it is a flexible medium to deal with and it answers the quality requirements of ultimate disposal as well as the manufacturing constraints, such as equipment corrosion, volatilization during fabrication, and suitability to casting into canisters

Influence of micro-additions of bismuth on structures, mechanical and electrical transport properties of rapidly solidified Sn-3.5% Ag Alloy from melt

International Nuclear Information System (INIS)

El Bahay, M.M.; Mady, H.A.

2005-01-01

The present study was undertaken to investigate the influence of the Bi addition in the Sn-3.5 Ag rapidly solidified binary system for use as a Pb-free solder. The resulting properties of the binary system were extended to the Sn based ternary systems Sn 9 6.5-X Ag 3 .5 Bi x (0≤ X ≤ 2.5) solder. The structure and electrical resistivity of rapidly solidified (melt spun) alloys have been investigated. With the addition of up to 2.5 mass % Bi, the melting temperature decreases from 221.1 to 214.8 degree C. Wetting contact angle of the six alloys on Cu Zn 3 0 substrate are carried out at 573 K. Microhardness evaluations were also performed on the Sn-Ag-Bi alloys. The measured values and other researcher's results were compared with the calculated data

Removal of radioactive cesium from surface soils solidified using polyion complex. Rapid communication for decontamination test at Iitate-mura in Fukushima Prefecture

International Nuclear Information System (INIS)

Naganawa, Hirochika; Yanase, Nobuyuki; Mitamura, Hisayoshi; Nagano, Tetsushi; Yoshida, Zenko; Kumazawa, Noriyuki; Saitoh, Hiroshi; Kashima, Kaoru; Fukuda, Tatsuya; Tanaka, Shun-ichi

2011-01-01

We tried the decontamination of surface soils for three types of agricultural land at Nagadoro district of Iitate-mura (village) in Fukushima Prefecture, which is highly contaminated by deposits of radionuclides from the plume released from the Fukushima Daiichi nuclear power plant. The decontamination method consisted of the peeling of surface soils solidified using a polyion complex, which was formed from a salt solution of polycations and polyanions. Two types of polyion complex solution were applied to an upland field in a plastic greenhouse, a pasture, and a paddy field. The decontamination efficiency of the surface soils reached 90%, and dust release was effectively suppressed during the removal of surface soils. (author)

Alternative containers for low-level wastes containing large amounts of tritium

International Nuclear Information System (INIS)

Gause, E.P.; Lee, B.S.; MacKenzie, D.R.; Wiswall, R. Jr.

1984-11-01

High-activity tritiated waste generated in the United States is mainly composed of tritium gas and tritium-contaminated organic solvents sorbed onto Speedi-Dri which are packaged in small glass bulbs. Low-activity waste consists of solidified and adsorbed liquids. In this report, current packages for high-activity gaseous and low-activity adsorbed liquid wastes are emphasized with regard to containment potential. Containers for low-level radioactive waste containing large amounts of tritium need to be developed. An integrity may be threatened by: physical degradation due to soil corrosion, gas pressure build-up (due to radiolysis and/or biodegradation), rapid permeation of tritium through the container, and corrosion from container contents. Literature available on these points is summarized in this report. 136 references, 20 figures, 40 tables

Preliminary analysis of engineered barrieer performances in geological disposal of high level waste

International Nuclear Information System (INIS)

Ohe, Toshiaki; Maki, Yasuo; Tanaka, Hiroshi; Kawanishi, Motoi.

1988-01-01

This report represents preliminary results of safety analysis of a engineered barrier system in geological disposal of high level radioactive waste. Three well-known computer codes; ORIGEN 2, TRUMP, and SWIFT were used in the simulation. Main conceptual design of the repository was almost identical to that of SKB in Sweden and NAGRA in Switzerland; the engineered barrier conasists glass solidified waste, steel overpack, and compacted bentonite. Two different underground formations are considered; granite and neogene sedimentary rock, which are typically found in Japan. We first determined the repository configuration, particularly the space between disposal pitts. The ORIGEN 2 was used to estimate heat generation in the waste glass reprocessed at 4 years after removal from PWR. Then, temperature distribution was calculated by the TRUMP. The results of two or three dimensional calculation indicated that the pit interval should be kept more than 5 m in the case of granite formation at 500 m depth, according to the temperature criteria in the bentonite layer ( 90 Sr, 241 Am, 239 Pu, and 237 Np were chosen in one or two dimensional calculations. For both cases of steady release and instanteneous release, the maximum concentration in the pore water at the boundary between bentonite and surrounding rock had the following order; 237 Np> 239 Pu> 90 Sr> 241 Am. Sensitivity analysis showed that the order mainly due to the different adsorption characteristics of the nuclides in bentonite layer. (author)

Directionally Solidified Aluminum - 7 wt% Silicon Alloys: Comparison of Earth and International Space Station Processed Samples

Science.gov (United States)

Grugel, Richard N,; Tewari, Surendra; Rajamure, R. S.; Erdman, Robert; Poirier, David

2012-01-01

Primary dendrite arm spacings of Al-7 wt% Si alloy directionally solidified in low gravity environment of space (MICAST-6 and MICAST-7: Thermal gradient approx. 19 to 26 K/cm, Growth speeds varying from 5 to 50 microns/s show good agreement with the Hunt-Lu model. Primary dendrite trunk diameters of the ISS processed samples show a good fit with a simple analytical model based on Kirkwood s approach, proposed here. Natural convection, a) decreases primary dendrite arm spacing. b) appears to increase primary dendrite trunk diameter.

Effect of cooling rate and Mg addition on the structural evaluation of rapidly solidified Al-20wt%Cu-12wt%Fe alloy

Energy Technology Data Exchange (ETDEWEB)

Karaköse, Ercan, E-mail: ekarakose@karatekin.edu.tr [Çankırı Karatekin University, Faculty of Sciences, Department of Physics, 18100 Çankırı (Turkey); Çolak, Hakan [Çankırı Karatekin University, Faculty of Sciences, Department of Chemistry, 18100 Çankırı (Turkey)

2016-11-15

The present work examines the effect of Mg contents and cooling rate on the morphology and mechanical properties of Al{sub 20}Cu{sub 12}Fe quasicrystalline alloy. The microstructure of the alloys was analyzed by scanning electron microscopy and the phase composition was identified by X-ray diffractometry. The melting characteristics were studied by differential thermal analysis under an Ar atmosphere. The mechanical features of the melt-spun and conventionally solidified alloys were tested by tensile-strength test and Vickers micro-hardness test. It was found that the final microstructure of the Al{sub 20}Cu{sub 12}Fe samples mainly depends on the cooling rate and Mg contents, which suggests that different cooling rates and Mg contents produce different microstructures and properties. The average grain sizes of the melt spun samples were about 100–300 nm at 35 m/s. The nanosize, dispersed, different shaped quasicrystal particles possessed a remarkable effect to the mechanical characteristics of the rapidly solidified ribbons. The microhardness values of the melt spun samples were approximately 18% higher than those of the conventionally counterparts. - Highlights: •Quasicrystal-creating materials have high potential for applications. •Different shaped nanosize quasicrystal particles were observed. •The addition of Mg has an important impact on the mechanical properties. •H{sub V} values of the MS0, MS3 and MS5 samples at 35 m/s were 8.56, 8.66 and 8.80 GPa. •The volume fraction of IQC increases with increasing cooling rates.

Leachability and heavy metal speciation of 17-year old stabilised/solidified contaminated site soils

Energy Technology Data Exchange (ETDEWEB)

Wang, Fei, E-mail: fwtiffany@gmail.com [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom); Wang, Hailing, E-mail: wanghailing@njtech.edu.cn [College of Environment, Nanjing Tech University, Nanjing 210009 (China); Al-Tabbaa, Abir, E-mail: aa22@cam.ac.uk [Department of Engineering, University of Cambridge, Trumpington Street, Cambridge CB2 1PZ (United Kingdom)

2014-08-15

Highlights: • The effectiveness of the cement-based S/S at 17 years in West Drayton site is still satisfactory. • Major leaching of Cu, Zn, Ni, Cd and Pb in all mixes took place in the Fe/Mn oxides phase. • The hydration process has been fully completed and further carbonation took place at 17 years. • Microstructure analyses show that unreacted PFA exists. - Abstract: The long-term leachability, heavy metal speciation transformation and binding mechanisms in a field stabilised/solidified contaminated soil (made ground) from West Drayton site were recently investigated following in situ auger mixing treatment with a number of cement-based binders back in 1996. Two batch leaching tests (TCLP and BS EN 12457) and a modified five step sequential extraction procedure along with X-ray diffraction (XRD) and scanning electron microscopy (SEM) analyses were employed for the testing of the 17-year-old field soil. The results of batch leaching tests show that the treatment employed remained effective at 17 years of service time, with all BS EN 12457 test samples and most of TCLP test samples satisfied drinking water standards. Sequential extraction results illustrate that the leaching of Cu, Ni, Zn, Pb and Cd in all mixes mainly occurred at the Fe/Mn phase, ranging from 43% to 83%. Amongst the five metals tested, Ni was the most stable with around 40% remained in the residual phase for all the different cement-based binder stabilised/solidified samples. XRD and SEM analyses show that the hydration process has been fully completed and further carbonation took place. In summary, this study confirms that such cement-based stabilisation/solidification (S/S) treatment can achieve satisfactory durability and thus is a reliable technique for long-term remediation of heavy metal contaminated soil.

Valence electron structure analysis of the cubic silicide intermetallics in rapidly solidified Al-Fe-V-Si alloy

International Nuclear Information System (INIS)

Wang, J.Q.; Qian, C.F.; Zhang, B.J.; Tseng, M.K.; Xiong, S.W.

1996-01-01

The application of rapid solidification for the development of elevated temperature aluminum alloys has resulted in the emergence of several alloys based on the Al-Fe alloy system. Of particular interest are Al-Fe-V-Si alloys which have excellent room temperature and high temperature mechanical properties. In a pioneering study, Skinner et al. showed the stabilization of the cubic phase in ternary Al-Fe-Si alloy by the addition of a quaternary element, vanadium. The evolution of the microstructure in these alloys both during rapid solidification and subsequent processing is of crucial importance. Kim has demonstrated that the composition of the silicide phase in rapidly solidified Al-Fe-V-Si alloy is very close to Al 12 (Fe,V) 3 Si with the body centered cubic (bcc) structure. The structure is closely related to that of quasicrystals.In view of the structural features and the relationship between the α 12 and α 13 phases, the researching emphasis should firstly be put on the α 12 phase. In this paper the authors analyzed the α -(AlFeSi)(α 12 -type) phase from the angle of atomic valence electron structure other than the traditional methods of obtaining the diffraction spots of the phase. Several pieces of information were obtained about the hybrid levels and bond natures of every kind of atom in the α -(AlFeSi) phase. Finally the authors explained the phenomenon which V atom can substitute for Fe atom in the α 12 phase and improve the thermal stability of the phase in Al-Fe-V-Si alloy

Overview of the geochemical code MINTEQ: applications to performance assessment for low-level wastes

International Nuclear Information System (INIS)

Graham, M.J.; Peterson, S.R.

1985-09-01

The MINTEQ geochemical computer code, developed at Pacific Northwest Laboratory, integrates many of the capabilities of its two immediate predecessors, WATEQ3 and MINEQL. MINTEQ can be used to perform the calculations necessary to simulate (model) the contact of low-level waste solutions with heterogeneous sediments or the interaction of ground water with solidified low-level wastes. The code is capable of performing calculations of ion speciation/solubility, adsorption, oxidation-reduction, gas phase equilibria, and precipitation/dissolution of solid phases. Under the Special Waste Form Lysimeters-Arid program, the composition of effluents (leachates) from column and batch experiments, using laboratory-scale waste forms, will be used to develop a geochemical model of the interaction of ground water with commercial solidified low-level wastes. The wastes being evaluated include power reactor waste streams that have been solidified in cement, vinyl ester-styrene, and bitumen. The thermodynamic database for the code is being upgraded before the geochemical modeling is performed. Thermodynamic data for cobalt, antimony, cerium, and cesium solid phases and aqueous species are being added to the database. The need to add these data was identified from the characterization of the waste streams. The geochemical model developed from the laboratory data will then be applied to predict the release from a field-lysimeter facility that contains full-scale waste samples. The contaminant concentrations migrating from the wastes predicted using MINTEQ will be compared to the long-term lysimeter data. This comparison will constitute a partical field validation of the geochemical model. 28 refs

Alternative solidified forms for nuclear wastes

International Nuclear Information System (INIS)

McElroy, J.L.; Ross, W.A.

1976-01-01

Radioactive wastes will occur in various parts of the nuclear fuel cycle. These wastes have been classified in this paper as high-level waste, intermediate and low-level waste, cladding hulls, and residues. Solidification methods for each type of waste are discussed in a multiple barrier context of primary waste form, applicable coatings or films, matrix encapsulation, canister, engineered structures, and geological storage. The four major primary forms which have been most highly developed are glass for HLW, cement for ILW, organics for LLW, and metals for hulls

Printing low-melting-point alloy ink to directly make a solidified circuit or functional device with a heating pen.

Science.gov (United States)

Wang, Lei; Liu, Jing

2014-12-08

A new method to directly print out a solidified electronic circuit through low-melting-point metal ink is proposed. A functional pen with heating capability was fabricated. Several typical thermal properties of the alloy ink Bi 35 In 48.6 Sn 16 Zn 0.4 were measured and evaluated. Owing to the specifically selected melting point of the ink, which is slightly higher than room temperature, various electronic devices, graphics or circuits can be manufactured in a short period of time and then rapidly solidified by cooling in the surrounding air. The liquid-solid phase change mechanism of the written lines was experimentally characterized using a scanning electron microscope. In order to determine the matching substrate, wettability between the metal ink Bi 35 In 48.6 Sn 16 Zn 0.4 and several materials, including mica plate and silicone rubber, was investigated. The resistance-temperature curve of a printed resistor indicated its potential as a temperature control switch. Furthermore, the measured reflection coefficient of a printed double-diamond antenna accords well with the simulated result. With unique merits such as no pollution, no requirement for encapsulation and easy recycling, the present printing approach is an important supplement to current printed electronics and has enormous practical value in the future.

Final report, Task 4: options for on-site management of Nuclear Fuel Services, Inc. high level waste

International Nuclear Information System (INIS)

1978-01-01

Two on-site management options for handling the NFS high-level waste were analyzed: in-tank cement solidification and perpetual tank storage of the liquid waste. The cost of converting the 8D4 plus 8D2 waste to a cementitious solid, including mixing, grout preparation, and transfer to tank 8D1 would require $3,651,000; the cost of cooling the solidified solid for 15 years, plus the cost of filling the rest of the tank space and annulus with grout, plus the cost of minimum surveillance are $10,002,000. Modification of tank 8D2 would be required; prior to transfer of the waste, tank 8D1 would also be modified for cooling of the grout mass. Estimated costs of perpetual tank storage (replacing the existing neutralized waste tank after 10 years, then transferring contents at 50-y intervals for 1000 y, with replacement of ventilation system and auxiliaries at 30-y intervals) would require a sinking fund of $11,039,000. The acidic 8D4 waste would be transferred at 50-y intervals. The sinking fund requirements are sensitive to the difference between the interest rate and the escalation rate, and also to the time assumed from present to the first tank replacement

An overview of the geochemical code MINTEQ: Applications to performance assessment for low-level wastes

International Nuclear Information System (INIS)

Peterson, S.R.; Opitz, B.E.; Graham, M.J.; Eary, L.E.

1987-03-01

The MINTEQ geochemical computer code, developed at the Pacific Northwest Laboratory (PNL), integrates many of the capabilities of its two immediate predecessors, MINEQL and WATEQ3. The MINTEQ code will be used in the Special Waste Form Lysimeters-Arid program to perform the calculations necessary to simulate (model) the contact of low-level waste solutions with heterogeneous sediments of the interaction of ground water with solidified low-level wastes. The code can calculate ion speciation/solubilitya, adsorption, oxidation-reduction, gas phase equilibria, and precipitation/dissolution of solid phases. Under the Special Waste Form Lysimeters-Arid program, the composition of effluents (leachates) from column and batch experiments, using laboratory-scale waste forms, will be used to develop a geochemical model of the interaction of ground water with commercial, solidified low-level wastes. The wastes being evaluated include power-reactor waste streams that have been solidified in cement, vinyl ester-styrene, and bitumen. The thermodynamic database for the code was upgraded preparatory to performing the geochemical modeling. Thermodynamic data for solid phases and aqueous species containing Sb, Ce, Cs, or Co were added to the MINTEQ database. The need to add these data was identified from the characterization of the waste streams. The geochemical model developed from the laboratory data will then be applied to predict the release from a field-lysimeter facility that contains full-scale waste samples. The contaminant concentrations migrating from the waste forms predicted using MINTEQ will be compared to the long-term lysimeter data. This comparison will constitute a partial field validation of the geochemical model

Radiant energy dissipation during final storage of high-level radioactive waste in rock salt

International Nuclear Information System (INIS)

Ramthun, H.

1981-08-01

A final disposal concept is assumed where the high-active waste from 1400 t of uranium, remaining after conditioning, is solidified in borosilicate glass and distributed in 1.760 waste casks. These containers 1.2 m in height and 0.3 m in diameter are to be buried 10 years after the fuel is removed from the reactor in the 300 m deep boreholes of a salt dome. For this design the mean absorbed dose rates are calculated in the glass die (3.9 Gy/s), the steel mantle (0.26 Gy/s) and in the salt rock (0.12 Gy/s at a distance of 1 cm and 0.034 Gy/s at a distance of 9 cm from the container surface) valid at the beginning of disposal. The risk involved with these amounts of stored lattice energy is shortly discussed. (orig.) [de

Strategic review on management and disposal of low- and intermediate-level solid radwastes

International Nuclear Information System (INIS)

Li Xuequn

1993-01-01

An overview on the actual status of solid low- and intermediate-level wastes (L/ILW) management in China is described. Some of the main problems at present are analysed. The strategies on management and disposal of the wastes are discussed in light of systematology. A large amount of solid L/ILW and distilled residual solution to be solidified have been accumulated during the past 30 years development of nuclear industry in China. These wastes, containing fission products, activated products, and uranium and transuranium elements respectively, mainly come from nuclear reactors, spent fuel reprocessing plants, and nuclear fuel fabrication plants. In the century, solid L/ILW and solidified wastes are produced mainly by nuclear industry; but in the next century, solid wastes will be steadily produced mainly from nuclear power plants

Thermoelectric properties by high temperature annealing

Science.gov (United States)

Ren, Zhifeng (Inventor); Chen, Gang (Inventor); Kumar, Shankar (Inventor); Lee, Hohyun (Inventor)

2009-01-01

The present invention generally provides methods of improving thermoelectric properties of alloys by subjecting them to one or more high temperature annealing steps, performed at temperatures at which the alloys exhibit a mixed solid/liquid phase, followed by cooling steps. For example, in one aspect, such a method of the invention can include subjecting an alloy sample to a temperature that is sufficiently elevated to cause partial melting of at least some of the grains. The sample can then be cooled so as to solidify the melted grain portions such that each solidified grain portion exhibits an average chemical composition, characterized by a relative concentration of elements forming the alloy, that is different than that of the remainder of the grain.

Experimental study of directionally solidified ferromagnetic shape memory alloy under multi-field coupling

Energy Technology Data Exchange (ETDEWEB)

Zhu, Yuping, E-mail: zhuyuping@126.com [Institute of Geophysics, China Earthquake Administration, Beijing 100081 (China); Chen, Tao; Teng, Yao [Faculty of Civil Engineering and Mechanics, Jiangsu University, Zhenjiang 212013 (China); Liu, Bingfei [Airport College, Civil Aviation University of China, Tianjin 300300 (China); Xue, Lijun [Tianjin Key Laboratory of the Design and Intelligent Control of the Advanced Mechatronical System, School of Mechanical Engineering, Tianjin University of Technology, Tianjin 300384 (China)

2016-11-01

Directionally solidified, polycrystalline Ni–Mn–Ga is studied in this paper. The polycrystalline Ni–Mn–Ga samples were cut at different angles to solidification direction. The magnetic field induced strain under constant stress and the temperature-induced strain under constant magnetic field during the loading–unloading cycle were measured. The experimental results show that the mechanical behavior during the loading–unloading cycle of the material is nonlinear and anisotropic. Based on the experimental results, the effects of multi-field coupling factors, such as stress, magnetic field, temperature and cutting angle on the mechanical behaviors were analyzed. Some useful conclusions were obtained, which will provide guidance for practical applications. - Highlights: • The magnetic-induced strains in different directions are tested. • The temperature-induced strains in different directions are tested. • The effects of coupling factors on directional solidification samples are studied.

High-level language computer architecture

CERN Document Server

Chu, Yaohan

1975-01-01

High-Level Language Computer Architecture offers a tutorial on high-level language computer architecture, including von Neumann architecture and syntax-oriented architecture as well as direct and indirect execution architecture. Design concepts of Japanese-language data processing systems are discussed, along with the architecture of stack machines and the SYMBOL computer system. The conceptual design of a direct high-level language processor is also described.Comprised of seven chapters, this book first presents a classification of high-level language computer architecture according to the pr

INEL studies concerning solidification of low-level waste in cement

International Nuclear Information System (INIS)

Mandler, J.W.

1989-01-01

The Idaho National Engineering Laboratory (INEL) has performed numerous studies addressing issues concerning the solidification of low-level radioactive waste in cement. These studies have been performed for both the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE). This short presentation will only outline the major topics addressed in some of these studies, present a few conclusions, and identify some of the technical concerns we have. More details of the work and pertinent results will be given in the Working Group sessions. The topics that have been addressed at the INEL which are relevant to this Workshop include (1) solidification of ion-exchange resins and evaporator waste in cement at commercial nuclear power plants, (2) leachability and compressive strength of power plant waste solidified in cement, (3) suggested guidelines for preparation of a solid waste process control program (PCP), (4) cement solidification of EPICOR-II resin wastes, and (5) performance testing of cement-solidified EPICOR-II resin wastes

Liquid-fed ceramic melter: a general description report

International Nuclear Information System (INIS)

Buelt, J.L.; Chapman, C.C.

1978-10-01

The Pacific Northwest Laboratory is conducting several research and development programs for the solidification of high-level wastes. The liquid-fed ceramic melter (LFCM) is a major component in the solidification process. This melter can solidify liquid high-level waste, as well as melt calcined waste with glass additives and then solidify the mixture. This report describes the LFCM system and shows the main features of the refractories, electrodes and power systems, melter box and lid, draining system, feeding system, and off-gas system

Temperature calculations on different configurations for disposal of high-level reprocessing waste in a salt dome model

International Nuclear Information System (INIS)

Hamstra, J.; Kevenaar, J.W.A.M.

1978-06-01

A medium size salt dome is considered as a structure in which a repository can be located for all radioactive wastes to be produced within the scope of a postulated nuclear power program. A dominating design factor for the lay-out of such a waste repository is the temperature distribution in the salt dome resulting from decay heat released from the buried solidified high-level reprocessing waste. Two numerical models are presented for the calculation of both global and local rock salt temperatures. The results of calculations performed with these models are demonstrated to be compatible. Rock salt temperatures related to several types of burial configurations, ranging from two layer configurations with various vertical distances between the layers via a three and a four layer repository to deep bore hole concepts varying from 100 to 600 m bore hole depth, can therefore be calculated with one rather simple unit cell model. The results of these calculations indicate that rock salt temperatures can be kept within acceptable limits to realize a repository using standard mining techniques. The temperatures at mine galery level prove to be a dominating factor in the selection of a repository configuration. More detailed calculations of these temperatures taking into account the loading sequence and the cooling capacity of the mine ventilation are recommended. Finally the apparent advantages of a deep bore hole concept emphasize the need for R and D work with respect to advanced drilling techniques in order to achieve deep dry disposal bore holes that can be realized from a burial mine in the salt dome. (Auth.)

Potential assessment of using fly ash as a binding agent for stabilization and solidification of dredged material; Potentialbedoemning av flygaskor som bindemedelskomponent foer stabilisering och solidifiering (s/s) av muddermassor

Energy Technology Data Exchange (ETDEWEB)

Wilhelmsson, Anna; Holm, Goeran; Lagerlund, Johan; Maijala, Aino; Macsik, Josef

2010-04-15

Over the next few years, about 200 000-800 000 m3 of contaminated sediments, with a muddy, slimy texture, high water ratio and low strength, shall be dredged annually in the development of ports and maintenance dredging of navigable waterways. Dumping at sea is limited since the dredged materials are contaminated. Land disposal requires transports and land area and is thus high in costs. In the construction of new port areas, large volumes of crushed rock, etc. are normally used as construction filling materials. These materials can be replaced by stabilised and solidified dredged materials, with modified geotechnical properties. The method of stabilising/solidifying (s/s) contaminated dredged materials has been used internationally for a long period of time, and, in more recent years, also in the Nordic countries. In Sweden, for instance, the Port of Gaevle and the Port of Oxeloesund have received permissions to reuse s/s-treated contaminated dredged materials in the port structures. Reuse of the stabilised/solidified masses in a geotechnical structure is supported by the new Framework Directive (2008/98/EC) on waste where great emphasis is placed on recycling. Within the project, fly ashes were inventoried with respect to suitability and availability. Five fly ashes, both individual fly ashes and mixtures of different fly ashes, were investigated in the laboratory as a binder component in a binder mix consisting of 50% cement, 20% Merit 5000 and 30% fly ash. Sediment from the Port of Gaevle were stabilised with a binder mixture amount of 150 kg/m3. Produced samples were examined in terms of strength, permeability and leaching. An assessment of the fly ashes' potential was performed based on technological, environmental and economical aspects, as well as market demand and the acceptance of stabilised and solidified dredged materials as construction material. The results show that fly ash, together in a binder mixture with construction cement and slag cement

Relationship between critical current properties and microstructure in cylindrical RE123 melt-solidified bulks

International Nuclear Information System (INIS)

Nakashima, T.; Shimoyama, J.; Honzumi, M.; Tazaki, Y.; Horii, S.; Kishio, K.

2005-01-01

We report the synthesis of cylindrical melt-solidified bulks in REBa 2 Cu 3 O y (RE = Sm, Gd, Dy, Ho, Y and Er), and their critical current properties and microstructures of the a- and the c-growth regions. It was found from the microstructure analysis that the volume fractions of RE211 particles in the c-growth region were always lower than those in the a-growth region. Moreover, those in the c-growth region were increased with distance from the seed crystal. Interestingly, the second peak effects in J c -B curves were prominently enhanced for the c-growth region. J c values at zero field for the c-growth region through the appropriate oxygen post-annealing reached approximately 95 kA cm -2 for RE = Ho, Dy and Y

The influence of Si and V on the kinetics of phase transformation and microstructure of rapidly solidified Al-Fe-Zr alloys

OpenAIRE

Karpe B.; Kosec B.; Nagode A.; Bizjak M.

2013-01-01

The influence of Si and V on the precipitation kinetics of the rapidly solidified (RS) Al-Fe-Zr alloys is presented. Precipitation kinetics and microstructural development of RS Al-Fe-Zr alloys with Si or V addition have been investigated by the combination of four point electrical resistance measurement, optical microscopy, transmition electron microscopy (TEM) and scanning electron microscopy (SEM). For verification of the electrical resistivity measurement results differential scanni...

Solidified package-storage device

International Nuclear Information System (INIS)

Takakura, Masahide

1998-01-01

Vitrification products such as high level radioactive liquid wastes are contained in a solidification package. A containing tube for vertically containing the solidification packages in multi-stages is disposed such that it passes through a ceiling slab. A shielding plug for preventing leakage of radiation from the solidification packages is fitted to an upper opening thereof. A lid of the containing tube is fitted above the plug. The lid is a carbon steel plate having a thickness of 10cm or more. A heat insulation layer comprising glass wool or rock wool is formed on the lower surface of the ceiling slab. A radiation shielding layer comprising such as an iron plate is formed on the lower surface of the heat insulation layer. Then, deterioration of the ceiling slug by heat can be prevented by the heat insulation layer even if high temperature cooling air flown from the upper opening of a ventilation tube should reach the lower surface of the ceiling slab. (I.N.)

Systems analysis approach to the disposal of high-level waste in deep ocean sediments

International Nuclear Information System (INIS)

Marsily, G. de; Hill, M.D.; Murray, C.N.; Talbert, D.M.; Van Dorp, F.; Webb, G.A.M.

1980-01-01

Among the different options being studied for disposal of high-level solidified waste, increasing attention is being paid to that of emplacement of glasses incorporating the radioactivity in deep oceanic sediments. This option has the advantage that the areas of the oceans under investigation appear to be relatively unproductive biologically, are relatively free from cataclysmic events, and are areas in which the natural processes are slow. Thus the environment is stable and predictable so that a number of barriers to the release and dispersion of radioactivity can be defined. Task Groups set up in the framework of the International Seabed Working Group have been studying many aspects of this option since 1976. In order that the various parts of the problem can be assessed within an integrated framework, the methods of systems analysis have been applied. In this paper the Systems Analysis Task Group members report the development of an overall system model. This will be used in an iterative process in which a preliminary analysis, together with a sensitivity analysis, identifies the parameters and data of most importance. The work of the other task groups will then be focussed on these parameters and data requirements so that improved results can be fed back into an improved overall systems model. The major requirements for the development of a preliminary overall systems model are that the problem should be separated into identified elements and that the interfaces between the elements should be clearly defined. The model evolved is deterministic and defines the problem elements needed to estimate doses to man

Storage facility for solid medium level waste at Eurochemic

International Nuclear Information System (INIS)

Balseyro-Castro, M.

1976-01-01

An engineered surface storage facility is described; it will serve for the interim storage of solid and solidified medium-level waste resulting from the reprocessing of irradiated fuels. Up till now, two storage bunkers have been constructed. Each of them is 64 m long, 12 m wide and 8 m high and can take up to about 5,000 drums of 220 1 volume. The drums are stored in a vertical position and in four layers. The waste product drums are transported by a wagon to the entrance of the bunkers from where they are transferred in to the bunker by an overhead crane which is remotely controlled by high-frequency modulated laser beams. A closed-circuit camera is used to watch the handling operations. The waste stored is fully retrievable, either by means of an overhead crane of a lift-truck and can then be transported to an ultimate storage site

Prediction of as-cast grain size of inoculated aluminum alloys melt solidified under non-isothermal conditions

International Nuclear Information System (INIS)

Du, Qiang; Li, Yanjun

2015-01-01

In this paper, a multi-scale as-cast grain size prediction model is proposed to predict as-cast grain size of inoculated aluminum alloys melt solidified under non-isothermal condition, i.e., the existence of temperature gradient. Given melt composition, inoculation and heat extraction boundary conditions, the model is able to predict maximum nucleation undercooling, cooling curve, primary phase solidification path and final as-cast grain size of binary alloys. The proposed model has been applied to two Al-Mg alloys, and comparison with laboratory and industrial solidification experimental results have been carried out. The preliminary conclusion is that the proposed model is a promising suitable microscopic model used within the multi-scale casting simulation modelling framework. (paper)

Preparation and Stability of Inorganic Solidified Foam for Preventing Coal Fires

Directory of Open Access Journals (Sweden)

Botao Qin

2014-01-01

Full Text Available Inorganic solidified foam (ISF is a novel material for preventing coal fires. This paper presents the preparation process and working principle of main installations. Besides, aqueous foam with expansion ratio of 28 and 30 min drainage rate of 13% was prepared. Stability of foam fluid was studied in terms of stability coefficient, by varying water-slurry ratio, fly ash replacement ratio of cement, and aqueous foam volume alternatively. Light microscope was utilized to analyze the dynamic change of bubble wall of foam fluid and stability principle was proposed. In order to further enhance the stability of ISF, different dosage of calcium fluoroaluminate was added to ISF specimens whose stability coefficient was tested and change of hydration products was detected by scanning electron microscope (SEM. The outcomes indicated that calcium fluoroaluminate could enhance the stability coefficient of ISF and compact hydration products formed in cell wall of ISF; naturally, the stability principle of ISF was proved right. Based on above-mentioned experimental contents, ISF with stability coefficient of 95% and foam expansion ratio of 5 was prepared, which could sufficiently satisfy field process requirements on plugging air leakage and thermal insulation.

RPython high-level synthesis

Science.gov (United States)

Cieszewski, Radoslaw; Linczuk, Maciej

2016-09-01

The development of FPGA technology and the increasing complexity of applications in recent decades have forced compilers to move to higher abstraction levels. Compilers interprets an algorithmic description of a desired behavior written in High-Level Languages (HLLs) and translate it to Hardware Description Languages (HDLs). This paper presents a RPython based High-Level synthesis (HLS) compiler. The compiler get the configuration parameters and map RPython program to VHDL. Then, VHDL code can be used to program FPGA chips. In comparison of other technologies usage, FPGAs have the potential to achieve far greater performance than software as a result of omitting the fetch-decode-execute operations of General Purpose Processors (GPUs), and introduce more parallel computation. This can be exploited by utilizing many resources at the same time. Creating parallel algorithms computed with FPGAs in pure HDL is difficult and time consuming. Implementation time can be greatly reduced with High-Level Synthesis compiler. This article describes design methodologies and tools, implementation and first results of created VHDL backend for RPython compiler.

Evaluation of leaching behavior and immobilization of zinc in cement-based solidified products

Directory of Open Access Journals (Sweden)

Krolo Petar

2012-01-01

Full Text Available This study has examined leaching behavior of monolithic stabilized/solidified products contaminated with zinc by performing modified dynamic leaching test. The effectiveness of cement-based stabilization/solidification treatment was evaluated by determining the cumulative release of Zn and diffusion coefficients, De. The experimental results indicated that the cumulative release of Zn decreases as the addition of binder increases. The values of the Zn diffusion coefficients for all samples ranged from 1.210-8 to 1.1610-12 cm2 s-1. The samples with higher amounts of binder had lower De values. The test results showed that cement-based stabilization/solidification treatment was effective in immobilization of electroplating sludge and waste zeolite. A model developed by de Groot and van der Sloot was used to clarify the controlling mechanisms. The controlling leaching mechanism was found to be diffusion for samples with small amounts of waste material, and dissolution for higher waste contents.

Dendritic coarsening of γ' phase in a directionally solidified superalloy during 24,000 h of exposure at 1173 K

International Nuclear Information System (INIS)

Li, H.; Wang, L.; Lou, L.H.

2010-01-01

Dendritic coarsening of γ' was investigated in a directionally solidified Ni-base superalloy during exposure at 1173 K for 24,000 h. Chemical homogeneity along different directions and residual internal strain in the experimental superalloy were measured by electronic probe microanalysis (EPMA) and electron back-scattered diffraction (EBSD) technique. It was indicated that the gradient of element distribution was anisotropic and the inner strain between dendrite core and interdendritic regions was different even after 24,000 h of exposure at 1173 K, which influenced the kinetics for the dendrite coarsening of γ' phase.

Solidification of highly active fission products by a thermite reaction. Pt. 1

International Nuclear Information System (INIS)

Rudolph, G.; Hild, W.

1976-07-01

To solidify high-level fission products a process was developed according to which a high-melting ceramic product is obtained as a solidification matrix in a thermite reaction. With a constant content of fission product oxides reaction mixtures consisting of 35 to 55 wt.% of manganese dioxide, 24 to 32 wt.% of aluminum shot and 17 to 36 wt.% of sand give suitable products. In the thermite reactiom some components contained in the reactic mixture volatilize partly by evaporation (alkali oxides, manganese oxide, and others) and partly by the formation of volatile oxides having lower valencies (silicon and aluminum oxide). The smoke generated can be easily collected in filters made of glass wool fibers. (orig./HR) [de

Assessment of phase constitution on the Al-rich region of rapidly solidified Al-Co-Fe-Cr alloys

International Nuclear Information System (INIS)

Wolf, W.; Bolfarini, C.; Kiminami, C.S.; Botta, W.J.

2016-01-01

The formation of quasicrystalline approximants in rapidly solidified Al-Co-Fe-Cr alloys was investigated. Alloys of atomic composition Al 71 Co 13 Fe 8 Cr 8 , Al 77 Co 11 Fe 6 Cr 6 and Al 76 Co 19 Fe 4 Cr 1 were produced using melt spinning and arc melting methods and their microstructural characterization was carried out by X-ray diffraction, scanning electron microscopy and transmission electron microscopy. Up to the present there is no consensus in the literature regarding the formation of quasicrystalline phase or quasicrystalline approximants in the Al 71 Co 13 Fe 8 Cr 8 alloy. This work presents, for the first time, a detailed structural characterization of selected alloys in the Al-Co-Fe-Cr system close to the atomic composition Al 71 Co 13 Fe 8 Cr 8 . The results indicated the samples to be composed, mostly, by two intermetallic phases, which are quaternary extensions of Al 5 Co 2 and Al 13 Co 4 and are quasicrystalline approximants. Although the Al 5 Co 2 phase has already been reported in the Al 71 Co 13 Fe 8 Cr 8 alloy, the presence of the monoclinic Al 13 Co 4 is now identified for the first time in the as cast state. In the binary Al-Co system a quasicrystalline phase is known to form in a rapidly solidified alloy with composition close to the monoclinic and orthorhombic Al 13 Co 4 phases. This binary quasicrystalline phase presents an average valence electron per atom (e/a) between 1.7 and 1.9; thus, in addition to the Al 71 Co 13 Fe 8 Cr 8 alloy, the compositions Al 77 Co 11 Fe 6 Cr 6 and Al 76 Co 19 Fe 4 Cr 1 were chosen to be within the region of formation of the quaternary extension of the Al 13 Co 4 phase and also within the (e/a) of 1.7 to 1.9. However, no quasicrystalline phase is present in any of the studied alloys. The Al-Co-Fe-Cr system, around the compositions studied, is composed of quaternary extensions of Al-Co intermetallic phases, which present solubility of Fe and Cr at Co atomic sites. - Highlights: •The Al rich region of the Al

Microbial degradation of low-level radioactive waste. Volume 2, Annual report for FY 1994

International Nuclear Information System (INIS)

Rogers, R.D.; Hamilton, M.A.; Veeh, R.H.; McConnell, J.W. Jr.

1995-08-01

The Nuclear Regulatory Commission stipulates in 10 CFR 61 that disposed low-level radioactive waste (LLW) be stabilized. To provide guidance to disposal vendors and nuclear station waste generators for implementing those requirements, the NRC developed the Technical Position on Waste Form, Revision 1. That document details a specified set of recommended testing procedures and criteria, including several tests for determining the biodegradation properties of waste forms. Cement has been widely used to solidify LLW; however, the resulting waste forms are sometimes susceptible to failure due to the actions of waste constituents, stress, and environment. The purpose of this research program is to develop modified microbial degradation test procedures that will be more appropriate than the existing procedures for evaluating the effects of microbiologically influenced chemical attack on cement-solidified LLW. Groups of microorganisms indigenous to LLW disposal sites are being employed that can metabolically convert organic and inorganic substrates into organic and mineral acids. Such acids aggressively react with cement and can ultimately lead to structural failure. Results over the past year on the application of mechanisms inherent in microbially influenced degradation of cement-based material are the focus of the annual report. Data-validated evidence of the potential for microbially influenced deterioration of cement-solidified LLW and subsequent release of radionuclides has been developed during this study

Experimental study on thermal interaction between a high-temperature molten jet and plates

International Nuclear Information System (INIS)

Sato, K.; Saito, M.; Furutani, A.; Isozaki, M.; Imahori, S.; Konishi, K.

1994-01-01

This paper summarizes the recent simulant experiments to study molten corium-structure interactions under postulated core disruptive accident (CDA) conditions in liquid-metal fast breeder reactors (LMFMRs). These experiments were conducted in the MELT-II facility generating high-temperature molten simulants by an induction heating technique. From a series of molten jet-structure interaction experiments, the effects of the solidified crust layer and molten layer on the erosion behavior were identified, and analytical models were developed to assess the structure erosion rate with and without crust formation. Especially, we revealed the inherent mitigation mechanism that when the molten oxide jet with high melting point falls down onto the structure plate, solidified crust of the oxide can significantly reduce the erosion rate. (author)

Rapid Solidification of Sn-Cu-Al Alloys for High-Reliability, Lead-Free Solder: Part I. Microstructural Characterization of Rapidly Solidified Solders

Science.gov (United States)

Reeve, Kathlene N.; Choquette, Stephanie M.; Anderson, Iver E.; Handwerker, Carol A.

2016-12-01

Particles of Cu x Al y in Sn-Cu-Al solders have previously been shown to nucleate the Cu6Sn5 phase during solidification. In this study, the number and size of Cu6Sn5 nucleation sites were controlled through the particle size refinement of Cu x Al y via rapid solidification processing and controlled cooling in a differential scanning calorimeter. Cooling rates spanning eight orders of magnitude were used to refine the average Cu x Al y and Cu6Sn5 particle sizes down to submicron ranges. The average particle sizes, particle size distributions, and morphologies in the microstructures were analyzed as a function of alloy composition and cooling rate. Deep etching of the samples revealed the three-dimensional microstructures and illuminated the epitaxial and morphological relationships between the Cu x Al y and Cu6Sn5 phases. Transitions in the Cu6Sn5 particle morphologies from faceted rods to nonfaceted, equiaxed particles were observed as a function of both cooling rate and composition. Initial solidification cooling rates within the range of 103 to 104 °C/s were found to be optimal for realizing particle size refinement and maintaining the Cu x Al y /Cu6Sn5 nucleant relationship. In addition, little evidence of the formation or decomposition of the ternary- β phase in the solidified alloys was noted. Solidification pathways omitting the formation of the ternary- β phase agreed well with observed room temperature microstructures.

Giant Enhancement of Magnetostrictive Response in Directionally-Solidified Fe83Ga17Erx Compounds

Directory of Open Access Journals (Sweden)

Radhika Barua

2018-06-01

Full Text Available We report, for the first time, correlations between crystal structure, microstructure and magnetofunctional response in directionally solidified [110]-textured Fe83Ga17Erx (0 < x < 1.2 alloys. The morphology of the doped samples consists of columnar grains, mainly composed of a matrix phase and precipitates of a secondary phase deposited along the grain boundary region. An enhancement of more than ~275% from ~45 to 170 ppm is observed in the saturation magnetostriction value (λs of Fe83Ga17Erx alloys with the introduction of small amounts of Er. Moreover, it was noted that the low field derivative of magnetostriction with respect to an applied magnetic field (i.e., dλs/dHapp for Happ up to 1000 Oe increases by ~230% with Er doping (dλs/dHapp,FeGa= 0.045 ppm/Oe; dλs/dHapp,FeGaEr= 0.15 ppm/Oe. The enhanced magnetostrictive response of the Fe83Ga17Erx alloys is ascribed to an amalgamation of microstructural and electronic factors, namely: (i improved grain orientation and local strain effects due to deposition of Er in the intergranular region; and (ii strong local magnetocrystalline anisotropy, due to the highly anisotropic localized nature of the 4f electronic charge distribution of the Er atom. Overall, this work provides guidelines for further improving galfenol-based materials systems for diverse applications in the power and energy sector.

High potassium level

Science.gov (United States)

... level is very high, or if you have danger signs, such as changes in an ECG . Emergency ... Seifter JL. Potassium disorders. In: Goldman L, Schafer AI, eds. Goldman-Cecil Medicine . 25th ed. Philadelphia, PA: ...

High-level Petri Nets

DEFF Research Database (Denmark)

various journals and collections. As a result, much of this knowledge is not readily available to people who may be interested in using high-level nets. Within the Petri net community this problem has been discussed many times, and as an outcome this book has been compiled. The book contains reprints...... of some of the most important papers on the application and theory of high-level Petri nets. In this way it makes the relevant literature more available. It is our hope that the book will be a useful source of information and that, e.g., it can be used in the organization of Petri net courses. To make......High-level Petri nets are now widely used in both theoretical analysis and practical modelling of concurrent systems. The main reason for the success of this class of net models is that they make it possible to obtain much more succinct and manageable descriptions than can be obtained by means...

Thermal characteristics of highly compressed bentonite

International Nuclear Information System (INIS)

Sueoka, Tooru; Kobayashi, Atsushi; Imamura, S.; Ogawa, Terushige; Murata, Shigemi.

1990-01-01

In the disposal of high level radioactive wastes in strata, it is planned to protect the canisters enclosing wastes with buffer materials such as overpacks and clay, therefore, the examination of artificial barrier materials is an important problem. The concept of the disposal in strata and the soil mechanics characteristics of highly compressed bentonite as an artificial barrier material were already reported. In this study, the basic experiment on the thermal characteristics of highly compressed bentonite was carried out, therefore, it is reported. The thermal conductivity of buffer materials is important because the possibility that it determines the temperature of solidified bodies and canisters is high, and the buffer materials may cause the thermal degeneration due to high temperature. Thermophysical properties are roughly divided into thermodynamic property, transport property and optical property. The basic principle of measured thermal conductivity and thermal diffusivity, the kinds of the measuring method and so on are explained. As for the measurement of the thermal conductivity of highly compressed bentonite, the experimental setup, the procedure, samples and the results are reported. (K.I.)

A study on crystalline phases present in the as-solidified and crystallized microstructures in Zr53Cu21Al10Ni8Ti8 alloy

International Nuclear Information System (INIS)

Neogy, S.; Tewari, R.; Srivastava, D.; Dey, G.K.

2011-01-01

In the present study the as-solidified and crystallized microstructures of Zr 53 Cu 21 Al 10 Ni 8 Ti 8 alloy have been examined in detail. Solidification was carried out by melt spinning, suction casting and copper mould casting techniques. The last technique yielded a partially crystalline microstructure, whereas, the other two techniques resulted in amorphous microstructures. (author)

Development of high temperature fasteners using directionally solidified eutectic alloys

Science.gov (United States)

George, F. D.

1972-01-01

The suitability of the eutectics for high temperature fasteners was investigated. Material properties were determined as a function of temperature, and included shear parallel and perpendicular to the growth direction and torsion parallel to it. Techniques for fabricating typical fastener shapes included grinding, creep forming, and direct casting. Both lamellar Ni3Al-Ni3Nb and fibrous (Co,Cr,Al)-(Cr,Co)7C3 alloys showed promise as candidate materials for high temperature fastener applications. A brief evaluation of the performance of the best fabricated fastener design was made.

Cement encapsulation of low-level radioactive slurries of complex chemistry

International Nuclear Information System (INIS)

Cau Dit Coumes, C.

2000-01-01

Investigations have been carried out to solidify in cement a low-level radioactive waste of complex chemistry which should be produced in a new plant designed to process radioactive effluents from CEA Cadarache Research Center. Direct cementation comes up against a major problem: a very long setting time of cement due to strong inhibition by borates from the waste. A two-stage process, including a chemical treatment prior to immobilization, has been elaborated and the resulted material characterized. (authors)

Effect of Bi-content on hardness and micro-creep behavior of Sn-3.5Ag rapidly solidified alloy

Energy Technology Data Exchange (ETDEWEB)

Kamal, M. [Metal Physics Laboratory, Faculty of Science, Mansoura University (Egypt); Gouda, El Said [Metal Physics Laboratory, Department of Solid State Physics, Physics Division, National Research Center, Dokki, Giza (Egypt); Marei, L.K. [Faculty of Petroleum and Mining Engineering, Suez Canal University, Suez (Egypt)

2009-12-15

In the present paper, the influence of 1, 3, 5 and 10 % Bi (weight %) as ternary additions on structure, melting and mechanical properties of rapidly solidified Sn-3.5Ag alloy has been investigated. The effect of Bi was discussed based on the experimental results. The experimental results showed that the alloys of Sn-3.5Ag, Sn-3.5Ag-1Bi and Sn-3.5Ag-3Bi are composed of two phases; Ag{sub 3}Sn IMC embedded in Sn matrix phase, which indicated that the solubility of Bi phase in Sn-matrix was extended to 3 % as a result of rapid solidification. Bi precipitation in Sn matrix was only observed in Sn-3.5Ag-5Bi and Sn-3.5Ag-10Bi alloys. Also, addition of Bi decreased continuously the melting point of the eutectic Sn-3.5Ag alloy to 202.6 C at 10 % Bi. Vickers hardness of Sn-3.5Ag rapidly solidified alloy increased with increasing Bi content up to 3 % due to supersaturated solid solution strengthening hardening mechanism of Bi phase in Sn matrix, while the alloys contain 5 and 10 % Bi exhibited lower values of Vickers hardness. The lower values can be attributed to the precipitation of Bi as a secondary phase which may form strained regions due to the embrittlement of Bi atom. In addition, the effect of Bi addition on the micro-creep behavior of Sn-3.5Ag alloy as well as the creep rate have been described and has been calculated at room temperature. (copyright 2009 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim) (orig.)

Development of methods to extralt and solidify highly radioactive waste

International Nuclear Information System (INIS)

Arnek, R.; Persson, A.; Faelth, L.; Annehed, H.

1977-06-01

Zeolites are proposed as selective ion exchange materials to extract highly radioactive fission products as cesium 137 and strontium 90, and corrosion products. The zeolites 13X, F and PC showed a high adsorption capacity for cesium and strontium. A heat treatment at 800-1300 degrees C for about two hours gave a vitrified material. The chemical resistance of the heat treated zeolites was tested in a soxhlets-apparatus, were a streaming solution at 100 degrees C was in contact with the zeolite for 1-2 days. For all cases, the amount of dissolved strontium was below the detection threshold.(L.K.)

Evaluation of systems incorporating transmutation for the reduction of the long term toxicity of high-level waste

International Nuclear Information System (INIS)

Davidson, J.W.

1979-01-01

One of the alternative high-level nuclear waste (HLW) management/disposal concepts proposed involves the separation from HLW of the elements with isotopes which dominate the radiotoxicity and the transmutation of these nuclides to shortlived or stable products. The waste management system required for transmutation employs chemical processing of HLW to recover waste nuclides for irradiation with neutrons in a transmutation device. The transmuter periodically requires replenishment of the target nuclides and chemical processing to remove the transmutation products. The waste streams from HLW processing and product recovery together comprise the discharge from the system. An imploding liner fusion reactor (ILFR) is assumed for the transmuter with the waste nuclides dissolved in a molten lead-lithium alloy blanket. The potential transmutation candidates are defined as the elements with toxicities per unit volume (toxicity indexes) in solidified HLW at 1000 years which are greater than that for 0.2% uranium ore (carnotite). The candidates which require separation for transmutation are the actinides; Np, Pu, Am, and Cu and the fission products; I and Tc. Certain assumptions were made for the parameters for the ILFR and its operating conditions, and a system evaluation was done. System evaluations indicate that blanket waste loadings on the order of several percent of the total concentration result in attractive performance in terms of high transmutation capacities and low blanket processing requirments. It appears that transmutation system goals in terms of toxicity reduction are achievable with a modest number of transmuters. In addition, requirements for transmuter performance, chemical processing capacity and chemical separation efficiency appear to be within projected values for this technology

Analysis of nuclear waste management

International Nuclear Information System (INIS)

Center, J.L.; Crawford, B.S.; Ross, B.; Sutherland, A.A. Jr.

1976-01-01

An event tree is developed, outlining ways which radioactivity can be accidentally released from high level solidified wastes. Probabilities are assigned to appropriate events in the tree and the major contributors to dose to the general population are identified. All doses are computed on a per megawatt electric-year basis. Sensitivity relations between the expected dose and key characteristics of the solidified wasted are developed

The integrated melter off-gas treatment systems at the West Valley Demonstration Project

International Nuclear Information System (INIS)

Vance, R.F.

1991-12-01

The West Valley Demonstration project was established by an act of Congress in 1980 to solidify the high level radioactive liquid wastes produced from operation of the Western New York Nuclear Services Center from 1966 to 1972. The waste will be solidified as borosilicate glass. This report describes the functions, the controlling design criteria, and the resulting design of the melter off-gas treatment systems

Electrochemical properties of rapidly solidified Si-Ti-Ni(-Cu) base anode for Li-ion rechargeable batteries

Science.gov (United States)

Kwon, Hye Jin; Sohn, Keun Yong; Park, Won-Wook

2013-11-01

In this study, rapidly solidified Si-Ti-Ni-Cu alloys have been investigated as high capacity anodes for Li-ion secondary batteries. To obtain nano-sized Si particles dispersed in the inactive matrix, the alloy ribbons were fabricated using the melt spinning process. The thin ribbons were pulverized using ball-milling to make a fine powder of ˜ 4 µm average size. Coin-cell assembly was carried out under an argon gas in a glove box, in which pure lithium was used as a counter-electrode. The cells were cycled using the galvanostatic method in the potential range of 0.01 V and 1.5 V vs. Li/Li+. The microstructure and morphology were examined using an x-ray diffractometer, Field-Emission Scanning Electron Microscopy and High Resolution Transmission Electron Microscopy. Among the anode alloys, the Si70Ti15Ni15 electrodes had the highest discharge capacity (974.1 mAh/g) after the 50th cycle, and the Si60Ti16Ni16Cu8 electrode showed the best coulombic efficiency of ˜95.9% in cyclic behavior. It was revealed that the Si7Ni4Ti4 crystal phase coexisting with an amorphous phase, could more efficiently act as a buffer layer than the fully crystallized Si7Ni4Ti4 phase. Consequently, the electrochemical properties of the anode materials pronouncedly improved when the nano-sized primary Si particle was dispersed in the inactive Si7Ni4Ti4-based matrix mixed with an amorphous structure.

The Characterization of Filtration Waste Solidified Product from Baghouse Filter of the Incineration Process

International Nuclear Information System (INIS)

Sutoto

2000-01-01

To increase of the safety, quality and to easy maintenance of the incinerator media of bag house filter, coating of the surface filter media by CaCO 3 powder were done. In the incinerator process, the CaCO 3 powder will scrub of fly ash as secondary waste. And finally, both of the secondary waste and CaCO 3 will immobilized by cement matrix. The research has an objective to study and characterizing of the CaCO 3 as secondary waste on their cemented product. The research were done on block samples with content of CaCO 3 and the properties characterized by compressive strength and density. From this research known that on their solidified, each quantity of CaCO 3 will be impact to decreasing of the quality cementation product. The optimum formula for solidification of bag house filter scrubbed is CaCO 3 : cement: water is 3 : 10 : 7. (author)

Radioactivity evaluation method for pre-packed concrete packages of low-level dry active wastes

International Nuclear Information System (INIS)

Sakai, Toshiaki; Funahashi, Tetsuo; Watabe, Kiyomi; Ozawa, Yukitoshi; Kashiwagi, Makoto

1998-01-01

Low-level dry active wastes of nuclear power plants are grouted with cement mortal in a container and planned to disposed into the shallow land disposal site. The characteristics of radionuclides contained in dry active wastes are same as homogeneous solidified wastes. In the previous report, we reported the applicability of the radioactivity evaluation methods established for homogeneous solidified wastes to pre-packed concrete packages. This report outlines the developed radioactivity evaluation methods for pre-packed concrete packages based upon recent data. Since the characteristics of dry active wastes depend upon the plant system in which dry active wastes originate and the types of contamination, sampling of wastes and activity measurement were executed to derive scaling factors. The radioactivity measurement methods were also verified. The applicability of non-destructive methods to measure radioactivity concentration of pre-packed concrete packages was examined by computer simulation. It is concluded that those methods are accurate enough to measure actual waste packages. (author)

Variation of long-period stacking order structures in rapidly solidified Mg97Zn1Y2 alloy

International Nuclear Information System (INIS)

Matsuda, M.; Ii, S.; Kawamura, Y.; Ikuhara, Y.; Nishida, M.

2005-01-01

The long-period stacking order (LPSO) structures in rapidly solidified Mg 97 Zn 1 Y 2 alloy have been studied by conventional and high-resolution transmission electron microscopes (HRTEMs). There are four kinds of stacking sequences in the LPSO structures, i.e., 18R of ABABABCACACABCBCBC, 14H of ACBCBABABABCBC, 10H of ABACBCBCAB and 24R of ABABABABCACACACABCBCBCBC. The 18R structure is dominantly observed in the present study. The rest three are occasionally observed in places. The 10H and 24R structures are recently discovered. The lattice constants of 18R(1-bar 1-bar -bar 1-bar 1-bar -bar 2) 3 , 14H(2-bar -bar 1-bar 2-bar -bar 1-bar 1-bar -bar 1-bar 1-bar -bar 2-bar 1-bar -bar 2), 10H(1-bar 3-bar -bar 1-bar 1-bar -bar 3-bar 1-bar ) and 24R(1-bar 1-bar -bar 1-bar 1-bar -bar 1-bar 1-bar -bar 2) 3 structures are estimated to be a=0.320nm and c=4.678nm, a=0.325nm and c=3.694nm, a=0.325nm and c=2.603nm, a=0.322nm and c=6.181nm for the hexagonal structure, respectively

Low-level radioactive waste management

Energy Technology Data Exchange (ETDEWEB)

Ishihara, T [Radioactive Waste Management Center, Tokyo (Japan)

1980-08-01

In the development and utilization of nuclear energy, variety of radioactive wastes arise. A largest part is low level radioactive wastes. In Japan, they are concentrated and solidified, and stored in drums. However, no low level wastes have yet been finally disposed of; there are now about 260,000 drums of such wastes stored on the sites. In Japan, the land is narrow, and its structure is geologically unstable, so that the sea disposal is sought. On the other hand, the development of technology for the ground disposal has lagged behind the sea disposal until recently because of the law concerned. The following matters are described: for the sea disposal, preparatory technology studies, environment safety assessment, administrative measures, and international control; for the ground disposal, experiments, surveys, disposal site selection, and the concept of island repositories.

Assessment of phase constitution on the Al-rich region of rapidly solidified Al-Co-Fe-Cr alloys

Energy Technology Data Exchange (ETDEWEB)

Wolf, W., E-mail: witorw@gmail.com [Programa de Pós-Graduação em Ciência e Engenharia de Materiais, Universidade Federal de São Carlos, Rod. Washington Luiz, Km 235, 13565-905 São Carlos, SP (Brazil); Bolfarini, C., E-mail: cbolfa@ufscar.br [Departamento de Engenharia de Materiais, Universidade Federal de São Carlos, Rod. Washington Luiz, Km 235, 13565-905 São Carlos, SP (Brazil); Kiminami, C.S., E-mail: kiminami@ufscar.br [Departamento de Engenharia de Materiais, Universidade Federal de São Carlos, Rod. Washington Luiz, Km 235, 13565-905 São Carlos, SP (Brazil); Botta, W.J., E-mail: wjbotta@ufscar.br [Departamento de Engenharia de Materiais, Universidade Federal de São Carlos, Rod. Washington Luiz, Km 235, 13565-905 São Carlos, SP (Brazil)

2016-12-15

The formation of quasicrystalline approximants in rapidly solidified Al-Co-Fe-Cr alloys was investigated. Alloys of atomic composition Al{sub 71}Co{sub 13}Fe{sub 8}Cr{sub 8}, Al{sub 77}Co{sub 11}Fe{sub 6}Cr{sub 6} and Al{sub 76}Co{sub 19}Fe{sub 4}Cr{sub 1} were produced using melt spinning and arc melting methods and their microstructural characterization was carried out by X-ray diffraction, scanning electron microscopy and transmission electron microscopy. Up to the present there is no consensus in the literature regarding the formation of quasicrystalline phase or quasicrystalline approximants in the Al{sub 71}Co{sub 13}Fe{sub 8}Cr{sub 8} alloy. This work presents, for the first time, a detailed structural characterization of selected alloys in the Al-Co-Fe-Cr system close to the atomic composition Al{sub 71}Co{sub 13}Fe{sub 8}Cr{sub 8}. The results indicated the samples to be composed, mostly, by two intermetallic phases, which are quaternary extensions of Al{sub 5}Co{sub 2} and Al{sub 13}Co{sub 4} and are quasicrystalline approximants. Although the Al{sub 5}Co{sub 2} phase has already been reported in the Al{sub 71}Co{sub 13}Fe{sub 8}Cr{sub 8} alloy, the presence of the monoclinic Al{sub 13}Co{sub 4} is now identified for the first time in the as cast state. In the binary Al-Co system a quasicrystalline phase is known to form in a rapidly solidified alloy with composition close to the monoclinic and orthorhombic Al{sub 13}Co{sub 4} phases. This binary quasicrystalline phase presents an average valence electron per atom (e/a) between 1.7 and 1.9; thus, in addition to the Al{sub 71}Co{sub 13}Fe{sub 8}Cr{sub 8} alloy, the compositions Al{sub 77}Co{sub 11}Fe{sub 6}Cr{sub 6} and Al{sub 76}Co{sub 19}Fe{sub 4}Cr{sub 1} were chosen to be within the region of formation of the quaternary extension of the Al{sub 13}Co{sub 4} phase and also within the (e/a) of 1.7 to 1.9. However, no quasicrystalline phase is present in any of the studied alloys. The Al-Co-Fe-Cr system

Preliminary engineering and economic analysis of the fixation of high-level radioactive wastes in concrete

International Nuclear Information System (INIS)

Weeren, H.O.; Perona, J.J.

1979-07-01

This feasibility study was based on a waste fixation facility that would serve a reprocessing plant with a capacity of 5 metric tons of uranium per day (MTU/day). Postirradiation cooling times of 3 to 10 years prior to waste solidification were assumed. The waste solution would be concentrated, denitrated, mixed with cement, and cast under pressure in cylindrical canisters similar to those envisioned for a glass facility. The solidified waste grout would be vented, to allow the free water to escape, and then sealed. The filled canisters would be shipped to a geologic repository for permanent storage. Recent work with concretes formed under elevated temperatures and pressures (FUETAP) indicates that they are highly leach resistant. The operating costs were estimated for a waste fixation facility under several conditions. Operating costs for a glass fixation facility were also estimated and compared with the operating costs for a concrete fixation facility. The principal conclusion is that concrete could be an alternative to glass as a matrix for fixation of wastes with high heat-generation rates. The operating costs of an optimized concrete fixation process would probably not be greatly higher than the operating costs of a glass plant, and the capital costs would almost surely be lower. In addition, the concrete process is not a high-temperature process and would not have the consequent operating problems

Radionuclide release from solidified high level waste Task 3 Characterization of radioactive waste forms a series of final reports (1985-89) No 19

International Nuclear Information System (INIS)

Boult, K.A.; Dalton, J.T.; Hough, A.; Marples, J.A.C.; Robertson, G.P.; Wilkins, R.I.

1991-01-01

Samples of glass were made up containing a full inactive simulant of the high-level waste. These were doped with isotopes of the four radioelements Am, Pu, Np, Tc and after crushing were mixed with possible components of the repository and with water and loaded into capsules. The capsules were held in an oven at, normally, 60 0 C for periods of up to a year before they were opened and the water overlying the solids sampled and analyzed. After a series of similar such experiments, the following conclusions were obtained: (a) In the presence of a backfill containing ordinary Portland cement (OPC) and under reducing conditions, the steady-state concentrations of Tc and the actinides Np and Am, measured using doped glasses, were respectively ca. 0.3, 1 and 5 times the limiting concentration. (b) Under the same conditions, the steady-state concentration of Pu increased from 0.03 times the limiting concentration to 15 times it when the Pu concentration in the glass was increased from 6 X 10 -5 wt% to 0.12 wt%. (c) Bentonite did not absorb Np and Am as efficiently as the cements. (d) Under oxidizing conditions, Tc was quite soluble, the steady-state concentration being about 1 000 times the limiting concentration. Further results concerning steady-state concentrations of Np, Pu, Am and Tc under varying conditions as well as with various barrier materials and leachants are discussed in this report

Field lysimeter investigations: Low-level waste data base development program for fiscal year 1995. Volume 8, Annual report, October 1994-- September 1995

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Rogers, R.D.; Larsen, I.L.; Jastrow, J.D.; Sanford, W.E.; Sullivan, T.M.

1996-06-01

The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the U.S. Nuclear Regulatory Commission, is (a) studying the degradation effects in organic ion-exchange resins caused by radiation, (b) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified ion-exchange resins, (c) obtaining performance information on solidified ion- exchange resins in a disposal environment, and (d) determining the condition of liners used to dispose the ion-exchange resins. Compressive test results of 12-year-old cement and vinyl ester- styrene solidified waste form samples are presented, which show effects of aging and self-irradiation. Results of the tenth year of data acquisition from the field testing are presented and discussed. During the continuing field testing, both portland type I-II cement and Dow vinyl ester-styrene waste form samples are being tested in lysimeter arrays located at Argonne National Laboratory-East in Illinois and at Oak Ridge National Laboratory. The study is designed to provide continuous data on nuclide release and movement, as well as environmental conditions, over a 20-year period

An Abnormal Increase of Fatigue Life with Dwell Time during Creep-Fatigue Deformation for Directionally Solidified Ni-Based Superalloy DZ445

Science.gov (United States)

Ding, Biao; Ren, Weili; Deng, Kang; Li, Haitao; Liang, Yongchun

2018-03-01

The paper investigated the creep-fatigue behavior for directionally solidified nickel-based superalloy DZ445 at 900 °C. It is found that the fatigue life shows an abnormal increase when the dwell time exceeds a critical value during creep-fatigue deformation. The area of hysteresis loop and fractograph explain the phenomenon quite well. The shortest life corresponds to the maximal area of hysteresis loop, i. e. the maximum energy to be consumed during the creep-fatigue cycle. The fractographic observation of failed samples further supports the abnormal behavior of fatigue life.

EAP high-level product architecture

DEFF Research Database (Denmark)

Guðlaugsson, Tómas Vignir; Mortensen, Niels Henrik; Sarban, Rahimullah

2013-01-01

EAP technology has the potential to be used in a wide range of applications. This poses the challenge to the EAP component manufacturers to develop components for a wide variety of products. Danfoss Polypower A/S is developing an EAP technology platform, which can form the basis for a variety...... of EAP technology products while keeping complexity under control. High level product architecture has been developed for the mechanical part of EAP transducers, as the foundation for platform development. A generic description of an EAP transducer forms the core of the high level product architecture...... the function of the EAP transducers to be changed, by basing the EAP transducers on a different combination of organ alternatives. A model providing an overview of the high level product architecture has been developed to support daily development and cooperation across development teams. The platform approach...

High Level Radioactive Waste Management

International Nuclear Information System (INIS)

1991-01-01

The proceedings of the second annual international conference on High Level Radioactive Waste Management, held on April 28--May 3, 1991, Las Vegas, Nevada, provides information on the current technical issue related to international high level radioactive waste management activities and how they relate to society as a whole. Besides discussing such technical topics as the best form of the waste, the integrity of storage containers, design and construction of a repository, the broader social aspects of these issues are explored in papers on such subjects as conformance to regulations, transportation safety, and public education. By providing this wider perspective of high level radioactive waste management, it becomes apparent that the various disciplines involved in this field are interrelated and that they should work to integrate their waste management activities. Individual records are processed separately for the data bases

Disposal of high level and intermediate level radioactive wastes

International Nuclear Information System (INIS)

Flowers, R.H.

1991-01-01

The waste products from the nuclear industry are relatively small in volume. Apart from a few minor gaseous and liquid waste streams, containing readily dispersible elements of low radiotoxicity, all these products are processed into stable solid packages for disposal in underground repositories. Because the volumes are small, and because radioactive wastes are latecomers on the industrial scene, a whole new industry with a world-wide technological infrastructure has grown up alongside the nuclear power industry to carry out the waste processing and disposal to very high standards. Some of the technical approaches used, and the Regulatory controls which have been developed, will undoubtedly find application in the future to the management of non-radioactive toxic wastes. The repository site outlined would contain even high-level radioactive wastes and spent fuels being contained without significant radiation dose rates to the public. Water pathway dose rates are likely to be lowest for vitrified high-level wastes with spent PWR fuel and intermediate level wastes being somewhat higher. (author)

Solute redistribution and Rayleigh number in the mushy zone during directional solidifi cation of Inconel 718

Directory of Open Access Journals (Sweden)

Wang Ling

2009-08-01

Full Text Available The interdendritic segregation along the mushy zone of directionally solidifi ed superalloy Inconel 718 has been measured by scanning electron microscope (SEM and energy dispersion analysis spectrometry (EDAXtechniques and the corresponding liquid composition profile was presented. The liquid density and Rayleigh number (Ra profi les along the mushy zone were calculated as well. It was found that the liquid density difference increased from top to bottom in the mushy zone and there was no density inversion due to the segregation of Nb and Mo. However carbide formation in the freezing range and the preferred angle of the orientated dendrite array could prompt the fl uid fl ow in the mushy zone although there was no liquid density inversion. The largest relative Rayleigh number appeared at 1,326 篊 for Inconel 718 where the fl uid fl ow most easily occurred.

Large magnetoresistance in a directionally solidified Ni44.5Co5.1Mn37.1In13.3 magnetic shape memory alloy

Science.gov (United States)

Li, Zongbin; Hu, Wei; Chen, Fenghua; Zhang, Mingang; Li, Zhenzhuang; Yang, Bo; Zhao, Xiang; Zuo, Liang

2018-04-01

Polycrystalline Ni44.5Co5.1Mn37.1In13.3 alloy with coarse columnar-shaped grains and 〈0 0 1〉A preferred orientation was prepared by directional solidification. Due to the strong magnetostructural coupling, inverse martensitic transformation can be induced by the magnetic field, resulting in large negative magnetoresistance up to -58% under the field of 3 T. Such significant field controlled functional behaviors should be attributed to the coarse grains and strong preferred orientation in the directionally solidified alloy.

Idaho Chemical Processing Plant low-level waste grout stabilization development program FY-96 status report

International Nuclear Information System (INIS)

Herbst, A.K.

1996-09-01

The general purpose of the Grout Stabilization Development Program is to solidify and stabilize the liquid low-level wastes (LLW) generated at the Idaho Chemical Processing Plant (ICPP). It is anticipated that LLW will be produced from the following: (1) chemical separation of the tank farm high-activity sodium-bearing waste; (2) retrieval, dissolution, and chemical separation of the aluminum, zirconium, and sodium calcines; (3) facility decontamination processes; and (4) process equipment waste. The main tasks completed this fiscal year as part of the program were chromium stabilization study for sodium-bearing waste and stabilization and solidification of LLW from aluminum and zirconium calcines. The projected LLW will be highly acidic and contain high amounts of nitrates. Both of these are detrimental to Portland cement chemistry; thus, methods to precondition the LLW and to cure the grout were explored. A thermal calcination process, called denitration, was developed to solidify the waste and destroy the nitrates. A three-way blend of Portland cement, blast furnace slag, and fly ash was successfully tested. Grout cubes were prepared at various waste loadings to maximize loading while meeting compressive strength and leach resistance requirements. For the sodium LLW, a 25% waste loading achieves a volume reduction of 3.5 and a compressive strength of 2,500 pounds per square inch while meeting leach, mix, and flow requirements. It was found that the sulfur in the slag reduces the chromium leach rate below regulatory limits. For the aluminum LLW, a 15% waste loading achieves a volume reduction of 8.5 and a compressive strength of 4,350 pounds per square inch while meeting leach requirements. Likewise for zirconium LLW, a 30% waste loading achieves a volume reduction of 8.3 and a compressive strength of 3,570 pounds per square inch

Microstructure investigation of NiAl-Cr(Mo) interface in a directionally solidified NiAl-Cr(Mo) eutectic alloyed with refractory metal

International Nuclear Information System (INIS)

Chen, Y.X.; Cui, C.Y.; Guo, J.T.; Li, D.X.

2004-01-01

The microstructure of a directionally solidified NiAl-Cr(Mo) eutectic alloyed with refractory metal in as-processed and heat-treated states has been studied by means of scanning electron microscopy and high resolution electron microscopy (HREM). The microstructure of the NiAl-Cr(Mo) eutectic was characterized by lamellar Cr(Mo) phases embedded within NiAl matrix with common growth direction of . The interface between NiAl and lamellar Cr(Mo) did not have any transition layers. Misfit dislocations were observed at the NiAl-Cr(Mo) interface. In addition to lamellar Cr(Mo) phases, coherent Cr(Mo, Ni, Al) precipitates and NiAl precipitates were also observed in the NiAl matrix and lamellar Cr(Mo) phases, respectively. After hot isostatic pressing and heat treatment, the NiAl-Cr(Mo) interfaces became smooth and straight. Square array of misfit dislocations was directly observed at the (0 0 1) interface between NiAl and Cr(Mo, Ni, Al) precipitate. The configuration of misfit dislocation network showed a generally good agreement with prediction based on the geometric O-lattice model

Incorporating an extended dendritic growth model into the CAFE model for rapidly solidified non-dilute alloys

International Nuclear Information System (INIS)

Ma, Jie; Wang, Bo; Zhao, Shunli; Wu, Guangxin; Zhang, Jieyu; Yang, Zhiliang

2016-01-01

We have extended the dendritic growth model first proposed by Boettinger, Coriell and Trivedi (here termed EBCT) for microstructure simulations of rapidly solidified non-dilute alloys. The temperature-dependent distribution coefficient, obtained from calculations of phase equilibria, and the continuous growth model (CGM) were adopted in the present EBCT model to describe the solute trapping behaviors. The temperature dependence of the physical properties, which were not used in previous dendritic growth models, were also considered in the present EBCT model. These extensions allow the present EBCT model to be used for microstructure simulations of non-dilute alloys. The comparison of the present EBCT model with the BCT model proves that the considerations of the distribution coefficient and physical properties are necessary for microstructure simulations, especially for small particles with high undercoolings. Finally, the EBCT model was incorporated into the cellular automaton-finite element (CAFE) model to simulate microstructures of gas-atomized ASP30 high speed steel particles that were then compared with experimental results. Both the simulated and experimental results reveal that a columnar dendritic microstructure preferentially forms in small particles and an equiaxed microstructure forms otherwise. The applications of the present EBCT model provide a convenient way to predict the microstructure of non-dilute alloys. - Highlights: • A dendritic growth model was developed considering non-equilibrium distribution coefficient. • The physical properties with temperature dependence were considered in the extended model. • The extended model can be used to non-dilute alloys and the extensions are necessary in small particles. • Microstructure of ASP30 steel was investigated using the present model and verified by experiment.

Incorporating an extended dendritic growth model into the CAFE model for rapidly solidified non-dilute alloys

Energy Technology Data Exchange (ETDEWEB)

Ma, Jie; Wang, Bo [State Key Laboratory of Advanced Special Steel, Shanghai University, Shanghai 200072 (China); Shanghai Engineering Technology Research Center of Special Casting, Shanghai 201605 (China); Zhao, Shunli [Research Institute, Baoshan Iron & Steel Co., Ltd, Shanghai 201900 (China); Wu, Guangxin [State Key Laboratory of Advanced Special Steel, Shanghai University, Shanghai 200072 (China); Shanghai Engineering Technology Research Center of Special Casting, Shanghai 201605 (China); Zhang, Jieyu, E-mail: zjy6162@staff.shu.edu.cn [State Key Laboratory of Advanced Special Steel, Shanghai University, Shanghai 200072 (China); Shanghai Engineering Technology Research Center of Special Casting, Shanghai 201605 (China); Yang, Zhiliang [State Key Laboratory of Advanced Special Steel, Shanghai University, Shanghai 200072 (China); Shanghai Engineering Technology Research Center of Special Casting, Shanghai 201605 (China)

2016-05-25

We have extended the dendritic growth model first proposed by Boettinger, Coriell and Trivedi (here termed EBCT) for microstructure simulations of rapidly solidified non-dilute alloys. The temperature-dependent distribution coefficient, obtained from calculations of phase equilibria, and the continuous growth model (CGM) were adopted in the present EBCT model to describe the solute trapping behaviors. The temperature dependence of the physical properties, which were not used in previous dendritic growth models, were also considered in the present EBCT model. These extensions allow the present EBCT model to be used for microstructure simulations of non-dilute alloys. The comparison of the present EBCT model with the BCT model proves that the considerations of the distribution coefficient and physical properties are necessary for microstructure simulations, especially for small particles with high undercoolings. Finally, the EBCT model was incorporated into the cellular automaton-finite element (CAFE) model to simulate microstructures of gas-atomized ASP30 high speed steel particles that were then compared with experimental results. Both the simulated and experimental results reveal that a columnar dendritic microstructure preferentially forms in small particles and an equiaxed microstructure forms otherwise. The applications of the present EBCT model provide a convenient way to predict the microstructure of non-dilute alloys. - Highlights: • A dendritic growth model was developed considering non-equilibrium distribution coefficient. • The physical properties with temperature dependence were considered in the extended model. • The extended model can be used to non-dilute alloys and the extensions are necessary in small particles. • Microstructure of ASP30 steel was investigated using the present model and verified by experiment.

Reference values on safety regulation of land disposal of low level radioactive solid waste (the second interim report) and its incorporation into legal regulations

International Nuclear Information System (INIS)

Aoki, Terumi

1994-01-01

Safety regulation of land disposal of low level radioactive solid waste in Japan is based on 'the basic philosophy on the safety regulation of land disposal of low level radioactive solid waste' determined by the Nuclear safety Committee (October 1985). The basic philosophy on the upper limit of radioactivity of disposed wastes was published as the reference values in the interim report (February 1987) and in the second interim report (June 1992). In the second interim report, the upper limits of radioactivity are established for three types of solid radioactive wastes: 1) metals, incombustible or flame resistant wastes generated nuclear reactor facilities and solidified in vessels, 2) large metallic structures generated from decommissioning of reactor facilities and difficult to solidify in vessels, and 3) radioactive concrete waste generated from decommissioning of reactor facilities. The upper limits of radioactivity are presented for C-14, Co-60, Ni-63, Sr-90, Cs-137, alfa-emmitters, Ca-41 (for concrete) and Eu-152 (for concrete). Related laws and regulations in Japan on safe disposal of low level wastes are explained. (T.H.)

Disposal and long-term storage in geological formations of solidified radioactive wastes

International Nuclear Information System (INIS)

Shischits, I.

1996-01-01

The special depository near Krasnoyarsk contains temporarily about 1,100 tons of spent nuclear fuel (SNF) from WWR- should be solidified and for the most part buried in geological formations. Solid wastes and SNF from RBMK reactors are assumed to be buried as well. For this purpose special technologies and underground constructions are required. They are to be created in the geological plots within the territory of Russian Federation and adjacent areas of CIS, meeting the developed list of requirements. The burial structures will vary greatly depending on the geological formation, the amount of wastes and their isotope composition. The well-known constructions such as deep wells, shafts, mines and cavities can be mentioned. There is a need to design constructions, which have no analog in the world practice. In the course of the Project fulfillment the following work will be conducted: -theoretical work followed by code creation for mathematical simulation of processes; - modelling on the base of prototypes made from equivalent materials with the help of simulators; - bench study; - experiments in real conditions; - examination of massif properties in particular plots using achievements of geophysics, including gamma-gamma density detectors and geo locators. Finally, ecological-economical model will be given for designing burial sites

Segregation and microstructure evolution in chill cast and directionally solidified Ni-Mn-Sn metamagnetic shape memory alloys

Science.gov (United States)

Czaja, P.; Wierzbicka-Miernik, A.; Rogal, Ł.

2018-06-01

A multiphase solidification behaviour is confirmed for a range of Ni-rich and Ni-deficient Ni-Mn-Sn induction cast and directionally solidified (Bridgman) alloys. The composition variation is primarily linked to the changing Mn/Sn ratio, whereas the content of Ni remains largely stable. The partitioning coefficients for the Ni50Mn37Sn13 and Ni46Mn41.5Sn12.5 Bridgman alloys were obtained according to the Scheil equation based on the composition distribution along the longitudinal cross section of the ingots. Homogenization heat treatment performed for 72 h at 1220 K turned out sufficient for ensuring chemical uniformity on the macro- and microscale. It is owed to a limited segregation length scale due to slow cooling rates adopted for the directional solidification process.

On oscillatory microstructure during cellular growth of directionally solidified Sn–36at.%Ni peritectic alloy

Science.gov (United States)

Peng, Peng; Li, Xinzhong; Li, Jiangong; Su, Yanqing; Guo, Jingjie

2016-01-01

An oscillatory microstructure has been observed during deep-cellular growth of directionally solidified Sn–36at.%Ni hyperperitectic alloy containing intermetallic compounds with narrow solubility range. This oscillatory microstructure with a dimension of tens of micrometers has been observed for the first time. The morphology of this wave-like oscillatory structure is similar to secondary dendrite arms, and can be observed only in some local positions of the sample. Through analysis such as successive sectioning of the sample, it can be concluded that this oscillatory microstructure is caused by oscillatory convection of the mushy zone during solidification. And the influence of convection on this oscillatory microstructure was characterized through comparison between experimental and calculations results on the wavelength. Besides, the change in morphology of this oscillatory microstructure has been proved to be caused by peritectic transformation during solidification. Furthermore, the melt concentration increases continuously during solidification of intermetallic compounds with narrow solubility range, which helps formation of this oscillatory microstructure. PMID:27066761

On oscillatory microstructure during cellular growth of directionally solidified Sn-36at.%Ni peritectic alloy.

Science.gov (United States)

Peng, Peng; Li, Xinzhong; Li, Jiangong; Su, Yanqing; Guo, Jingjie

2016-04-12

An oscillatory microstructure has been observed during deep-cellular growth of directionally solidified Sn-36at.%Ni hyperperitectic alloy containing intermetallic compounds with narrow solubility range. This oscillatory microstructure with a dimension of tens of micrometers has been observed for the first time. The morphology of this wave-like oscillatory structure is similar to secondary dendrite arms, and can be observed only in some local positions of the sample. Through analysis such as successive sectioning of the sample, it can be concluded that this oscillatory microstructure is caused by oscillatory convection of the mushy zone during solidification. And the influence of convection on this oscillatory microstructure was characterized through comparison between experimental and calculations results on the wavelength. Besides, the change in morphology of this oscillatory microstructure has been proved to be caused by peritectic transformation during solidification. Furthermore, the melt concentration increases continuously during solidification of intermetallic compounds with narrow solubility range, which helps formation of this oscillatory microstructure.

Processing method of radiation concrete waste and manufacturing method for radioactive waste solidifying filling mortar

International Nuclear Information System (INIS)

Sukekiyo, Mitsuaki; Okamoto, Masamichi

1998-01-01

Radioactive concrete wastes are crushed and pulverized. Fine solid granular materials caused by the pulverization are classified and the grain size is controlled so that the maximum grain size is 2.5mm, with the grains having a grain size of up to 0.15mm being up to 30% by weight to form fine aggregates. Separated and recovered fine concrete powders are classified and the size of the powder is controlled within a range of from 3,000 to 15,000cm 2 /g which is smaller than cement particles to form fine powders having a stable quality suitable as a mixing agent. The fine aggregates and the mixing agent are mixed to form a filling mortar (filler) for solidifying radioactive wastes. The filling mortar is filled together with other radioactive wastes in a drum to form a waste body in a drum. With such a constitution, crushed radioactive concrete wastes can be reutilized completely. (I.N.)

Particle Engulfment and Pushing by Solidifying Interfaces. Pt. 2; Micro-Gravity Experiments and Theoretical Analysis

Science.gov (United States)

Stefanescu, Doru M.; Juretzko, Frank R.; Dhindaw, Brij K.; Catalina, Adrian; Sen, Subhayu; Curreri, Peter A.

1998-01-01

Results of the directional solidification experiments on Particle Engulfment and Pushing by Solidifying Interfaces (PEP) conducted on the space shuttle Columbia during the Life and Microgravity Science Mission are reported. Two pure aluminum (99.999%) 9 mm cylindrical rods, loaded with about 2 vol.% 500 micrometers diameter zirconia particles were melted and resolidified in the microgravity (microg) environment of the shuttle. One sample was processed at step-wise increased solidification velocity, while the other at step-wise decreased velocity. It was found that a pushing-to-engulfment transition (PET) occurred in the velocity range of 0.5 to 1 micrometers. This is smaller than the ground PET velocity of 1.9 to 2.4 micrometers. This demonstrates that natural convection increases the critical velocity. A previously proposed analytical model for PEP was further developed. A major effort to identify and produce data for the surface energy of various interfaces required for calculation was undertaken. The predicted critical velocity for PET was of 0.775 micrometers/s.

Sandia solidification process: a broad range aqueous waste solidification method

International Nuclear Information System (INIS)

Lynch, R.W.; Dosch, R.G.; Kenna, B.T.; Johnstone, J.K.; Nowak, E.J.

1976-01-01

New ion-exchange materials of the hydrous oxide type were developed for solidifying aqueous radioactive wastes. These materials have the general formula M[M'/sub x/O/sub y/H/sub z/]/sub n/, where M is an exchangeable cation of charge +n and M' may be Ti; Nb; Zr, or Ta. Affinities for polyvalent cations were found to be very high and ion-exchange capacities large (e.g., 4.0--4.5 meq/g for NaTi 2 O 5 H depending on moisture content). The effectiveness of the exchangers for solidifying high-level waste resulting from reprocessing light-water reactor fuel was demonstrated in small-scale tests. Used in conjunction with anion exchange resin, these materials reduced test solution radioactivity from approximately 0.2 Ci/ml to as low as approximately 2 nCi/ml. The residual radioactivity was almost exclusively due to 106 Ru and total α-activity was only a few pCi/ml. Alternative methods of consolidating the solidified waste were evaluated using nonradioactive simulants. Best results were obtained by pressure-sintering which yielded essentially fully dense ceramics, e.g., titanate/titania ceramics with bulk density as high as 4.7 g/cm 3 , waste oxide content as high as 1.2 g/cm 3 , and leach resistance comparable to good borosilicate glass. Based on the above results, a baseline process for solidifying high-level waste was defined and approximate economic analyses indicated costs were not prohibitive. Additional tests have demonstrated that, if desired, operating conditions could be modified to allow recovery of radiocesium (and perhaps other isotopes) during solidification of the remaining constituents of high-level waste. Preliminary tests have also shown that these materials offer promise for treating tank-stored neutralized wastes

Alternatives for managing wastes from reactors and post-fission operations in the LWR fuel cycle. Volume 3. Alternatives for interim storage and transportation

International Nuclear Information System (INIS)

1976-05-01

Volume III of the five-volume report contains information on alternatives for interim storage and transportation. Section titles are: interim storage of spent fuel elements; interim storage of chop-leach fuel bundle residues; tank storage of high-level liquid waste; interim storage of solid non-high-level wastes; interim storage of solidified high-level waste; and, transportation alternatives

Containment of solidified liquid hazardous waste in domal salt

International Nuclear Information System (INIS)

Domenico, P.A.; Lerman, A.

1992-01-01

In recent years, the solidification of hazardous liquid waste has become a viable option in waste management. The solidification process results in an increased volume but more stable waste form that must be disposed of or stored in a dry environment. An environment of choice in south central Texas is domal salt. The salt dome currently under investigation has a water content of 0.002 percent by weight and a permeability less than one nanodarcy. A question that must be addressed is whether a salt dome has a particular set of attributes that will prevent the release of contaminants to the environment. From a regulatory perspective, a ''no migration'' petition must be approved by the U.S.E.P.A. for the containment facility. By ''no migration'' it is implied that the waste must be contained for 10,000 years. A demonstration that this condition will be met will require model calculations and such models must be based on the physical and chemical characteristics of the waste form and the geologic environment. In particular, the models must address the rate of brine infiltration into the caverns, providing information on how fast an immobile solid waste form could convert to a more mobile liquid state. Additionally, the potential for migration by both diffusion and advection is of concern. Lastly, given a partially saturated cavern, the question of how far gaseous waste will be transported over the 10,000 year containment period must also be addressed. Results indicate that the containment capabilities of domal salt are exceptional. A nominal volume of brine will seep into the cavern and most voids between the injected solidified waste pellets will remain unsaturated. Very small quantities of hazardous constituents will be leached from the waste pellets

Synthesis of laser beam rapidly solidified novel surfaces on D2 tool steel

International Nuclear Information System (INIS)

Ahmed, B.A.; Rizwan, K.F.; Minhas, J.A.; Waheed-ul-Haq, S.; Shahid, M.

2011-01-01

Surface layer of D2 tool steel was subjected to laser surface melting using continuous wave 2.5 kW CO/sub 2/ laser in point source melting mode. The processing parameters were varied to achieve a uniform depth of around 2 mm. Microstructural study revealed epitaxial growth of fine dendritic structure with secondary dendrite arm spacing in the range of 20-25 mu m. The phases in the parent annealed sample were BCC ferrite and chromium rich M7C3 carbide. The major phase after laser treatment was austenite and M7C3. The average hardness of annealed sample was 195 HV which increased to 410 HV after laser melting. Corrosion studies in 2% HCl solution exhibited a drastic improvement in corrosion resistance in laser treated samples. Improvement in properties is attributed to the refinement and uniformity of microstructure in the rapidly solidified surface. The case of a moving heat source was subjected to computer aided simulation to predict the melt depth at different processing conditions in point source melting mode. The calculated depths using the model, in ABAQUS software was found in good agreement with the experimental data. (author)

Leaching behaviour and mechanical properties of copper flotation waste in stabilized/solidified products.

Science.gov (United States)

Mesci, Başak; Coruh, Semra; Ergun, Osman Nuri

2009-02-01

This research describes the investigation of a cement-based solidification/stabilization process for the safe disposal of copper flotation waste and the effect on cement properties of the addition of copper flotation waste (CW) and clinoptilolite (C). In addition to the reference mixture, 17 different mixtures were prepared using different proportions of CW and C. Physical properties such as setting time, specific surface area and compressive strength were determined and compared to a reference mixture and Turkish standards (TS). Different mixtures with the copper flotation waste portion ranging from 2.5 to 12.5% by weight of the mixture were tested for copper leachability. The results show that as cement replacement materials especially clinoptilolite had clear effects on the mechanical properties. Substitution of 5% copper flotation waste for Portland cement gave a similar strength performance to the reference mixture. Higher copper flotation waste addition such as 12.5% replacement yielded lower strength values. As a result, copper flotation waste and clinoptilolite can be used as cementitious materials, and copper flotation waste also can be safely stabilized/solidified in a cement-based solidification/stabilization system.

Comparison of ice particle morphology crushed from ice chunk and directly solidified from droplet

Energy Technology Data Exchange (ETDEWEB)

Yang, J.H.; Yoon, Y.S.; Bang, S.Y. [Dongguk Univ., Pil-dong, Chung-gu, Seoul (Korea, Republic of). Dept. of Mechanical Engineering

2008-07-01

In order to investigate the transition kinetics of ice to hydrate and to produce standard specimens of hydrate pellet from prepared hydrate powders, fine ice beads with uniform diameters must be fabricated. This paper discussed the construction of several experimental setups for the fabrication of fine ice particle generation. The ultrasonic nozzle was used to produce fine mist which solidified near the free surface of liquid nitrogen bath. The shape and population distribution of ice bead diameters was analyzed. The study also compared ice particles produced by crushing. The surface morphology of ice particles produced with a ball mill was also examined. Experimental results were obtained for an ice shaver, ball mill, bowl for grinding medicine, and ultrasonic nozzle. It was concluded that the information generated from the study was useful in estimating the macroscopic flow characteristics such as permeability of bulk powder and in determining mean effective diameter of irregular shaped particles. Future work was also noted as being underway with different experiments for other cases with different operating conditions. 5 refs., 5 figs.

Characterization of solidified radioactive wastes produced at Montalto di Castro BWR plant with reference to the site storage

International Nuclear Information System (INIS)

Donato, A.; Ricci, G.; Pace, A.

1985-01-01

The cement solidification of the Montalto di Castro BWR plant radwastes has been studied both from the point of view of the mixtures of formulation and of the product characterization. Five radwaste types and mixtures of them have been taken into consideration, determining the best chemical formulations starting from the compressive strenght as leading parameter. The solidified products have been characterized from the point of view of the freeze and thawing resistance, the water immersion resistance, the leachability, the dimensional changes and the free standing water. All the tests have been performed taking into account the real site conditions, so the leaching tests and the water immersion tests have been carried out using sea water and table water as leachant

High-level waste solidification: why we chose glass

International Nuclear Information System (INIS)

Grover, J.R.

1980-01-01

This paper considers the desirable properties and factors to be assessed in the selection of a solidified waste product, surveys the possible product options and then analyzes in detail their suitability in meeting the criteria. It concludes that glasses are currently the preferred choice for the following reasons: their ability to fix the full spectrum of elements contained in the waste; their tolerance of the composition variations that will occur on a day to day basis in practice; their relatively low formation temperatures that lead to simpler and hence safer processing; their radiation stability; and their adequate leach rates. Suitable compositions are available for the wastes that will arise in the UK and techniques are available for manufacture on a production scale. Lower leach rates might be obtained by choosing glasses with higher formation temperatures or ceramics. However, these latter generally also have higher formation temperatures, have less tolerance for composition variations and their radiation stability is unproven. Supercalcines and synthetic rocks (SYNROC) may eventually be demonstrated to have some advantageous properties, but present indications are that these could be major disadvantages which more than offset any gains. Other advanced concepts (for example, the dispersion of glass beads in a metal matrix) have lower leach rates, but lead to additional complexity in manufacture

High-level-waste immobilization

International Nuclear Information System (INIS)

Crandall, J.L.

1982-01-01

Analysis of risks, environmental effects, process feasibility, and costs for disposal of immobilized high-level wastes in geologic repositories indicates that the disposal system safety has a low sensitivity to the choice of the waste disposal form

Method for enhancing stability of high explosives, for purposes of transport or storage, and the stabilized high explosives

International Nuclear Information System (INIS)

Nutt, G.L.

1991-01-01

This papent describes a method for suppressing the tendency of a porous solid high explosive to ignite and detonate. It comprises: filling substantially all the press of the solid high explosive material with a predetermined pore radius of at least 10μm with a relatively inert, stable, pore filling material in liquid form, the pore filling being selected from gallium, rubidium-potassium eutectic, and Wood's metal; and solidifying the pore filling material in the pores of the explosive material

Microstructural characterization of a rapidly solidified ultrahigh strength Al94.5Cr3Co1.5Ce1 alloy

International Nuclear Information System (INIS)

Ping, D.H.; Hono, K.; Inoue, A.

2000-01-01

The microstructure of a rapidly solidified Al 94.5 Cr 3 Co 1.5 Ce 1 alloy has been examined in detail by means of high resolution transmission electron microscopy (HRTEM) and atom probe field ion microscopy (APFIM). In the as-quenched microstructure, nanoscale particles of a solute-enriched amorphous phase and an Al-Cr compound are dispersed in randomly oriented fine grains of α-Al ( 200nm ). The interface between the Al grains and the amorphous particles is not smooth but irregular with atomic protrusions and concavities, suggesting that interfacial instability occurs during the solidification process. Nanoscale amorphous particles are formed as a result of solute trapping within the rapidly grown Al grains. After annealing at 400 C for 15 minutes grain growth occurs, and the interface of the Al grains is smoothed. The amorphous region trapped within the grains if crystallized to an Al-Cr compound, but no icosahedral phase has been confirmed. The APFIM results have revealed that Cr and Ce atoms have a similar partitioning behavior, i.e., they are rejected from the α-Al phase and partitioned into the trapped amorphous regions. On the other hand, Co atoms are not partitioned between the two phases in the as-quenched state but are partitioned into the α-Al grains in the annealed alloys being rejected from the Al compounds and finally form Al-Co compounds. Based on these microstructural characterization results, the origins of high strength of this alloy are discussed

Low- and intermediate-level waste management practices in Japan

International Nuclear Information System (INIS)

Tsuchiya, M.

1982-01-01

At present, disposal of low-level radioactive wastes is yet to be carried out in Japan. Liquid wastes, except for the diluted discharge of very low-level waste into the environment, are mostly solidified with cement or bitumen to be packed in 200 litre drums and put in storage. Solid wastes, on the other hand, are mostly put into in 200 litre drums, some of them being incinerated beforehand. Efforts are being made to develop technology for reducing the production of wastes. Regarding sea disposal, a test dumping program has been forestalled by the opposition of South Pacific islanders, but we are endeavoring to promote their understandings on this matter. Regarding land disposal, first we are going to start centralized storage, then shift to underground disposal

Crack initiation modeling of a directionally-solidified nickel-base superalloy

Science.gov (United States)

Gordon, Ali Page

Combustion gas turbine components designed for application in electric power generation equipment are subject to periodic replacement as a result of cracking, damage, and mechanical property degeneration that render them unsafe for continued operation. In view of the significant costs associated with inspecting, servicing, and replacing damaged components, there has been much interest in developing models that not only predict service life, but also estimate the evolved microstructural state of the material. This thesis explains manifestations of microstructural damage mechanisms that facilitate fatigue crack nucleation in a newly-developed directionally-solidified (DS) Ni-base superalloy components exposed to elevated temperatures and high stresses. In this study, models were developed and validated for damage and life prediction using DS GTD-111 as the subject material. This material, proprietary to General Electric Energy, has a chemical composition and grain structure designed to withstand creep damage occurring in the first and second stage blades of gas-powered turbines. The service conditions in these components, which generally exceed 600°C, facilitate the onset of one or more damage mechanisms related to fatigue, creep, or environment. The study was divided into an empirical phase, which consisted of experimentally simulating service conditions in fatigue specimens, and a modeling phase, which entailed numerically simulating the stress-strain response of the material. Experiments have been carried out to simulate a variety of thermal, mechanical, and environmental operating conditions endured by longitudinally (L) and transversely (T) oriented DS GTD-111. Both in-phase and out-of-phase thermo-mechanical fatigue tests were conducted. In some cases, tests in extreme environments/temperatures were needed to isolate one or at most two of the mechanisms causing damage. Microstructural examinations were carried out via SEM and optical microscopy. A continuum

A High-Voltage Level Tolerant Transistor Circuit

NARCIS (Netherlands)

Annema, Anne J.; Geelen, Godefridus Johannes Gertrudis Maria

2001-01-01

A high-voltage level tolerant transistor circuit, comprising a plurality of cascoded transistors, including a first transistor (T1) operatively connected to a high-voltage level node (3) and a second transistor (T2) operatively connected to a low-voltage level node (2). The first transistor (T1)

Morphological variants of carbides of solidification origin in the rapidly solidified powder particles of hypereutectic iron alloy

International Nuclear Information System (INIS)

Kusy, M.; Grgac, P.; Behulova, M.; Vyrostkova, A.; Miglierini, M.

2004-01-01

The paper deals with the analysis of the morphological variants of solidification microstructures and vanadium rich M 4 C 3 carbide phases in the rapidly solidified (RS) powder particles from hypereutectic Fe-C-Cr-V alloy prepared by the nitrogen gas atomisation. Five main types of solidification microstructures were identified in RS particles: microstructure with globular carbides, microstructure with globular and star-like carbides, microstructure with primary carbides in the centres of eutectic colonies, microstructure with eutectic colonies without primary carbides and microstructure with eutectic spherulites. Based on the morphological features of carbide phases and the thermal history of RS particles, the microstructures were divided into two groups - microstructures morphologically affected and non-affected during the post-recalescence period of solidification. Thermophysical reasons for the morphologically different M 4 C 3 carbide phases development in the RS powder particles are discussed

A novel solidified floating organic drop microextraction method for preconcentration and determination of copper ions by flow injection flame atomic absorption spectrometry

International Nuclear Information System (INIS)

Sahin, Cigdem Arpa; Tokgoez, Ilknur

2010-01-01

A rapid, simple and cost effective solidified floating organic drop microextraction (SFODME) and flow injection flame atomic absorption spectrometric determination (FI-FAAS) method for copper was developed. In this method, a free microdrop of 1-undecanol containing 1,5-diphenyl carbazide (DPC) as the complexing agent was transferred to the surface of an aqueous sample including Cu(II) ions, while being agitated by a stirring bar in the bulk of the solution. Under the proper stirring conditions, the suspended microdrop can remain at the top-center position of the aqueous sample. After the completion of the extraction, the sample vial was cooled by placing it in a refrigerator for 10 min. The solidified microdrop was then transferred into a conical vial, where it melted immediately and diluted to 300 μL with ethanol. Finally, copper ions in 200 μL of diluted solution were determined by FI-FAAS. Several factors affecting the microextraction efficiency, such as type of extraction solvent, pH, complexing agent concentration, extraction time, stirring rate, sample volume and temperature were investigated and optimized. Under optimized conditions for 100 mL of solution, the preconcentration factor was 333 and the enrichment factor was 324. The limit of detection (3 s) was 0.4 ng mL -1 , the limit of quantification (10 s) was 1.1 ng mL -1 and the relative standard deviation (RSD) for 10 replicate measurements of 10 ng mL -1 copper was 0.9%. The proposed method was successfully applied to the determination of copper in different water samples.

Effect of thermal cycling on the microstructure of a directionally solidified Fe, Cr, Al-TaC eutectic alloy

Science.gov (United States)

Harf, F. H.; Tewari, S. N.

1977-01-01

Cylindrical bars (1.2 cm diameter) of Fe-13.6Cr-3.7Al-9TaC (wt %) eutectic alloy were directionally solidified in a modified Bridgman type furnace at 1 cm/h. The alloy microstructure consisted of aligned TaC fibers imbedded in a bcc Fe-Cr-Al matrix. Specimens of the alloy were thermally cycled from 1100 to 425 C in a burner rig. The effects of 1800 thermal cycles on the microstructure was examined by scanning electron microscopy, revealing a zig-zag shape of TaC fibers aligned parallel to the growth direction. The mechanism of carbide solution and reprecipitation on the (111) easy growth planes, suggested previously to account for the development of irregular serrations in Co-Cr-Ni matrix alloys, is believed to be responsible for these zig-zag surfaces.

Effect of processing on the microstructural development in a rapidly solidified Al-Fe-V-Si alloy

International Nuclear Information System (INIS)

Srivastava, A.K.; Ranganathan, S.; Ojha, S.N.

1993-01-01

An Al 80 Fe 10 Si 6 alloy has been rapidly solidified using melt spinning, gas atomization and spray forming processes. The effect of processing techniques on the microstructural characteristics of the alloy has ben evaluated. The melt spun alloy has shown an icosahedral quasicrystalline phase surrounded by a rational approximant structure of the icosahedral phase. The rational approximant structure has been identified as a crystalline cubic silicide phase. The atomized powders have exhibited cellular and dendritic morphology depending on the size of particles. In addition, the second phase particles of the silicide phase are observed to decorate the cell boundaries and interdendritic regions. In contrast, the alloy processed by spray deposition has revealed an equiaxed solidification morphology with a uniform dispersion of find silicon phase inside the grain. The origin of the microstructure in the alloy processed by these techniques is discussed. The results are compared wherever possible with the commercially available Al-Fe-V-Si alloys

Radial macrosegregation and dendrite clustering in directionally solidified Al-7Si and Al-19Cu alloys

Science.gov (United States)

Ghods, M.; Johnson, L.; Lauer, M.; Grugel, R. N.; Tewari, S. N.; Poirier, D. R.

2016-05-01

Hypoeutectic Al-7 wt% Si and Al-19 wt% Cu alloys were directionally solidified upward in a Bridgman furnace through a range of constant growth speeds and thermal gradients. Though processing is thermo-solutally stable, flow initiated by gravity-independent advection at, slightly leading, central dendrites moves rejected solute out ahead and across the advancing interface. Here any lagging dendrites are further suppressed which promotes a curved solid-liquid interface and the eventual dendrite "clustering" seen in transverse sections (dendrite "steepling" in longitudinal orientations) as well as extensive radial macrosegregation. Both aluminum alloys showed considerable macrosegregation at the low growth speeds (10 and 30 μm s-1) but not at higher speed (72 μm s-1). Distribution of the fraction eutectic-constituent on transverse sections was determined in order to quantitatively describe radial macrosegregation. The convective mechanisms leading to dendrite-steepling were elucidated with numerical simulations, and their results compared with the experimental observations.

High-level waste processing and disposal

International Nuclear Information System (INIS)

Crandall, J.L.; Krause, H.; Sombret, C.; Uematsu, K.

1984-01-01

The national high-level waste disposal plans for France, the Federal Republic of Germany, Japan, and the United States are covered. Three conclusions are reached. The first conclusion is that an excellent technology already exists for high-level waste disposal. With appropriate packaging, spent fuel seems to be an acceptable waste form. Borosilicate glass reprocessing waste forms are well understood, in production in France, and scheduled for production in the next few years in a number of other countries. For final disposal, a number of candidate geological repository sites have been identified and several demonstration sites opened. The second conclusion is that adequate financing and a legal basis for waste disposal are in place in most countries. Costs of high-level waste disposal will probably add about 5 to 10% to the costs of nuclear electric power. The third conclusion is less optimistic. Political problems remain formidable in highly conservative regulations, in qualifying a final disposal site, and in securing acceptable transport routes

High levels of Porphyromonas gingivalis-induced immunoglobulin G2 are associated with lower high-density lipoprotein levels in chronic periodontitis.

Science.gov (United States)

Ardila, Carlos M; Guzmán, Isabel C

2016-11-01

To investigate the association between the presence of Porphyromonas gingivalis-induced immunoglobulin G antibodies and the high-density lipoprotein (HDL) level. A total of 108 individuals were examined. The presence of P. gingivalis was detected using primers designed to target the 16S rRNA gene sequence. Peripheral blood was collected from each subject to determine the levels of P. gingivalis-induced IgG1 and IgG2 serum antibodies. The HDL levels were determined using fully enzymatic methods. A higher proportion of periodontitis patients had high levels of P. gingivalis-induced IgG1 and IgG2, and the proportion of subjects with a HDL level of chronic periodontitis patients. In the unadjusted regression model, the presence of high levels of P. gingivalis-induced IgG2 was associated with a HDL level of periodontitis patients with high levels of P. gingivalis-induced IgG2 showed 3.2 more chances of having pathological HDL levels (odds ratio = 3.2, 95% confidence interval = 1.2-9.8). High levels of P. gingivalis-induced IgG2 were associated with low HDL concentrations in patients with periodontitis, which suggests that the response of the host to periodontal infection may play an important role in the pathogenesis of cardiovascular diseases. © 2015 Wiley Publishing Asia Pty Ltd.

Measurement of the leaching rate of radionuclide 134Cs from the solidified radioactive sources in Portland cement mixed with microsilica and barite matrixes

International Nuclear Information System (INIS)

Shaaban, Ismail; Assi, Nasim

2011-01-01

Portland cement was mixed with radionuclide 134 Cs to produce low-level radioactive sources. These sources were surrounded with cement mixed with different materials like microsilica and barite. The leaching rate of 134 Cs from the solidified radioactive source in Portland cement alone was found to be 4.481 x 10 -4 g/cm 2 per day. Mixing this Portland cement with microsilica and with barite reduced significantly the leaching rate to 1.091 x 10 -4 g/cm 2 per day and 3.153 x 10 -4 g/cm 2 per day for 1 wt.% mixing, and to 1.401 x 10 -5 g/cm 2 per day and 1.703 x 10 -4 g/cm 2 per day for 3 wt.% mixing, respectively. It was also found that the application of a latex paint reduced these leaching rates by about 6.5%, 20.3% and 13.3% for Portland cement, cement mixed with microsilica and with barite, respectively. The leaching data were also analyzed using the polynomial method. The obtained results showed that cement mixed with microsilica and with barite can be effectively used for radioactive sources solidification.

Short-term thermal response of rapidly solidified Type 304 stainless steel containing helium

International Nuclear Information System (INIS)

Clark, D.E.

1988-06-01

Type 304 stainless steel was heat treated for short times near its melting point in order to determine its microstructural response to thermal cycles typical of the near heat-affected zones of welding processes. The material was rapidly solidified as a powder by centrifugal atomization in a helium environment and consolidated by hot extrusion. Along with the ingot metallurgy material used for canning the powder prior to hot extrusion, it was heat treated using a Gleeble at temperatures of 1200 and 1300 degree C for times ranging from <1 to 1000 s, and the samples were examined for microstructure and the existence of porosity due to entrapped helium. At higher test temperatures and longer treatment times, the material developed extensive porosity, which was stabilized by the presence of helium and which may also have a role in anchoring grain boundaries and inhibiting grain growth. The powder material. At lower test temperatures and shorter treatment times, grain growth in the γ phase appeared to be restricted in the powder material, possible by the presence of helium. An intermediate temperatures and times, a γ-δ duplex microstructure also restricted grain growth again occurred in the δ microstructure. 9 refs., 14 figs., 3 tabs

Field lysimeter facility for evaluating the performance of commercial solidified low-level waste

International Nuclear Information System (INIS)

Walter, M.B.; Graham, M.J.; Gee, G.W.

1984-11-01

Analyzing the potential migration of radionuclides from sites containing solid low-level wastes requires knowledge of contaminant concentrations in the soil solution surrounding the waste. This soil solution concentration is generally referred to as the source term and is determined by such factors as the concentration of radionuclides in the solid waste, the rate of leachate formation, the concentration of dissolved species in the leachate, any solubility reactions occurring when the leachate contacts the soil, and the rate of water flow in the soil surrounding the waste. A field lysimeter facility established at the Hanford site is being used to determine typical source terms in arid climates for commercial low-level wastes solidifed with cement, Dow polymer (vinyl ester-styrene), and bitumen. The field lysimeter facility consists of 10, 3-m-deep by 1.8-m-dia closed-bottom lysimeters situated around a 4-m-deep by 4-m-dia central instrument caisson. Commercial cement and Dow polymer waste samples were removed from 210-L drums and placed in 8 of the lysimeters. Two bitumen samples are planned to be emplaced in the facility's remaining 2 lysimeters during 1984. The central caisson provides access to the instrumentation in the individual lysimeters and allows selective sampling of the soil and waste. Suction candles (ceramic cups) placed around the waste forms will be used to periodically collect soil-water samples for chemical analysis. Meteorological data, soil moisture content, and soil temperature are automatically monitored at the facility. Characterization of the soils and waste forms have been partially completed. These data consist of moisture release characteristics, particle-size distribution, and distributions and concentrations of radionuclides in the waste forms. 11 references, 12 figures, 5 tables

High-Level Application Framework for LCLS

Energy Technology Data Exchange (ETDEWEB)

Chu, P; Chevtsov, S.; Fairley, D.; Larrieu, C.; Rock, J.; Rogind, D.; White, G.; Zalazny, M.; /SLAC

2008-04-22

A framework for high level accelerator application software is being developed for the Linac Coherent Light Source (LCLS). The framework is based on plug-in technology developed by an open source project, Eclipse. Many existing functionalities provided by Eclipse are available to high-level applications written within this framework. The framework also contains static data storage configuration and dynamic data connectivity. Because the framework is Eclipse-based, it is highly compatible with any other Eclipse plug-ins. The entire infrastructure of the software framework will be presented. Planned applications and plug-ins based on the framework are also presented.

Evaluation of low and intermediate level radioactive solidified waste forms and packages

International Nuclear Information System (INIS)

1990-10-01

Evaluation of low and intermediate level radioactive waste forms and packages with respect to compliance with quality and safety requirements for transport, interim storage and disposal has become a very important part of the radioactive waste management strategy in many countries. The evaluation of waste forms and packages provides precise basic data for regulatory bodies to establish safety requirements, and implement quality control and quality assurance procedures for radioactive waste management programmes. The requirements depend very much upon the disposal option selected, treatment technology used, waste form characteristics, package quality and other factors. The regulatory requirements can also influence the methodology of waste form/package evaluation together with selection and analysis of data for quality control and safety assurance. A coordinated research programme started at the end of 1985 and brought together 12 participants from 11 countries. The results of the programme and each particular project were discussed at three Research Coordination Meetings held in Cairo, Egypt, in May, 1986; in Beijing, China, in April, 1998; and at Harwell Laboratory, United Kingdom, in November, 1989. This document summarises the salient features and results achieved during the four year investigation and a recommendation for future work in this area. Refs, figs and tabs

Concept and Idea-Project for Yugoslav Low and Intermediate level Radioactive Waste Materials Final Disposal Facility

International Nuclear Information System (INIS)

Peric, A.

1997-01-01

Encapsulation of rad waste in a mortar matrix and displacement of such solidified waste forms into the shallow land burial system, engineered trench system type is suggested concept for the final disposal of low and intermediate level rad waste. The mortar-rad waste mixtures are cured in containers of either concrete or metal for an appropriate period of time, after which solidified rad waste-mortar monoliths are then placed in the engineered trench system, parallelepiped honeycomb structure. Trench consists of vertical barrier-walls, bottom barrier-floors, surface barrier-caps and permeable-reactive walls. Surroundings of the trench consists of buffer barrier materials, mainly clay. Each segment of the trench is equipped with the independent drainage system, as a part of the main drainage. Encapsulation of each filled trench honeycomb segment is performed with concrete cap. Completed trench is covered with impermeable plastic foil and soil leaner, preferably clay. Paper presents an overview of the final disposal facility engineered trench system type. Advantages in comparison with other types of final disposal system are given. (author)

General Algorithm (High level)

Indian Academy of Sciences (India)

First page Back Continue Last page Overview Graphics. General Algorithm (High level). Iteratively. Use Tightness Property to remove points of P1,..,Pi. Use random sampling to get a Random Sample (of enough points) from the next largest cluster, Pi+1. Use the Random Sampling Procedure to approximate ci+1 using the ...

Partitioning of high level liquid waste: experiences in plant level adoption

International Nuclear Information System (INIS)

Manohar, Smitha; Kaushik, C.P.

2016-01-01

High Level Radioactive Wastes are presently vitrified in borosilicate matrices in all our back end facilities in our country. This is in accordance with internationally endorsed methodology for the safe management of high level radioactive wastes. Recent advancements in the field of partitioning technology in our group, has presented us with an opportunity to have a fresh perspective on management of high level liquid radioactive wastes streams, that emanate from reprocessing operations. This paper will highlight our experiences with respect to both partitioning studies and vitrification practices, with a focus on waste volume reduction for final disposal. Incorporation of this technique has led to the implementation of the concept of recovering wealth from waste, a marked decrease on the load of disposal in deep geological repositories and serve as a step towards the vision of transmutation of long lived radionuclides

Dimensional stability under wet curing of mortars containing high amounts of nitrates and phosphates

International Nuclear Information System (INIS)

Benard, P.; Cau Dit Coumes, C.; Garrault, S.; Nonat, A.; Courtois, S.

2008-01-01

Investigations were carried out in order to solidify in cement some aqueous streams resulting from nuclear decommissioning processes and characterized by a high salinity (300 g/L), as well as important concentrations of nitrate (150-210 g/L) and phosphate ions (0-50 g/L). Special attention was paid to the influence of these compounds on the dimensional variations under wet curing of simulated solidified waste forms. The length changes of mortars containing nitrate salts only (KNO 3 , NaNO 3 ) were shown to be governed by a concentration effect which involved osmosis: the higher their concentration in the mixing solution, the higher the swelling. The expansion of mortars containing high amounts of phosphates (≥ 30 g/L in the mixing solution) was preceded by a shrinkage which increased with the phosphate concentration, and which could be suppressed by seeding the cement used with hydroxyapatite crystals. This transitory shrinkage was attributed to the conversion into hydroxyapatite of a precursor readily precipitated in the cement paste after mixing

Predictors of Placement in Lower Level versus Higher Level High School Mathematics

Science.gov (United States)

Archbald, Doug; Farley-Ripple, Elizabeth N.

2012-01-01

Educators and researchers have long been interested in determinants of access to honors level and college prep courses in high school. Factors influencing access to upper level mathematics courses are particularly important because of the hierarchical and sequential nature of this subject and because students who finish high school with only lower…

High-Level Development of Multiserver Online Games

Directory of Open Access Journals (Sweden)

Frank Glinka

2008-01-01

Full Text Available Multiplayer online games with support for high user numbers must provide mechanisms to support an increasing amount of players by using additional resources. This paper provides a comprehensive analysis of the practically proven multiserver distribution mechanisms, zoning, instancing, and replication, and the tasks for the game developer implied by them. We propose a novel, high-level development approach which integrates the three distribution mechanisms seamlessly in today's online games. As a possible base for this high-level approach, we describe the real-time framework (RTF middleware system which liberates the developer from low-level tasks and allows him to stay at high level of design abstraction. We explain how RTF supports the implementation of single-server online games and how RTF allows to incorporate the three multiserver distribution mechanisms during the development process. Finally, we describe briefly how RTF provides manageability and maintenance functionality for online games in a grid context with dynamic resource allocation scenarios.

Long-term high-level waste technology program

International Nuclear Information System (INIS)

1980-04-01

The Department of Energy (DOE) is conducting a comprehensive program to isolate all US nuclear wastes from the human environment. The DOE Office of Nuclear Energy - Waste (NEW) has full responsibility for managing the high-level wastes resulting from defense activities and additional responsiblity for providing the technology to manage existing commercial high-level wastes and any that may be generated in one of several alternative fuel cycles. Responsibilities of the Three Divisions of DOE-NEW are shown. This strategy document presents the research and development plan of the Division of Waste Products for long-term immobilization of the high-level radioactive wastes resulting from chemical processing of nuclear reactor fuels and targets. These high-level wastes contain more than 99% of the residual radionuclides produced in the fuels and targets during reactor operations. They include essentially all the fission products and most of the actinides that were not recovered for use

Feasibility of large volume casting cementation process for intermediate level radioactive waste

International Nuclear Information System (INIS)

Chen Zhuying; Chen Baisong; Zeng Jishu; Yu Chengze

1988-01-01

The recent tendency of radioactive waste treatment and disposal both in China and abroad is reviewed. The feasibility of the large volume casting cementation process for treating and disposing the intermediate level radioactive waste from spent fuel reprocessing plant in shallow land is assessed on the basis of the analyses of the experimental results (such as formulation study, solidified radioactive waste properties measurement ect.). It can be concluded large volume casting cementation process is a promising, safe and economic process. It is feasible to dispose the intermediate level radioactive waste from reprocessing plant it the disposal site chosen has resonable geological and geographical conditions and some additional effective protection means are taken

High-Level Radioactive Waste.

Science.gov (United States)

Hayden, Howard C.

1995-01-01

Presents a method to calculate the amount of high-level radioactive waste by taking into consideration the following factors: the fission process that yields the waste, identification of the waste, the energy required to run a 1-GWe plant for one year, and the uranium mass required to produce that energy. Briefly discusses waste disposal and…

High-level radioactive wastes

International Nuclear Information System (INIS)

Grissom, M.C.

1982-10-01

This bibliography contains 812 citations on high-level radioactive wastes included in the Department of Energy's Energy Data Base from January 1981 through July 1982. These citations are to research reports, journal articles, books, patents, theses, and conference papers from worldwide sources. Five indexes are provided: Corporate Author, Personal Author, Subject, Contract Number, and Report Number

Leaching behavior of various low-level waste solids

International Nuclear Information System (INIS)

Ito, Akihiko; Ouchi, Yasuyoshi; Matsuzuru, Hideo; Wadachi, Yoshiki

1985-01-01

This report deals with the leaching of radioactive nuclides from low-level wastes solidified with cement, bitumen or plastics. Considerations are made on the effects of type of solidification matrix and waste; type, amount and exchange frequency of leachate; type and conditions of embedding soil; temperature and pressure; and secular deterioration. It is assumed that a waste composite is entirely immersed in leachate and that the amount of the leachate is large compared to the surface area of the waste. Cement solid is characterized by its high alkalinity and porosity while plastic and bitumen solids are dense and neutral. The content of waste in a composite is low for cement and high for plastics. It is generally high in bitumen solid though it should be reduced if the solid is likely to bulge. The leaching of 137 Cs from cement solid is slightly dependent on the waste-cement ratio while it increases with increasing waste content in the case of plastic or bitumen solid. For 60 Co, the leaching from cement solid depends on the alkalinity of the cement material used though it is not affected by the waste-cement ratio. In the case of plastics and bitumen, on the other hand, the pH value of the waste have some effects on the leaching of 60 Co; the leaching decreases with increasing pH. (Nogami, K.)

Treatment options of low level liquid waste of ETP origin by synthetic zeolites

International Nuclear Information System (INIS)

Singh, I.J.; Jain, Savita; Sathi Sasidharan, N.; Deshingkar, D.S.

2001-08-01

Mixture of synthetic zeolites, AR1, 4A and 13X of Indian origin were tested in a single fixed bed column operation for the treatment of low level liquid waste received at Effluents Treatment Plant (ETP) Trombay, under dynamic conditions. The mixed bed of zeolites was highly effective in decontaminating thousands of bed volumes of waste stream from radio cesium, radio strontium and gross beta gamma activity. High volume reduction factors, upwards of 10,000 are available in this process compared to less than 100 available with chemical precipitation process, currently followed. Containment of entrapped activity in zeolite bed was studied by solidifying them in Portland cement matrix as stable waste form. Incorporation of minerals like vermiculite as minor additive for improving the leaching characteristics of the final waste form was evaluated. Zeolite incorporated cement blocks were subjected to leach tests in distilled water for over 200 days to assess the incremental and cumulative leach rates of individual activity components. Leachability index of radio cesium and strontium were computed, which indicated the suitability of the matrix for safe shallow land burial. (author)

High-level waste immobilization program: an overview

International Nuclear Information System (INIS)

Bonner, W.R.

1979-09-01

The High-Level Waste Immobilization Program is providing technology to allow safe, affordable immobilization and disposal of nuclear waste. Waste forms and processes are being developed on a schedule consistent with national needs for immobilization of high-level wastes stored at Savannah River, Hanford, Idaho National Engineering Laboratory, and West Valley, New York. This technology is directly applicable to high-level wastes from potential reprocessing of spent nuclear fuel. The program is removing one more obstacle previously seen as a potential restriction on the use and further development of nuclear power, and is thus meeting a critical technological need within the national objective of energy independence

National high-level waste systems analysis report

Energy Technology Data Exchange (ETDEWEB)

Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

1995-09-01

This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy.

National high-level waste systems analysis report

International Nuclear Information System (INIS)

Kristofferson, K.; Oholleran, T.P.; Powell, R.H.

1995-09-01

This report documents the assessment of budgetary impacts, constraints, and repository availability on the storage and treatment of high-level waste and on both existing and pending negotiated milestones. The impacts of the availabilities of various treatment systems on schedule and throughput at four Department of Energy sites are compared to repository readiness in order to determine the prudent application of resources. The information modeled for each of these sites is integrated with a single national model. The report suggests a high-level-waste model that offers a national perspective on all high-level waste treatment and storage systems managed by the Department of Energy

Development of an improved ion-exchange process for removing cesium and strontium from high-level radioactive liquid wastes

International Nuclear Information System (INIS)

Wallace, R.M.; Ferguson, R.B.

1980-11-01

Processes are being developed to solidify and isolate the biologically hazardous radionuclides from approximately 23 million gallons of alkaline high-level waste accumulated at the Savannah River Plant. The waste consists mainly of a liquid supernate, a damp salt cake, and a gelatinous, insoluble sludge. The reference solidification process involves separation of the water soluble fraction (supernate) from the insoluble fraction, removal of cesium and traces of strontium from the supernate, incorporation of the sludge and the radionuclides from the supernate in glass, and incorporation of the residual salt in concrete. A new process, now being developed, involves sorbing cesium on phenolic resins that contain no strongly acidic sulfonate groups. These resins can then be eluted with formic acid which is not possible with Duolite ARC-359. Duolite CS-100, a phenol-carboxylate resin, was chosen for further development because of its greater breakthrough capacity and because it also sorbs strontium to some extent. Strontium sorption by CS-100 was not sufficient to eliminate the need for Amberlite IRC-718. However, the latter resin can also be eluted with formic acid because its functional groups are weakly acidic. Formic acid elution permits several options to be considered. The preferred option consists simply of mixing the eluate with sludge prior to calcination. Sodium formate, which is formed when the resins in the sodium form are eluted, decomposes rapidly between 450 0 C and 500 0 C and will be destroyed in either the calciner or the melter. The resulting sodium oxide would be incorporated into glass. The principal advantage of the new process is the elimination of a number of process steps

Comparative Study on the Grain Refinement of Al-Si Alloy Solidified under the Impact of Pulsed Electric Current and Travelling Magnetic Field

Directory of Open Access Journals (Sweden)

Yunhu Zhang

2016-07-01

Full Text Available It is high of commercial importance to generate the grain refinement in alloys during solidification by means of electromagnetic fields. Two typical patterns of electromagnetic fields, pulsed electric currents (ECP and traveling magnetic field (TMF, are frequently employed to produce the finer equiaxed grains in solidifying alloys. Various mechanisms were proposed to understand the grain refinement in alloys caused by ECP and TMF. In this paper, a comparative study is carried out in the same solidification regime to investigate the grain refinement of Al-7 wt. %Si alloy driven by ECP and TMF. Experimental results show that the application of ECP or TMF can cause the same grain refinement occurrence period, during which the refinement of primary Al continuously occurs. In addition, the related grain refinement mechanisms are reviewed and discussed, which shows the most likely one caused by ECP and TMF is the promoted dendrite fragmentation as the result of the ECP-induced or TMF-induced forced flow. It suggests that the same grain refinement process in alloys is provoked when ECP and TMF are applied in the same solidification regime, respectively.

Overview: Defense high-level waste technology program

International Nuclear Information System (INIS)

Shupe, M.W.; Turner, D.A.

1987-01-01

Defense high-level waste generated by atomic energy defense activities is stored on an interim basis at three U.S. Department of Energy (DOE) operating locations; the Savannah River Plant in South Carolina, the Hanford Site in Washington, and the Idaho National Engineering Laboratory in Idaho. Responsibility for the permanent disposal of this waste resides with DOE's Office of Defense Waste and Transportation Management. The objective of the Defense High-Level Wast Technology Program is to develop the technology for ending interim storage and achieving permanent disposal of all U.S. defense high-level waste. New and readily retrievable high-level waste are immobilized for disposal in a geologic repository. Other high-level waste will be stabilized in-place if, after completion of the National Environmental Policy Act (NEPA) process, it is determined, on a site-specific basis, that this option is safe, cost effective and environmentally sound. The immediate program focus is on implementing the waste disposal strategy selected in compliance with the NEPA process at Savannah River, while continuing progress toward development of final waste disposal strategies at Hanford and Idaho. This paper presents an overview of the technology development program which supports these waste management activities and an assessment of the impact that recent and anticipated legal and institutional developments are expected to have on the program

Performance of cement solidification with barium for high activity liquid waste including sulphate

International Nuclear Information System (INIS)

Waki, Toshikazu; Yamada, Motoyuki; Horikawa, Yoshihiko; Kaneko, Masaaki; Saso, Michitaka; Haruguchi, Yoshiko; Yamashita, Yu; Sakai, Hitoshi

2009-01-01

The target liquid waste to be solidified is generated from PWR primary loop spent resin treatment with sulphate acid, so, its main constituent is sodium sulphate and the activity of this liquid is relatively high. Waste form of this liquid waste is considered to be a candidate for the subsurface disposal. The disposed waste including sulphate is anticipated to rise a concentration of sulphate ion in the ground water around the disposal facility and it may cause degradation of materials such as cement and bentonite layer and comprise the disposal facility. There could be two approaches to avoid this problem, the strong design of the disposal facility and the minimization of sulphaste ion migration from the solidified waste. In this study, the latter approach was examined. In order to keep the low concentration of sulphate ion in the ground water, it is effective to make barium sulphate by adding barium compound into the liquid waste in solidification. However, adding equivalent amount of barium compound with sulphate ion causes difficulty of mixing, because production of barium sulphate causes high viscosity. In this study, mixing condition after and before adding cement into the liquid waste was estimated. The mixing condition was set with consideration to keep anion concentration low in the ground water and of mixing easily enough in practical operation. Long term leaching behavior of the simulated solidified waste was also analyzed by PHREEQC. And the concentration of the constitution affected to the disposal facility was estimated be low enough in the ground water. (author)

Disposal of high-level radioactive waste

International Nuclear Information System (INIS)

Glasby, G.P.

1977-01-01

Although controversy surrounding the possible introduction of nuclear power into New Zealand has raised many points including radiation hazards, reactor safety, capital costs, sources of uranium and earthquake risks on the one hand versus energy conservation and alternative sources of energy on the other, one problem remains paramount and is of global significance - the storage and dumping of the high-level radioactive wastes of the reactor core. The generation of abundant supplies of energy now in return for the storage of these long-lived highly radioactive wastes has been dubbed the so-called Faustian bargain. This article discusses the growth of the nuclear industry and its implications to high-level waste disposal particularly in the deep-sea bed. (auth.)

Effect of Trace Ce on Microstructure and Properties of Near-rapidly Solidified Al-Zn-Mg-Cu Alloys

Directory of Open Access Journals (Sweden)

HUANG Gao-ren

2018-03-01

Full Text Available Through using DSC, XRD, SEM, EDS, static tensile test and other analysis methods of materials, the effect of trace Ce on microstructure and properties of near-rapidly solidified Al-Zn-Mg-Cu alloy was studied in order to find out rational homogenizing heat treatment process. The results show that Ce plays a role of refining grain and purifying molten alloy. The addition of Ce reduces dendritic spacing, refines the grain structures, eliminates dispersed shrinkage. The addition of Ce reduces the initial melting point of low melting eutectic phases by 3℃, under the same homogenization conditions. Trace Ce promotes the dissolution of low melting eutectic phases into the matrix, which improves the effect of homogenization. Homogenization temperatures of alloy A should be lower than 480℃and alloy B should be lower than 470℃; the addition of Ce decreases the homogenization temperature and improves the homogenization effect. The addition of Ce also greatly increases the tensile strength of the alloys.

Shipment and Disposal of Solidified Organic Waste (Waste Type IV) to the Waste Isolation Pilot Plant (WIPP)

International Nuclear Information System (INIS)

D'Amico, E. L; Edmiston, D. R.; O'Leary, G. A.; Rivera, M. A.; Steward, D. M.

2006-01-01

In April of 2005, the last shipment of transuranic (TRU) waste from the Rocky Flats Environmental Technology Site to the WIPP was completed. With the completion of this shipment, all transuranic waste generated and stored at Rocky Flats was successfully removed from the site and shipped to and disposed of at the WIPP. Some of the last waste to be shipped and disposed of at the WIPP was waste consisting of solidified organic liquids that is identified as Waste Type IV in the Contact-Handled Transuranic Waste Authorized Methods for Payload Control (CH-TRAMPAC) document. Waste Type IV waste typically has a composition, and associated characteristics, that make it significantly more difficult to ship and dispose of than other Waste Types, especially with respect to gas generation. This paper provides an overview of the experience gained at Rocky Flats for management, transportation and disposal of Type IV waste at WIPP, particularly with respect to gas generation testing. (authors)

The integrated melter off-gas treatment systems at the West Valley Demonstration Project

Energy Technology Data Exchange (ETDEWEB)

Vance, R.F. [West Valley Nuclear Services Co., Inc., NY (United States)

1995-02-01

The West Valley Demonstration Project was established by Public Law 96-368, the {open_quotes}West Valley Demonstration Project Act, {close_quotes} on October 1, l980. Under this act, Congress directed the Department of Energy to carry out a high level radioactive waste management demonstration project at the Western New York Nuclear Service Center in West Valley, New York. The purpose of this project is to demonstrate solidification techniques which can be used for preparing high level radioactive waste for disposal. In addition to developing this technology, the West Valley Demonstration Project Act directs the Department of Energy to: (1) develop containers suitable for permanent disposal of the high level waste; (2) transport the solidified high level waste to a Federal repository; (3) dispose of low level and transuranic waste produced under the project; and (4) decontaminate and decommission the facilities and materials associated with project activities and the storage tanks originally used to store the liquid high level radioactive waste. The process of vitrification will be used to solidify the high level radioactive liquid wastes into borosilicate glass. This report describes the functions, the controlling design criteria, and the resulting design of the melter off-gas treatment systems which are used in the vitrification process.

Effect of pH on the release of radionuclides and chelating agents from cement-solidified decontamination ion-exchange resins collected from operating nuclear power stations

International Nuclear Information System (INIS)

McIsaac, C.V.; Akers, D.W.; McConnell, J.W.

1991-06-01

Data are presented on the physical stability and leachability of radionuclides and chelating agents from cement-solidified decontamination ion-exchange resin wastes collected from two operating commercial light water reactors. Small-scale waste--form specimens collected during solidifications performed at the Brunswick Steam Electric Plant Unit 1 and at the James A. FitzPatrick Nuclear Power Station were leach-tested and subjected to compressive strength testing in accordance with the Nuclear Regulatory Commission's ''Technical Position on Waste Form'' (Revision 1). Samples of untreated resin waste collected from each solidification vessel before the solidification process were analyzed for concentrations of radionuclides, selected transition metals, and chelating agents to determine the quantities of these chemicals in the waste-form specimens. The chelating agents included oxalic, citric, and picolinic acids. In order to determine the effect of leachant chemical composition and pH on the stability and leachability of the waste forms, waste-form specimens were leached in various leachants. Results of this study indicate that differences in pH do not affect releases from cement-solidified decontamination ion-exchange resin waste forms, but that differences in leachant chemistry and the presence of chelating agents may affect the releases of radionuclides and chelating agents. Also, this study indicates that the cumulative releases of radionuclides and chelating agents are similar for waste- form specimens that decomposed and those that retained their general physical form. 36 refs., 60 figs., 28 tabs

40 CFR 227.30 - High-level radioactive waste.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 24 2010-07-01 2010-07-01 false High-level radioactive waste. 227.30...-level radioactive waste. High-level radioactive waste means the aqueous waste resulting from the operation of the first cycle solvent extraction system, or equivalent, and the concentrated waste from...

Some thermal analysis aspects of metal encapsulated waste

International Nuclear Information System (INIS)

Jardine, L.J.; Steindler, M.J.

1978-01-01

This paper is to summarize two waste management schemes: (1) packaging for extended storage of LWR spent fuel assemblies, with the capability for simple conversion either to terminal storage if a ''throwaway'' fuel cycle is ultimately adopted or to a form that can be reprocessed and (2) packaging for the terminal storage of solidified high-level wastes when the reprocessing of spent fuel is initiated. Only concepts utilizing metals or metal alloys to encapsulate either spent fuel or solidified high-level waste forms have been considered. Conceptual process flow sheets have been constructed to allow potential advantages and disadvantages of encapsulation alternatives to be identified in comparison with more conventional reference processes. Identification is also made of uncertainties of the analysis due to a lack of fundamental data required to perform evaluations. 3 tables

Hot dewatering and resin encapsulation of intermediate level radioactive waste

International Nuclear Information System (INIS)

Rickman, J.; Birch, D.

1985-01-01

The chemistry of the processes involved in the hot dewatering and encapsulation of alumino-ferric hydroxide floc in epoxide resin have been studied. Pretreatment of the floc to reduce resin attack and hydrolysis and to increase the dimensional stability of the solidified wasteform has been evaluated. It has been demonstrated that removal of ammonium nitrate from the floc and control of the residual water in the resin are important factors in ensuring dimensional stability of the solidified resin. Resin systems have been identified which, together with the appropriate waste pretreatment have successfully encapsulated a simulated magnox sludge producing a stable wasteform having mechanical and physical properties comparable with the basic resin. (author)

Highly flexible and all-solid-state paperlike polymer supercapacitors.

Science.gov (United States)

Meng, Chuizhou; Liu, Changhong; Chen, Luzhuo; Hu, Chunhua; Fan, Shoushan

2010-10-13

In recent years, much effort have been dedicated to achieve thin, lightweight and even flexible energy-storage devices for wearable electronics. Here we demonstrate a novel kind of ultrathin all-solid-state supercapacitor configuration with an extremely simple process using two slightly separated polyaniline-based electrodes well solidified in the H(2)SO(4)-polyvinyl alcohol gel electrolyte. The thickness of the entire device is much comparable to that of a piece of commercial standard A4 print paper. Under its highly flexible (twisting) state, the integrate device shows a high specific capacitance of 350 F/g for the electrode materials, well cycle stability after 1000 cycles and a leakage current of as small as 17.2 μA. Furthermore, due to its polymer-based component structure, it has a specific capacitance of as high as 31.4 F/g for the entire device, which is more than 6 times that of current high-level commercial supercapacitor products. These highly flexible and all-solid-state paperlike polymer supercapacitors may bring new design opportunities of device configuration for energy-storage devices in the future wearable electronic area.

The development of basic glass formulations for solidifying HLW from nuclear fuel reprocessing plant

International Nuclear Information System (INIS)

Jiang Yaozhong; Tang Baolong; Zhang Baoshan; Zhou Hui

1995-01-01

Basic glass formulations 90U/19, 90U/20, 90Nd/7 and 90Nd/10 applied in electric melting process are developed by using the mathematical model of the viscosity and electric resistance of waste glass. The yellow phase does not occur for basic glass formulations 90U/19 and 90U/20 solidifying HLW from nuclear fuel reprocessing plant when the waste loading is 20%. Under the waste loading is 16%, the process and product properties of glass 90U/19 and 90U/20 come up to or surpass the properties of the same kind of foreign waste glasses, and other properties are about the same to them of foreign waste glasses. The process and product properties of basic glass formulations 90Nd/7 and 90Nd/10 used for the solidification of 'U replaced by Nd' liquid waste are almost similar to them of 90U/19 and 90U/20. These properties fairly meet the requirements of 'joint test' (performed at KfK-INE, Germany). Among these formulations, 90Nd/7 is applied in cold engineering scale electric melting test performed at KfK-INE in Germany. The main process properties of cold test is similar to laboratory results

Evaluating Local Primary Dendrite Arm Spacing Characterization Techniques Using Synthetic Directionally Solidified Dendritic Microstructures

Science.gov (United States)

Tschopp, Mark A.; Miller, Jonathan D.; Oppedal, Andrew L.; Solanki, Kiran N.

2015-10-01

Microstructure characterization continues to play an important bridge to understanding why particular processing routes or parameters affect the properties of materials. This statement certainly holds true in the case of directionally solidified dendritic microstructures, where characterizing the primary dendrite arm spacing is vital to developing the process-structure-property relationships that can lead to the design and optimization of processing routes for defined properties. In this work, four series of simulations were used to examine the capability of a few Voronoi-based techniques to capture local microstructure statistics (primary dendrite arm spacing and coordination number) in controlled (synthetically generated) microstructures. These simulations used both cubic and hexagonal microstructures with varying degrees of disorder (noise) to study the effects of length scale, base microstructure, microstructure variability, and technique parameters on the local PDAS distribution, local coordination number distribution, bulk PDAS, and bulk coordination number. The Voronoi tesselation technique with a polygon-side-length criterion correctly characterized the known synthetic microstructures. By systematically studying the different techniques for quantifying local primary dendrite arm spacings, we have evaluated their capability to capture this important microstructure feature in different dendritic microstructures, which can be an important step for experimentally correlating with both processing and properties in single crystal nickel-based superalloys.

Operation for Rokkasho Low Level Radioactive Waste Disposal Center

International Nuclear Information System (INIS)

Kamizono, Hideki

2008-01-01

The Rokkasho Low Level Radioactive Waste (LLW) Disposal Center is located in Oishitai, Rokkasho-mura, Kamikitagun, of Aomori Prefecture. This district is situated in the southern part of Shimohita Peninsula in the northeastern corner of the prefecture, which lies at the northern tip of Honshu, Japan's main island. The Rokkasho LLW Disposal Center deals with only LLW generated by operating of nuclear power plants. The No.1 and No.2 disposal facility are now in operation. The disposal facilities in operation have a total dispose capacity of 80,000m 3 (equivalent to 400,000 drums). Our final business scope is to dispose of radioactive waste corresponding to 600,000 m 3 (equivalent to 3000,000 drums). For No.1 disposal facility, we have been disposing of homogeneous waste, including condensed liquid waste, spent resin, solidified with cement and asphalt, etc. For No.2 disposal facility, we can bury a solid waste solidified with mortar, such as activated metals and plastics, etc. Using an improved construction technology for an artificial barrier, the concrete pits in No.2 disposal facility could be constructed more economical and spacious than that of No.1. Both No.1 and No.2 facility will be able to bury about 200,000 waste packages (drums) each corresponding to 40,000 m 3 . As of March 17, 2008, Approximately 200,00 waste drums summing up No.1 and No.2 disposal facility have been received from Nuclear power plants and buried. (author)

Growth and microstructure formation of isothermally-solidified Zircaloy-4 joints brazed by a Zr-Ti-Cu-Ni amorphous alloy ribbon

Science.gov (United States)

Kim, K. H.; Lim, C. H.; Lee, J. G.; Lee, M. K.; Rhee, C. K.

2013-10-01

The microstructure and growth characteristics of Zircaloy-4 joints brazed by a Zr48Ti16Cu17Ni19 (at.%) amorphous filler metal have been investigated with regard to the controlled isothermal solidification and intermetallic formation. Two typical joints were produced depending on the isothermal brazing temperature: (1) a dendritic growth structure including bulky segregation in the central zone (at 850 °C), and (2) a homogeneous dendritic structure throughout the joint without segregation (at 890 °C). The primary α-Zr phase was solidified isothermally, nucleating to grow into a joint with a cellular or dendritic structure. Also, the continuous Zr2Ni and particulate Zr2Cu phases were formed in the segregated center zone and at the intercellular region, respectively, owing to the different solubility and atomic mobility of the solute elements (Ti, Cu, and Ni) in the α-Zr matrix. A disappearance of the central Zr2Ni phase was also rate-controlled by the outward diffusion of the Cu and Ni elements. When the detrimental Zr2Ni intermetallic phase was eliminated by a complete isothermal solidification at 890 °C, the strengths of the joints were high enough to cause yielding and fracture in the base metal, exceeding those of the bulk Zircaloy-4, at room temperature as well as at elevated temperatures (up to 400 °C).

Results after nine years of field testing low-level radioactive waste forms using lysimeters

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Rogers, R.D.; Jastrow, J.D.; Sanford, W.E.; Sullivan, T.M.

1995-01-01

The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program is obtaining information on the performance of radioactive waste forms. Ion-exchange resins from a nuclear power station were solidified into waste forms using Portland cement and vinyl ester-styrene. These waste forms are being tested to develop a low-level waste data base and to obtain information on survivability of waste forms in a disposal environment. This paper reviews radionuclide releases from those waste forms in the first 9 years of sampling. Included is a discussion of the recently discovered upward migration of radionuclides. Also, lysimeter data are applied to a performance assessment source term model, and initial results are presented

Report on achievements in fiscal 1999. Research and development of immediately effective and innovative energy environment technology (Development of immediately effective and high-efficiency solar cell technology, development of high-quality ingot manufacturing technology, and development of high-efficiency cell making technology); 1999 nendo sokkoteki kakushinteki energy kankyo gijutsu kenkyu kaihatsu seika hokokusho. Sokkogata kokoritsu taiyo denchi gijutsu kaihatsu (kohinshitsu ingot seizo gijutsu kaihatsu / kohinshitsu cell ka gijutsu kaihatsu)

Energy Technology Data Exchange (ETDEWEB)

NONE

2000-03-01

Research and development has been made on improving quality of ingots for substrates, manufacturing high-quality thin type substrates, and making high-efficiency cells. This paper summarizes the achievements in fiscal 1999. In developing the high-quality ingot manufacturing technology, discussions were given on a method for assessing impurities and crystal defects by using the total reflection scattering type infrared tomography, and on the optimal solidifying and cooling conditions during the ingot manufacturing by using simulation calculation for solidification. As a result of analyses and discussions, such findings were found effective that the ingot should be solidified through making the solid-liquid interface shape flat, and the temperature falling rate in an ingot should be maintained constant. In developing the high-efficiency cell making technology, discussions were given on the optimal construction based on a simulation that assumes the light sealing structure using the RIE method, and on the optimal construction of polycrystalline silicon solar cells by using a device simulator (PCID). The important factors in achieving a conversion efficiency of 20% are the light sealing structure, surface passivation, and substrate thickness. (NEDO)

Layered Composite of TiC-TiB2 to Ti-6Al-4V in Graded Composition by Combustion Synthesis in High-gravity Field

International Nuclear Information System (INIS)

Huang Xuegang; Zhao Zhongmin; Zhang Long

2013-01-01

By taking combustion synthesis to prepare solidified TiB 2 matrix ceramic in high-gravity field, the layered composite of TiC-TiB 2 ceramic to Ti-6Al-4V substrate in graded composition was achieved. XRD, FESEM and EDS results showed that the bulk full-density solidified TiC-TiB 2 composite was composed of fine TiB 2 platelets, TiC irregular grains, a few of α-Al 2 O 3 inclusions and Cr alloy phases, and α'-Ti phases alternating with Ti-enriched carbides constituted the matrix of the joint in which fine TiB platelets were embedded, whereas some C, B atoms were also detected at the heat-affected zone of Ti-6A1-4V substrate. The layered composite of the solidified ceramic to Ti-6Al-4V substrate in graded composition with continuous microstructure was considered a result of fused joint and inter-diffusion between liquid ceramic and surface-molten Ti alloy, followed by TiB 2 -Ti peritectic reaction and subsequent eutectic reaction in TiC-TiB-Ti ternary system.

Modeling of zinc solubility in stabilized/solidified electric arc furnace dust

International Nuclear Information System (INIS)

Fernandez-Olmo, Ignacio; Lasa, Cristina; Irabien, Angel

2007-01-01

Equilibrium models which attempt for the influence of pH on the solubility of metals can improve the dynamic leaching models developed to describe the long-term behavior of waste-derived forms. In addition, such models can be used to predict the concentration of metals in equilibrium leaching tests at a given pH. The aim of this work is to model the equilibrium concentration of Zn from untreated and stabilized/solidified (S/S) electric arc furnace dust (EAFD) using experimental data obtained from a pH-dependence leaching test (acid neutralization capacity, ANC). EAFD is a hazardous waste generated in electric arc furnace steel factories; it contains significant amounts of heavy metals such as Zn, Pb, Cr or Cd. EAFD from a local factory was characterized by X-ray fluorescence (XRF), acid digestion and X-ray diffraction (XRD). Zn and Fe were the main components while the XRD analysis revealed that zincite, zinc ferrite and hematite were the main crystalline phases. Different cement/EAFD formulations ranging from 7 to 20% dry weight of cement were prepared and subjected to the ANC leaching test. An amphoteric behavior of Zn was found from the pH dependence test. To model this behavior, the geochemical model Visual MINTEQ (VMINTEQ) was used. In addition to the geochemical model, an empirical model based on the dissolution of Zn in the acidic zone and the re-dissolution of zinc compounds in the alkaline zone was considered showing a similar prediction than that obtained with VMINTEQ. This empirical model seems to be more appropriate when the metal speciation is unknown, or when if known, the theoretical solid phases included in the database of VMINTEQ do not allow to describe the experimental data

Handling and storage of conditioned high-level wastes

International Nuclear Information System (INIS)

1983-01-01

This report deals with certain aspects of the management of one of the most important wastes, i.e. the handling and storage of conditioned (immobilized and packaged) high-level waste from the reprocessing of spent nuclear fuel and, although much of the material presented here is based on information concerning high-level waste from reprocessing LWR fuel, the principles, as well as many of the details involved, are applicable to all fuel types. The report provides illustrative background material on the arising and characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The report introduces the principles important in conditioned high-level waste storage and describes the types of equipment and facilities, used or studied, for handling and storage of such waste. Finally, it discusses the safety and economic aspects that are considered in the design and operation of handling and storage facilities

Handling and storage of conditioned high-level wastes

International Nuclear Information System (INIS)

Heafield, W.

1984-01-01

This paper deals with certain aspects of the management of one of the most important radioactive wastes arising from the nuclear fuel cycle, i.e. the handling and storage of conditioned high-level wastes. The paper is based on an IAEA report of the same title published during 1983 in the Technical Reports Series. The paper provides illustrative background material on the characteristics of high-level wastes and, qualitatively, their requirements for conditioning. The principles important in the storage of high-level wastes are reviewed in conjunction with the radiological and socio-political considerations involved. Four fundamentally different storage concepts are described with reference to published information and the safety aspects of particular storage concepts are discussed. Finally, overall conclusions are presented which confirm the availability of technology for constructing and operating conditioned high-level waste storage facilities for periods of at least several decades. (author)

Development of radioactive waste treatment system for nuclear power stations by Toshiba (III)

International Nuclear Information System (INIS)

Irie, H.; Takahara, T.; Matsuda, T.; Matsuura, H.; Yasumura, K.; Nakayama, Y.

1989-01-01

This paper describes a solidification process with thermosetting resin to satisfy both requirements of volume reduction and quality of solidified products. Volumes of solidified products in drums generated from spent resins and concentrated wastes were reduced respectively to 1/4 and less than 1/6 of those in the conventional cement solidification process. In plants using a simple demineralizing system for condensate polishing, a large amount of waste water with regenerant chemicals is generated from the condensate demineralizer. In general, radioactivity concentration of wastes from this type of nuclear power plant is comparatively high, so the dose rate at the surface of drums containing solidified wastes exceeds 200mR/h. A pelletizing system for radioactive wastes was developed to reduce their volumes and allow their interim storage until the radioactivity decays down to a level at which they can be handled easily

Phase Composition of a CrMo0.5NbTa0.5TiZr High Entropy Alloy: Comparison of Experimental and Simulated Data

Directory of Open Access Journals (Sweden)

Fan Zhang

2013-09-01

Full Text Available Microstructure and phase composition of a CrMo0.5NbTa0.5TiZr high entropy alloy were studied in the as-solidified and heat treated conditions. In the as-solidified condition, the alloy consisted of two disordered BCC phases and an ordered cubic Laves phase. The BCC1 phase solidified in the form of dendrites enriched with Mo, Ta and Nb, and its volume fraction was 42%. The BCC2 and Laves phases solidified by the eutectic-type reaction, and their volume fractions were 27% and 31%, respectively. The BCC2 phase was enriched with Ti and Zr and the Laves phase was heavily enriched with Cr. After hot isostatic pressing at 1450 °C for 3 h, the BCC1 dendrites coagulated into round-shaped particles and their volume fraction increased to 67%. The volume fractions of the BCC2 and Laves phases decreased to 16% and 17%, respectively. After subsequent annealing at 1000 °C for 100 h, submicron-sized Laves particles precipitated inside the BCC1 phase, and the alloy consisted of 52% BCC1, 16% BCC2 and 32% Laves phases. Solidification and phase equilibrium simulations were conducted for the CrMo0.5NbTa0.5TiZr alloy using a thermodynamic database developed by CompuTherm LLC. Some discrepancies were found between the calculated and experimental results and the reasons for these discrepancies were discussed.

The ATLAS High-Level Calorimeter Trigger in Run-2

CERN Document Server

Wiglesworth, Craig; The ATLAS collaboration

2018-01-01

The ATLAS Experiment uses a two-level triggering system to identify and record collision events containing a wide variety of physics signatures. It reduces the event rate from the bunch-crossing rate of 40 MHz to an average recording rate of 1 kHz, whilst maintaining high efficiency for interesting collision events. It is composed of an initial hardware-based level-1 trigger followed by a software-based high-level trigger. A central component of the high-level trigger is the calorimeter trigger. This is responsible for processing data from the electromagnetic and hadronic calorimeters in order to identify electrons, photons, taus, jets and missing transverse energy. In this talk I will present the performance of the high-level calorimeter trigger in Run-2, noting the improvements that have been made in response to the challenges of operating at high luminosity.

Combined storage system for LWR spent fuel and high-level waste

International Nuclear Information System (INIS)

Baxter, B.J.; Ganley, J.T.; Washington, J.A.

1983-01-01

The MODREX storage system consists of four basic elements: (1) the storage canister, (2) the storage module, (3) the storage cask, and (4) the transport cask. The storage canister is the heart of the system and, when used in combination with the module or either of the casks, allows the MODREX system to respond quickly to varied storage system requirements. The MODREX system can be used to hold either spent fuel assemblies or canistered solidified HLW. The ability to combine a basic storage canister with either a concrete module or a metal cask provides flexibility to meet a wide range of storage requirements. The spent fuel is stored in a dry, inert atmosphere, which essentially eliminates corrosion or deterioration of the cladding during extended storage periods. The storage canister and concrete storage module provide additional barriers against radioactivity release, enhancing long-term safety. Heat dissipation is passive, eliminating the need for additional emergency cooling systems or special redundancy. Modular, expandable construction permits minimum initial investment and capital carrying charges; additional capacity is built and paid for only as it is needed, retaining flexibility. 6 references, 2 figures, 1 table

The effects of aging on compressive strength of low-level radioactive waste form samples

International Nuclear Information System (INIS)

McConnell, J.W. Jr.; Neilson, R.M. Jr.

1996-06-01

The Field Lysimeter Investigations: Low-Level Waste Data Base Development Program, funded by the US Nuclear Regulatory Commission (NRC), is (a) studying the degradation effects in organic ion-exchange resins caused by radiation, (b) examining the adequacy of test procedures recommended in the Branch Technical Position on Waste Form to meet the requirements of 10 CFR 61 using solidified ion-exchange resins, (c) obtaining performance information on solidified ion-exchange resins in a disposal environment, and (d) determining the condition of liners used to dispose ion-exchange resins. Compressive tests were performed periodically over a 12-year period as part of the Technical Position testing. Results of that compressive testing are presented and discussed. During the study, both portland type I-II cement and Dow vinyl ester-styrene waste form samples were tested. This testing was designed to examine the effects of aging caused by self-irradiation on the compressive strength of the waste forms. Also presented is a brief summary of the results of waste form characterization, which has been conducted in 1986, using tests recommended in the Technical Position on Waste Form. The aging test results are compared to the results of those earlier tests. 14 refs., 52 figs., 5 tabs

High-Level Waste System Process Interface Description

International Nuclear Information System (INIS)

D'Entremont, P.D.

1999-01-01

The High-Level Waste System is a set of six different processes interconnected by pipelines. These processes function as one large treatment plant that receives, stores, and treats high-level wastes from various generators at SRS and converts them into forms suitable for final disposal. The three major forms are borosilicate glass, which will be eventually disposed of in a Federal Repository, Saltstone to be buried on site, and treated water effluent that is released to the environment

High-Level Waste Vitrification Facility Feasibility Study

Energy Technology Data Exchange (ETDEWEB)

D. A. Lopez

1999-08-01

A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035.

High-Level Waste Vitrification Facility Feasibility Study

International Nuclear Information System (INIS)

D. A. Lopez

1999-01-01

A ''Settlement Agreement'' between the Department of Energy and the State of Idaho mandates that all radioactive high-level waste now stored at the Idaho Nuclear Technology and Engineering Center will be treated so that it is ready to be moved out of Idaho for disposal by a compliance date of 2035. This report investigates vitrification treatment of the high-level waste in a High-Level Waste Vitrification Facility based on the assumption that no more New Waste Calcining Facility campaigns will be conducted after June 2000. Under this option, the sodium-bearing waste remaining in the Idaho Nuclear Technology and Engineering Center Tank Farm, and newly generated liquid waste produced between now and the start of 2013, will be processed using a different option, such as a Cesium Ion Exchange Facility. The cesium-saturated waste from this other option will be sent to the Calcine Solids Storage Facilities to be mixed with existing calcine. The calcine and cesium-saturated waste will be processed in the High-Level Waste Vitrification Facility by the end of calendar year 2035. In addition, the High-Level Waste Vitrification Facility will process all newly-generated liquid waste produced between 2013 and the end of 2035. Vitrification of this waste is an acceptable treatment method for complying with the Settlement Agreement. This method involves vitrifying the waste and pouring it into stainless-steel canisters that will be ready for shipment out of Idaho to a disposal facility by 2035. These canisters will be stored at the Idaho National Engineering and Environmental Laboratory until they are sent to a national geologic repository. The operating period for vitrification treatment will be from the end of 2015 through 2035

Effects of Microalloying on the Microstructures and Mechanical Properties of Directionally Solidified Ni-33(at.%)Al-31Cr-3Mo Eutectic Alloys Investigated

Science.gov (United States)

Whittenberger, J. Daniel; Raj, Sai V.; Locci, Ivan E.; Salem, Jonathan A.

2002-01-01

Despite nickel aluminide (NiAl) alloys' attractive combination of oxidation and thermophysical properties, their development as replacements for superalloy airfoils in gas turbine engines has been largely limited by difficulties in developing alloys with an optimum combination of elevated-temperature creep resistance and room-temperature fracture toughness. Alternatively, research has focused on developing directionally solidified NiAl-based in situ eutectic composites composed of NiAl and (Cr,Mo) phases in order to obtain a desirable combination of properties a systematic investigation was undertaken at the NASA Glenn Research Center to examine the effects of small additions of 11 alloying elements (Co, Cu, Fe, Hf, Mn, Nb, Re, Si, Ta, Ti, and Zr) in amounts varying from 0.25 to 1.0 at.% on the elevated-temperature strength and room-temperature fracture toughness of directionally solidified Ni-33Al-31Cr-3Mo eutectic alloy. The alloys were grown at 12.7 mm/hr, where the unalloyed eutectic base alloy exhibited a planar eutectic microstructure. The different microstructures that formed because of these fifth-element additions are included in the table. The additions of these elements even in small amounts resulted in the formation of cellular microstructures, and in some cases, dendrites and third phases were observed. Most of these elemental additions did not improve either the elevated-temperature strength or the room-temperature fracture toughness over that of the base alloy. However, small improvements in the compression strength were observed between 1200 and 1400 K when 0.5 at.% Hf and 0.25 at.% Ti were added to the base alloy. The results of this study suggest that the microalloying of Ni-33Al-31Cr-3Mo will not significantly improve either its elevatedtemperature strength or its room-temperature fracture toughness. Thus, any improvements in these properties must be acquired by changing the processing conditions.

High level nuclear wastes

International Nuclear Information System (INIS)

Lopez Perez, B.

1987-01-01

The transformations involved in the nuclear fuels during the burn-up at the power nuclear reactors for burn-up levels of 33.000 MWd/th are considered. Graphs and data on the radioactivity variation with the cooling time and heat power of the irradiated fuel are presented. Likewise, the cycle of the fuel in light water reactors is presented and the alternatives for the nuclear waste management are discussed. A brief description of the management of the spent fuel as a high level nuclear waste is shown, explaining the reprocessing and giving data about the fission products and their radioactivities, which must be considered on the vitrification processes. On the final storage of the nuclear waste into depth geological burials, both alternatives are coincident. The countries supporting the reprocessing are indicated and the Spanish programm defined in the Plan Energetico Nacional (PEN) is shortly reviewed. (author) 8 figs., 4 tabs

High spin levels in 151Ho

International Nuclear Information System (INIS)

Gizon, J.; Gizon, A.; Andre, S.; Genevey, J.; Jastrzebski, J.; Kossakowski, R.; Moszinski, M.; Preibisz, Z.

1981-02-01

We report here on the first study of the level structure of 151 Ho. High spin levels in 151 Ho have been populated in the 141 Pr + 16 O and 144 Sm + 12 C reactions. The level structure has been established up to 6.6 MeV energy and the spins and particles determined up to 49/2 - . Most of the proposed level configurations can be explained by the coupling of hsub(11/2) protons to fsub(7/2) and/or hsub(9/2) neutrons. An isomer with 14 +- 3 ns half-life and a delayed gamma multiplicity equal to 17 +- 2 has been found. Its spin is larger than 57/2 h units

The potential of solidified molasses-based blocks for the correction of multinutritional deficiencies in buffaloes and other ruminants fed low-quality agro-industrial byproducts

International Nuclear Information System (INIS)

Leng, R.A.

1984-01-01

The main principles for formulating diets for ruminant animals in developing countries are outlined and examples provided of the successful application of these principles for feeding buffaloes and cattle in India, Philippines and Australia. It is concluded that the provision of a continuous supply of urea in the form of solidified feed blocks to increase the intake and digestibility of roughage-based diets is a management tool that could be used by small farmers in developing countries to improve weight gains and milk yields. Since such blocks can be easily supplemented with macro- and micro-elements needed by ruminants, they could also be useful for correcting multi-nutritional deficiencies. (author)

Radiation transport in high-level waste form

International Nuclear Information System (INIS)

Arakali, V.S.; Barnes, S.M.

1992-01-01

The waste form selected for vitrifying high-level nuclear waste stored in underground tanks at West Valley, NY is borosilicate glass. The maximum radiation level at the surface of a canister filled with the high-level waste form is prescribed by repository design criteria for handling and disposition of the vitrified waste. This paper presents an evaluation of the radiation transport characteristics for the vitreous waste form expected to be produced at West Valley and the resulting neutron and gamma dose rates. The maximum gamma and neutron dose rates are estimated to be less than 7500 R/h and 10 mRem/h respectively at the surface of a West Valley canister filled with borosilicate waste glass

Low-level waste cement solidification design, installation, and start-up

International Nuclear Information System (INIS)

Jezek, G.R.

1988-08-01

This report describes the design, installation, and start-up activities of the Cement Solidification System (CSS) at the West Valley Demonstration Project (WVDP), West Valley, New York. The CSS, designed to operate within an existing process cell, automatically and remotely solidifies low-level nuclear waste by mixing it with Portland Type I cement. The qualified waste form mixture is placed into square, 270-litre (71-gallon) metal drums. The drums have an integral polyethylene liner to protect the carbon-steel material from potential corrosion. The CSS produces drums at a continuous operation rate of four drums per hour. All system processing data is monitored by a computerized Data Acquisition System (DAS). 6 figs

Macrosegregation During Re-melting and Holding of Directionally Solidified Al-7 wt.% Si Alloy in Microgravity

Science.gov (United States)

Lauer, M.; Ghods, M.; Angart, S. G.; Grugel, R. N.; Tewari, S. N.; Poirier, D. R.

2017-08-01

As-cast aluminum-7 wt.% ailicon alloy sample rods were re-melted and directionally solidified on Earth which resulted in uniform dendritically aligned arrays. These arrays were then partially back-melted through an imposed, and constant, temperature gradient in the microgravity environment aboard the International Space Station. The mushy zones that developed in the seed crystals were held for different periods prior to initiating directional solidification. Upon return, examination of the initial mushy-zone regions exhibited significant macrosegregation in terms of a solute-depleted zone that increased as a function of the holding time. The silicon (solute) content in these regions was measured on prepared longitudinal sections by electron microprobe analysis as well as by determining the fraction eutectic on several transverse sections. The silicon content was found to increase up the temperature gradient resulting in significant silicon concentration immediately ahead of the mushy-zone tips. The measured macrosegregation agrees well with calculations from a mathematical model developed to simulate the re-melting and holding process. The results, due to processing in a microgravity environment where buoyancy and thermosolutal convection are minimized, serve as benchmark solidification data.

The ALICE Dimuon Spectrometer High Level Trigger

CERN Document Server

Becker, B; Cicalo, Corrado; Das, Indranil; de Vaux, Gareth; Fearick, Roger; Lindenstruth, Volker; Marras, Davide; Sanyal, Abhijit; Siddhanta, Sabyasachi; Staley, Florent; Steinbeck, Timm; Szostak, Artur; Usai, Gianluca; Vilakazi, Zeblon

2009-01-01

The ALICE Dimuon Spectrometer High Level Trigger (dHLT) is an on-line processing stage whose primary function is to select interesting events that contain distinct physics signals from heavy resonance decays such as J/psi and Gamma particles, amidst unwanted background events. It forms part of the High Level Trigger of the ALICE experiment, whose goal is to reduce the large data rate of about 25 GB/s from the ALICE detectors by an order of magnitude, without loosing interesting physics events. The dHLT has been implemented as a software trigger within a high performance and fault tolerant data transportation framework, which is run on a large cluster of commodity compute nodes. To reach the required processing speeds, the system is built as a concurrent system with a hierarchy of processing steps. The main algorithms perform partial event reconstruction, starting with hit reconstruction on the level of the raw data received from the spectrometer. Then a tracking algorithm finds track candidates from the recon...

Current high-level waste solidification technology

International Nuclear Information System (INIS)

Bonner, W.F.; Ross, W.A.

1976-01-01

Technology has been developed in the U.S. and abroad for solidification of high-level waste from nuclear power production. Several processes have been demonstrated with actual radioactive waste and are now being prepared for use in the commercial nuclear industry. Conversion of the waste to a glass form is favored because of its high degree of nondispersibility and safety

Effect of swaging on the 1000 C compressive slow plastic flow characteristics of the directionally solidified eutectic alloy gamma/gamma prime-alpha

Science.gov (United States)

Whittenberger, J. D.; Wirth, G.

1983-01-01

Swaging between 750 and 1050 C has been investigated as a means to introduce work into the directionally solidified eutectic alloy gamma/gamma prime-alpha (Ni-32.3 wt percent Mo-6.3 wt percent Al) and increase the elevated temperature creep strength. The 1000 C slow plastic compressive flow stress-strain rate properties in air of as-grown, annealed, and worked nominally 10 and 25 percent materials have been determined. Swaging did not improve the slow plastic behavior. In fact large reductions tended to degrade the strength and produced a change in the deformation mechanism from uniform flow to one involving intense slip band formation. Comparison of 1000 C tensile and compressive strength-strain rate data reveals that deformation is independent of the stress state.

The CMS High-Level Trigger

International Nuclear Information System (INIS)

Covarelli, R.

2009-01-01

At the startup of the LHC, the CMS data acquisition is expected to be able to sustain an event readout rate of up to 100 kHz from the Level-1 trigger. These events will be read into a large processor farm which will run the 'High-Level Trigger'(HLT) selection algorithms and will output a rate of about 150 Hz for permanent data storage. In this report HLT performances are shown for selections based on muons, electrons, photons, jets, missing transverse energy, τ leptons and b quarks: expected efficiencies, background rates and CPU time consumption are reported as well as relaxation criteria foreseen for a LHC startup instantaneous luminosity.

The CMS High-Level Trigger

CERN Document Server

Covarelli, Roberto

2009-01-01

At the startup of the LHC, the CMS data acquisition is expected to be able to sustain an event readout rate of up to 100 kHz from the Level-1 trigger. These events will be read into a large processor farm which will run the "High-Level Trigger" (HLT) selection algorithms and will output a rate of about 150 Hz for permanent data storage. In this report HLT performances are shown for selections based on muons, electrons, photons, jets, missing transverse energy, tau leptons and b quarks: expected efficiencies, background rates and CPU time consumption are reported as well as relaxation criteria foreseen for a LHC startup instantaneous luminosity.

The CMS High-Level Trigger

Science.gov (United States)

Covarelli, R.

2009-12-01

At the startup of the LHC, the CMS data acquisition is expected to be able to sustain an event readout rate of up to 100 kHz from the Level-1 trigger. These events will be read into a large processor farm which will run the "High-Level Trigger" (HLT) selection algorithms and will output a rate of about 150 Hz for permanent data storage. In this report HLT performances are shown for selections based on muons, electrons, photons, jets, missing transverse energy, τ leptons and b quarks: expected efficiencies, background rates and CPU time consumption are reported as well as relaxation criteria foreseen for a LHC startup instantaneous luminosity.

Overview of the performance objectives and scenarios of TWRS Low-Level Waste Disposal Program. Revision 1

International Nuclear Information System (INIS)

1995-01-01

As a result of past Department of Energy (DOE) weapons material production operations, Hanford now stores nuclear waste from processing facilities in underground tanks on the 200 area plateau. An agreement between the DOE, the Environmental Protection Agency (EPA), and the Washington state Department of Ecology (the Tri-Party Agreement, or TPA) establishes an enforceable schedule and a technical framework for recovering, processing, solidifying, and disposing of the Hanford tank wastes. The present plan includes retrieving the tank waste, pre-treating the waste to separate into low level and high level streams, and converting both streams to a glass waste form. The low level glass will represent by far the largest volume and lowest quantity of radioactivity (i.e., large volume of waste chemicals) of waste requiring disposal. The low level glass waste will be retrievably stored in sub-surface disposal vaults for several decades. Assuming the low level disposal system proves to be acceptable, the disposal site will be closed with the low level waste in place. If the disposal system is not acceptable, then the waste will be subject to possible retrieval followed by some other disposal solution. Westinghouse Hanford Company is also planning to emplace the waste so that it is retrievable for up to 50 years after completion of the tank waste processing

SIGWX Charts - High Level Significant Weather

Data.gov (United States)

National Oceanic and Atmospheric Administration, Department of Commerce — High level significant weather (SIGWX) forecasts are provided for the en-route portion of international flights. NOAA's National Weather Service Aviation Center...

Ramifications of defining high-level waste

International Nuclear Information System (INIS)

Wood, D.E.; Campbell, M.H.; Shupe, M.W.

1987-01-01

The Nuclear Regulatory Commission (NRC) is considering rule making to provide a concentration-based definition of high-level waste (HLW) under authority derived from the Nuclear Waste Policy Act (NWPA) of 1982 and the Low Level Waste Policy Amendments Act of 1985. The Department of Energy (DOE), which has the responsibility to dispose of certain kinds of commercial waste, is supporting development of a risk-based classification system by the Oak Ridge National Laboratory to assist in developing and implementing the NRC rule. The system is two dimensional, with the axes based on the phrases highly radioactive and requires permanent isolation in the definition of HLW in the NWPA. Defining HLW will reduce the ambiguity in the present source-based definition by providing concentration limits to establish which materials are to be called HLW. The system allows the possibility of greater-confinement disposal for some wastes which do not require the degree of isolation provided by a repository. The definition of HLW will provide a firm basis for waste processing options which involve partitioning of waste into a high-activity stream for repository disposal, and a low-activity stream for disposal elsewhere. Several possible classification systems have been derived and the characteristics of each are discussed. The Defense High Level Waste Technology Lead Office at DOE - Richland Operations Office, supported by Rockwell Hanford Operations, has coordinated reviews of the ORNL work by a technical peer review group and other DOE offices. The reviews produced several recommendations and identified several issues to be addressed in the NRC rule making. 10 references, 3 figures

Solidification of low-level waste - a dilemma for the small user

International Nuclear Information System (INIS)

Harris, S.; Gilmore, A.

1980-01-01

The requirement that radioactive waste for sea disposal must be solidified by the originator is discussed. Attempts to solidify small quantities of radioactive waste such as contaminated oils and labelled benzyopyrene with other solvents are described. Encapsulation media tested were concrete and interior and exterior grade Polyfilla (a plaster and cellulose based filler). Problems were presented by the difficulty of mixing the materials and by the maximum uptake of solvents while still allowing solidification. In all cases a soft crumbling material resulted. It is concluded that solidification processing on a small scale does not make economic or scientific sense and that if solidification is necessary it would be better carried out as a national operation by collecting liquids from users. (U.K.)

Scenarios of the TWRS low-level waste disposal program

International Nuclear Information System (INIS)

1994-10-01

As a result of past Department of Energy (DOE) weapons material production operations, Hanford now stores nuclear waste from processing facilities in underground tanks on the 200 Area plateau. An agreement between the DOE, the Environmental Protection Agency (EPA), and the Washington state Department of Ecology (the Tri-Party Agreement, or TPA) establishes an enforceable schedule and a technical framework for recovering, processing, solidifying, and disposing of the Hanford tank wastes. The present plan includes retrieving the tank waste, pretreating the waste to separate into low level and high level streams, and converting both streams to a glass waste form. The low level glass will represent by far the largest volume and lowest quantity of radioactivity (i.e., large volume of waste chemicals) of waste requiring disposal. The low level glass waste will be retrievably stored in sub-surface disposal vaults for several decades. If the low level disposal system proves to be acceptable, the disposal site will be closed with the low level waste in place. If, however, at some time the disposal system is found to be unacceptable, then the waste can be retrieved and dealt with in some other manner. WHC is planning to emplace the waste so that it is retrievable for up to 50 years after completion of the tank waste processing. Acceptability of disposal of the TWRS low level waste at Hanford depends on technical, cultural, and political considerations. The Performance Assessment is a major part of determining whether the proposed disposal action is technically defensible. A Performance Assessment estimates the possible future impact to humans and the environment for thousands of years into the future. In accordance with the TPA technical strategy, WHC plans to design a near-surface facility suitable for disposal of the glass waste

Design of high-temperature high-strength Al-Ti-V-Zr alloys

International Nuclear Information System (INIS)

Lee, H.M.

1990-01-01

This paper reports that it seems plausible to develop high-strength Al-base alloys useful up to 698K in view of the behavior of nickel base superalloys which resist degradation of mechanical properties to 75 pct of their absolute melting temperature. For high temperature Al alloys, the dispersed hardening phase must not undergo phase transformation to an undesirable phase during long time exposure at the temperature of interest. An additional factor to be considered is the stability of the hardening phase with respect to Ostwald ripening. This coarsening resistance is necessary so that the required strength level can be maintained after the long-time service at high temperatures. The equilibrium crystal structures of Al 3 Ti, Al 3 V and Al 3 Zr are tetragonal D0 22 , D0 22 and D0 23 , respectively. At the temperatures of interest, around 698K, vanadium and titanium are mutually substitutable in the form of Al 3 (Ti, V). Much of titanium and vanadium can be substituted for zirconium in the D0 23 - type Al 3 Zr compound, creating Al 3 (Ti, Zr) and Al 3 (V, Zr), respectively. In particular, it has been reported that fcc L1 2 -structured Al 3 M dispersoids form in the rapidly solidified Al-V-Zr and Al-Ti-Zr systems and both L1 2 and D0 23 -structured Al 3 M phases showed slow coarsening kinetics

Translation of a High-Level Temporal Model into Lower Level Models: Impact of Modelling at Different Description Levels

DEFF Research Database (Denmark)

Kraft, Peter; Sørensen, Jens Otto

2001-01-01

given types of properties, and examine how descriptions on higher levels translate into descriptions on lower levels. Our example looks at temporal properties where the information is concerned with the existence in time. In a high level temporal model with information kept in a three-dimensional space...... the existences in time can be mapped precisely and consistently securing a consistent handling of the temporal properties. We translate the high level temporal model into an entity-relationship model, with the information in a two-dimensional graph, and finally we look at the translations into relational...... and other textual models. We also consider the aptness of models that include procedural mechanisms such as active and object databases...

Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

Energy Technology Data Exchange (ETDEWEB)

ELsourougy, M R; Zaki, A A; Aly, H F [Atomic energy authority, hot laboratory center, Cairo, (Egypt); Khalil, M Y [Nuclear engineering department, Alexandria university. Alexandria, (Egypt)

1995-10-01

Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs.

Development of new treatment process for low level radioactive waste at Tokai reprocessing plant

International Nuclear Information System (INIS)

Horiguchi, Kenichi; Sugaya, Atsushi; Saito, Yasuo; Tanaka, Kenji; Akutsu, Shigeru; Hirata, Toshiaki

2009-01-01

The Low-level radioactive Waste Treatment Facility (LWTF) was constructed at the Tokai Reprocessing Plant (TRP) and cold testing has been carried out since 2006. The waste which will be treated in the LWTF is combustible/incombustible solid waste and liquid waste. In the LWTF, the combustible/incombustible solid waste will be incinerated. The liquid waste will be treated by a radio-nuclides removal process and subsequently solidified in cement. This report describes the essential technologies of the LWTF and results of R and D work for the nitrate-ion decomposition technology for the liquid waste. (author)

Investigations on cement/polymer Waste packages containing intermediate level waste and organic exchange resins

International Nuclear Information System (INIS)

ELsourougy, M.R.; Zaki, A.A.; Aly, H.F.; Khalil, M.Y.

1995-01-01

Polymers can be added to cements to improve its nuclear waste immobilization properties. This trend in cementation processes is attracting attention and requiring through investigations. In this work, polymers of different kinds were added to ordinary portland cement for the purpose of solidifying intermediate level liquid wastes and organic ion exchange resins. Epoxy polymer such as Kemapoxy-150 reduced the leaching rate of cesium compared to cement alone. Latex to cement ratio less than 4% caused an increase in leaching rate of cesium. When cesium was absorbed to an organic resin its leachability was improved. 5 figs., 4 tabs

Evaluation of radionuclide concentrations in high-level radioactive wastes

International Nuclear Information System (INIS)

Fehringer, D.J.

1985-10-01

This report describes a possible approach for development of a numerical definition of the term ''high-level radioactive waste.'' Five wastes are identified which are recognized as being high-level wastes under current, non-numerical definitions. The constituents of these wastes are examined and the most hazardous component radionuclides are identified. This report suggests that other wastes with similar concentrations of these radionuclides could also be defined as high-level wastes. 15 refs., 9 figs., 4 tabs

Timing of High-level Waste Disposal

International Nuclear Information System (INIS)

2008-01-01

This study identifies key factors influencing the timing of high-level waste (HLW) disposal and examines how social acceptability, technical soundness, environmental responsibility and economic feasibility impact on national strategies for HLW management and disposal. Based on case study analyses, it also presents the strategic approaches adopted in a number of national policies to address public concerns and civil society requirements regarding long-term stewardship of high-level radioactive waste. The findings and conclusions of the study confirm the importance of informing all stakeholders and involving them in the decision-making process in order to implement HLW disposal strategies successfully. This study will be of considerable interest to nuclear energy policy makers and analysts as well as to experts in the area of radioactive waste management and disposal. (author)

High-Level Waste Melter Study Report

Energy Technology Data Exchange (ETDEWEB)

Perez, Joseph M.; Bickford, Dennis F.; Day, Delbert E.; Kim, Dong-Sang; Lambert, Steven L.; Marra, Sharon L.; Peeler, David K.; Strachan, Denis M.; Triplett, Mark B.; Vienna, John D.; Wittman, Richard S.

2001-07-13

At the Hanford Site in Richland, Washington, the path to site cleanup involves vitrification of the majority of the wastes that currently reside in large underground tanks. A Joule-heated glass melter is the equipment of choice for vitrifying the high-level fraction of these wastes. Even though this technology has general national and international acceptance, opportunities may exist to improve or change the technology to reduce the enormous cost of accomplishing the mission of site cleanup. Consequently, the U.S. Department of Energy requested the staff of the Tanks Focus Area to review immobilization technologies, waste forms, and modifications to requirements for solidification of the high-level waste fraction at Hanford to determine what aspects could affect cost reductions with reasonable long-term risk. The results of this study are summarized in this report.

Rapid Solidification of Sn-Cu-Al Alloys for High-Reliability, Lead-Free Solder: Part II. Intermetallic Coarsening Behavior of Rapidly Solidified Solders After Multiple Reflows

Science.gov (United States)

Reeve, Kathlene N.; Choquette, Stephanie M.; Anderson, Iver E.; Handwerker, Carol A.

2016-12-01

Controlling the size, dispersion, and stability of intermetallic compounds in lead-free solder alloys is vital to creating reliable solder joints regardless of how many times the solder joints are melted and resolidified (reflowed) during circuit board assembly. In this article, the coarsening behavior of Cu x Al y and Cu6Sn5 in two Sn-Cu-Al alloys, a Sn-2.59Cu-0.43Al at. pct alloy produced via drip atomization and a Sn-5.39Cu-1.69Al at. pct alloy produced via melt spinning at a 5-m/s wheel speed, was characterized after multiple (1-5) reflow cycles via differential scanning calorimetry between the temperatures of 293 K and 523 K (20 °C and 250 °C). Little-to-no coarsening of the Cu x Al y particles was observed for either composition; however, clustering of Cu x Al y particles was observed. For Cu6Sn5 particle growth, a bimodal size distribution was observed for the drip atomized alloy, with large, faceted growth of Cu6Sn5 observed, while in the melt spun alloy, Cu6Sn5 particles displayed no significant increase in the average particle size, with irregularly shaped, nonfaceted Cu6Sn5 particles observed after reflow, which is consistent with shapes observed in the as-solidified alloys. The link between original alloy composition, reflow undercooling, and subsequent intermetallic coarsening behavior was discussed by using calculated solidification paths. The reflowed microstructures suggested that the heteroepitaxial relationship previously observed between the Cu x Al y and the Cu6Sn5 was maintained for both alloys.

Growth and microstructure formation of isothermally-solidified Zircaloy-4 joints brazed by a Zr–Ti–Cu–Ni amorphous alloy ribbon

Energy Technology Data Exchange (ETDEWEB)

Kim, K.H. [University of Science and Technology, Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of); Lim, C.H. [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of); Lee, J.G., E-mail: jglee88@kaeri.re.kr [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of); Lee, M.K.; Rhee, C.K. [Nuclear Materials Development Division, Korea Atomic Energy Research Institute (KAERI), Yuseong, Daejeon 305-353 (Korea, Republic of)

2013-10-15

The microstructure and growth characteristics of Zircaloy-4 joints brazed by a Zr{sub 48}Ti{sub 16}Cu{sub 17}Ni{sub 19} (at.%) amorphous filler metal have been investigated with regard to the controlled isothermal solidification and intermetallic formation. Two typical joints were produced depending on the isothermal brazing temperature: (1) a dendritic growth structure including bulky segregation in the central zone (at 850 °C), and (2) a homogeneous dendritic structure throughout the joint without segregation (at 890 °C). The primary α-Zr phase was solidified isothermally, nucleating to grow into a joint with a cellular or dendritic structure. Also, the continuous Zr{sub 2}Ni and particulate Zr{sub 2}Cu phases were formed in the segregated center zone and at the intercellular region, respectively, owing to the different solubility and atomic mobility of the solute elements (Ti, Cu, and Ni) in the α-Zr matrix. A disappearance of the central Zr{sub 2}Ni phase was also rate-controlled by the outward diffusion of the Cu and Ni elements. When the detrimental Zr{sub 2}Ni intermetallic phase was eliminated by a complete isothermal solidification at 890 °C, the strengths of the joints were high enough to cause yielding and fracture in the base metal, exceeding those of the bulk Zircaloy-4, at room temperature as well as at elevated temperatures (up to 400 °C)

Vitrification of liquid waste from nuclear power plants

International Nuclear Information System (INIS)

Sheng Jiawei; Choi, Kwansik; Song, Myung-Jae

2001-01-01

Glass is an acceptable waste form to solidify the low-level waste from nuclear power plants (NPPs) because of the simplicity of processing and its unique ability to accept a wide variety of waste streams. Vitrification is being considered to solidify the high-boron-containing liquid waste generated from Korean NPPs. This study dealt with the development of a glass formulation to solidify the liquid waste. Studies were conducted in a borosilicate glass system. Crucible studies have been performed with surrogate waste. Several developed glass frits were evaluated to determine their suitability for vitrifying the liquid waste. The results indicated that the 20 wt% waste oxides loading required could not be obtained using these glass frits. Flyash produced from coal-burning electric power stations, whose major components are SiO 2 and Al 2 O 3 , is a desirable glass network former. Detailed product evaluations including waste loading, homogeneity, chemical durability and viscosity, etc., were carried out on selected formulations using flyash. Up to 30 wt% of the waste oxides was successfully solidified into the flyash after the addition of 5-10 wt% Na 2 O at 1200 deg. C

Nondestructive testing of the low-level radioactive waste drums for uni-axial compressive strength and free liquid content

International Nuclear Information System (INIS)

Yu Geping; Chang Mingyu; Wang Yeajeng; Chu, David S.L.; Ju Yihzen

1992-01-01

This paper summarizes the nondestructive test to determine the uni-axial compressive strength and free water content of solidified low level radioactive waste. The uni-axial compressive strength is determined by ultrasonic wave propagation speed, and the results are compared with those of compressive tests. Three methods of detecting the surface free water by ultrasonic testing are established, the ultrasonic wave speed, wave form and pulse height are used to determine the existence and amount of the surface free liquid. Possible difficulties are discussed. (author)

Evolution of the microstructure and hardness of a rapidly solidified/melt-spun AZ91 alloy upon aging at different temperatures

International Nuclear Information System (INIS)

Wang Baishu; Liu Yongbing; An Jian; Li Rongguang; Su Zhenguo; Su Guihua; Lu You; Cao Zhanyi

2009-01-01

The effect of aging at different temperatures on a rapidly solidified/melt-spun AZ91 alloy has been investigated in depth. The microstructures of as-spun and aged ribbons with a thickness of approximately 60 μm were characterized using X-ray diffraction, transmission electron microscopy and laser optical microscopy; microhardness measurements were also conducted. It was found that the commercial AZ91 alloy undergoes a cellular/dendritic transition during melt-spinning at a speed of 34 m/s. A strengthening effect due to aging was observed: a maximum hardness of 110 HV/0.05 and an age-hardenability of 50% were obtained when the ribbon was aged at 200 deg. C for 20 min. The β-Mg 17 Al 12 phase exhibits net and dispersion types of distribution during precipitation. The dispersion of precipitates in dendritic grains or cells is the main source of strengthening

Sterilization, high-level disinfection, and environmental cleaning.

Science.gov (United States)

Rutala, William A; Weber, David J

2011-03-01

Failure to perform proper disinfection and sterilization of medical devices may lead to introduction of pathogens, resulting in infection. New techniques have been developed for achieving high-level disinfection and adequate environmental cleanliness. This article examines new technologies for sterilization and high-level disinfection of critical and semicritical items, respectively, and because semicritical items carry the greatest risk of infection, the authors discuss reprocessing semicritical items such as endoscopes and automated endoscope reprocessors, endocavitary probes, prostate biopsy probes, tonometers, laryngoscopes, and infrared coagulation devices. In addition, current issues and practices associated with environmental cleaning are reviewed. Copyright © 2011. Published by Elsevier Inc.

West Valley Demonstration Project, West Valley, New York: Annual report

International Nuclear Information System (INIS)

1989-01-01

Under the West Valley Demonstration Project Act, Public Law 96-368, liquid high-level radioactive waste stored at the Western New York Nuclear Services Center, West Valley, New York, that resulted from spent nuclear fuel reprocessing operations conducted between 1966 and 1972, is to be solidified in borosilicate glass and transported to a federal repository for geologic disposal. A major milestone was reached in May 1988 when the Project began reducing the volume of the liquid high-level waste. By the end of 1988, approximately 15 percent of the initial inventory had been processed into two waste streams. The decontaminated low-level liquid waste is being solidified in cement. The high-level waste stream is being stored in an underground tank pending its incorporation into borosilicate glass. Four tests of the waste glass melter system were completed. These tests confirmed equipment operability, control system reliability, and provided samples of waste glass for durability testing. In mid-1988, the Department validated an integrated cost and schedule plan for activities required to complete the production of the waste borosilicate glass. Design of the radioactive Vitrification Facility continued

High level of CA 125 due to large endometrioma.

Science.gov (United States)

Phupong, Vorapong; Chen, Orawan; Ultchaswadi, Pornthip

2004-09-01

CA 125 is a tumor-associated antigen. Its high levels are usually associated with ovarian malignancies, whereas smaller increases in the levels were associated with benign gynecologic conditions. The authors report a high level of CA 125 in a case of large ovarian endometrioma. A 45-year-old nulliparous Thai woman, presented with an increase of her abdominal girth for 7 months. Transabdominal ultrasonogram demonstrated a large ovarian cyst and multiple small leiomyoma uteri, and serum CA 125 level was 1,006 U/ml. The preoperative diagnosis was ovarian cancer with leiomyoma uteri. Exploratory laparotomy was performed. There were a large right ovarian endometrioma, small left ovarian endometrioma and multiple small leiomyoma. Total abdominal hysterectomy and bilateral salpingo-oophorectomy was performed and histopathology confirmed the diagnosis of endometrioma and leiomyoma. The serum CA 125 level declined to non-detectable at the 4th week. She was well at discharge and throughout her 4th week follow-up period Although a very high level of CA 125 is associated with a malignant process, it can also be found in benign conditions such as a large endometrioma. The case emphasizes the association of high levels of CA 125 with benign gynecologic conditions.

Extended storage of low-level radioactive waste: potential problem areas

International Nuclear Information System (INIS)

Siskind, B.; Dougherty, D.R.; MacKenzie, D.R.

1985-01-01

If a state or state compact does not have adequate disposal capacity for low-level radioactive waste (LLRW) by 1986 as required by the Low-Level Waste Policy Act, then extended storage of certain LLRW may be necessary. The issue of extended storage of LLRW is addressed in order to determine for the Nuclear Regulatory Commission the areas of concern and the actions recommended to resolve these concerns. The focus is on the properties and behavior of the waste form and waste container. Storage alternatives are considered in order to characterize the likely storage environments for these wastes. The areas of concern about extended storage of LLRW are grouped into two categories: 1. Behavior of the waste form and/or container during storage, e.g., radiolytic gas generation, radiation-enhanced degradation of polymeric materials, and corrosion. 2. Effects of extended storage on the properties of the waste form and/or container that are important after storage (e.g., radiation-induced oxidative embrittlement of high-density polyethylene and the weakening of steel containers resulting from corrosion by the waste). The additional information and actions required to address these concerns are discussed and, in particular, it is concluded that further information is needed on the rates of corrosion of container material by Class A wastes and on the apparent dose-rate dependence of radiolytic processes in Class B and C waste packages. Modifications to the guidance for solidified wastes and high-integrity containers in NRC's Technical Position on Waste Form are recommended. 27 references

Modeling of microstructure evolution of magnesium alloy during the high pressure die casting process

International Nuclear Information System (INIS)

Wu Mengwu; Xiong Shoumei

2012-01-01

Two important microstructure characteristics of high pressure die cast magnesium alloy are the externally solidified crystals (ESCs) and the fully divorced eutectic which form at the filling stage of the shot sleeve and at the last stage of solidification in the die cavity, respectively. Both of them have a significant influence on the mechanical properties and performance of magnesium alloy die castings. In the present paper, a numerical model based on the cellular automaton (CA) method was developed to simulate the microstructure evolution of magnesium alloy during cold-chamber high pressure die casting (HPDC) process. Modeling of dendritic growth of magnesium alloy with six-fold symmetry was achieved by defining a special neighbourhood configuration and calculating of the growth kinetics from complete solution of the transport equations. Special attention was paid to establish a nucleation model considering both of the nucleation of externally solidified crystals in the shot sleeve and the massive nucleation in the die cavity. Meanwhile, simulation of the formation of fully divorced eutectic was also taken into account in the present CA model. Validation was performed and the capability of the present model was addressed by comparing the simulated results with those obtained by experiments.

High-Mn steel weldment mechanical properties at 4 K

International Nuclear Information System (INIS)

Chan, J.W.; Sunwoo, A.J.; Morris, J.W. Jr.

1988-06-01

Advanced high-field superconducting magnets of the next generation of magnetic confinement fusion devices will require structural alloys with high yield strength and high toughness at cryogenic temperatures. Commercially available alloys used in the current generation of magnets, such as 300 series stainless steels, do not have the required properties. N-strengthened, high-Mn alloys meet base plate requirements in the as-rolled condition. However, the property changes associated with weld microstructural and chemical changes in these alloys have not been well characterized. In this work welding induced cryogenic mechanical property changes of an 18Mn-16Cr-5Ni-0.2N alloy are correlated with as-solidified weld microstructures and chemistries. 30 refs., 12 figs., 3 tabs

The sleep of elite athletes at sea level and high altitude: a comparison of sea-level natives and high-altitude natives (ISA3600).

Science.gov (United States)

Roach, Gregory D; Schmidt, Walter F; Aughey, Robert J; Bourdon, Pitre C; Soria, Rudy; Claros, Jesus C Jimenez; Garvican-Lewis, Laura A; Buchheit, Martin; Simpson, Ben M; Hammond, Kristal; Kley, Marlen; Wachsmuth, Nadine; Gore, Christopher J; Sargent, Charli

2013-12-01

Altitude exposure causes acute sleep disruption in non-athletes, but little is known about its effects in elite athletes. The aim of this study was to examine the effects of altitude on two groups of elite athletes, that is, sea-level natives and high-altitude natives. Sea-level natives were members of the Australian under-17 soccer team (n=14). High-altitude natives were members of a Bolivian under-20 club team (n=12). Teams participated in an 18-day (19 nights) training camp in Bolivia, with 6 nights at near sea level in Santa Cruz (430 m) and 13 nights at high altitude in La Paz (3600 m). Sleep was assessed on every day/night using activity monitors. The Australians' sleep was shorter, and of poorer quality, on the first night at altitude compared with sea level. Sleep quality returned to normal by the end of the first week at altitude, but sleep quantity had still not stabilised at its normal level after 2 weeks. The quantity and quality of sleep obtained by the Bolivians was similar, or greater, on all nights at altitude compared with sea level. The Australians tended to obtain more sleep than the Bolivians at sea level and altitude, but the quality of the Bolivians' sleep tended to be better than that of the Australians at altitude. Exposure to high altitude causes acute and chronic disruption to the sleep of elite athletes who are sea-level natives, but it does not affect the sleep of elite athletes who are high-altitude natives.

The measurement for level of marine high-temperature and high-pressure vessels

International Nuclear Information System (INIS)

Lin Jie.

1986-01-01

The various error factors in measurement for level of marine high-temperature and high-pressure vessels are anslysed. The measuring method of error self compensation and its simplification for land use are shown

Time-dependent performance of soil mix technology stabilized/solidified contaminated site soils.

Science.gov (United States)

Wang, Fei; Wang, Hailing; Al-Tabbaa, Abir

2015-04-09

This paper presents the strength and leaching performance of stabilized/solidified organic and inorganic contaminated site soil as a function of time and the effectiveness of modified clays applied in this project. Field trials of deep soil mixing application of stabilization/solidification (S/S) were performed at a site in Castleford in 2011. A number of binders and addictives were applied in this project including Portland cement (PC), ground granulated blastfurnace slag (GGBS), pulverised fuel ash (PFA), MgO and modified clays. Field trial samples were subjected to unconfined compressive strength (UCS), BS CN 12457 batch leaching test and the extraction of total organics at 28 days and 1.5 years after treatment. The results of UCS test show that the average strength values of mixes increased from 0-3250 kPa at 28 days to 250-4250 kPa at 1.5 years curing time. The BS EN 12457 leachate concentrations of all metals were well below their drinking water standard, except Ni in some mixes exceed its drinking water standard at 0.02 mg/l, suggesting that due to varied nature of binders, not all of them have the same efficiency in treating contaminated soil. The average leachate concentrations of total organics were in the range of 20-160 mg/l at 28 days after treatment and reduced to 18-140 mg/l at 1.5 years. In addition, organo clay (OC)/inorgano-organo clay (IOC) slurries used in this field trial were found to have a negative effect on the strength development, but were very effective in immobilizing heavy metals. The study also illustrates that the surfactants used to modify bentonite in this field trail were not suitable for the major organic pollutants exist in the site soil in this project. Copyright © 2015 Elsevier B.V. All rights reserved.

Final repositories for high-level radioactive waste; Endlagerung hochradioaktiver Abfaelle

Energy Technology Data Exchange (ETDEWEB)

NONE

2015-10-15

The brochure on final repositories for high-level radioactive waste covers the following issues: What is the origin of radioactive wastes? How large are the waste amounts? What is going to happen with the wastes? What is the solution for the Waste disposal? A new site search is started. Safety requirements for the final disposal of high-level radioactive wastes. Comparison of host rocks. Who is responsible and who will pay? Final disposal of high-level radioactive wastes worldwide. Short summary: History of the search for a final repository for high-level radioactive wastes in Germany.

The high level vibration test program

International Nuclear Information System (INIS)

Hofmayer, C.H.; Curreri, J.R.; Park, Y.J.; Kato, W.Y.; Kawakami, S.

1989-01-01

As part of cooperative agreements between the US and Japan, tests have been performed on the seismic vibration table at the Tadotsu Engineering Laboratory of Nuclear Power Engineering Test Center (NUPEC) in Japan. The objective of the test program was to use the NUPEC vibration table to drive large diameter nuclear power piping to substantial plastic strain with an earthquake excitation and to compare the results with state-of-the-art analysis of the problem. The test model was subjected to a maximum acceleration well beyond what nuclear power plants are designed to withstand. A modified earthquake excitation was applied and the excitation level was increased carefully to minimize the cumulative fatigue damage due to the intermediate level excitations. Since the piping was pressurized, and the high level earthquake excitation was repeated several times, it was possible to investigate the effects of ratchetting and fatigue as well. Elastic and inelastic seismic response behavior of the test model was measured in a number of test runs with an increasing excitation input level up to the limit of the vibration table. In the maximum input condition, large dynamic plastic strains were obtained in the piping. Crack initiation was detected following the second maximum excitation run. Crack growth was carefully monitored during the next two additional maximum excitation runs. The final test resulted in a maximum crack depth of approximately 94% of the wall thickness. The HLVT (high level vibration test) program has enhanced understanding of the behavior of piping systems under severe earthquake loading. As in other tests to failure of piping components, it has demonstrated significant seismic margin in nuclear power plant piping

Discovery of high-level tasks in the operating room

NARCIS (Netherlands)

Bouarfa, L.; Jonker, P.P.; Dankelman, J.

2010-01-01

Recognizing and understanding surgical high-level tasks from sensor readings is important for surgical workflow analysis. Surgical high-level task recognition is also a challenging task in ubiquitous computing because of the inherent uncertainty of sensor data and the complexity of the operating

Spent fuel and high-level radioactive waste storage

International Nuclear Information System (INIS)

Trigerman, S.

1988-06-01

The subject of spent fuel and high-level radioactive waste storage, is bibliographically reviewed. The review shows that in the majority of the countries, spent fuels and high-level radioactive wastes are planned to be stored for tens of years. Sites for final disposal of high-level radioactive wastes have not yet been found. A first final disposal facility is expected to come into operation in the United States of America by the year 2010. Other final disposal facilities are expected to come into operation in Germany, Sweden, Switzerland and Japan by the year 2020. Meanwhile , stress is placed upon the 'dry storage' method which is carried out successfully in a number of countries (Britain and France). In the United States of America spent fuels are stored in water pools while the 'dry storage' method is still being investigated. (Author)

Development of radioactive waste management at Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Miyanaga, I.; Sakata, S.; Ito, A.; Amano, H.

1977-01-01

For low- and medium-level waste treatment, main efforts have been put on the reduction of waste volume. For high-level wastes, studies are being carried out on the solidification and partitioning techniques in preparation for completion of the fuel cycle in Japan. For sea disposal of low-level wastes planned by the JAEC, significant information has been obtained regarding integrity and leaching behavior of cement solidified wastes. This paper describes the present status of development of the techniques in the following sections; 1. Treatment of Low- and Medium-Level Wastes; an incinerator with two stage ceramic filters has been tested, and the decontamination factor was found to be 10 4 for various nuclides; reverse osmosis method with a cellulose acetate membrane has been tested for laundry liquid waste, and 60 Co was removed more than 99% together with detergents; and solidification products of spent ion-exchange resin with polyethylene have been proved to be superior in mechanical properties, water resistance and volume reduction to asphalt products. 2. Safety Evaluation of Cement Solidified Wastes for Sea Disposal; homogeneous cement-solidified wastes in 200 l sealed drums did not show any cracks or defects under high hydrostatic pressure; the leaching ratio of 137 Cs for the first one year was estimated to be lower than 0.3%. 3. Treatment of High-Level Wastes; vitrification using natural zeolite has been developed and properties of the products were proved to be excellent; and a partitioning procedure consisting mainly of solvent extraction and ion-exchange method has been studied; reduction of the amount of alkaline agent by introducing a denitration technique, and reduction of resin volume by adopting a porous type resin were achieved

Formation of equiaxed crystal structures in directionally solidified Al-Si alloys using Nb-based heterogeneous nuclei

Science.gov (United States)

Bolzoni, Leandro; Xia, Mingxu; Babu, Nadendla Hari

2016-01-01

The design of chemical compositions containing potent nuclei for the enhancement of heterogeneous nucleation in aluminium, especially cast alloys such as Al-Si alloys, is a matter of importance in order to achieve homogeneous properties in castings with complex geometries. We identified that Al3Nb/NbB2 compounds are effective heterogeneous nuclei and are successfully produced in the form of Al-2Nb-xB (x = 0.5, 1 and 2) master alloys. Our study shows that the inoculation of Al-10Si braze alloy with these compounds effectively promotes the heterogeneous nucleation of primary α-Al crystals and reduces the undercooling needed for solidification to take place. Moreover, we present evidences that these Nb-based compounds prevent the growth of columnar crystals and permit to obtain, for the first time, fine and equiaxed crystals in directionally solidified Al-10Si braze alloy. As a consequence of the potent heterogeneous particles, the size of the α-Al crystals was found to be less dependent on the processing conditions, especially the thermal gradient. Finally, we also demonstrate that the enhanced nucleation leads to the refinement of secondary phases such as eutectic silicon and primary silicon particles. PMID:28008967

Properties and characteristics of high-level waste glass

International Nuclear Information System (INIS)

Ross, W.A.

1977-01-01

This paper has briefly reviewed many of the characteristics and properties of high-level waste glasses. From this review, it can be noted that glass has many desirable properties for solidification of high-level wastes. The most important of these include: (1) its low leach rate; (2) the ability to tolerate large changes in waste composition; (3) the tolerance of anticipated storage temperatures; (4) its low surface area even after thermal shock or impact

Stable superconducting magnet. [high current levels below critical temperature

Science.gov (United States)

Boom, R. W. (Inventor)

1967-01-01

Operation of a superconducting magnet is considered. A method is described for; (1) obtaining a relatively high current in a superconducting magnet positioned in a bath of a gas refrigerant; (2) operating a superconducting magnet at a relatively high current level without training; and (3) operating a superconducting magnet containing a plurality of turns of a niobium zirconium wire at a relatively high current level without training.

ALICE High Level Trigger

CERN Multimedia

Alt, T

2013-01-01

The ALICE High Level Trigger (HLT) is a computing farm designed and build for the real-time, online processing of the raw data produced by the ALICE detectors. Events are fully reconstructed from the raw data, analyzed and compressed. The analysis summary together with the compressed data and a trigger decision is sent to the DAQ. In addition the reconstruction of the events allows for on-line monitoring of physical observables and this information is provided to the Data Quality Monitor (DQM). The HLT can process event rates of up to 2 kHz for proton-proton and 200 Hz for Pb-Pb central collisions.

The sleep of elite athletes at sea level and high altitude: a comparison of sea-level natives and high-altitude natives (ISA3600)

Science.gov (United States)

Roach, Gregory D; Schmidt, Walter F; Aughey, Robert J; Bourdon, Pitre C; Soria, Rudy; Claros, Jesus C Jimenez; Garvican-Lewis, Laura A; Buchheit, Martin; Simpson, Ben M; Hammond, Kristal; Kley, Marlen; Wachsmuth, Nadine; Gore, Christopher J; Sargent, Charli

2013-01-01

Background Altitude exposure causes acute sleep disruption in non-athletes, but little is known about its effects in elite athletes. The aim of this study was to examine the effects of altitude on two groups of elite athletes, that is, sea-level natives and high-altitude natives. Methods Sea-level natives were members of the Australian under-17 soccer team (n=14). High-altitude natives were members of a Bolivian under-20 club team (n=12). Teams participated in an 18-day (19 nights) training camp in Bolivia, with 6 nights at near sea level in Santa Cruz (430 m) and 13 nights at high altitude in La Paz (3600 m). Sleep was assessed on every day/night using activity monitors. Results The Australians’ sleep was shorter, and of poorer quality, on the first night at altitude compared with sea level. Sleep quality returned to normal by the end of the first week at altitude, but sleep quantity had still not stabilised at its normal level after 2 weeks. The quantity and quality of sleep obtained by the Bolivians was similar, or greater, on all nights at altitude compared with sea level. The Australians tended to obtain more sleep than the Bolivians at sea level and altitude, but the quality of the Bolivians’ sleep tended to be better than that of the Australians at altitude. Conclusions Exposure to high altitude causes acute and chronic disruption to the sleep of elite athletes who are sea-level natives, but it does not affect the sleep of elite athletes who are high-altitude natives. PMID:24282197

Analysis of Cyberbullying Sensitivity Levels of High School Students and Their Perceived Social Support Levels

Science.gov (United States)

Akturk, Ahmet Oguz

2015-01-01

Purpose: The purpose of this paper is to determine the cyberbullying sensitivity levels of high school students and their perceived social supports levels, and analyze the variables that predict cyberbullying sensitivity. In addition, whether cyberbullying sensitivity levels and social support levels differed according to gender was also…

Sodalite-type radioactive waste solidification product and method of synthesizing the same

International Nuclear Information System (INIS)

Koyama, Masashi; Yoshida, Takumasa.

1995-01-01

Radioactive waste solidification products formed by solidifying radioactive wastes comprising halides such as chlorides of alkali metal elements, alkaline earth metal elements, rare earth elements, noble metal elements generated upon dry-type reprocessing of nuclear fuels and separation of dry-type high level liquid wastes, are solidified to stable products by incorporating radioactive wastes in the form of halides into a cavity of sodalite condensation cage of aluminosilicates having three dimensional skeleton structure. Alternatively, NaOH, Al 2 O 3 , SiO 2 are mixed and heated to 600 to 900degC to form an intermediate reaction products, and then the reaction products are mixed with the halides and heated to form sodalite-type radioactive water solidification products. Thus, the halides in fission products can be held by the three dimensional skeleton structure similar with that of sodalite which is a sort of natural minerals containing chlorides, thereby enabling to solidify them stably. (N.H.)

Managing the nation's commercial high-level radioactive waste

International Nuclear Information System (INIS)

1985-03-01

This report presents the findings and conclusions of OTA's analysis of Federal policy for the management of commercial high-level radioactive waste. It represents a major update and expansion of the Analysis presented to Congress in our summary report, Managing Commercial High-Level Radioactive Waste, published in April of 1982 (NWPA). This new report is intended to contribute to the implementation of NWPA, and in particular to Congressional review of three major documents that DOE will submit to the 99th Congress: a Mission Plan for the waste management program; a monitored retrievable storage (MRS) proposal; and a report on mechanisms for financing and managing the waste program. The assessment was originally focused on the ocean disposal of nuclear waste. OTA later broadened the study to include all aspects of high-level waste disposal. The major findings of the original analysis were published in OTA's 1982 summary report

Primary Dendrite Arm Spacing and Trunk Diameter in Al-7-Weight-Percentage Si Alloy Directionally Solidified Aboard the International Space Station

Science.gov (United States)

Ghods, M.; Tewari, S. N.; Lauer, M.; Poirier, D. R.; Grugel, R. N.

2016-01-01

Under a NASA-ESA collaborative research project, three Al-7-weight-percentage Si samples (MICAST-6, MICAST-7 and MICAST 2-12) were directionally solidified aboard the International Space Station to determine the effect of mitigating convection on the primary dendrite array. The samples were approximately 25 centimeters in length with a diameter of 7.8 millimeter-diameter cylinders that were machined from [100] oriented terrestrially grown dendritic Al-7Si samples and inserted into alumina ampoules within the Sample Cartridge Assembly (SCA) inserts of the Low Gradient Furnace (LGF). The feed rods were partially remelted in space and directionally solidified to effect the [100] dendrite-orientation. MICAST-6 was grown at 5 microns per second for 3.75 centimeters and then at 50 microns per second for its remaining 11.2 centimeters of its length. MICAST-7 was grown at 20 microns per second for 8.5 centimeters and then at 10 microns per second for 9 centimeters of its remaining length. MICAST2-12 was grown at 40 microns per second for 11 centimeters. The thermal gradient at the liquidus temperature varied from 22 to 14 degrees Kelvin per centimeter during growth of MICAST-6, from 26 to 24 degrees Kelvin per centimeter for MICAST-7 and from 33 to 31 degrees Kelvin per centimeter for MICAST2-12. Microstructures on the transverse sections along the sample length were analyzed to determine nearest-neighbor spacing of the primary dendrite arms and trunk diameters of the primary dendrite-arrays. This was done along the lengths where steady-state growth prevailed and also during the transients associated with the speed-changes. The observed nearest-neighbor spacings during steady-state growth of the MICAST samples show a very good agreement with predictions from the Hunt-Lu primary spacing model for diffusion controlled growth. The observed primary dendrite trunk diameters during steady-state growth of these samples also agree with predictions from a coarsening-based model

High level cognitive information processing in neural networks

Science.gov (United States)

Barnden, John A.; Fields, Christopher A.

1992-01-01

Two related research efforts were addressed: (1) high-level connectionist cognitive modeling; and (2) local neural circuit modeling. The goals of the first effort were to develop connectionist models of high-level cognitive processes such as problem solving or natural language understanding, and to understand the computational requirements of such models. The goals of the second effort were to develop biologically-realistic model of local neural circuits, and to understand the computational behavior of such models. In keeping with the nature of NASA's Innovative Research Program, all the work conducted under the grant was highly innovative. For instance, the following ideas, all summarized, are contributions to the study of connectionist/neural networks: (1) the temporal-winner-take-all, relative-position encoding, and pattern-similarity association techniques; (2) the importation of logical combinators into connection; (3) the use of analogy-based reasoning as a bridge across the gap between the traditional symbolic paradigm and the connectionist paradigm; and (4) the application of connectionism to the domain of belief representation/reasoning. The work on local neural circuit modeling also departs significantly from the work of related researchers. In particular, its concentration on low-level neural phenomena that could support high-level cognitive processing is unusual within the area of biological local circuit modeling, and also serves to expand the horizons of the artificial neural net field.

Experimental evaluation of cement materials for solidifying sodium nitrate

International Nuclear Information System (INIS)

Sasaki, Tadashi; Numata, Mamoru; Suzuki, Yasuhiro; Kubo, Yoshikazu

2003-03-01

Low-level liquid waste containing sodium nitrate is planned to be transformed to salt block by evaporation with sodium borate in the Low-level Waste Treatment Facility (LWTF), then salt block will be stored temporally. It should be important to investigate the method how to treat these liquid waste suitable to final disposal criteria that will be settled in future. Cement solidification is one of promising candidates because it has been achieved as the solidification material for the shallow land disposal. The research was conducted to evaluate applicability of various cement materials to solidification of sodium nitrate. The following cements were tested. Ordinary Portland Cement (OPC). Portland Blast-furnace Slag Cement; C type (PBFSC). Alkali Activated Slag Cement (AASC, supplied by JGC). The test results are as follows; (1) AASC is characterized by a high sodium nitrate loading (-70 wt%) compared with other types of cement material. High fluidity of the cement paste, high strength after solidification, and minimization of free water on the cement paste are achieved under all test conditions. (2) OOPC and PBFSC produced free water on the cement paste in the early days and delayed the hardening period. 3 or more days are required to harden evan with 30 wt% content of sodium nitrate. (3) Though PBFSC contains blast furnace slag similar to AASC, there is no advantage prior to OPC. To design an ideal cement conditioning system for sodium nitrate liquid waste in the LWTF, the further studies are necessary such as the simulated waste test, Kd test, pilot test, and layout design. (author)

The disposal of low-level radioactive waste into the sea

International Nuclear Information System (INIS)

Saruhashi, Katsuko

1979-01-01

Disposal of low-level radioactive wastes is made both on land and in sea. Though the land disposal has been already carried out in the U.S.A. and the U.S.S.R., it is impossible in the narrow land of Japan. In the United States, the wastes solidified with cement in drums were previously abandoned in deep seas of the Pacific and the Atlantic. This is no longer done presently; instead, the land disposal is employed due to its lower costs. In European countries, the sea disposal is performed under OECDNEA, trial disposal in 1961 and full-scale disposal since 1967, in the Atlantic. Meanwhile, in Japan, test sea disposal will be carried out in the near future in deep sea of the northern Pacific, the important sea area for fisheries. The international trends of the deep sea disposal of low-level wastes and the correspondent trends of the same in Japan, in the past years are described. (J.P.N.)

Evaluation of conditioned high-level waste forms

International Nuclear Information System (INIS)

Mendel, J.E.; Turcotte, R.P.; Chikalla, T.D.; Hench, L.L.

1983-01-01

The evaluation of conditioned high-level waste forms requires an understanding of radiation and thermal effects, mechanical properties, volatility, and chemical durability. As a result of nuclear waste research and development programs in many countries, a good understanding of these factors is available for borosilicate glass containing high-level waste. The IAEA through its coordinated research program has contributed to this understanding. Methods used in the evaluation of conditioned high-level waste forms are reviewed. In the US, this evaluation has been facilitated by the definition of standard test methods by the Materials Characterization Center (MCC), which was established by the Department of Energy (DOE) in 1979. The DOE has also established a 20-member Materials Review Board to peer-review the activities of the MCC. In addition to comparing waste forms, testing must be done to evaluate the behavior of waste forms in geologic repositories. Such testing is complex; accelerated tests are required to predict expected behavior for thousands of years. The tests must be multicomponent tests to ensure that all potential interactions between waste form, canister/overpack and corrosion products, backfill, intruding ground water and the repository rock, are accounted for. An overview of the status of such multicomponent testing is presented

High-level waste management technology program plan

Energy Technology Data Exchange (ETDEWEB)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs.

High-level waste management technology program plan

International Nuclear Information System (INIS)

Harmon, H.D.

1995-01-01

The purpose of this plan is to document the integrated technology program plan for the Savannah River Site (SRS) High-Level Waste (HLW) Management System. The mission of the SRS HLW System is to receive and store SRS high-level wastes in a see and environmentally sound, and to convert these wastes into forms suitable for final disposal. These final disposal forms are borosilicate glass to be sent to the Federal Repository, Saltstone grout to be disposed of on site, and treated waste water to be released to the environment via a permitted outfall. Thus, the technology development activities described herein are those activities required to enable successful accomplishment of this mission. The technology program is based on specific needs of the SRS HLW System and organized following the systems engineering level 3 functions. Technology needs for each level 3 function are listed as reference, enhancements, and alternatives. Finally, FY-95 funding, deliverables, and schedules are s in Chapter IV with details on the specific tasks that are funded in FY-95 provided in Appendix A. The information in this report represents the vision of activities as defined at the beginning of the fiscal year. Depending on emergent issues, funding changes, and other factors, programs and milestones may be adjusted during the fiscal year. The FY-95 SRS HLW technology program strongly emphasizes startup support for the Defense Waste Processing Facility and In-Tank Precipitation. Closure of technical issues associated with these operations has been given highest priority. Consequently, efforts on longer term enhancements and alternatives are receiving minimal funding. However, High-Level Waste Management is committed to participation in the national Radioactive Waste Tank Remediation Technology Focus Area. 4 refs., 5 figs., 9 tabs

Department of Energy pretreatment of high-level and low-level wastes

International Nuclear Information System (INIS)

McGinnis, C.P.; Hunt, R.D.

1995-01-01

The remediation of the 1 x 10 8 gal of highly radioactive waste in the underground storage tanks (USTs) at five US Department of Energy (DOE) sites is one of DOE's greatest challenges. Therefore, the DOE Office of Environmental Management has created the Tank Focus Area (TFA) to manage an integrated technology development program that results in the safe and efficient remediation of UST waste. The TFA has divided its efforts into five areas, which are safety, characterization, retrieval/closure, pretreatment, and immobilization. All DOE pretreatment activities are integrated by the Pretreatment Technical Integration Manager of the TFA. For FY 1996, the 14 pretreatment tasks are divided into 3 systems: supernate separations, sludge treatment, and solid/liquid separation. The plans and recent results of these TFA tasks, which include two 25,000-gal demonstrations and two former TFA tasks on Cs removal, are presented. The pretreatment goals are to minimize the volume of high-level waste and the radioactivity in low-level waste

Low-level radioactive waste research program plan

International Nuclear Information System (INIS)

O'Donnell, E.; Lambert, J.

1989-11-01

The Waste Management Branch, Division of Engineering, Office of Nuclear Regulatory Research, has developed a strategy for conducting research on issues of concern to the US Nuclear Regulatory Commission (NRC) in its efforts to ensure safe disposal of low-level radioactive waste (LLW). The resulting LLW research program plan provides an integrated framework for planning the LLW research program to ensure that the program and its products are responsive and timely for use in NRC's LLW regulatory program. The plan discusses technical and scientific issues and uncertainties associated with the disposal of LLW, presents programmatic goals and objectives for resolving them, establishes a long-term strategy for conducting the confirmatory and investigative research needed to meet these goals and objectives, and includes schedules and milestones for completing the research. Areas identified for investigation include waste form and other material concerns, failure mechanisms and radionuclide releases, engineered barrier performance, site characterization and monitoring, and performance assessment. The plan proposes projects that (1) analyze and test actual LLW and solidified LLW under laboratory and field conditions to determine leach rates and radionuclide releases, (2) examine the short- and long-term performance of concrete-enhanced LLW burial structures and high-integrity containers, and (3) attempt to predict water movement and contaminant transport through low permeability saturated media and unsaturated porous media. 4 figs., 3 tabs

High level radioactive wastes: Considerations on final disposal

International Nuclear Information System (INIS)

Ciallella, Norberto R.

2000-01-01

When at the beginnings of the decade of the 80 the National Commission on Atomic Energy (CNEA) in Argentina decided to study the destination of the high level radioactive wastes, was began many investigations, analysis and multidisciplinary evaluations that be origin to a study of characteristics never before carried out in Argentina. For the first time in the country was faced the study of an environmental eventual problem, several decades before that the problem was presented. The elimination of the high level radioactive wastes in the technological aspects was taken in advance, avoiding to transfer the problems to the future generations. The decision was based, not only in technical evaluations but also in ethical premises, since it was considered that the future generations may enjoy the benefits of the nuclear energy and not should be solve the problem. The CNEA in Argentina in 1980 decided to begin a feasibility study and preliminary engineering project for the construction of the final disposal of high level radioactive wastes

Techniques for the solidification of high-level wastes

International Nuclear Information System (INIS)

1977-01-01

The problem of the long-term management of the high-level wastes from the reprocessing of irradiated nuclear fuel is receiving world-wide attention. While the majority of the waste solutions from the reprocessing of commercial fuels are currently being stored in stainless-steel tanks, increasing effort is being devoted to developing technology for the conversion of these wastes into solids. A number of full-scale solidification facilities are expected to come into operation in the next decade. The object of this report is to survey and compare all the work currently in progress on the techniques available for the solidification of high-level wastes. It will examine the high-level liquid wastes arising from the various processes currently under development or in operation, the advantages and disadvantages of each process for different types and quantities of waste solutions, the stages of development, the scale-up potential and flexibility of the processes

Transmission Level High Temperature Superconducting Fault Current Limiter

Energy Technology Data Exchange (ETDEWEB)

Stewart, Gary [SuperPower, Inc., Schenectady, NY (United States)

2016-10-05

The primary objective of this project was to demonstrate the feasibility and reliability of utilizing high-temperature superconducting (HTS) materials in a Transmission Level Superconducting Fault Current Limiter (SFCL) application. During the project, the type of high-temperature superconducting material used evolved from 1^st generation (1G) BSCCO-2212 melt cast bulk high-temperature superconductors to 2^nd generation (2G) YBCO-based high-temperature superconducting tape. The SFCL employed SuperPower's “Matrix” technology, that offers modular features to enable scale up to transmission voltage levels. The SFCL consists of individual modules that contain elements and parallel inductors that assist in carrying the current during the fault. A number of these modules are arranged in an m x n array to form the current-limiting matrix.

Materials for high-level waste containment

International Nuclear Information System (INIS)

Marsh, G.P.

1982-01-01

The function of the high-level radioactive waste container in storage and of a container/overpack combination in disposal is considered. The consequent properties required from potential fabrication materials are discussed. The strategy adopted in selecting containment materials and the experimental programme underway to evaluate them are described. (U.K.)

Method of controlling radioactive waste processing systems

International Nuclear Information System (INIS)

Mikawa, Hiroji; Sato, Takao.

1981-01-01

Purpose: To minimize the pellet production amount, maximize the working life of a solidifying device and maintaining the mechanical strength of pellets to a predetermined value irrespective of the type and the cycle of occurrence of the secondary waste in the secondary waste solidifying device for radioactive waste processing systems in nuclear power plants. Method: Forecasting periods for the type, production amount and radioactivity level of the secondary wastes are determined in input/output devices connected to a control system and resulted signals are sent to computing elements. The computing elements forecast the production amount of regenerated liquid wastes after predetermined days based on the running conditions of a condensate desalter and the production amounts of filter sludges and liquid resin wastes after predetermined days based on the liquid waste processing amount or the like in a processing device respectively. Then, the mass balance between the type and the amount of the secondary wastes presently stored in a tank are calculated and the composition and concentration for the processing liquid are set so as to obtain predetermined values for the strength of pellets that can be dried to solidify, the working life of the solidifying device itself and the radioactivity level of the pellets. Thereafter, the running conditions for the solidifying device are determined so as to maximize the working life of the solidifying device. (Horiuchi, T.)

High temperature creep properties of directionally solidified CM-247LC Ni-based superalloy

Energy Technology Data Exchange (ETDEWEB)

Chiou, Mau-Sheng [Department of Materials Science and Engineering, I-Shou University, Kaohsiung 840, Taiwan (China); Jian, Sheng-Rui, E-mail: srjian@gmail.com [Department of Materials Science and Engineering, I-Shou University, Kaohsiung 840, Taiwan (China); Yeh, An-Chou [Department of Materials Science and Engineering, National Tsing Hua University, Hsinchu 300, Taiwan (China); Kuo, Chen-Ming [Department of Mechanical and Automation Engineering, I-Shou University, Kaohsiung 840, Taiwan (China); Juang, Jenh-Yih [Department of Electrophysics, National Chiao Tung University, Hsinchu 300, Taiwan (China)

2016-02-08

This study explores the effects of cooling rate after solution heat treatment on the high temperature/low stress (982 °C/200 MPa) creep properties of CM-247LC Nickel base superalloy. Cooling rate was controlled by blowing argon gas, air cooling, and furnace cooling, which, in turn, gave rise to corresponding cooling rates (from 1260 °C to 800 °C) of 18.7, 7.4, and 0.19 °C/s, respectively. The results indicated that higher cooling rate from the solution heat treatment temperature led to finer γ′ precipitates and much improved tertiary creep as well as rupture life time in high-temperature creep test. The microstructural analyses using both scanning electron microscopy (SEM) and transmission electron microscopy (TEM) revealed that finer γ′ precipitates and narrower γ channel width could result in denser rafting structure which might have hindered the climb of dislocations across the precipitates rafts.

Psilocybin impairs high-level but not low-level motion perception.

Science.gov (United States)

Carter, Olivia L; Pettigrew, John D; Burr, David C; Alais, David; Hasler, Felix; Vollenweider, Franz X

2004-08-26

The hallucinogenic serotonin(1A&2A) agonist psilocybin is known for its ability to induce illusions of motion in otherwise stationary objects or textured surfaces. This study investigated the effect of psilocybin on local and global motion processing in nine human volunteers. Using a forced choice direction of motion discrimination task we show that psilocybin selectively impairs coherence sensitivity for random dot patterns, likely mediated by high-level global motion detectors, but not contrast sensitivity for drifting gratings, believed to be mediated by low-level detectors. These results are in line with those observed within schizophrenic populations and are discussed in respect to the proposition that psilocybin may provide a model to investigate clinical psychosis and the pharmacological underpinnings of visual perception in normal populations.

Scenarios of the TWRS low-level waste disposal program. Revision 1

International Nuclear Information System (INIS)

1995-01-01

As a result of past Department of Energy (DOE) weapons material production operations, Hanford now stores nuclear waste from processing facilities in underground tanks on the 200 area plateau. An agreement between the DOE, the Environmental Protection Agency (EPA), and the Washington state Department of Ecology (the Tri-Party Agreement, or TPA) establishes an enforceable schedule and a technical framework for recovering, processing, solidifying, and disposing of the Hanford tank wastes. The present plan includes retrieving the tank waste, pre-treating the waste to separate into low level and high level streams, and converting both streams to a glass waste form. The low level glass will represent by far the largest volume and lowest quantity of radioactivity (i.e., large volume of waste chemicals) of waste requiring disposal. The low level glass waste will be retrievably stored in sub-surface disposal vaults for several decades. If the low level disposal system proves to be acceptable, the disposal site will be closed with the low level waste in place. If, however, at some time the disposal system is found to be unacceptable, then the waste can be retrieved and dealt with in some other manner. WHC is planning to emplace the waste so that it is retrievable for up to 50 years after completion of the tank waste processing. Acceptability of disposal of the TWRS low level waste at Hanford depends on technical, cultural, and political considerations. The Performance Assessment is a major part of determining whether the proposed disposal action is technically defensible. A Performance Assessment estimates the possible future impact to humans and the environment for thousands of years into the future. In accordance with the TPA technical strategy, WHC plans to design a near-surface facility suitable for disposal of the glass waste

Cooling thermal parameters and microstructure features of directionally solidified ternary Sn–Bi–(Cu,Ag) solder alloys

Energy Technology Data Exchange (ETDEWEB)

Silva, Bismarck L., E-mail: bismarck_luiz@yahoo.com.br [Department of Materials Engineering, Federal University of São Carlos, UFSCar, 13565-905 São Carlos, SP (Brazil); Garcia, Amauri [Department of Manufacturing and Materials Engineering, University of Campinas, UNICAMP, 13083-860 Campinas, SP (Brazil); Spinelli, José E. [Department of Materials Engineering, Federal University of São Carlos, UFSCar, 13565-905 São Carlos, SP (Brazil)

2016-04-15

Low temperature soldering technology encompasses Sn–Bi based alloys as reference materials for joints since such alloys may be molten at temperatures less than 180 °C. Despite the relatively high strength of these alloys, segregation problems and low ductility are recognized as potential disadvantages. Thus, for low-temperature applications, Bi–Sn eutectic or near-eutectic compositions with or without additions of alloying elements are considered interesting possibilities. In this context, additions of third elements such as Cu and Ag may be an alternative in order to reach sounder solder joints. The length scale of the phases and their proportions are known to be the most important factors affecting the final wear, mechanical and corrosions properties of ternary Sn–Bi–(Cu,Ag) alloys. In spite of this promising outlook, studies emphasizing interrelations of microstructure features and solidification thermal parameters regarding these multicomponent alloys are rare in the literature. In the present investigation Sn–Bi–(Cu,Ag) alloys were directionally solidified (DS) under transient heat flow conditions. A complete characterization is performed including experimental cooling thermal parameters, segregation (XRF), optical and scanning electron microscopies, X-ray diffraction (XRD) and length scale of the microstructural phases. Experimental growth laws relating dendritic spacings to solidification thermal parameters have been proposed with emphasis on the effects of Ag and Cu. The theoretical predictions of the Rappaz-Boettinger model are shown to be slightly above the experimental scatter of secondary dendritic arm spacings for both ternary Sn–Bi–Cu and Sn–Bi–Ag alloys examined. - Highlights: • Dendritic growth prevailed for the ternary Sn–Bi–Cu and Sn–Bi–Ag solder alloys. • Bi precipitates within Sn-rich dendrites were shown to be unevenly distributed. • Morphology and preferential region for the Ag{sub 3}Sn growth depend on Ag

Approximate Bisimulation for High-Level Datapaths in Intelligent Transportation Systems

Directory of Open Access Journals (Sweden)

Hui Deng

2013-01-01

Full Text Available A relation called approximate bisimulation is proposed to achieve behavior and structure optimization for a type of high-level datapath whose data exchange processes are expressed by nonlinear polynomial systems. The high-level datapaths are divided into small blocks with a partitioning method and then represented by polynomial transition systems. A standardized form based on Ritt-Wu's method is developed to represent the equivalence relation for the high-level datapaths. Furthermore, we establish an approximate bisimulation relation within a controllable error range and express the approximation with an error control function, which is processed by Sostools. Meanwhile, the error is controlled through tuning the equivalence restrictions. An example of high-level datapaths demonstrates the efficiency of our method.

High-Level Waste (HLW) Feed Process Control Strategy

International Nuclear Information System (INIS)

STAEHR, T.W.

2000-01-01

The primary purpose of this document is to describe the overall process control strategy for monitoring and controlling the functions associated with the Phase 1B high-level waste feed delivery. This document provides the basis for process monitoring and control functions and requirements needed throughput the double-shell tank system during Phase 1 high-level waste feed delivery. This document is intended to be used by (1) the developers of the future Process Control Plan and (2) the developers of the monitoring and control system

Treatment technologies for non-high-level wastes (USA)

International Nuclear Information System (INIS)

Cooley, C.R.; Clark, D.E.

1976-06-01

Non-high-level waste arising from operations at nuclear reactors, fuel fabrication facilities, and reprocessing facilities can be treated using one of several technical alternatives prior to storage. Each alternative and the associated experience and status of development are summarized. The technology for treating non-high-level wastes is generally available for industrial use. Improved techniques applicable to the commercial nuclear fuel cycle are being developed and demonstrated to reduce the volume of waste and to immobilize it for storage. 36 figures, 59 references

Removing high-level contaminants

International Nuclear Information System (INIS)

Wallace, Paula

2013-01-01

Full text: Using biomimicry, an Australian cleantech innovation making inroads intoChinas's industrial sector offers multiple benefits to miners and processors in Australia. Stephen Shelley, the executive chairman of Creative Water Technology (CWT), was on hand at a recent trade show to explain how his Melbourne company has developed world-class techniques in zero liquid discharge and fractional crystallization of minerals to apply to a wide range of water treatment and recycling applications. “Most existing technologies operate with high energy distillation, filters or biological processing. CWT's appliance uses a low temperature, thermal distillation process known as adiabatic recovery to desalinate, dewater and/or recycle highly saline and highly contaminated waste water,” said Shelley. The technology has been specifically designed to handle the high levels of contaminant that alternative technologies struggle to process, with proven water quality results for feed water samples with TDS levels over 300,000ppm converted to clean water with less than 20ppm. Comparatively, reverse osmosis struggles to process contaminant levels over 70,000ppm effectively. “CWT is able to reclaim up to 97% clean usable water and up to 100% of the contaminants contained in the feed water,” said Shelley, adding that soluble and insoluble contaminants are separately extracted and dried for sale or re-use. In industrial applications CWT has successfully processed feed water with contaminant levels over 650,000 mg/1- without the use of chemicals. “The technology would be suitable for companies in oil exploration and production, mining, smelting, biofuels, textiles and the agricultural and food production sectors,” said Shelley. When compared to a conventional desalination plant, the CWT system is able to capture the value in the brine that most plants discard, not only from the salt but the additional water it contains. “If you recover those two commodities... then you

Elevated level of polysaccharides in a high level UV-B tolerant cell ...

African Journals Online (AJOL)

Jane

2011-04-26

Apr 26, 2011 ... A cell line of Bupleurum scorzonerifolium Willd with high level ... mechanisms to repair UV-induced damages via repairing ... for treatment or prevention of solar radiation. ..... working as both UV-B absorbing compounds and.