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Sample records for solidifying waste items

  1. 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.)

  2. 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.)

  3. 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.)

  4. 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.)

  5. 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.)

  6. 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.)

  7. 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.)

  8. 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.)

  9. 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.)

  10. 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.)

  11. 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.)

  12. 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)

  13. 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.)

  14. Method of solidifying radioactive wastes

    International Nuclear Information System (INIS)

    Tomita, Toshihide; Minami, Yuji; Matsuura, Hiroyuki; Kageyama, Hisashi; Kobori, Junzo.

    1984-01-01

    Purpose: To perform the curing sufficiently even when copper hydroxide that interferes the curing reaction is contained in radioactive wastes. Method: Solidification of radioactive wastes containing copper hydroxide using thermoset resins is carried out under the presence of an alkaline material. The thermoset resin used herein is an polyester resin comprising unsaturated polyester and a polymerizable monomer. The alkaline substance usable herein can include powder or an aqueous solution of hydroxides or oxides of sodium, magnesium, calcium or the like. (Yoshino, Y.)

  15. Method of solidifying radioactive wastes

    International Nuclear Information System (INIS)

    Tomita, Toshihide; Minami, Yuji; Matsuura, Hiroyuki

    1984-01-01

    Purpose: To enable complete curing even when radioactive wastes contain those materials hindering the curing reaction, for example, copper hydroxide. Method: After admixing an alkaline substance to radioactive concentrated liquid wastes containing copper hydroxide or other amphoteric substances, they are dried, powderized and then cured with thermosetting resins. The thermosetting resins usable herein include, for example, those prepared by mixing an unsaturated polyester with a monomer such as styrene. When a polymerization initiator such as methyl ethyl ketone peroxide and a polymerization promotor are added to the mixture, it takes places curing reaction at normal temperature. Suitable alkaline substances usable herein are those which are insoluble to the liquid wastes and do not change the chemical form under heating and drying. (Yoshihara, H.)

  16. 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.)

  17. 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.)

  18. 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.)

  19. 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.)

  20. 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

  1. 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)

  2. Biodegradation testing of solidified low-level waste streams

    International Nuclear Information System (INIS)

    Piciulo, P.L.; Shea, C.E.; Barletta, R.E.

    1985-05-01

    The NRC Technical Position on Waste Form (TP) specifies that waste should be resistant to biodegradation. The methods recommended in the TP for testing resistance to fungi, ASTM G21, and for testing resistance to bacteria, ASTM G22, were carried out on several types of solidified simulated wastes, and the effect of microbial activity on the mechanical strength of the materials tested was examined. The tests are believed to be sufficient for distinguishing between materials that are susceptible to biodegradation and those that are not. It is concluded that failure of these tests should not be regarded of itself as an indication that the waste form will biodegrade to an extent that the form does not meet the stability requirements of 10 CFR Part 61. In the case of failure of ASTM G21 or ASTM G22 or both, it is recommended that additional data be supplied by the waste generator to demonstrate the resistance of the waste form to microbial degradation. To produce a data base on the applicability of the biodegradation tests, the following simulated laboratory-scale waste forms were prepared and tested: boric acid and sodium sulfate evaporator bottoms, mixed-bed bead resins and powdered resins each solidified in asphalt, cement, and vinyl ester-styrene. Cement solidified wastes supported neither fungal nor bacterial growth. Of the asphalt solidified wastes, only the forms of boric acid evaporator bottoms did not support fungal growth. Bacteria grew on all of the asphalt solidified wastes. Cleaning the surface of these waste forms did not affect bacterial growth and had a limited effect on the fungal growth. Only vinyl esterstyrene solidified sodium sulfate evaporator bottoms showed viable fungi cultures, but surface cleaning with solvents eliminated fungal growth in subsequent testing. Some forms of all the waste streams solidified in vinyl ester-styrene showed viable bacteria cultures. 13 refs., 12 tabs

  3. Evaluation of solidified high-level waste forms

    International Nuclear Information System (INIS)

    1981-01-01

    One of the objectives of the IAEA waste management programme is to coordinate and promote development of improved technology for the safe management of radioactive wastes. The Agency accomplished this objective specifically through sponsoring Coordinated Research Programmes on the ''Evaluation of Solidified High Level Waste Products'' in 1977. The primary objectives of this programme are to review and disseminate information on the properties of solidified high-level waste forms, to provide a mechanism for analysis and comparison of results from different institutes, and to help coordinate future plans and actions. This report is a summary compilation of the key information disseminated at the second meeting of this programme

  4. 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.

  5. Characteristics of solidified high-level waste products

    International Nuclear Information System (INIS)

    1979-01-01

    The object of the report is to contribute to the establishment of a data bank for future preparation of codes of practice and standards for the management of high-level wastes. The work currently in progress on measuring the properties of solidified high-level wastes is being studied

  6. 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.)

  7. 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

  8. Site suitability criteria for solidified high level waste repositories

    International Nuclear Information System (INIS)

    Heckman, R.A.; Holdsworth, T.; Towse, D.F.

    1979-01-01

    Activities devoted to development of regulations, criteria, and standards for storage of solidified high-level radioactive wastes are reported. The work is summarized in sections on site suitability regulations, risk calculations, geological models, aquifer models, human usage model, climatology model, and repository characteristics. Proposed additional analytical work is also summarized

  9. 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

  10. 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)

  11. Production and properties of solidified high-level waste

    International Nuclear Information System (INIS)

    Brodersen, K.

    1980-08-01

    Available information on production and properties of solidified high-level waste are presented. The review includes literature up to the end of 1979. The feasibility of production of various types of solidified high-level wast is investigated. The main emphasis is on borosilicate glass but other options are also mentioned. The expected long-term behaviour of the materials are discussed on the basis of available results from laboratory experiments. Examples of the use of the information in safety analysis of disposal in salt formations are given. The work has been made on behalf of the Danish utilities investigation of the possibilities of disposal of high-level waste in salt domes in Jutland. (author)

  12. Method for solidifying liquid radioactive wastes

    International Nuclear Information System (INIS)

    Berreth, J.R.

    1976-01-01

    The quantity of nitrous oxides produced during the solidification of liquid radioactive wastes containing nitrates and nitrites can be substantially reduced by the addition to the wastes of a stoichiometric amount of urea which, upon heating, destroys the nitrates and nitrites, liberating nontoxic N 2 , CO 2 and NH 3 . 5 claims, no drawings

  13. 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.

  14. Leaching properties of solidified TRU waste forms

    International Nuclear Information System (INIS)

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

    1979-01-01

    Safety analysis of waste forms requires an estimate of the ability of these forms to retain activity in the disposal environment. This program of leaching tests will determine the leaching properties of TRU contaminated incinerator ash waste forms using hydraulic cement, urea--formaldehyde, bitumen, and vinyl ester--styrene as solidification agents. Three types of leaching tests will be conducted, including both static and flow rate. Five generic groundwaters will be used. Equipment and procedures are described. Experiments have been conducted to determine plate out of 239 Pu, counter efficiency, and stability of counting samples

  15. Method of solidifying liquid radioactive wastes

    International Nuclear Information System (INIS)

    Pekar, A.; Petrovic, J.; Timulak, J.

    1987-01-01

    Liquid radioactive waste containing boric acid salts is mixed with zeolite tuff and neutralized by lime. Power plant fly ash containing single-component or mixed Portland cement is then added to the mixture. Prior to packaging, anion-active bitumen emulsion or an aqueous emulsion of fatty acid salts and of free fatty acids insoluble in water can be added. Examples are given listing accurate proportions of the individual components. The advantage of the said solidification method is the use of easily available raw materials and improved values of extractability of the resulting product radionuclides. (E.S.)

  16. 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)

  17. 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.)

  18. 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

  19. 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

  20. 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.

  1. 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.)

  2. 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)

  3. Site suitability criteria for solidified high level waste repositories

    International Nuclear Information System (INIS)

    Heckman, R.A.; Holdsworth, T.; Isherwood, D.; Towse, D.F.; Dayem, N.L.

    1979-01-01

    The NRC is developing a framework of regulations, criteria, and standards. Lawrence Livermore Laboratory provides broad technical support to the NRC for developing this regulatory framework, part of which involves site suitability criteria for solidified high-level wastes (SHLW). Both the regulatory framework and the technical base on which it rests have evolved in time. This document is the second report of the technical support project. It was issued as a draft working paper for a programmatic review held at LLL from August 16 to 18, 1977. It was printed and distributed solely as a briefing document on preliminary methodology and initial findings for the purpose of critical review by those in attendance. These briefing documents are being reprinted now in their original formats as UCID-series reports for the sake of the historical record. Analysis results have evolved as both the models and data base have changed. As a result, the methodology, models, and data base in this document are severely outmoded

  4. 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)

  5. 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

  6. 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.)

  7. 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.)

  8. 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.)

  9. 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)

  10. 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

  11. 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

  12. IAEA coordinated research program on the evaluation of solidified high-level radioactive waste products

    International Nuclear Information System (INIS)

    Grover, J.R.; Schneider, K.J.

    1979-01-01

    A coordinated research program on the evaluation of solidified high-level radioactive waste products has been active with the IAEA since 1976. The program's objectives are to integrate research and to provide a data bank on an international basis in this subject area. Results and considerations to date are presented

  13. Performance criteria for solidified high-level radioactive wastes. Environmental impact statement. Revision 1

    International Nuclear Information System (INIS)

    1977-09-01

    This draft Environmental Impact Statement on performance criteria for solidified high-level radioactive wastes (PCSHLW) covers: considerations for PCSHLW development, the proposed rulemaking, characteristics of the PCSHLW, environmental impacts of the proposed PCSHLW, alternatives to the PCSHLW criteria, and cost/benefit/risk evaluation. Five appendices are included to support the technical data required in the Environmental Impact Statement

  14. 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

  15. Determination of performance criteria for high-level solidified nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Heckman, R.A.; Holdsworth, T.

    1979-05-07

    To minimize radiological risk from the operation of a waste management system, performance limits on volatilization, particulate dispersion, and dissolution characteristics of solidified high level waste must be specified. The results show clearly that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. Absolute values of expected risk are very sensitive to modeling assumptions. The transportation and interim storage operations appear to be most limiting in determining the performance characteristics required. The expected values of risk do not rely upon the repositories remaining intact over the potentially hazardous lifetime of the waste.

  16. Determination of performance criteria for high-level solidified nuclear waste

    International Nuclear Information System (INIS)

    Heckman, R.A.; Holdsworth, T.

    1979-01-01

    To minimize radiological risk from the operation of a waste management system, performance limits on volatilization, particulate dispersion, and dissolution characteristics of solidified high level waste must be specified. The results show clearly that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. Absolute values of expected risk are very sensitive to modeling assumptions. The transportation and interim storage operations appear to be most limiting in determining the performance characteristics required. The expected values of risk do not rely upon the repositories remaining intact over the potentially hazardous lifetime of the waste

  17. 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

  18. Process for solidifying high-level nuclear waste

    Science.gov (United States)

    Ross, Wayne A.

    1978-01-01

    The addition of a small amount of reducing agent to a mixture of a high-level radioactive waste calcine and glass frit before the mixture is melted will produce a more homogeneous glass which is leach-resistant and suitable for long-term storage of high-level radioactive waste products.

  19. Chemical characterization, leach, and adsorption studies of solidified low-level wastes

    International Nuclear Information System (INIS)

    Walter, M.B.; Serne, R.J.; Jones, T.L.; McLaurine, S.B.

    1986-12-01

    Laboratory and field leaching experiments are beig conducted by Pacific Northwest Laboratory (PNL) to investigate the performance of solidified low-level nuclear waste in a typical, arid, near-surface disposal site. Under PNL's Special Waste Form Lysimeters-Arid Program, a field test facility was constructed to monitor the leaching of commercial solidified waste. Laboratory experiments were conducted to investigate the leaching and adsorption characteristics of the waste forms in contact with soil. Liquid radioactive wastes solidified in cement, vinyl ester-styrene, and bitumen were obtained from commercial boiling water and pressurized water reactors, and buried in a field leaching facility on the Hanford site in southeastern Washington State. Batch leaching, soil column adsorption, and soil/waste form column experiments were conducted in the laboratory, using small-scale cement waste forms and Hanford site ground water. The purpose of these experiments is to evaluate the ability of laboratory leaching tests to predict leaching under actual field conditions and to determine which mechanisms (i.e., diffusion, solubility, adsorption) actually control the concentration of radionuclides in the soil surrounding the waste form. Chemical and radionuclide analyses performed on samples collected from the field and laboratory experiments indicate strong adsorption of /sup 134,137/Cs and 85 Sr onto the Hanford site sediment. Small amounts of 60 Co are leached from the waste forms as very mobile species. Some 60 Co migrated through the soil at the same rate as water. Chemical constituents present in the reactor waste streams also found at elevated levels in the field and laboratory leachates include sodium, sulfate, magnesium, and nitrate. Plausible solid phases that could be controlling some of the chemical and radionuclide concentrations in the leachate were identified using the MINTEQ geochemical computer code

  20. 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

  1. Development and characterization of solidified forms for high-level wastes: 1978. Annual report

    International Nuclear Information System (INIS)

    Ross, W.A.; Mendel, J.E.

    1979-12-01

    Development and characterization of solidified high-level waste forms are directed at determining both process properties and long-term behaviors of various solidified high-level waste forms in aqueous, thermal, and radiation environments. Waste glass properties measured as a function of composition were melt viscosity, melt electrical conductivity, devitrification, and chemical durability. The alkali metals were found to have the greatest effect upon glass properties. Titanium caused a slight decrease in viscosity and a significant increase in chemical durability in acidic solutions (pH-4). Aluminum, nickel and iron were all found to increase the formation of nickel-ferrite spinel crystals in the glass. Four multibarrier advanced waste forms were produced on a one-liter scale with simulated waste and characterized. Glass marbles encapsulated in a vacuum-cast lead alloy provided improved inertness with a minimal increase in technological complexity. Supercalcine spheres exhibited excellent inertness when coated with pyrolytic carbon and alumina and put in a metal matrix, but the processing requirements are quite complex. Tests on simulated and actual high-level waste glasses continue to suggest that thermal devitrification has a relatively small effect upon mechanical and chemical durabilities. Tests on the effects radiation has upon waste forms also continue to show changes to be relatively insignificant. Effects caused by decay of actinides can be estimated to saturate at near 10 19 alpha-events/cm 3 in homogeneous solids. Actually, in solidified waste forms the effects are usually observed around certain crystals as radiation causes amorphization and swelling of th crystals

  2. Functions and requirements document for interim store solidified high-level and transuranic waste

    Energy Technology Data Exchange (ETDEWEB)

    Smith-Fewell, M.A., Westinghouse Hanford

    1996-05-17

    The functions, requirements, interfaces, and architectures contained within the Functions and Requirements (F{ampersand}R) Document are based on the information currently contained within the TWRS Functions and Requirements database. The database also documents the set of technically defensible functions and requirements associated with the solidified waste interim storage mission.The F{ampersand}R Document provides a snapshot in time of the technical baseline for the project. The F{ampersand}R document is the product of functional analysis, requirements allocation and architectural structure definition. The technical baseline described in this document is traceable to the TWRS function 4.2.4.1, Interim Store Solidified Waste, and its related requirements, architecture, and interfaces.

  3. Embedding methods of solidified waste in metal matrices

    International Nuclear Information System (INIS)

    Neumann, W.

    1979-01-01

    The embedding of simulated waste calcines by three different methods (vacuum-pressure casting, centrifugal casting, and metal stirred with the calcines) was investigated. The experimental performance is described and advantages and disadvantages noted. The feasibility of embedding fines by stirring in metal was shown. In addition, an estimation of the influence of porosity on the properties of composites was carried out

  4. Safety analysis of sea transportation of solidified reactor wastes

    International Nuclear Information System (INIS)

    Devell, L.; Edlund, O.; Kjellbert, N.; Grundfelt, B.; Milchert, T.

    1980-06-01

    A central handling and storage facility (ALMA) for low- and medium-level reactor waste from Swedish nuclear power plants is being planned and the transportation to it will be by sea. A safety assessment devoted to the potential environmental impacts from the transportation is presented. (Auth.)

  5. 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

  6. 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)

  7. 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.

  8. 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

  9. 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

  10. Retrievable surface storage: interim storage of solidified high-level waste

    International Nuclear Information System (INIS)

    LaRiviere, J.R.; Nelson, D.C.

    1976-01-01

    Studies have been conducted on retrievable-surface-storage concepts for the interim storage of solidified high-level wastes. These studies have been reviewed by the Panel on Engineered Storage, convened by the Committee on Radioactive Waste Management of the National Research Council-National Academy of Sciences. The Panel has concluded that ''retrievable surface storage is an acceptable interim stage in a comprehensive system for managing high-level radioactive wastes.'' The scaled storage cask concept, which was recommended by the Panel on Engineered Storage, consists of placing a canister of waste inside a carbon-steel cask, which in turn is placed inside a thick concrete cylinder. The waste is cooled by natural convection air flow through an annulus between the cask and the inner wall of the concrete cylinder. The complete assembly is placed above ground in an outdoor storage area

  11. 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

  12. 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

  13. 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

  14. 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

  15. 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

  16. 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

  17. 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)

  18. 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)

  19. Review of metal-matrix encapsulation of solidified radioactive high-level waste

    International Nuclear Information System (INIS)

    Jardine, L.J.; Steindler, M.J.

    1978-05-01

    Literature describing previous and current work on the encapsulation of solidified high-level waste forms in a metal matrix was reviewed. Encapsulation of either stabilized calcine pellets or glass beads in alloys by casting techniques was concluded to be the most developed and direct approach to fabricating solid metal-matrix waste forms. Further characterizations of the physical and chemical properties of metal-matrix waste forms are still needed to assess the net attributes of metal-encapsulation alternatives. Steady-state heat transfer properties of waste canisters in air and water environments were calculated for four reference waste forms: (1) calcine, (2) glass monoliths, (3) metal-encapsulated calcine, and (4) metal-encapsulated glass beads. A set of criteria for the maximum allowable canister centerline and surface temperatures and heat generation rates per canister at the time of shipment to a Federal repository was assumed, and comparisons were made between canisters of these reference waste forms of the shortest time after reactor discharge that canisters could be filled and the subsequent ''interim'' storage times prior to shipment to a Federal repository for various canister diameters and waste ages. A reference conceptual flowsheet based on existing or developing technology for encapsulation of stabilized calcine pellets is discussed. Conclusions and recommendations are presented

  20. 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.

  1. 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

  2. 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

  3. 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

  4. Testing and evaluation of solidified high-level waste forms. Joint annual progress report 1983

    International Nuclear Information System (INIS)

    Malow, G.

    1985-01-01

    A second joint programme of the European Atomic Community was started in 1981 under the indirect action programme (1980-84), Action No 5 'Testing and evaluation of the properties of various potential materials for immobilizing high activity waste'. The overall objective of the research is to test various European potential solidified high-level radioactive waste forms so as to predict their behaviour after disposal. The most important aspect is to produce data to calculate the activity release from the waste products under the attack of various aqueous solutions. The experiments were partly performed under waste repository relevant conditions and partly under simplified conditions for investigating basic activity release mechanisms. The topics of the programme were: (i) studies of basic leaching mechanisms; (ii) studies of hydrothermal leaching and surface attack of waste glasses; (iii) leach test carried out in contact with granite at low water flow rates; (iv) static leach tests with specimen surrounded by canister and backfill materials; (v) specific isotope leach tests in slowly flowing water; (vi) leach test of actinide spiked samples; (vii) leach tests of highly radioactive samples; (viii) leach tests of alpha radiation stability; (ix) studies of mechanical stability; (x) studies of mineral phases as model compounds and phase relations

  5. 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

  6. 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

  7. 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

  8. 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

  9. SWEPP PAN assay system uncertainty analysis: Active mode measurements of solidified aqueous sludge waste

    International Nuclear Information System (INIS)

    Blackwood, L.G.; Harker, Y.D.; Meachum, T.R.

    1997-12-01

    The Idaho National Engineering and Environmental Laboratory is being used as a temporary storage facility for transuranic waste generated by the US Nuclear Weapons program at the Rocky Flats Plant (RFP) in Golden, Colorado. Currently, there is a large effort in progress to prepare to ship this waste to the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. In order to meet the TRU Waste Characterization Quality Assurance Program Plan nondestructive assay compliance requirements and quality assurance objectives, it is necessary to determine the total uncertainty of the radioassay results produced by the Stored Waste Examination Pilot Plant (SWEPP) Passive Active Neutron (PAN) radioassay system. This paper is one of a series of reports quantifying the results of the uncertainty analysis of the PAN system measurements for specific waste types and measurement modes. In particular this report covers active mode measurements of weapons grade plutonium-contaminated aqueous sludge waste contained in 208 liter drums (item description codes 1, 2, 7, 800, 803, and 807). Results of the uncertainty analysis for PAN active mode measurements of aqueous sludge indicate that a bias correction multiplier of 1.55 should be applied to the PAN aqueous sludge measurements. With the bias correction, the uncertainty bounds on the expected bias are 0 ± 27%. These bounds meet the Quality Assurance Program Plan requirements for radioassay systems

  10. Properties of backfilling material for solidifying miscellaneous waste using recycled cement from waste concrete

    International Nuclear Information System (INIS)

    Matsuda, Atsuo; Yamamoto, Kazuo; Konishi, Masao; Iwamoto, Yoshiaki; Yoshikane, Toru; Koie, Toshio; Nakashima, Yoshio.

    1997-01-01

    A large reduction of total radioactive waste is expected, if recycled cement from the waste concrete of decommissioned nuclear power plants would be able to be used the material for backfilling mortar among the miscellaneous waste. In this paper, we discuss the hydration, strength and consistency of recycled cement compared with normal portland cement. The strength of recycled cement mortar is lower than that of normal portland cement mortar on the same water to cement ratio. It is possible to obtain the required strength to reduce the water to cement ratio by using of high range water-reducing AE agent. According to reducing of water to cement ratio, the P-type funnel time of mortar increase with the increase of its viscosity. However, in new method of self-compactability for backfilling mortar, it became evident that there was no difference between the recycled cement and normal portland cement on the self-compactability. (author)

  11. Assessing radioactive concentrates and waste vapor condensate in solidifying radioactive wastes by bituminization

    International Nuclear Information System (INIS)

    Tibensky, L.; Krejci, F.; Breza, M.; Timulak, J.; Hladky, E.

    1986-01-01

    A brief overview is presented of chemical and radiochemical methods used in the world for the analysis of the concentrate of liquid radioactive wastes from nuclear power plants destined for bituminization. Most methods are also suitable for an analysis of the condensate of waste vapors produced in bituminization. The methods of analysis of the radioactive concentrate from the V-1 nuclear power plant in Jaslovske Bohunice and of the waste vapors condensate were developed and tested in practice. Gross gamma activity was measured using a well-type Na(Tl) scintillation detector, the content of radionuclides was determined using semiconductor Ge(Li) spectrometry. The concentration of boric acid in the concentrate was determined by titration with mannite; in the condensate, using spectrophotometry with curcumine. The content of nitrates in both the concentrate and the condensate was determined spectrophotometrically using salicylic acid, the content of nitrites was determined by spectrophotometry using sulfanilic acid and α-naphthylamine. Carbonates and chlorides were determined by titration, sodium and potassium by flame photometry. The content of organic acids was measured by gravimetry of extracted methyl esters, the content of surfactants by spectrophotometry. Infrared spectrophotometry was used in determining hydrocarbons in the waste vapor condensate. The measured value range and the measurement errors are shown for each method. (A.K.)

  12. 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

  13. 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.)

  14. Long-term reactive transport modelling of stabilized/solidified waste: from dynamic leaching tests to disposal scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Windt, Laurent de [Ecole des Mines de Paris, CG-Hydrodynamics and Reaction Group, 35 R. St-Honore, 77300 Fontainebleau (France)]. E-mail: laurent.dewindt@ensmp.fr; Badreddine, Rabia [INERIS, Direction des Risques Chroniques, Unite Dechets et Sites Pollues, Parc Technologique Alata BP 2, 60550 Verneuil-en-Halatte (France); Lagneau, Vincent [Ecole des Mines de Paris, CG-Hydrodynamics and Reaction Group, 35 R. St-Honore, 77300 Fontainebleau (France)

    2007-01-31

    Environmental impact assessment of hazardous waste disposal relies, among others, on standardized leaching tests characterized by a strong coupling between diffusion and chemical processes. In that respect, this study shows that reactive transport modelling is a useful tool to extrapolate laboratory results to site conditions characterized by lower solution/solid (L/S) ratios, site specific geometry, infiltration, etc. A cement solidified/stabilized (S/S) waste containing lead is investigated as a typical example. The reactive transport model developed in a previous study to simulate the initial state of the waste as well as laboratory batch and dynamic tests is first summarized. Using the same numerical code (HYTEC), this model is then integrated to a simplified waste disposal scenario assuming a defective cover and rain water infiltration. The coupled evolution of the S/S waste chemistry and the pollutant plume migration are modelled assessing the importance of the cracking state of the monolithic waste. The studied configurations correspond to an undamaged and fully sealed system, a few main fractures between undamaged monoliths and, finally, a dense crack-network in the monoliths. The model considers the potential effects of cracking, first the increase of rain water and carbon dioxide infiltration and, secondly, the increase of L/S ratio and reactive surfaces, using either explicit fracture representation or dual porosity approaches.

  15. 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)

  16. Acceptance issues for large items and difficult waste

    International Nuclear Information System (INIS)

    Palmer, J.; Lock, Peter

    2002-01-01

    Peter Lock described some particular cases which had given rise to difficult acceptance issues at NIREX, ranging from large size items to the impacts of chemicals used during decontamination on the mobility of radionuclides in a disposal facility: The UK strategy for intermediate level and certain low level radioactive waste disposal is based on production of cementitious waste-forms packaged in a standard range of containers as follows: 500 litre Drum - the normal container for most operational ILW (0.8 m diameter x 1.2 m high); 3 m"3 Box - a larger container for solid wastes (1.72 m x 1.72 m plan x 1.2 m high); 3 m"3 Drum - a larger container for in-drum mixing and immobilisation of sludge waste-forms (1.72 m diameter x 1.2 m high); 4 m Box - for large items of waste, especially from decommissioning (4.0 m x 2.4 m plan x 2.2 m high); 2 m LLW Box - for higher-density wastes (2.0 m x 2.4 m plan x 2.2 m high). In addition the majority of LLW is packaged by supercompaction followed by grouting in modified ISO freight containers (6 m x 2.5 m x 2.5 m). Some wastes do not fit easily into this strategy. These wastes include: very large items, (too big for the 4 m box) which, if dealt with whole, pose transport and disposal problems. These items are discussed further in Section 2; waste whose characteristics make packaging difficult. Such wastes are described in more detail in Section 3

  17. 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.)

  18. Determination of performance criteria for high-level solidified nuclear waste from the commercial nuclear fuel cycle: a probabilistic safety analysis

    International Nuclear Information System (INIS)

    Heckman, R.A.

    1978-01-01

    To minimize the radiological risk from the operation of a waste management system for the safe disposal of high-level waste, performance characteristics of the solidified waste form must be specified. The minimum waste form characteristics that must be specified are the radionuclide volatilization fraction, airborne particulate dispersion fraction, and the aqueous dissolution characteristics. The results indicate that the pre-emplacement environs are more limiting in establishing the waste form performance criteria than the post-emplacement environs. The actual values of expected risk are sensitive to modeling assumptions and data base uncertainties. The transportation step appears to be the most limiting in determining the required performance characteristics

  19. 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)

  20. 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

  1. 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

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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.

  7. Radionuclide distributions in sediments of marine areas used for dumping solidified radioactive wastes

    International Nuclear Information System (INIS)

    Bowen, V.T.; Livingston, H.D.

    A number of sediment samples, collected both by coring and by grabbing, from the shallow Pacific solid waste radioactive dump site and from the Atlantic dump site, have been analyzed carefully for a number of long-lived radionuclides. Both dump sites yielded samples that were expected to serve as controls, collected at considerable distance from any visually-located waste containers, as well as other samples that were collected close to identified waste drums, some of which showed evidence of physical disintegration. The Atlantic site shows evidence of wide-spread, general contamination, with 137 Cs and possibly with 241 Am. The Pacific site is perhaps less generally contaminated with 137 Cs, but shows evidence of widespread general contamination with several transuranic nuclides. Samples collected near to identified waste containers, at both sites, show that significant portions of leached radioactivity ( 137 Cs, 239 240 Pu, 238 Pu, 241 Am, 242 Cm, and 244 Cm) are immobilized by the sediments within very short distances, possibly measured in meters or tens of meters. The data also suggest considerable differences among the horizontal trajectories of the various leached transuranic elements. It is argued that careful study of nuclide distributions around such old waste containers would provide data of great value in helping to predict long-term behavior of radionuclides released to marine environments

  8. Performance Demonstration Program Plan for RCRA Constituent Analysis of Solidified Wastes

    International Nuclear Information System (INIS)

    2006-01-01

    The Performance Demonstration Program (PDP) for Resource Conservation and Recovery Act (RCRA) constituents distributes test samples for analysis of volatile organic compounds (VOCs), semivolatile organic compounds (SVOCs), and metals in solid matrices. Each distribution of test samples is termed a PDP cycle. These evaluation cycles provide an objective measure of the reliability of measurements performed for transuranic (TRU) waste characterization. The primary documents governing the conduct of the PDP are the Quality Assurance Program Document (QAPD; DOE/CBFO-94-1012) and the Waste Isolation Pilot Plant (WIPP) Waste Analysis Plan (WAP) contained in the Hazardous Waste Facility Permit (NM4890139088-TSDF) issued by the New Mexico Environment Department. The WAP requires participation in the PDP; the PDP must comply with the QAPD and the WAP. This plan implements the general requirements of the QAPD and the applicable requirements of the WAP for the RCRA PDP. Participating laboratories demonstrate acceptable performance by successfully analyzing single-blind performance evaluation samples (subsequently referred to as PDP samples) according to the criteria established in this plan. PDP samples are used as an independent means to assess laboratory performance regarding compliance with the WAP quality assurance objectives (QAOs). The concentrations of analytes in the PDP samples address levels of regulatory concern and encompass the range of concentrations anticipated in waste characterization samples. The WIPP requires analyses of homogeneous solid wastes to demonstrate compliance with regulatory requirements. These analyses must be performed by laboratories that demonstrate acceptable performance in this PDP. These analyses are referred to as WIPP analyses, and the samples on which they are performed are referred to as WIPP samples. Participating laboratories must analyze PDP samples using the same procedures used for WIPP samples.

  9. Methodology of environmental evaluation of wastes stabilized/solidified by hydraulic binders; Methodologie d'evaluation environnementale des dechets stabilises / solidifies par liants hydrauliques

    Energy Technology Data Exchange (ETDEWEB)

    Imyim, A.

    2000-12-15

    The aim of this work is the formalization of a methodology of evaluation of the leaching behaviour of massive porous materials obtained by stabilization/solidification of wastes. In a first part, a set of simple leaching tests is proposed which allow the physico-chemical characterization of materials. In order to better understand the phenomena involved in the release process, the methodology has been applied to hydraulic binder-based and lead-bearing synthesized materials. In a second step, a mathematical model has been proposed for the description of the leaching behaviour. The development of the model is based on the observations and experimental results obtained with the synthesized materials. Finally, the methodology of evaluation of the leaching behaviour has been applied to two cases of real wastes: the fly ashes of a Danish municipal waste incineration facility, and the galvanic sludges from an industrial waste water processing facility from Netherlands. (J.S.)

  10. 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

  11. Waste treatment process by solidifying cementitious materials using hydrothermal hot-pressing

    International Nuclear Information System (INIS)

    Matsumoto, Y.; Kamakura, T.; Yamasaki, N.; Hashida, T.

    2001-01-01

    Solidification of low-level radioactive wastes containing Na 2 SO 4 with cement by hydrothermal hot-pressing (HHP) technique was examined. Relatively high mechanical strength, reduced leaching ratio of SO 3 , and higher resistance to the carbonation of the HHP product were attained in comparison with conventional concrete. The solidification by the HHP treatment may be proceeded by the rearrangement of particles and the bonding material formation among the particles by dissolution-deposition process. The possibility of developing the accelerated testing method for duration of cemented materials by hydrothermal method was discussed. (author)

  12. Electric melting furnace of solidifying radioactive waste by utilizing magnetic field and melting method

    International Nuclear Information System (INIS)

    Igarashi, Hiroshi.

    1990-01-01

    An electric melting furnace for solidification of radioactive wastes utilizing magnetic fields in accordance with the present invention comprises a plurality of electrodes supplying AC current to molten glass in a glass melting furnace and a plurality of magnetic poles for generating AC magnetic fields. Interactions between the current and the magnetic field, generated forces in the identical direction in view of time in the molten glass. That is, forces for promoting the flow of molten glass in the melting furnace are resulted due to the Fleming's left-hand rule. As a result, the following effects can be obtained. (1) The amount of heat ransferred from the molten glass to the starting material layer on the molten surface is increased to improve the melting performance. (2) For an identical melting performance, the size and the weight of the melting furnace can be reduced to decrease the amount of secondary wastes when the apparatus-life is exhausted. (3) Bottom deposits can be suppressed and prevented from settling and depositing to the reactor bottom by the promoted flow in the layer. (4) Further, the size of auxiliary electrodes for directly supplying electric current to heat the molten glass near the reactor bottom can be decreased. (I.S.)

  13. Influence of fracture networks on radionuclide transport from solidified waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Seetharam, S.C., E-mail: suresh.seetharam@sckcen.be [Performance Assessments Unit, Institute for Environment, Health and Safety, Belgian Nuclear Research Centre (SCK CEN), Boeretang 200, B-2400 Mol (Belgium); Perko, J.; Jacques, D. [Performance Assessments Unit, Institute for Environment, Health and Safety, Belgian Nuclear Research Centre (SCK CEN), Boeretang 200, B-2400 Mol (Belgium); Mallants, D. [CSIRO Land and Water, Waite Road – Gate 4, Glen Osmond, SA 5064 (Australia)

    2014-04-01

    Highlights: • Magnitude of peak radionuclide fluxes is less sensitive to the fracture network geometry. • Time of peak radionuclide fluxes is sensitive to the fracture networks. • Uniform flow model mimics a limiting case of a porous medium with large number of fine fractures. • Effect of fracture width on radionuclide flux depends on the ratio of fracture to matrix conductivity. • Effect of increased dispersivity in fractured media does not always result in a lower peak flux for specific fracture networks due to higher concentrations adjacent to the fracture plane. - Abstract: Analysis of the effect of fractures in porous media on fluid flow and mass transport is of great interest in many fields including geotechnical, petroleum, hydrogeology and waste management. This paper presents sensitivity analyses examining the effect of various hypothetical fracture networks on the performance of a planned near surface disposal facility in terms of radionuclide transport behaviour. As it is impossible to predict the initiation and evolution of fracture networks and their characteristics in concrete structures over time scales of interest, several fracture networks have been postulated to test the sensitivity of radionuclide release from a disposal facility. Fluid flow through concrete matrix and fracture networks are modelled via Darcy's law. A single species radionuclide transport equation is employed for both matrix and fracture networks, which include the processes advection, diffusion, dispersion, sorption/desorption and radioactive decay. The sensitivity study evaluates variations in fracture network configuration and fracture width together with different sorption/desorption characteristics of radionuclides in a cement matrix, radioactive decay constants and matrix dispersivity. The effect of the fractures is illustrated via radionuclide breakthrough curves, magnitude and time of peak mass flux, cumulative mass flux and concentration profiles. For the

  14. Influence of fracture networks on radionuclide transport from solidified waste forms

    International Nuclear Information System (INIS)

    Seetharam, S.C.; Perko, J.; Jacques, D.; Mallants, D.

    2014-01-01

    Highlights: • Magnitude of peak radionuclide fluxes is less sensitive to the fracture network geometry. • Time of peak radionuclide fluxes is sensitive to the fracture networks. • Uniform flow model mimics a limiting case of a porous medium with large number of fine fractures. • Effect of fracture width on radionuclide flux depends on the ratio of fracture to matrix conductivity. • Effect of increased dispersivity in fractured media does not always result in a lower peak flux for specific fracture networks due to higher concentrations adjacent to the fracture plane. - Abstract: Analysis of the effect of fractures in porous media on fluid flow and mass transport is of great interest in many fields including geotechnical, petroleum, hydrogeology and waste management. This paper presents sensitivity analyses examining the effect of various hypothetical fracture networks on the performance of a planned near surface disposal facility in terms of radionuclide transport behaviour. As it is impossible to predict the initiation and evolution of fracture networks and their characteristics in concrete structures over time scales of interest, several fracture networks have been postulated to test the sensitivity of radionuclide release from a disposal facility. Fluid flow through concrete matrix and fracture networks are modelled via Darcy's law. A single species radionuclide transport equation is employed for both matrix and fracture networks, which include the processes advection, diffusion, dispersion, sorption/desorption and radioactive decay. The sensitivity study evaluates variations in fracture network configuration and fracture width together with different sorption/desorption characteristics of radionuclides in a cement matrix, radioactive decay constants and matrix dispersivity. The effect of the fractures is illustrated via radionuclide breakthrough curves, magnitude and time of peak mass flux, cumulative mass flux and concentration profiles. For the

  15. Standard test method for accelerated leach test for diffusive releases from solidified waste and a computer program to model diffusive, fractional leaching from cylindrical waste forms

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 This test method provides procedures for measuring the leach rates of elements from a solidified matrix material, determining if the releases are controlled by mass diffusion, computing values of diffusion constants based on models, and verifying projected long-term diffusive releases. This test method is applicable to any material that does not degrade or deform during the test. 1.1.1 If mass diffusion is the dominant step in the leaching mechanism, then the results of this test can be used to calculate diffusion coefficients using mathematical diffusion models. A computer program developed for that purpose is available as a companion to this test method (Note 1). 1.1.2 It should be verified that leaching is controlled by diffusion by a means other than analysis of the leach test solution data. Analysis of concentration profiles of species of interest near the surface of the solid waste form after the test is recommended for this purpose. 1.1.3 Potential effects of partitioning on the test results can...

  16. 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)

  17. 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

  18. 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

  19. Identification of items and activities important to waste form acceptance by Westinghouse GoCo sites

    International Nuclear Information System (INIS)

    Plodinec, M.J.; Marra, S.L.; Dempster, J.; Randklev, E.H.

    1993-01-01

    The Department of Energy has established specifications (Waste Acceptance Product Specifications for Vitrified High-Level Waste Forms, or WAPS) for canistered waste forms produced at Hanford, Savannah River, and West Valley. Compliance with these specifications requires that each waste form producer identify the items and activities which must be controlled to ensure compliance. As part of quality assurance oversight activities, reviewers have tried to compare the methodologies used by the waste form producers to identify items and activities important to waste form acceptance. Due to the lack of a documented comparison of the methods used by each producer, confusion has resulted over whether the methods being used are consistent. This confusion has been exacerbated by different systems of nomenclature used by each producer, and the different stages of development of each project. The waste form producers have met three times in the last two years, most recently on June 28, 1993, to exchange information on each producer's program. These meetings have been sponsored by the Westinghouse GoCo HLW Vitrification Committee. This document is the result of this most recent exchange. It fills the need for a documented comparison of the methodologies used to identify items and activities important to waste form acceptance. In this document, the methodology being used by each waste form producer is summarized, and the degree of consistency among the waste form producers is determined

  20. Comparative study of test methods for bituminized and other low- and medium-level solidified waste materials

    International Nuclear Information System (INIS)

    Brodersen, K.; Mose Pedersen, B.; Vinther, A.

    1983-12-01

    Various aspects of the behaviour of bituminized or cemented simulated low- or medium-level radioactive waste in contact with water or salt solutions have been investigated. The solubility (approximately 0.5%) and the diffusion coefficient (approximately 5.10 -8 cm 2 /sec) determining transort of water in pure bitumen have been measured for Mexphalte 40/50 at room temperature. A weighing method has been used to study water uptake and swelling of bituminized sodium nitrate, sodium sulphate or cation-exchange resin. The swelling of samples in contact with water was in some cases very pronounced. In strong salt solutions the tendency to swell is much less. The particle size of the embedded waste material is an important parameter. Thermal pre-treatment of cation-exchange resin before bituminization does not seem to improve the quality of the final product. The interaction between bituminized-exchange resin and concrete barrier materials has been studied. Microbial degradation of bitumen and bituminized waste under aerobic conditions has been investigated. It is probably of minor importance as far as leaching is concerned. A method for measuring leaching from a plane surface of cemented waste has been developed. The method avoids the problem of cracks between the sample and the container. Leaching from cemented sodium nitrate or sulphate was investigated. Absorption of CO 2 from the atmosphere was found to have only minor effect on Cs- and Na-leaching but gave a pronounced decrease in Ca-leaching. The use of silica-fume as an additive to cemented sodium nitrate decreased the leach rate by a factor 4. (author)

  1. Cost analysis for application of solidified waste fission product canisters in U.S. Army steam plants

    International Nuclear Information System (INIS)

    Sande, W.E.; Bjorklund, W.J.; Brooks, N.A.

    1977-04-01

    The main objectives of the present study are to design steam plants using projected waste fission product canister characteristics, to analyze the overall impact and cost/benefit to the nuclear fuel cycle associated with these plants, and to develop plans for this application if the cost analysis so warrants it. The construction and operation of a steam plant fueled with waste fission product canisters would require the involvement and cooperation of various government agencies and private industry; thus the philosophies of these groups were studied. These philosophies are discussed, followed by a forecast of canister supply, canister characteristics, and strategies for Army canister use. Another section describes the safety and licensing of these steam plants since this affects design and capital costs. The discussion of steam plant design includes boiler concepts, boiler heat transfer, canister temperature distributions, steam plant size, and steam plant operation. Also, canister transportation is discussed since this influences operating costs. Details of economics of Army steam plants are provided including steam plant capital costs, operating costs, fuel reprocessor savings due to Army canister storage, and overall economics. Recommendations are made in the final section

  2. 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.

  3. 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.

  4. Radiation exposure estimates on production and utilization of recycled items using dismantling waste

    International Nuclear Information System (INIS)

    Nakamura, Hisashi; Nakashima, Mikio

    2002-03-01

    Radiation exposure was estimated on production and utilization of recycled items using dismantling wastes by assuming that their usage are restricted to nuclear facilities. The radiation exposure attributed to production of a steel-plate cast iron waste container, a receptacle for slag, and a drum reinforcement was calculated to be in the range of several μSv to several tens of μSv even in recycling contaminated metal waste of which radioactivity concentration of Co-60 is higher than the clearance level by a factor of two figures. It is also elucidated that casting of a multiple casting waste package meets the standards of dose equivalent rate for the transport of a radioactive package and the weight of the package will be able to kept around 20 tons for the convenience of the handling, in case of disposal of metal waste less than 37 MBq/g with the steel-plate cast iron waste container. As the results, from the radiological exposure's point of view, it should be possible to use slightly contaminated metal for recycled items in waste management. (author)

  5. Modelling informally collected quantities of bulky waste and reusable items in Austria

    International Nuclear Information System (INIS)

    Ramusch, R.; Pertl, A.; Scherhaufer, S.; Schmied, E.; Obersteiner, G.

    2015-01-01

    Highlights: • Informal collectors from Hungary collect bulky waste and reusable items in Austria. • Two methodologies were applied to estimate the informally collected quantities. • Both approaches lead to an estimation of roughly 100,000 t p.a. informally collected. • The formal Austrian system collects 72 kg/cap/yr of bulky waste, WEE & scrap metal. • Informal collection amounts to approx. 12 kg/cap/yr. - Abstract: Disparities in earnings between Western and Eastern European countries are the reason for a well-established informal sector actively involved in collection and transboundary shipment activities from Austria to Hungary. The preferred objects are reusable items and wastes within the categories bulky waste, WEEE and metals, intended to be sold on flea markets. Despite leading to a loss of recyclable resources for Austrian waste management, these informal activities may contribute to the extension of the lifetime of certain goods when they are reused in Hungary; nevertheless they are discussed rather controversially. The aim of this paper is to provide objective data on the quantities informally collected and transhipped. The unique activities of informal collectors required the development and implementation of a new set of methodologies. The concept of triangulation was used to verify results obtained by field visits, interviews and a traffic counting campaign. Both approaches lead to an estimation of approx. 100,000 t per year of reusable items informally collected in Austria. This means that in addition to the approx. 72 kg/cap/yr formally collected bulky waste, bulky waste wood, household scrap (excluding packaging) and WEEE, up to a further 12 kg/cap/yr might, in the case that informal collection is abandoned, end up as waste or in the second-hand sector

  6. Modelling informally collected quantities of bulky waste and reusable items in Austria

    Energy Technology Data Exchange (ETDEWEB)

    Ramusch, R., E-mail: roland.ramusch@boku.ac.at; Pertl, A.; Scherhaufer, S.; Schmied, E.; Obersteiner, G.

    2015-10-15

    Highlights: • Informal collectors from Hungary collect bulky waste and reusable items in Austria. • Two methodologies were applied to estimate the informally collected quantities. • Both approaches lead to an estimation of roughly 100,000 t p.a. informally collected. • The formal Austrian system collects 72 kg/cap/yr of bulky waste, WEE & scrap metal. • Informal collection amounts to approx. 12 kg/cap/yr. - Abstract: Disparities in earnings between Western and Eastern European countries are the reason for a well-established informal sector actively involved in collection and transboundary shipment activities from Austria to Hungary. The preferred objects are reusable items and wastes within the categories bulky waste, WEEE and metals, intended to be sold on flea markets. Despite leading to a loss of recyclable resources for Austrian waste management, these informal activities may contribute to the extension of the lifetime of certain goods when they are reused in Hungary; nevertheless they are discussed rather controversially. The aim of this paper is to provide objective data on the quantities informally collected and transhipped. The unique activities of informal collectors required the development and implementation of a new set of methodologies. The concept of triangulation was used to verify results obtained by field visits, interviews and a traffic counting campaign. Both approaches lead to an estimation of approx. 100,000 t per year of reusable items informally collected in Austria. This means that in addition to the approx. 72 kg/cap/yr formally collected bulky waste, bulky waste wood, household scrap (excluding packaging) and WEEE, up to a further 12 kg/cap/yr might, in the case that informal collection is abandoned, end up as waste or in the second-hand sector.

  7. Modelling informally collected quantities of bulky waste and reusable items in Austria.

    Science.gov (United States)

    Ramusch, R; Pertl, A; Scherhaufer, S; Schmied, E; Obersteiner, G

    2015-10-01

    Disparities in earnings between Western and Eastern European countries are the reason for a well-established informal sector actively involved in collection and transboundary shipment activities from Austria to Hungary. The preferred objects are reusable items and wastes within the categories bulky waste, WEEE and metals, intended to be sold on flea markets. Despite leading to a loss of recyclable resources for Austrian waste management, these informal activities may contribute to the extension of the lifetime of certain goods when they are reused in Hungary; nevertheless they are discussed rather controversially. The aim of this paper is to provide objective data on the quantities informally collected and transhipped. The unique activities of informal collectors required the development and implementation of a new set of methodologies. The concept of triangulation was used to verify results obtained by field visits, interviews and a traffic counting campaign. Both approaches lead to an estimation of approx. 100,000 t per year of reusable items informally collected in Austria. This means that in addition to the approx. 72 kg/cap/yr formally collected bulky waste, bulky waste wood, household scrap (excluding packaging) and WEEE, up to a further 12 kg/cap/yr might, in the case that informal collection is abandoned, end up as waste or in the second-hand sector. Copyright © 2015 Elsevier Ltd. All rights reserved.

  8. Waste management of Line Item projects at Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    Zill, D.S.

    1993-01-01

    With the growing number of companies involved with today's Line Item projects at the Oak Ridge National Laboratory (ORNL), there are ever increasing problems in the handling of Radioactive Solid Low-Level Waste (SLLW). The most important of these problems is who is going to do what with the waste and when are they going to do it. The who brings to mind training; the what, compliance; and the when, cost. At ORNL, the authors have found that the best way to address the challenges of waste handling where several contractors are involved is through communication, compromise and consistency. Without these elements, opportunities bred from waste handling are likely to bring the project to a halt

  9. Critical Protection Item classification for a waste processing facility at Savannah River Site

    International Nuclear Information System (INIS)

    Ades, M.J.; Garrett, R.J.

    1993-01-01

    This paper describes the methodology for Critical Protection Item (CPI) classification and its application to the Structures, Systems and Components (SSC) of a waste processing facility at the Savannah River Site (SRS). The WSRC methodology for CPI classification includes the evaluation of the radiological and non-radiological consequences resulting from postulated accidents at the waste processing facility and comparison of these consequences with allowable limits. The types of accidents considered include explosions and fire in the facility and postulated accidents due to natural phenomena, including earthquakes, tornadoes, and high velocity straight winds. The radiological analysis results indicate that CPIs are not required at the waste processing facility to mitigate the consequences of radiological release. The non-radiological analysis, however, shows that the Waste Storage Tank (WST) and the dike spill containment structures around the formic acid tanks in the cold chemical feed area and waste treatment area of the facility should be identified as CPIs. Accident mitigation options are provided and discussed

  10. 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

  11. Large Item Disposal At The Drigg Low Level Waste Repository, United Kingdom

    International Nuclear Information System (INIS)

    Griffiths, Steve

    2012-01-01

    Currently the UK operates only one repository for low level radioactive waste, the LLWR near Drigg in Cumbria. It is located on the West Cumbrian coast near the village of Drigg. LLWR is designed for the management of solid LLW and has operated as the principal national disposal facility for LLW since 1959. LLWR is managed and operated on behalf of the Nuclear Decommissioning Authority (NDA) by UK Nuclear Waste Management Ltd. (UKNWM), parent body of LLW Repository Ltd. UKNWM is a consortium led by URS, Studsvik and AREVA. Waste is accepted at LLWR based on conditions for acceptance (1). Although there is some history of disposal of non-containerised 'large items' at the Drigg site these are anecdotally described as 'not quite fitting into an ISO container (2)' and enquiries indicate that their disposal was restricted to the legacy times when items were tumble-tipped into open trenches at the site, a practise now long ceased. The feasibility of true single large item disposal at the LLWR presents complex problems arising from the poor suitability of both rail and road infrastructure in UK. LLWR is serviced both by road and rail links. The static weight of large items being taken nominally as up to ∼300 metric tons would not necessarily preclude transportation by rail but the practicalities of this route are limited. The ageing rail infrastructure includes numerous tunnels, bridges and sections of line with overhead electrification. All these would require either careful justification or significant work to ensure the safe transit of large loads. Nuclear facilities in UK are by design in remote locations, not all of which are serviced by rail connections and the rail network itself has evolved to service inter-city transportation rather than heavy freight and as such tends to route through town centres, exacerbating the tunnel, bridge and pantograph concerns already identified. Within only a few miles of the LLWR itself there are requirements to pass both over and

  12. Initial Q-list for the prospective Yucca Mountain repository based on items important to safety and waste isolation

    International Nuclear Information System (INIS)

    Laub, T.W.; Jardine, L.J.

    1987-01-01

    A method for identifying items important to safety based on a probabilistic risk assessment approach was developed and implemented for the conceptual design of the Yucca Mountain repository. No items were classified as important to safety; however, six items were classified as potentially important to safety. These were the shipping cask, the cranes and the truck or rail-care vehicle stops in the cask receiving and preparation area, the hot cell structure of the waste packaging hot cells, the cranes in the waste packaging hot cells, and the waste-handling building fire protection system. In addition, a method for identifying items important to waste isolation was developed and implemented. Two hydrogeologic units of the Yucca Mountain site were classified as important to waste isolation: the Calico Hills nonwelded zeolitic unit and the Calico Hills nonwelded vitric unit. The preliminary Q-list for the Yucca Mountain repository is comprised of the two units of the site classified as important to waste isolation and contains no items important to safety

  13. Initial Q-list for the prospective Yucca Mountain repository based on items important to safety and waste isolation

    International Nuclear Information System (INIS)

    Laub, T.W.; Jardine, L.J.

    1987-01-01

    A method for identifying items important to safety based on a probabilistic risk assessment approach was developed and implemented for the conceptual design of the Yucca Mountain repository. No items were classified as important to safety; however, six items were classified as potentially important to safety. These were the shipping cask, the cranes and the truck or rail-car vehicle stops in the cask receiving and preparation area, the hot cell structure of the waste packaging hot cells, the cranes in the waste packaging hot cells, and the waste-handling building fire protection system. In addition, a method for identifying items important to waste isolation was developed and implemented. Two hydrogeologic units of the Yucca Mountain site were classified as important to waste isolation: the Calico Hills nonwelded zeolitic unit and the Calico Hills nonwelded vitric unit. The preliminary Q-list for the Yucca Mountain repository is comprised of the two units of the site classified as important to waste isolation and contains no items important to safety

  14. Development of the Open Items Tracking System

    International Nuclear Information System (INIS)

    Riggi, V.

    1994-01-01

    The West Valley Demonstration Project, located on the site of the only commercial nuclear fuel reprocessing facility to have operated in USA, has the directed objectives of solidifying the high-level radioactive waste into a durable, solid form for shipment; decontaminating and decommissioning the tanks and facilities; and disposing of the resulting low-level and transuranic wastes. Since an escalating trend of open work items was noticed in the Fall of 1988, and there was no control mechanism for tracking and closing the open items, a Work Control System was developed for this purpose. It is self-contained system on a mainframe ARTEMIS 9000, which tracks, monitors, and closes out external commitments in a timely manner. Audits, surveillances, site appraisals, preventive maintenance, instrument calibration recall, and scheduling are covered

  15. Influence of hydrologic factors on leaching of solidified low-level waste forms at an arid site field-scale lysimeter facility

    International Nuclear Information System (INIS)

    Jones, T.L.; Skaggs, R.L.

    1987-04-01

    Most of the precipitation at the Hanford Site arrives as winter snow; this contributes to a strong seasonal pattern in water storage and drainage observed in the lysimeters. This seasonal pattern in storage corresponds to an annual range in the volumetric soil water content of 11% in late winter to 7% in the late summer and early fall. Annual changes in drainage rates cause pore water velocities to vary annually by nearly two orders of magnitude. Rapid snowmelt and frozen soils in February 1985 caused runoff water from areas adjacent to the lysimeter facility to flood three of the lysimeters. This resulted in a temporary increase in soil water storage, and an additional 5 to 10 cm of drainage for these three lysimeters. Measurable quantities of tritium and cobalt-60 are being collected in lysimeter drainage water. Approximately 30% of the original tritium inventory has been leached from both lysimeters containing samples of this waste form. Cobalt-60 is consistently being leached from five lysimeters representing three of the five waste forms. Total cobalt-60 collected from each of the five lysimeters varies, but in each case is less than 0.1% of the original cobalt inventory of the waste sample. Comparisons of cobalt release among flooded and non-flooded lysimeters show no significant difference caused by the extra drainage

  16. Investigation of the leaching behavior of lead in stabilized/solidified waste using a two-year semi-dynamic leaching test.

    Science.gov (United States)

    Xue, Qiang; Wang, Ping; Li, Jiang-Shan; Zhang, Ting-Ting; Wang, Shan-Yong

    2017-01-01

    Long-term leaching behavior of contaminant from stabilization/solidification (S/S) treated waste stays unclear. For the purpose of studying long-term leaching behavior and leaching mechanism of lead from cement stabilized soil under different pH environment, semi-dynamic leaching test was extended to two years to investigate leaching behaviors of S/S treated lead contaminated soil. Effectiveness of S/S treatment in different scenarios was evaluated by leachability index (LX) and effective diffusion coefficient (D e ). In addition, the long-term leaching mechanism was investigated at different leaching periods. Results showed that no significant difference was observed among the values of the cumulative release of Pb, D e and LX in weakly alkaline and weakly acidic environment (pH value varied from 5.00 to 10.00), and all the controlling leaching mechanisms of the samples immersed in weakly alkaline and weakly acidic environments turned out to be diffusion. Strong acid environment would significantly affect the leaching behavior and leaching mechanism of lead from S/S monolith. The two-year variation of D e appeared to be time dependent, and D e values increased after the 210 th day in weakly alkaline and weakly acidic environment. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. Data sharing report characterization of the surveillance and maintenance project miscellaneous process inventory waste items Oak Ridge National Laboratory, Oak Ridge, TN

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Phyllis C. [Oak Ridge Inst. for Science and Education (ORISE), Oak Ridge, TN (United States)

    2013-12-12

    The U.S. Department of Energy (DOE) Oak Ridge Office of Environmental Management (EM-OR) requested Oak Ridge Associated Universities (ORAU), working under the Oak Ridge Institute for Science and Education (ORISE) contract, to provide technical and independent waste management planning support under the American Recovery and Reinvestment Act (ARRA). Specifically, DOE EM-OR requested ORAU to plan and implement a sampling and analysis campaign to target certain items associated with URS|CH2M Oak Ridge, LLC (UCOR) surveillance and maintenance (S&M) process inventory waste. Eight populations of historical and reoccurring S&M waste at the Oak Ridge National Laboratory (ORNL) have been identified in the Waste Handling Plan for Surveillance and Maintenance Activities at the Oak Ridge National Laboratory, DOE/OR/01-2565&D2 (WHP) (DOE 2012) for evaluation and processing for final disposal. This waste was generated during processing, surveillance, and maintenance activities associated with the facilities identified in the process knowledge (PK) provided in Appendix A. A list of items for sampling and analysis were generated from a subset of materials identified in the WHP populations (POPs) 4, 5, 6, 7, and 8, plus a small number of items not explicitly addressed by the WHP. Specifically, UCOR S&M project personnel identified 62 miscellaneous waste items that would require some level of evaluation to identify the appropriate pathway for disposal. These items are highly diverse, relative to origin; composition; physical description; contamination level; data requirements; and the presumed treatment, storage, and disposal facility (TSDF). Because of this diversity, ORAU developed a structured approach to address item-specific data requirements necessary for acceptance in a presumed TSDF that includes the Environmental Management Waste Management Facility (EMWMF)—using the approved Waste Lot (WL) 108.1 profile—the Y-12 Sanitary Landfill (SLF) if appropriate; Energy

  18. GAMMA-PULSE-HEIGHT EVALUATION OF A USA SAVANNAH RIVER SITE BURIAL GROUND SPECIAL CONFIGURATION WASTE ITEM

    Energy Technology Data Exchange (ETDEWEB)

    Dewberry, R.; Sigg, R.; Salaymeh, S.

    2009-03-23

    The Savannah River Site (SRS) Burial Ground had a container labeled as Box 33 for which they had no reliable solid waste stream designation. The container consisted of an outer box of dimensions 48-inch x 46-inch x 66-inch and an inner box that contained high density and high radiation dose material. From the outer box Radiation Control measured an extremity dose rate of 22 mrem/h. With the lid removed from the outer box, the maximum dose rate measured from the inner box was 100 mrem/h extremity and 80 mrem/h whole body. From the outer box the material was sufficiently high in density that the Solid Waste Management operators were unable to obtain a Co-60 radiograph of the contents. Solid Waste Management requested that the Analytical Development Section of Savannah River National Laboratory perform a {gamma}-ray assay of the item to evaluate the radioactive content and possibly to designate a solid waste stream. This paper contains the results of three models used to analyze the measured {gamma}-ray data acquired in an unusual configuration.

  19. GAMMA-PULSE-HEIGHT EVALUATION OF A USA SAVANNAH RIVER SITE BURIAL GROUND SPECIAL CONFIGURATION WASTE ITEM

    International Nuclear Information System (INIS)

    Dewberry, R.; Sigg, R.; Salaymeh, S.

    2009-01-01

    The Savannah River Site (SRS) Burial Ground had a container labeled as Box 33 for which they had no reliable solid waste stream designation. The container consisted of an outer box of dimensions 48-inch x 46-inch x 66-inch and an inner box that contained high density and high radiation dose material. From the outer box Radiation Control measured an extremity dose rate of 22 mrem/h. With the lid removed from the outer box, the maximum dose rate measured from the inner box was 100 mrem/h extremity and 80 mrem/h whole body. From the outer box the material was sufficiently high in density that the Solid Waste Management operators were unable to obtain a Co-60 radiograph of the contents. Solid Waste Management requested that the Analytical Development Section of Savannah River National Laboratory perform a γ-ray assay of the item to evaluate the radioactive content and possibly to designate a solid waste stream. This paper contains the results of three models used to analyze the measured γ-ray data acquired in an unusual configuration

  20. Technical position on items and activities in the high-level waste geologic repository program subject to quality assurance requirements

    International Nuclear Information System (INIS)

    Duncan, A.B.; Bilhorn, S.G.; Kennedy, J.E.

    1988-04-01

    This document provides guidance on how to identify items and activities subject to Quality Assurance in the high-level nuclear waste repository program for pre-closure and post-closure phases of the repository. In the pre-closure phase, structures, systems and components essential to the prevention or mitigation of an accident that could result in an off-site radiation dose of 0.5rem or greater are termed ''important to safety''. In the post-closure phase, the barriers which are relied on to meet the containment and isolation requirements are defined as ''important to waste isolation''. These structures, systems, components, and barriers, and the activities related to their characterization, design, construction, and operation are required to meet quality assurance (QA) criteria to provide confidence in the performance of the geologic repository. The list of structures, systems, and components important to safety and engineered barriers important to waste isolation is referred to as the ''Q-List'' and lies within the scope of the QA program. 10 refs

  1. 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)

  2. 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.

  3. 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.)

  4. 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.

  5. Plastic solidification system for radioactive waste

    International Nuclear Information System (INIS)

    Kani, Jiro; Irie, Hiromitsu; Obu, Etsuji; Nakayama, Yasuyuki; Matsuura, Hiroyuki.

    1979-01-01

    The establishment of a new solidification system is an important theme for recent radioactive-waste disposal systems. The conditions required of new systems are: (1) the volume of the solidified product to be reduced, and (2) the property of the solidified product to be superior to the conventional ones. In the plastic solidification system developed by Toshiba, the waste is first dried and then solidified with thermosetting resin. It has been confirmed that the property of the plastic solidified product is superior to that of the cement-or bitumen-solidified product. Investigation from various phases is being carried on for the application of this method to commercial plants. (author)

  6. Properties of radioactive wastes and waste containers

    International Nuclear Information System (INIS)

    Morcos, N.; Dayal, R.

    1982-01-01

    This program is sponsored by the Nuclear Regulatory Commission to address basic concerns in assessing the performance of solidified radwaste. Experiments were initiated to address these concerns. In particular, leachability of solidified radwastes and the physical stability of the ensuing waste forms were evaluated. In addition, leaching experiments designed to address the effects of alternating wet/dry cycles and of varying the length of these cycles on the leach behavior of waste forms were initiated

  7. 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

  8. 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

  9. Method to increase the safety of a final storage site in a salt cavern filled with solidified radioactive waste with regard to unforeseen rock movements and/or water ingress into cavities of the final storage site

    International Nuclear Information System (INIS)

    Koester, R.; Rudolph, G.; Kroebel, R.

    1986-01-01

    The wastes of weak or average radio-activity (e.g. T) are stored in barrels in a salt mine. In order to prevent leaching of the waste after the ingress of water into the salt mine, the intermediate spaces between the barrels are filled with a concrete grout. This grout consists of a water/bentonite/cement mixture, to which sand may be added, and which hardens. It forms a monolithic block. (orig./PW)

  10. Method to increase the safety of a final storage site in a salt cavern filled with solidified radioactive waste with regard to unforeseen rock movements and/or water ingress into cavities of the final storage site

    International Nuclear Information System (INIS)

    Koester, R.; Rudolph, G.; Kroebel, R.

    1980-01-01

    The wastes of weak or average radio-activity (e.g. T) are stored in barrels in a salt mine. In order to prevent leaching of the waste after the ingress of water into the salt mine, the intermediate spaces between the barrels are filled with a concrete grout. This grout consists of a water/bentonite/cement mixture, to which sand may be added, and which hardens. It forms a monolithic block. (DG) [de

  11. Glass and nuclear wastes

    International Nuclear Information System (INIS)

    Sombret, C.

    1982-10-01

    Glass shows interesting technical and economical properties for long term storage of solidified radioactive wastes by vitrification or embedding. Glass composition, vitrification processes, stability under irradiation, thermal stability and aqueous corrosion are studied [fr

  12. 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.)

  13. 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.

  14. 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

  15. Method of storing radioactive wastes

    International Nuclear Information System (INIS)

    Adachi, Toshio; Hiratake, Susumu.

    1980-01-01

    Purpose: To reduce the radiation doses externally irradiated from treated radioactive waste and also reduce the separation of radioactive nuclide due to external environmental factors such as air, water or the like. Method: Radioactive waste adhered with radioactive nuclide to solid material is molten to mix and submerge the radioactive nuclide adhered to the surface of the solid material into molten material. Then, the radioactive nuclide thus mixed is solidified to store the waste in solidified state. (Aizawa, K.)

  16. ENVIRONMENTAL RESEARCH BRIEF: WASTE REDUCTION ACTIVITIES AND OPTIONS FOR A MANUFACTURER OF WIRE STOCK USED FOR PRODUCTION OF METAL ITEMS.

    Science.gov (United States)

    The U.S. Environmental Protection Agency (EPA) funded a project with the New Jersey Department of Environmental Protection and Energy (NJDEPE) to assist in conducting waste minimization assessments at thirty small- to medium-sized businesses in the state of New Jersey. One of th...

  17. Method for processing radioactive wastes containing sodium

    International Nuclear Information System (INIS)

    Kubota, Takeshi.

    1975-01-01

    Object: To bake, solidify and process even radioactive wastes highly containing sodium. Structure: H and or NH 4 zeolites of more than 90g per chemical equivalent of sodium present in the waste is added to and left in radioactive wastes containing sodium, after which they are fed to a baker such as rotary cylindrical baker, spray baker and the like to bake and solidify the wastes at 350 to 800 0 C. Thereby, it is possible to bake and solidify even radioactive wastes highly containing sodium, which has been impossible to do so previously. (Kamimura, M.)

  18. 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

  19. 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.

  20. Method for calcining radioactive wastes

    International Nuclear Information System (INIS)

    Bjorklund, W.J.; McElroy, J.L.; Mendel, J.E.

    1979-01-01

    A method for the preparation of radioactive wastes in a low leachability form involves calcining the radioactive waste on a fluidized bed of glass frit, removing the calcined waste to melter to form a homogeneous melt of the glass and the calcined waste, and then solidifying the melt to encapsulate the radioactive calcine in a glass matrix

  1. 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.

  2. Waste canister for storage of nuclear wastes

    Science.gov (United States)

    Duffy, James B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall.

  3. Waste canister for storage of nuclear wastes

    International Nuclear Information System (INIS)

    Duffy, J.B.

    1977-01-01

    A waste canister for storage of nuclear wastes in the form of a solidified glass includes fins supported from the center with the tips of the fins spaced away from the wall to conduct heat away from the center without producing unacceptable hot spots in the canister wall. 4 claims, 4 figures

  4. 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.)

  5. 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

  6. Performance assessment of stabilised/solidified waste-forms

    OpenAIRE

    Antemir, Aurora

    2010-01-01

    A method to treat contaminated land is stabilisation/solidification (S/S), which physically encapsulates and chemically stabilises the contaminants. The current knowledge on the behaviour of S/S systems is based upon scarce and incomplete data, mostly obtained from laboratory simulations or small scale trials of the technology. The field performance of S/S soils is largely unknown.\\ud \\ud The aim of this research was to improve the understanding of the long-term performance of S/S soils, by e...

  7. Assessment of the loss of radioactive isotopes from solidified wastes

    International Nuclear Information System (INIS)

    Godbee, H.W.; Kibbey, A.H.; Joy, D.S.

    1978-01-01

    Amounts of 137 Cs and 239 Pu entering the environment by leaching of grout products are predicted as functions of time, considering a 55-gal drum as a finite source and as a semi-infinite medium with and without decay. Significant differences between finite-cylinder and semi-infinite-medium calculations appear after a few years of leaching when the effective diffusivity (D) is 1.2 x 10 -9 cm 2 /s, but do not appear up to 3,000 years when D is 6.0 x 10 -16 cm 2 /s. The smaller D is and the larger the size is, the longer a finite specimen will behave as a semi-infinite medium

  8. Method of melting to solidify radioactive powder wastes

    International Nuclear Information System (INIS)

    Ootsuka, Katsuyuki; Miyazaki, Hitoshi.

    1981-01-01

    Purpose: To improve the microwave irradiation efficiency in a melting furnace. Constitution: Pelletization, sludgification and granularization are carried out as powderous dust reducing treatment. In the granularization, for example, radioactive burning ashes are sent from a hopper to a mixer and mixed with processing aids such as binders. Then, they are pelletized in a pelletizer into granular products and sent to a microwave melting furnace by way of a sieve screen. The granular products are melted by microwaves from a microwave guide tube and taken out through an exit. This can prevent powderous dusts from floating and scattering in the melting furnace and prevent the reduction in the microwave irradiation efficiency due to generation of electric discharges. (Seki, T.)

  9. 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.)

  10. 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

  11. 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.

  12. Hazardous Waste Code Determination for First/Second-Stage Sludge Waste Stream (IDCs 001, 002, 800)

    International Nuclear Information System (INIS)

    Arbon, R.E.

    2001-01-01

    This document, Hazardous Waste Code Determination for the First/Second-Stage Sludge Waste Stream, summarizes the efforts performed at the Idaho National Engineering and Environmental Laboratory (INEEL) to make a hazardous waste code determination on Item Description Codes (IDCs) 001, 002, and 800 drums. This characterization effort included a thorough review of acceptable knowledge (AK), physical characterization, waste form sampling, chemical analyses, and headspace gas data. This effort included an assessment of pre-Waste Analysis Plan (WAP) solidified sampling and analysis data (referred to as preliminary data). Seventy-five First/Second-Stage Sludge Drums, provided in Table 1-1, have been subjected to core sampling and analysis using the requirements defined in the Quality Assurance Program Plan (QAPP). Based on WAP defined statistical reduction, of preliminary data, a sample size of five was calculated. That is, five additional drums should be core sampled and analyzed. A total of seven drums were sampled, analyzed, and validated in compliance with the WAP criteria. The pre-WAP data (taken under the QAPP) correlated very well with the WAP compliant drum data. As a result, no additional sampling is required. Based upon the information summarized in this document, an accurate hazardous waste determination has been made for the First/Second-Stage Sludge Waste Stream

  13. 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.)

  14. 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

  15. Solidifying power station resins and sludges

    International Nuclear Information System (INIS)

    Willis, A.S.D.; Haigh, C.P.

    1984-01-01

    Radioactive ion exchange resins and sludges arise at nuclear power stations from various operations associated with effluent treatment and liquid waste management. As the result of an intensive development programme, the Central Electricity Generating Board (CEGB) has designed a process to convert power station resins and sludges into a shielded, packaged solid monolithic form suitable for final disposal. Research and development, the generic CEGB sludge/resin conditioning plant and the CEGB Active Waste Project are described. (U.K.)

  16. A Balance Sheet for Educational Item Banking.

    Science.gov (United States)

    Hiscox, Michael D.

    Educational item banking presents observers with a considerable paradox. The development of test items from scratch is viewed as wasteful, a luxury in times of declining resources. On the other hand, item banking has failed to become a mature technology despite large amounts of money and the efforts of talented professionals. The question of which…

  17. Method of solidifying radioactive ion exchange resin

    International Nuclear Information System (INIS)

    Minami, Yuji; Tomita, Toshihide

    1989-01-01

    Spent anion exchange resin formed in nuclear power plants, etc. generally catch only a portion of anions in view of the ion exchange resins capacity and most of the anions are sent while possessing activities to radioactive waste processing systems. Then, the anion exchange resins increase the specific gravity by the capture of the anions. Accordingly, anions are caused to be captured on the anion exchange resin wastes such that the specific gravity of the anion exchange resin wastes is greater than that of the thermosetting resins to be mixed. This enables satisfactory mixing with the thermosetting resins and, in addition, enables to form integral solidification products in which anion exchange resins and cation exchange resins are not locallized separately and which are homogenous and free from cracks. (T.M.)

  18. 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

  19. 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

  20. 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

  1. 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.

  2. 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.

  3. 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”.

  4. 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.

  5. Interim solidification of SRP waste with silica, bentonite, or phosphoric acid

    International Nuclear Information System (INIS)

    Thompson, G.H.

    1976-03-01

    One option for interim waste management at the Savannah River Plant is in-tank solidification of the liquid waste solutions. This would reduce the mobility of these highly radioactive solutions until techniques for their long-term immobilization and storage are developed and implemented. Interim treatments must permit eventual retrieval of waste and subsequent incorporation into a high-integrity form. This study demonstrated the solidification of simulated alkaline waste solutions by reaction with silica, bentonite, and phosphoric acid. Alkaline waste can be solidified by reaction with silica gel, silica flour, or sodium silicate solution. Solidified products containing waste salt can be retrieved by slurrying with water. Alkaline supernate (solution in equilibrium with alkaline sludge in SRP waste tanks) can be solidified by reaction with bentonite to form cancrinite powder. The solidified waste can be retrieved by slurrying with water. Alkaline supernate can be solidified by partial evaporation and reaction with phosphoric acid. Water is incorporated into hydrated complexes of trisodium phosphate. The product is soluble, but actual plant waste would not solidify completely because of decay heat. Reaction of simulated alkaline waste solutions with silica gel, silica flour, or bentonite increases the volume by a factor of approximately 6 over that of evaporated waste; reaction with phosphoric acid results in a volume 1.5 times that of evaporated waste. At present, the best method for in-tank solidification is by evaporation, a method that contributes no additional solids to the waste and does not compromise any waste management options

  6. 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

  7. Method for processing powdery radioactive wastes

    International Nuclear Information System (INIS)

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

    1978-01-01

    Purpose: To solidify radioactive wastes with ease and safety at a high reaction speed but with no boiling by impregnating the radioactive wastes with chlorostyrene. Method: Beads-like dried ion exchange resin, powdery ion exchange resin, filter sludges, concentrated dried waste liquor or the like are mixed or impregnated with a chlorostyrene monomer dissolving therein a polymerization initiator such as methyl ethyl ketone peroxide and benzoyl peroxide. Mixed or impregnated products are polymerized to solid after a predetermined of time through curing reaction to produce solidified radioactive wastes. Since inflammable materials are used, this process has a high safety. About 70% wastes can be incorporated. The solidified products have a strength as high as 300 - 400 kg/cm 3 and are suitable to ocean disposal. The products have a greater radioactive resistance than other plastic solidification products. (Seki, T.)

  8. Immobilisation of hazardous waste

    International Nuclear Information System (INIS)

    Cope, C.B.

    1983-01-01

    Hazardous waste, e.g. radioactive waste, particularly that containing caesium-137, is immobilised by mixing with cement and solidifiable organic polymeric material. When first mixed, the organic material is preferably liquid and at this time can be polymerisable or already polymerised. The hardening can result from cooling or further polymerisation e.g. cross-linking. The organic material may be wax, or a polyester which may be unsaturated and cross-linkable by reaction with styrene. (author)

  9. 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)

  10. HOW THE ROCKY FLATS ENVIRONMENTAL TECHNOLOGY SITE DEVELOPED A NEW WASTE PACKAGE USING A POLYUREA COATING THAT IS SAFELY AND ECONOMICALLY ELIMINATING SIZE REDUCTION OF LARGE ITEMS

    International Nuclear Information System (INIS)

    Dorr, Kent A.; Hogue, Richard S.; Kimokeo, Margaret K.

    2003-01-01

    One of the major challenges involved in closing the Rocky Flats Environmental Technology Site (RFETS) is the disposal of extremely large pieces of contaminated production equipment and building debris. Past practice has been to size reduce the equipment into pieces small enough to fit into approved, standard waste containers. Size reducing this equipment is extremely expensive, and exposes workers to high-risk tasks, including significant industrial, chemical, and radiological hazards. RFETS has developed a waste package using a Polyurea coating for shipping large contaminated objects. The cost and schedule savings have been significant

  11. 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

  12. 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.

  13. Development of drying and pelletizing system for concentrated waste

    International Nuclear Information System (INIS)

    Horiuchi, Susumu; Saito, Toru; Hirano, Mikio; Kikuchi, Makoto; Takamura, Yoshiyuki.

    1980-01-01

    Volume reduction is strongly required for the radioactive liquid waste generated in nuclear power plants because its storing space has increased with the operating years of the plants, though it has temporarily been stored in drum cans within the plant sites after concentrated by evaporation. The drying and pelletizing system developed by Hitachi, Ltd. in cooperation with Tokyo Electric Power Co. aims at the final disposal by solidifying stored waste after drying, pulverizing, and pelletizing concentrated liquid waste, and storing it in tanks to reduce its radioactivity for the predetermined period. The outstanding features of the system are to be capable of realizing drastic volume reduction and of storing waste as the stable solid in the form flexibly adaptable to any disposing method. The system, to which the new concepts of pulverizing by drying and pelletizing concentrated liquid waste were applied, has been subjected to various fundamental tests and the demonstration tests in a pilot plant during the research and development for 7-years, consequently it was confirmed that the system can be used practically, and the data for designing the equipment for practical use were collected. The items to be considered in designing the equipment for practical use are also mentioned. (Wakatsuki, Y.)

  14. Disposal of bead ion exchange resin wastes

    International Nuclear Information System (INIS)

    Gay, R.L.; Granthan, L.F.

    1985-01-01

    Bead ion exchange resin wastes are disposed of by a process which involves spray-drying a bead ion exchange resin waste in order to remove substantially all of the water present in such waste, including the water on the surface of the ion exchange resin beads and the water inside the ion exchange resin beads. The resulting dried ion exchange resin beads can then be solidified in a suitable solid matrix-forming material, such as a polymer, which solidifies to contain the dried ion exchange resin beads in a solid monolith suitable for disposal by burial or other conventional means

  15. Wastes

    International Nuclear Information System (INIS)

    Bovard, Pierre

    The origin of the wastes (power stations, reprocessing, fission products) is determined and the control ensuring the innocuity with respect to man, public acceptance, availability, economics and cost are examined [fr

  16. Properties of radioactive wastes and waste containers

    International Nuclear Information System (INIS)

    Arora, H.S.; Dayal, R.

    1984-01-01

    Major tasks in this NRC-sponsored program include: (1) an evaluation of the acceptability of low-level solidified wastes with respect to minimizing radionuclide releases after burial; and (2) an assessment of the influence of pertinent environmental stresses on the performance of high-integrity radwaste container (HIC) materials. The waste form performance task involves studies on small-scale laboratory specimens to predict and extrapolate: (1) leachability for extended time periods; (2) leach behavior of full-size forms; (3) performance of waste forms under realistic leaching conditions; and (4) leachability of solidified reactor wastes. The results show that leach data derived from testing of small-scale specimens can be extrapolated to estimate leachability of a full-scale specimen and that radionuclide release data derived from testing of simulants can be employed to predict the release behavior of reactor wastes. Leaching under partially saturated conditions exhibits lower releases of radionuclides than those observed under the conventional IAEA-type or ANS 16.1 leach tests. The HIC assessment task includes the characterization of mechanical properties of Marlex CL-100, a candidate radwaste high density polyethylene material. Tensile strength and creep rupture tests have been carried out to determine the influence of specific waste constituents as well as gamma irradiation on material performance. Emphasis in ongoing tests is being placed on studying creep rupture while the specimens are in contact with a variety of chemicals including radiolytic by-products of irradiated resin wastes. 12 references 6 figures, 2 tables

  17. Solidification of radioactive waste resins using cement mixed with organic material

    International Nuclear Information System (INIS)

    Laili, Zalina; Yasir, Muhamad Samudi; Wahab, Mohd Abdul

    2015-01-01

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins

  18. Solidification of radioactive waste resins using cement mixed with organic material

    Energy Technology Data Exchange (ETDEWEB)

    Laili, Zalina, E-mail: liena@nm.gov.my [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi [Nuclear Science Programme, School of Applied Physics, Faculty of Science and Technology, Universiti Kebangsaan Malaysia (UKM), Bangi, 43600, Selangor Malaysia (Malaysia); Wahab, Mohd Abdul [Waste and Environmental Technology Division, Malaysian Nuclear Agency (Nuclear Malaysia), Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Solidification of radioactive waste resins using cement mixed with organic material i.e. biochar is described in this paper. Different percentage of biochar (0%, 5%, 8%, 11%, 14% and 18%) was investigated in this study. The characteristics such as compressive strength and leaching behavior were examined in order to evaluate the performance of solidified radioactive waste resins. The results showed that the amount of biochar affect the compressive strength of the solidified resins. Based on the data obtained for the leaching experiments performed, only one formulation showed the leached of Cs-134 from the solidified radioactive waste resins.

  19. In-situ thermeolectric stabilization of radioactive wastes

    International Nuclear Information System (INIS)

    Brouns, R.A.; Timmerman, C.L.

    1982-01-01

    Current analysis indicates that in situ vitrification is applicable to many wastes and soil types at a cost an order of magnitude less than exhumation, processing, and transportation to a deep geological disposal site. Once the waste materials have been solidified, future ground subsidence, wind erosion and plant or animal intrusion are virtually eliminated. Furthermore, the waste form is extremely durable

  20. 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

  1. State of the art report on bituminized waste forms of radioactive wastes

    International Nuclear Information System (INIS)

    Kim, Tae Kook; Shon, Jong Sik; Kim, Kil Jeong; Lee, Kang Moo; Jung, In Ha

    1998-03-01

    In this report, research and development results on the bituminization of radioactive wastes are closely reviewed, especially those regarding waste treatment technologies, waste solidifying procedures and the characteristics of asphalt and solidified forms. A new concept of the bituminization method is suggested in this report which can improve the characteristics of solidified forms. Stable solid forms with high leach resistance, high thermal resistance and good compression strength were produced by the suggested bituminization method, in which spent polyethylene from agricultural farms was added. This report can help further research and development of improved bituminized forms of radioactive wastes that will maintain long term stabilities in disposal sites. (author). 59 refs., 19 tabs., 18 figs

  2. 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

  3. 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.

  4. 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....

  5. Radioactive waste material disposal

    Science.gov (United States)

    Forsberg, Charles W.; Beahm, Edward C.; Parker, George W.

    1995-01-01

    The invention is a process for direct conversion of solid radioactive waste, particularly spent nuclear fuel and its cladding, if any, into a solidified waste glass. A sacrificial metal oxide, dissolved in a glass bath, is used to oxidize elemental metal and any carbon values present in the waste as they are fed to the bath. Two different modes of operation are possible, depending on the sacrificial metal oxide employed. In the first mode, a regenerable sacrificial oxide, e.g., PbO, is employed, while the second mode features use of disposable oxides such as ferric oxide.

  6. Radioactive waste processing device

    International Nuclear Information System (INIS)

    Ikeda, Takashi; Funabashi, Kiyomi; Chino, Koichi.

    1992-01-01

    In a waste processing device for solidifying, pellets formed by condensing radioactive liquid wastes generated from a nuclear power plant, by using a solidification agent, sodium chloride, sodium hydroxide or sodium nitrate is mixed upon solidification. In particular, since sodium sulfate in a resin regenerating liquid wastes absorbs water in the cement upon cement solidification, and increases the volume by expansion, there is a worry of breaking the cement solidification products. This reaction can be prevented by the addition of sodium chloride and the like. Accordingly, integrity of the solidification products can be maintained for a long period of time. (T.M.)

  7. 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.

  8. 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.)

  9. Method of processing radioactive wastes

    International Nuclear Information System (INIS)

    Nomura, Ichiro; Hashimoto, Yasuo.

    1984-01-01

    Purpose: To improve the volume-reduction effect, as well as enable simultaneous procession for the wastes such as burnable solid wastes, resin wastes or sludges, and further convert the processed materials into glass-solidified products which are much less burnable and stable chemically and thermally. Method: Auxiliaries mainly composed of SiO 2 such as clays, and wastes such as burnable solid wastes, waste resins and sludges are charged through a waste hopper into an incinerating melting furnace comprising an incinerating and a melting furnace, while radioactive concentrated liquid wastes are sprayed from a spray nozzle. The wastes are burnt by the heat from the melting furnace and combustion air, and the sprayed concentrated wastes are dried by the hot air after the combustion into solid components. The solid matters from the concentrated liquid wastes and the incinerating ashes of the wastes are melted together with the auxiliaries in the melting furnace and converted into glass-like matters. The glass-like matters thus formed are caused to flow into a vessel and gradually cooled to solidify. (Horiuchi, T.)

  10. On the experience of the management of solid alpha-bearing wastes

    International Nuclear Information System (INIS)

    Kryuchkov, V.A.; Rakov, N.A.; Romanovskii, V.N.; Yakushev, M.F.

    1978-01-01

    Spent fuel reprocessing is studied in a pilot plant. Low and high level radioactive wastes handling is described. Liquid wastes are solidified. Combustible solid wastes are incinerated. Non-combustible and ashes are send to disposal site. Volume reduction of alpha-bearing wastes is obtained by optimisation of the reprocessing and development of remote control methods

  11. Cleanup Verification Package for the 600-259 Waste Site

    Energy Technology Data Exchange (ETDEWEB)

    J. M. Capron

    2006-02-09

    This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste.

  12. Cleanup Verification Package for the 600-259 Waste Site

    International Nuclear Information System (INIS)

    Capron, J.M.

    2006-01-01

    This cleanup verification package documents completion of remedial action for the 600-259 waste site. The site was the former site of the Special Waste Form Lysimeter, consisting of commercial reactor isotope waste forms in contact with soils within engineered caissons, and was used by Pacific Northwest National Laboratory to collect data regarding leaching behavior for target analytes. A Grout Waste Test Facility also operated at the site, designed to test leaching rates of grout-solidified low-level radioactive waste

  13. Solidification of radioactive wastes with thermosetting resin

    International Nuclear Information System (INIS)

    Hayashi, M.; Kobayashi, K.; Okamoto, O.; Kagawa, T.; Wakamatsu, K.; Irie, H.; Matsuura, H.; Yasumura, K.; Nakayama, Y.

    1982-01-01

    Dried simulated radioactive wastes were solidified with thermosetting resin and their properties were investigated with laboratory scale and real scale products through extensive testings, such as mechanical resistance, resistance to leaching and swelling in water, radiation resistance, fire resistance and resistance to temperature cycling. The typical results were as follows: over 600 kg/cm 2 of compressive strength, diffusion constant of approx. 10 - 5 cm 2 /day for 137 Cs leaching from solidified waste products, no significant change was found for up to 5 x 10 8 RAD irradiation, and damages were limited to the surface of the products after the thermal test and dropping impact test. 7 figures, 4 tables

  14. 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

  15. Method of processing radioactive waste

    International Nuclear Information System (INIS)

    Uehara, Susumu.

    1990-01-01

    Radioactive solid wastes generated from nuclear power plants are pressed and reduced in the volume by a compressor into compression products. Next, the compression products are put into a vessel in a tank and a solidifying material at low viscosity such as vinyl monomer is supplied and impregnated into the inner gaps of the compression products while the pressure in the tank is reduced by a vacuum pump. Subsequently, the compression products are heated and pressurized in the tank to polymerize and solidify the solidifying material. Then, a plurality of solidified compression products are placed in the inside of a drum can and fixed at the periphery thereof together with fixing material such as mortars and plastics. Accordingly, even when underground water should intrude after underground disposal, there is no more risk of causing swelling pressure due to water absorption. Accordingly, there is no more possiblity to cause cracks in the wastes due to the swelling pressure, and wastes of excellent stability and integrity can be obtained. (I.N.)

  16. 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

  17. 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.

  18. Radioactive waste management and disposal

    International Nuclear Information System (INIS)

    Simon, R.; Orlowski, S.

    1980-01-01

    The first European Community conference on Radioactive Waste Management and Disposal was held in Luxembourg, where twenty-five papers were presented by scientists involved in European Community contract studies and by members of the Commission's scientific staff. The following topics were covered: treatment and conditioning technology of solid intermediate level wastes, alpha-contaminated combustible wastes, gaseous wastes, hulls and dissolver residues and plutonium recovery; waste product evaluation which involves testing of solidified high level wastes and other waste products; engineering storage of vitrified high level wastes and gas storage; and geological disposal in salt, granite and clay formations which includes site characterization, conceptual repository design, waste/formation interactions, migration of radionuclides, safety analysis, mathematical modelling and risk assessment

  19. 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

  20. Leaching of nuclear power reactor wastes forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1986-01-01

    The leaching tests for power reactor wastes carried out at IPEN/CNEN-SP are described. These waste forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. 3 years leaching results are reported, determining cesium and strontium diffusivity coefficients for boric acid waste form and ion-exchange resins. (Author) [pt

  1. Waste inventory, waste characteristics and waste repositories in Japan

    International Nuclear Information System (INIS)

    Shimooka, K.

    1997-01-01

    There are two types of repositories for the low level radioactive wastes in Japan. One is a trench type repository only for concrete debris generated from the dismantling of the research reactor. According to the safety assurance system, Japan Atomic Energy Research Institute (JAERI) has disposed of the concrete debris arose from the dismantling of the Japan Power Demonstration Reactor (JPDR). The other type is the concreted pit with engineered barriers. Rokkasho Low Level Radioactive Waste Disposal Center has this type of repository mainly for the power plant wastes. Japan Nuclear Fuel Ltd. (JNFL) established by electric power companies is the operator of the LLW disposal project. JNFL began the storage operation in 1992 and buried approximately 60,000 drums there. Two hundred thousand drums of uniformly solidified, waste may be buried ultimately. 4 refs, 3 tabs

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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

  7. Method of melting solid waste

    International Nuclear Information System (INIS)

    Ootsuka, Katsuyuki; Mizuno, Ryokichi; Kuwana, Katsumi; Sawada, Yoshihisa; Komatsu, Fumiaki.

    1982-01-01

    Purpose: To enable the volume reduction treatment of a HEPA filter containing various solid wastes, particularly acid digestion residue, or an asbestos separator at a relatively low temperature range. Method: Solid waste to be heated and molten is high melting point material treated by ''acid digestion treatment'' for treating solid waste, e.g. a HEPA filter or polyvinyl chloride, etc. of an atomic power facility treated with nitric acid or the like. When this material is heated and molten by an electric furnace, microwave melting furnace, etc., boron oxide, sodium boride, sodium carbonate, etc. is added as a melting point lowering agent. When it is molten in this state, its melting point is lowered, and it becomes remarkably fluid, and the melting treatment is facilitated. Solidified material thus obtained through the melting step has excellent denseness and further large volume reduction rate of the solidified material. (Yoshihara, H.)

  8. Method of processing liquid wastes

    International Nuclear Information System (INIS)

    Naba, Katsumi; Oohashi, Takeshi; Kawakatsu, Ryu; Kuribayashi, Kotaro.

    1980-01-01

    Purpose: To process radioactive liquid wastes with safety by distillating radioactive liquid wastes while passing gases, properly treating the distillation fractions, adding combustible and liquid synthetic resin material to the distillation residues, polymerizing to solidify and then burning them. Method: Radioactive substance - containing liquid wastes are distillated while passing gases and the distillation fractions containing no substantial radioactive substances are treated in an adequate method. Synthetic resin material, which may be a mixture of polymer and monomer, is added together with a catalyst to the distillation residues containing almost of the radioactive substances to polymerize and solidify. Water or solvent in such an extent as not hindering the solidification may be allowed if remained. The solidification products are burnt for facilitating the treatment of the radioactive substances. The resin material can be selected suitably, methacrylate syrup (mainly solution of polymethylmethacrylate and methylmethacrylate) being preferred. (Seki, T.)

  9. Study of plastic solidification process on solid radioactive waste treatment

    International Nuclear Information System (INIS)

    Jing Weiguan; Zhang Yinsheng; Qian Wenju

    1994-01-01

    Comparisons between the plastic solidification conditions of incinerated ash and waste cation resin by using thermosetting plastic polyvinyl chloride (PVC), polystyrene (PS) and polyethylene (PE), and identified physico-chemical properties and irradiation resistance of solidified products were presented. These solidified products have passed through different tests as compression strength, leachability, durability, stability, permeability and irradiation resistance (10 6 Gy) etc. The result showed that the solidified products possessed stable properties and met the storage requirement. The waste tube of radioimmunoassay, being used as solidification medium to contain incinerated ash, had good mechanical properties and satisfactory volume reduction. This process may develop a new way for disposal solid radioactive waste by means of re-using waste

  10. Study on cementation of simulated radioactive borated liquid wastes

    International Nuclear Information System (INIS)

    Sun Qina; Li Junfeng; Wang Jianlong

    2010-01-01

    To compare sulfoaluminate cement with ordinary Portland cement on their cementation of radioactive borated liquid waste and to provide more data for formula optimization, simulated radioactive borated liquid waste were solidified by the two cements. 28 d compressive strength and strength losses after water/freezing/irradiation resistance tests were investigated. Leaching test and X-ray diffraction analysis were also conducted. The results show that it is feasible to solidify borated liquid wastes with sulfoaluminate cement and ordinary Portland cement with formulas used in the study. The 28 d compressive strengths, strength losses after tests and simulated nuclides leaching rates of the solidified waste forms meet the demand of GB 14569.1-93. The sulfoaluminate cement formula show better retention of Cs + than ordinary Portland cement formula. Boron, in form of B (OH) 4 - , incorporate in ettringite as solid solutions. (authors)

  11. 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

  12. 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

  13. Treatment of radioactive wastes

    International Nuclear Information System (INIS)

    Machida, Chuji

    1976-01-01

    Japan Atomic Energy Research Institute (JAERI) is equipped with such atomic energy facilities as a power test reactor, four research reactors, a hot laboratory, and radioisotope-producing factory. All the radioactive wastes but gas generated from these facilities are treated by the waste treatment facilities established in JAERI. The wastes carried into JAERI through Japan Radioisotope Association are also treated there. Low level water solution is treated with an evaporating apparatus, an ion-exchange apparatus, and a cohesive precipitating apparatus, while medium level solution is treated with an evaporating apparatus, and low level combustible solid is treated with an incinerating apparatus. These treated wastes and sludges are mixed with Portland cement in drum cans to solidify, and stored in a concrete pit. The correct classification and its indication as well as the proper packing for the wastes are earnestly demanded by the treatment facilities. (Kobatake, H.)

  14. Cementation of wastes with boric acid

    International Nuclear Information System (INIS)

    Tello, Cledola C.O.; Haucz, Maria Judite A.; Alves, Lilian J.L.; Oliveira, Arno H.

    2000-01-01

    In nuclear power plants (PWR) are generated wastes, such as concentrate, which comes from the evaporation of liquid radioactive wastes, and spent resins. Both have boron in their composition. The cementation process is one of the options to solidify these wastes, but the boron has a negative effect on the setting of the cement mixture. In this paper are presented the experiments that are being carried out in order to overcome this problem and also to improve the efficiency of the process. Simulated wastes were cemented using additives (clays, admixtures etc.). In the process and product is being evaluated the effect of the amount, type and addition order of the materials. The mixtures were selected in accordance with their workability and incorporated waste. The solidified products are monolithic without free water with a good mechanical resistance. (author)

  15. Low-level waste management - suggested solutions for problem wastes

    International Nuclear Information System (INIS)

    Pechin, W.H.; Armstrong, K.M.; Colombo, P.

    1984-01-01

    Problem wastes are those wastes which are difficult or require unusual expense to place into a waste form acceptable under the requirements of 10 CFR 61 or the disposal site operators. Brookhaven National Laboratory has been investigating the use of various solidification agents as part of the DOE Low-Level Waste Management Program for several years. Two of the leading problem wastes are ion exchange resins and organic liquids. Ion exchange resins can be solidified in Portland cement up to about 25 wt % resin, but waste forms loaded to this degree exhibit significantly reduced compressive strength and may disintegrate when immersed in water. Ion exchange resins can also be incorporated into organic agents. Mound Laboratory has been investigating the use of a joule-heated glass melter as a means of disposing of ion exchange resins and organic liquids in addition to other combustible wastes

  16. System for disposing of radioactive waste

    International Nuclear Information System (INIS)

    Gablin, K.A.; Hansen, L.J.

    1980-01-01

    A system is disclosed for disposing of radioactive mixed liquid and particulate waste material from nuclear reactors by solidifying the liquid components into a free standing hardened mass with a syrup of partially polymerized particles of urea formaldehyde in water and a liquid curing agent

  17. Method of processing radioactive liquid waste

    International Nuclear Information System (INIS)

    Motojima, Kenji; Kawamura, Fumio.

    1981-01-01

    Purpose: To increase the efficiency of removing radioactive cesium from radioactive liquid waste by employing zeolite affixed to metallic compound ferrocyanide as an adsorbent. Method: Regenerated liquid waste of a reactor condensation desalting unit, floor drain and so forth are collected through respective supply tubes to a liquid waste tank, and the liquid waste is fed by a pump to a column filled with zeolite containing a metallic compound ferrocyanide, such as with copper, zinc, manganese, iron, cobalt, nickel or the like. The liquid waste from which radioactive cesium is removed is dried and pelletized by volume reducing and solidifying means. (Yoshino, Y.)

  18. 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)

  19. 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...

  20. Method of disposing radioactive wastes

    International Nuclear Information System (INIS)

    Isozaki, Kei.

    1983-01-01

    Purpose : To enable safety ocean disposal of radioactive wastes by decreasing the leaching rate of radioactive nucleides, improving the quick-curing nature and increasing the durability. Method : A mixture comprising 2 - 20 parts by weight of alkali metal hydroxide and 100 parts by weight of finely powdered aqueous slags from a blast furnace is added to radioactive wastes to solidify them. In the case of medium or low level radioactive wastes, the solidification agent is added by 200 parts by weight to 100 parts by weight of the wastes and, in the case of high level wastes, the solidification agent is added in such an amount that the wastes occupy about 20% by weight in the total of the wastes and the solidification agent. Sodium hydroxide used as the alkali metal hydroxide is partially replaced with sodium carbonate, a water-reducing agent such as lignin sulfonate is added to improve the fluidity and suppress the leaching rate and the wastes are solidified in a drum can. In this way, corrosions of the vessel can be suppressed by the alkaline nature and the compression strength, heat stability and the like of the product also become excellent. (Sekiya, K.)

  1. Special waste form lysimeters-arid. Annual report, 1985

    International Nuclear Information System (INIS)

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

    1985-09-01

    The Special Waste Form Lysimeters-Arid program was initiated to determine typical source terms generated by commercial solidified low-level nuclear waste in an arid climate. Waste-form leaching tests are being conducted at a field facility at the Hanford site near Richland, Washington. A similar program is being conducted at a humid site. The field facility consists of 10 lysimeters placed around a central instrument caisson. The waste samples from boiling water and pressurized water reactors were emplaced in 1984, and the lysimeters are being monitored for movement of contaminants and water. Solidifying agents being tested include vinyl ester-styrene, bitumen, and cement. Laboratory leaching and geochemical modeling studies are being conducted to predict expected leach rates at the field site and to aid field-data interpretation. Small samples of the solidified waste forms were made for use in the laboratory leaching studies that include standard leach tests and leaching of solidified waste forms in soil columns. Complete chemical and radionuclide analyses are being conducted on the solid and liquid portions of the wastes. 2 refs

  2. Special Waste Form Lysimeters-Arid: annual report 1985

    International Nuclear Information System (INIS)

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

    1986-01-01

    The Special Waste Form Lysimeters-Arid program was initiated to determine typical source terms generated by commercial solidified low-level nuclear waste in an arid climate. Waste-form leaching tests are being conducted at a field facility at the Hanford site near Richland, Washington. A similar program is being conducted at a humid site. The field facility consists of 10 lysimeters placed around a central instrument caisson. The waste samples from boiling water and pressurized water reactors were emplaced in 1984, and the lysimeters are being monitored for movement of contaminants and water. Solidifying agents being tested include vinyl ester-styrene, bitumen, and cement. Laboratory leaching and geochemical modeling studies are being conducted to predict expected leach rates at the field site and to aid field-data interpretation. Small samples of the solidified waste forms were made for use in the laboratory leaching studies that include standard leach tests and leaching of solidified waste forms in soil columns. Complete chemical and radionuclide analyses are being conducted on the solid and liquid portions of the wastes

  3. 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.

  4. Proposed waste form performance criteria and testing methods for low-level mixed waste

    International Nuclear Information System (INIS)

    Franz, E.M.; Fuhrmann, M.; Bowerman, B.

    1995-01-01

    Proposed waste form performance criteria and testing methods were developed as guidance in judging the suitability of solidified waste as a physico-chemical barrier to releases of radionuclides and RCRA regulated hazardous components. The criteria follow from the assumption that release of contaminants by leaching is the single most important property for judging the effectiveness of a waste form. A two-tier regimen is proposed. The first tier consists of a leach test designed to determine the net, forward leach rate of the solidified waste and a leach test required by the Environmental Protection Agency (EPA). The second tier of tests is to determine if a set of stresses (i.e., radiation, freeze-thaw, wet-dry cycling) on the waste form adversely impacts its ability to retain contaminants and remain physically intact. In the absence of site-specific performance assessments (PA), two generic modeling exercises are described which were used to calculate proposed acceptable leachates

  5. 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.''

  6. Special waste-form lysimeters-arid: Three-year monitoring report

    International Nuclear Information System (INIS)

    Jones, T.L.; Serne, R.J.; Toste, A.P.

    1988-04-01

    Regulations governing the disposal of commercial low-level waste require all liquid waste to be solidified before burial. Most waste must be solidified into a rigid matrix such as cement or plastic to prevent waste consolidation and site slumping after burial. These solidification processes affect the rate at which radionuclides and other solutes are released into the soil. In 1983, a program was initiated at Pacific Northwest Laboratory to study the release of waste from samples of low-level radioactive waste that had been commercially solidified. The primary method used by this program is to bury sample waste forms in field lysimeters and monitor leachate composition from the release and transport of solutes. The lysimeter facility consists of 10 lysimeters, each containing one sample of solidified waste. Five different waste forms are being tested, allowing duplicate samples of each one to be evaluated. The samples were obtained from operating nuclear power plants and are actual waste forms routinely generated at these facilities. All solidification was accomplished by commercial processes. Sample size is a partially filled 210-L drum. All containers were removed prior to burial leaving the bare waste form in contact with the lysimeter soil. 11 refs., 14 figs., 16 tabs

  7. 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)

  8. Method of processing radioactive liquid wastes

    International Nuclear Information System (INIS)

    Kurumada, Norimitsu; Shibata, Setsuo; Wakabayashi, Toshikatsu; Kuribayashi, Hiroshi.

    1984-01-01

    Purpose: To facilitate the procession of liquid wastes containing insoluble salts of boric acid and calcium in a process for solidifying under volume reduction of radioactive liquid wastes containing boron. Method: A soluble calcium compound (such as calcium hydroxide, calcium oxide and calcium nitrate) is added to liquid wastes whose pH value is adjusted neutral or alkaline such that the molar ratio of calcium to boron in the liquid wastes is at least 0.2. Then, they are agitated at a temperature between 40 - 70 0 C to form insoluble calcium salt containing boron. Thereafter, the liquid is maintained at a temperature less than the above-mentioned forming temperature to age the products and, thereafter, the liquid is evaporated to condensate into a liquid concentrate containing 30 - 80% by weight of solid components. The concentrated liquid is mixed with cement to solidify. (Ikeda, J.)

  9. Waste volume reduction by spray drying

    Energy Technology Data Exchange (ETDEWEB)

    Toscano, Rodrigo A.; Tello, Clédola C. O. de, E-mail: Rodrigotoscano1@gmail.com, E-mail: tellocc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The operation of nuclear facilities generates liquid wastes which require treatment to control the chemical compounds and removal of radioactive contaminants. These wastes can come from the cooling of the primary reactor system, from the reactor pool decontamination, washing of contaminated clothing, among others. The ion exchange resin constitutes the largest fraction of this waste, classified as low and intermediate level of radiation. According to CNEN Standard 8.01, the minimization of the volume and activity of the radioactive waste generated in the operation of a nuclear installation, radiative installation, industrial mining installation or radioactive waste deposit should be ensured. In addition, one of the acceptance criteria for wastes in repositories required by CNEN NN 6.09 is that it be solid or solidified. Thus, these wastes must be reduced in volume and solidified to meet the standards and the safety of the population and the environment. The objective of this work is to find a solution that associates the least generation of packaged waste and the acceptance criteria of waste for the deposition in the national repository. This work presents a proposal of reduction of the volume of the liquid wastes generated by nuclear facilities by drying by for reduction of volume for a greater incorporation of wastes in cement. Using spray dryer, an 18% reduction in the production of cemented waste products was observed in relation to the method currently used with compressive strength measurement above the standard, and it is believed that this value may increase in future tests. (author)

  10. Waste volume reduction by spray drying

    International Nuclear Information System (INIS)

    Toscano, Rodrigo A.; Tello, Clédola C. O. de

    2017-01-01

    The operation of nuclear facilities generates liquid wastes which require treatment to control the chemical compounds and removal of radioactive contaminants. These wastes can come from the cooling of the primary reactor system, from the reactor pool decontamination, washing of contaminated clothing, among others. The ion exchange resin constitutes the largest fraction of this waste, classified as low and intermediate level of radiation. According to CNEN Standard 8.01, the minimization of the volume and activity of the radioactive waste generated in the operation of a nuclear installation, radiative installation, industrial mining installation or radioactive waste deposit should be ensured. In addition, one of the acceptance criteria for wastes in repositories required by CNEN NN 6.09 is that it be solid or solidified. Thus, these wastes must be reduced in volume and solidified to meet the standards and the safety of the population and the environment. The objective of this work is to find a solution that associates the least generation of packaged waste and the acceptance criteria of waste for the deposition in the national repository. This work presents a proposal of reduction of the volume of the liquid wastes generated by nuclear facilities by drying by for reduction of volume for a greater incorporation of wastes in cement. Using spray dryer, an 18% reduction in the production of cemented waste products was observed in relation to the method currently used with compressive strength measurement above the standard, and it is believed that this value may increase in future tests. (author)

  11. Radioactive waste management

    International Nuclear Information System (INIS)

    Blomek, D.

    1980-01-01

    The prospects of nuclear power development in the USA up to 2000 and the problems of the fuel cycle high-level radioactive waste processing and storage are considered. The problems of liquid and solidified radioactive waste transportation and their disposal in salt deposits and other geologic formations are discussed. It is pointed out that the main part of the high-level radioactive wastes are produced at spent fuel reprocessing plants in the form of complex aqueous mixtures. These mixtures contain the decay products of about 35 isotopes which are the nuclear fuel fission products, about 18 actinides and their daughter products as well as corrosion products of fuel cans and structural materials and chemical reagents added in the process of fuel reprocessing. The high-level radioactive waste management includes the liquid waste cooling which is necessary for the short and middle living isotope decay, separation of some most dangerous components from the waste mixture, waste solidification, their storage and disposal. The conclusion is drawn that the seccessful solution of the high-level radioactive waste management problem will permit to solve the problem of the fuel cycle radioactive waste management as a whole. The salt deposits, shales and clays are the most suitable for radioactive waste disposal [ru

  12. Leaching of solidified TRU-contaminated incinerator ash

    International Nuclear Information System (INIS)

    Fuhrmann, M.; Colombo, P.

    1984-01-01

    Leach rate and cumulative fractional releases of plutonium were determined for a series of laboratory-scale waste forms containing transuranic (TRU) contaminated incinerator ash. The solidification agents from which these waste forms were produced are commercially available materials for radioactive waste disposal. The leachants simulate groundwaters with chemical compositions that are indiginous to different geological media proposed for repositories. In this study TRU-contaminated ash was incorporated into waste forms fabricated with portland type I cement, urea-formaldehyde, polyester-styrene or Pioneer 221 bitumen. The ash was generated at the dual-chamber incinerator at the Rocky Flats Plant. These waste forms contained between 1.25 x 10 -2 and 4.4 x 10 -2 Ci (depending on the solidification agent) of mixed TRU isotopes comprised primarily of 239 Pu and 240 Pu. Five leachant solutions were prepared consisting of: (1) demineralized water, (2) simulated brine, (3) simplified sodium-dominated groundwater (30 meq NaCl/liter), (4) simplified calcium-dominated groundwater (30 meq CaCl 2 /liter), and (5) simplified bicarbonate-dominated groundwater (30 meq NaHCO 3 /liter). Cumulative fractional releases were found to vary significantly with different leachants and solidification agents. In all cases waste forms leached in brine gave the lowest leach rates. Urea-formaldehyde had the greatest release of radionuclides while polyester-styrene and portland cement had approximately equivalent fractional releases. Cement cured for 210 days retained radionuclides three times more effectively than cement cured only 30 days

  13. 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

  14. 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.

  15. Solidification of problem wastes: Annual progress report, October 1985-September 1986

    International Nuclear Information System (INIS)

    Franz, E.M.; Heiser, J.H. III; Colombo, P.

    1987-02-01

    This report describes initial work on the development of solidification systems for sodium nitrate waste and compacted waste. Sodium nitrate waste has been solidified in three types of materials: polyethylene, polyester-styrene (PES), and latex cement. Evaluations of the properties of the waste form, such as the ANS 16.1 leaching test, water immersion test and compressive strength measurements were performed on the waste forms containing various amounts of sodium nitrate. 9 refs., 9 figs., 7 tabs

  16. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-07-01

    The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Vault design, run-on/run-off control design, and asphalt compatibility with 90-degree celsius double-shell slurry feed

  17. Hazardous metals in yellow items used in RCAs

    International Nuclear Information System (INIS)

    Brown, K.F.; Rankin, W.N.

    1992-01-01

    Yellow items used in Radiologically Controlled Areas (RCAs) that could contain hazardous metals were identified. X-ray fluorescence analyses indicated that thirty of the fifty-two items do contain hazardous metals. It is important to minimize the hazardous metals put into the wastes. The authors recommend that the specifications for all yellow items stocked in Stores be changed to specify that they contain no hazardous metals

  18. Safety Evaluation for Packaging (onsite) T Plant Canyon Items

    International Nuclear Information System (INIS)

    OBRIEN, J.H.

    2000-01-01

    This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments

  19. Safety Evaluation for Packaging (onsite) T Plant Canyon Items

    Energy Technology Data Exchange (ETDEWEB)

    OBRIEN, J.H.

    2000-07-14

    This safety evaluation for packaging (SEP) evaluates and documents the ability to safely ship mostly unique inventories of miscellaneous T Plant canyon waste items (T-P Items) encountered during the canyon deck clean off campaign. In addition, this SEP addresses contaminated items and material that may be shipped in a strong tight package (STP). The shipments meet the criteria for onsite shipments as specified by Fluor Hanford in HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments.

  20. Hanford Waste Vitrification Plant Dangerous Waste Permit Application

    International Nuclear Information System (INIS)

    1991-10-01

    The Hanford Facility currently stores mixed waste, resulting from various processing operations, in underground storage tanks. The Hanford Waste Vitrification Plant will be constructed and operated to process the high-activity fraction of mixed waste stored in these underground tanks. The Hanford Waste Vitrification Plant will solidify pretreated tank waste into a glass product that will be packaged for disposal in a national repository. This Vitrification Plant Dangerous Waste Permit Application, Revision 2, consists of both a Part A and a Part B permit application. An explanation of the Part A revisions, including Revision 4 submitted with this application, is provided at the beginning of the Part A section. The Part B consists of 15 chapters addressing the organization and content of the Part B Checklist prepared by the Washington State Department of Ecology (Ecology 1987)

  1. Method of processing radioactive liquid wastes by using zeolites

    Energy Technology Data Exchange (ETDEWEB)

    Kanno, T; Mimura, H

    1975-09-18

    The object is to processing radioactive liquid waste by zeolites to be fixed to a solidified body having a very small lixiviation property. The nuclide in radioactive liquid waste is exchanged and adsorbed into natural or synthetic zeolites, which are then burnt to a temperature lower than 1000/sup 0/C -- melting point. Thus, the zeolite structure is broken to form fine amorphous silicate aluminate or silicate aluminate of the nuclide exchanged and adsorbed. Both are very hard to be soluble in water. Further, the lixiviation from the solidified body is limited to the surface thereof, and it will no longer be detected in a few days.

  2. 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

  3. Household hazardous waste

    DEFF Research Database (Denmark)

    Fjelsted, Lotte; Christensen, Thomas Højlund

    2007-01-01

    .) comprised 15-25% and foreign items comprised 10-20%. Water-based paint was the dominant part of the paint waste. The chemical composition of the paint waste and the paint-like waste was characterized by an analysis of 27 substances in seven waste fractions. The content of critical substances was tow......'Paint waste', a part of the 'household hazardous waste', amounting to approximately 5 tonnes was collected from recycling stations in two Danish cities. Sorting and analyses of the waste showed paint waste comprised approximately 65% of the mass, paint-like waste (cleaners, fillers, etc...... and the paint waste was less contaminated with heavy metals than was the ordinary household waste. This may suggest that households no longer need to source-segregate their paint if the household waste is incinerated, since the presence of a small quantity of solvent-based paint will not be harmful when...

  4. Disposal facility for radioactive wastes

    International Nuclear Information System (INIS)

    Utsunomiya, Toru.

    1985-01-01

    Purpose: To remove heat generated from radioactive wastes thereby prevent the working circumstances from being worsened in a disposal-facility for radioactive wastes. Constitution: The disposal-facility comprises a plurality of holes dug out into the ground inside a tunnel excavated for the storage of radioactive wastes. After placing radioactive wastes into the shafts, re-filling materials are directly filled with a purpose of reducing the dosage. Further, a plurality of heat pipes are inserted into the holes and embedded within the re-filling materials so as to gather heat from the radioactive wastes. The heat pipes are connected to a heat exchanger disposed within the tunnel. As a result, heating of the solidified radioactive wastes itself or the containing vessel to high temperature can be avoided, as well as thermal degradation of the re-filling materials and the worsening in the working circumstance within the tunnel can be overcome. (Moriyama, K.)

  5. Processing and discarding method for contaminated concrete wastes

    International Nuclear Information System (INIS)

    Yamamoto, Kazuo; Konishi, Masao; Matsuda, Atsuo; Iwamoto, Yoshiaki; Yoshikane, Toru; Koie, Toshio; Nakajima, Yoshiro

    1998-01-01

    Contaminated concrete wastes are crashed into granular concrete wastes having a successive grain size distribution. They are filled in a contamination processing vessel and made hardenable in the presence of a water-hardenable material in the granular concrete wastes. When underground water intrudes into the contamination processing vessel filled with the granular concrete wastes upon long-term storage, the underground water reacts with the water-hardenable material to be used for the solidification effect. Accordingly, leaching of contaminated materials due to intrusion of underground water can be suppressed. Since the concrete wastes have a successive grain size distribution, coarse grains can be used as coarse aggregates, medium grains can be used as fine aggregates and fine grains can be used as a solidifying material. Accordingly, the amount of wastes after processing can be remarkably reduced, with no supply of a solidifying material from outside. (T.M.)

  6. Permeability of Consolidated Incinerator Facility Wastes Stabilized with Portland Cement

    International Nuclear Information System (INIS)

    Walker, B.W.

    1999-01-01

    The Consolidated Incinerator Facility (CIF) at the Savannah River Site (SRS) burns low-level radioactive wastes and mixed wastes as method of treatment and volume reduction. The CIF generates secondary waste, which consists of ash and off-gas scrubber solution. Currently the ash is stabilized/solidified in the Ashcrete process. The scrubber solution (blowdown) is sent to the SRS Effluent Treatment Facility (ETF) for treatment as waste water. In the past, the scrubber solution was also stabilized/solidified in the Ashcrete process as blowcrete and will continue to be treated this way for listed waste burns and scrubber solution that do not meet the Effluent Treatment Facility (ETF) Waste Acceptance Criteria (WAC)

  7. 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

  8. Weldon Spring, Missouri, Raffinate Pits 1, 2, 3, and 4: Preliminary grout development screening studies for in situ waste immobilization

    International Nuclear Information System (INIS)

    McDaniel, E.W.; Gilliam, T.M.; Dole, L.R.; West, G.A.

    1987-04-01

    Results of Oak Ridge National Laboratory's initial support program to develop a preliminary grout formula to solidify in situ the Weldon Spring waste are presented. The screening study developed preliminary formulas based on a simulated composite waste and then tested the formulas on actual waste samples. Future data needs are also discussed. 1 ref., 6 figs., 9 tabs

  9. 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)

  10. 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

  11. Conditioning of uranium-containing technological radioactive waste

    International Nuclear Information System (INIS)

    Smodis, B.; Tavcar, G.; Stepisnik, M.; Pucelj, B.

    2006-01-01

    Conditioning of mostly liquid uranium containing technological radioactive waste emerging from the past research activities at the Jozef Stefan Institute is described. The waste was first thoroughly characterised, then the radionuclides present solidified by appropriate chemical treatment, and the final product separated and prepared for storage in compliance with the legislation. The activities were carried out within the recently renewed Hot Cells Facility of the Jozef Stefan Institute and the overall process resulted in substantial volume reduction of the waste initially present. (author)

  12. Leaching of nuclear power reactor waste forms

    International Nuclear Information System (INIS)

    Endo, L.S.; Villalobos, J.P.; Miyamoto, H.

    1987-01-01

    The leaching tests for immobilized power reactor wastes carried out at IPEN are described. These wastes forms consist mainly of spent resins and boric acid concentrates solidified in ordinary Portland cement. All tests were conducted according to the ISO and IAEA recommendations. Three years leaching results are reported. The cesium diffuvity coefficients determined out of these results are about 1 x 10 -8 cm 2 /s for boric acid waste form and 9 x 10 -9 cm 2 /s for ion-exchange resin waste. Strontium diffusivity coefficients found are about 3 x 10 -11 cm 2 /s and 9 x 10 -11 cm 2 /s respectively. (Author) [pt

  13. Process and device for processing radioactive wastes

    International Nuclear Information System (INIS)

    1974-01-01

    A method is described for processing liquid radioactive wastes. It includes the heating of the liquid wastes so that the contained liquids are evaporated and a practically anhydrous mass of solid particles inferior in volume to that of the wastes introduced is formed, then the transformation of the solid particles into a monolithic structure. This transformation includes the compressing of the particles and sintering or fusion. The solidifying agent is a mixture of polyethylene and paraffin wax or a styrene copolymer and a polyester resin. The device used for processing the radioactive liquid wastes is also described [fr

  14. Evolution of a Test Item

    Science.gov (United States)

    Spaan, Mary

    2007-01-01

    This article follows the development of test items (see "Language Assessment Quarterly", Volume 3 Issue 1, pp. 71-79 for the article "Test and Item Specifications Development"), beginning with a review of test and item specifications, then proceeding to writing and editing of items, pretesting and analysis, and finally selection of an item for a…

  15. Some techniques for the solidification of radioactive wastes in concrete

    International Nuclear Information System (INIS)

    Colombo, P.; Neilson, R. Jr.

    1976-06-01

    Some techniques for the solidification of radioactive wastes in concrete are discussed. The sources, storage, volume reduction, and solidification of liquid wastes at Brookhaven National Laboratory (BNL) using the cement-vermiculite process is described. Solid waste treatment, shipping containers, and off-site shipments of solid wastes at BNL are also considered. The properties of low-heat-generating, high-level wastes, simulating those in storage at the Savannah River Plant (SRP), solidified in concrete were determined. Polymer impregnation was found to further decrease the leachability and improve the durability of these concrete waste forms

  16. 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)

  17. 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.)

  18. Radioactive waste management

    International Nuclear Information System (INIS)

    Alfredson, P.G.; Levins, D.M.

    1975-08-01

    Present and future methods of managing radioactive wastes in the nuclear industry are reviewed. In the stages from uranium mining to fuel fabrication, the main purpose of waste management is to limit and control dispersal into the environment of uranium and its decay products, particularly radium and radon. Nuclear reactors produce large amounts of radioactivity but release rates from commercial power reactors have been low and well within legal limits. The principal waste from reprocessing is a high activity liquid containing essentially all the fission products along with the transuranium elements. Most high activity wastes are currently stored as liquids in tanks but there is agreement that future wastes must be converted into solids. Processes to solidify wastes have been demonstrated in pilot plant facilities in the United States and Europe. After solidification, wastes may be stored for some time in man-made structures at or near the Earth's surface. The best method for ultimate disposal appears to be placing solid wastes in a suitable geological formation on land. (author)

  19. Solid waste containing method and solid waste container

    International Nuclear Information System (INIS)

    Sawai, Takeshi.

    1997-01-01

    Solid wastes are filled in a sealed vessel, and support spacers are inserted to the gap between the inner wall of a vessel main body and the solid wastes. The solid wastes comprise shorn pieces (crushed pieces) of spent fuel rod cladding tubes, radioactively contaminated metal pieces and miscellaneous solids pressed into a disk-like shape. The sealed vessel comprises, for example, a stainless steel. The solid wastes are filled while being stacked in a plurality of stages. A solidifying filler is filled into the gap between the inner wall and the solid wastes in the vessel main body by way of an upper opening, and the upper opening is closed by a closing lid to provide an entirely sealed state. Alumina particles having high heat conductivity and excellent heat durability are used for the solid filler. It is preferable to fill an inert gas such as a dried nitrogen gas in the sealed vessel. (I.N.)

  20. 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.

  1. 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.

  2. Liquid emulsion scintillators which solidify for facile disposal

    International Nuclear Information System (INIS)

    O'Brien, R.E.; Krieger, J.K.

    1981-01-01

    A liquid organic scintillation cocktail is described which counts solutions of radiolabelled compounds containing up to ten % by volume of water with high efficiency and is readily polymerizable to a solid for easy disposal. The cocktail comprises a polymerizable organic solvent, a solubilizing agent, an intermediate solvent, and an organic scintillator. A method of disposing of liquid organic scintillation cocktail waste and a kit useful for practising the method are also described. (U.K.)

  3. Particle Engulfment and Pushing by Solidifying Interfaces (PEPSI)

    Science.gov (United States)

    Stefanescu, Doru Michael; Curreri, Peter A.; Juretsko, F.; Pang, H.; Phalnikar, R.

    1993-01-01

    The preliminary definition phase included the following actions: producing a science requiring document (draft), producing a science requirements document (preliminary), updating the flight program proposal, project review at NASA Marshall Space Flight Center, and research work as defined in the statement of work. The first three items of this plan have been delivered by the University of Alabama to NASA according to schedule. A project review meeting was held at MSFC on June 29, 1993. Consequently, this part of the report will address the results of the research work performed in the Solidification Laboratory at the University of Alabama during the first six months of the project.

  4. 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.

  5. Microstructure and mechanical properties of an Al–Mg alloy solidified under high pressures

    International Nuclear Information System (INIS)

    Jie, J.C.; Zou, C.M.; Brosh, E.; Wang, H.W.; Wei, Z.J.; Li, T.J.

    2013-01-01

    Highlights: •Al–42.2Mg alloy was solidified under pressures of 1, 2, and 3 GPa and the microstructure analyzed. •A thermodynamic calculation of the Al–Mg phase diagram at high pressures was performed. •The phase content changes from predominantly γ-Al 12 Mg 17 at 1 GPa to FCC solid solution at 3 GPa. •The β-Al 3 Mg 2 is predicted to remain stable at low temperatures but is not observed. •The alloy solidified at high pressure has remarkably enhanced ultimate tensile strength. -- Abstract: Phase formation, the microstructure and its evolution, and the mechanical properties of an Al–42.2 at.% Mg alloy solidified under high pressures were investigated. After solidification at pressures of 1 GPa and 2 GPa, the main phase is the γ phase, richer in Al than in equilibrium condition. When the pressure is further increased to 3 GPa, the main phase is the supersaturated Al(Mg) solid solution with Mg solubility up to 41.6 at.%. Unlike in similar alloys solidified at ambient pressure, the β phase does not appear. Calculated high-pressure phase diagrams of the Al–Mg system show that although the stability range of the β phase is diminished with pressure, it is still thermodynamically stable at room temperature. Hence, the disappearance of the β phase is interpreted as kinetic suppression, due to the slow diffusion rate at high pressures, which inhibits solid–solid reactions. The Al–42.2 at.% Mg alloy solidified under 3 GPa has remarkably enhanced ultimate tensile strength compared to the alloy solidified under normal atmospheric pressure

  6. SHIPPING OF RADIOACTIVE ITEMS

    CERN Multimedia

    TIS/RP Group

    2001-01-01

    The TIS-RP group informs users that shipping of small radioactive items is normally guaranteed within 24 hours from the time the material is handed in at the TIS-RP service. This time is imposed by the necessary procedures (identification of the radionuclides, determination of dose rate and massive objects require a longer procedure and will therefore take longer.

  7. Spare Items validation

    International Nuclear Information System (INIS)

    Fernandez Carratala, L.

    1998-01-01

    There is an increasing difficulty for purchasing safety related spare items, with certifications by manufacturers for maintaining the original qualifications of the equipment of destination. The main reasons are, on the top of the logical evolution of technology, applied to the new manufactured components, the quitting of nuclear specific production lines and the evolution of manufacturers quality systems, originally based on nuclear codes and standards, to conventional industry standards. To face this problem, for many years different Dedication processes have been implemented to verify whether a commercial grade element is acceptable to be used in safety related applications. In the same way, due to our particular position regarding the spare part supplies, mainly from markets others than the american, C.N. Trillo has developed a methodology called Spare Items Validation. This methodology, which is originally based on dedication processes, is not a single process but a group of coordinated processes involving engineering, quality and management activities. These are to be performed on the spare item itself, its design control, its fabrication and its supply for allowing its use in destinations with specific requirements. The scope of application is not only focussed on safety related items, but also to complex design, high cost or plant reliability related components. The implementation in C.N. Trillo has been mainly curried out by merging, modifying and making the most of processes and activities which were already being performed in the company. (Author)

  8. Selecting Lower Priced Items.

    Science.gov (United States)

    Kleinert, Harold L.; And Others

    1988-01-01

    A program used to teach moderately to severely mentally handicapped students to select the lower priced items in actual shopping activities is described. Through a five-phase process, students are taught to compare prices themselves as well as take into consideration variations in the sizes of containers and varying product weights. (VW)

  9. 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.

  10. The Role of Item Models in Automatic Item Generation

    Science.gov (United States)

    Gierl, Mark J.; Lai, Hollis

    2012-01-01

    Automatic item generation represents a relatively new but rapidly evolving research area where cognitive and psychometric theories are used to produce tests that include items generated using computer technology. Automatic item generation requires two steps. First, test development specialists create item models, which are comparable to templates…

  11. Item information and discrimination functions for trinary PCM items

    NARCIS (Netherlands)

    Akkermans, Wies; Muraki, Eiji

    1997-01-01

    For trinary partial credit items the shape of the item information and the item discrimination function is examined in relation to the item parameters. In particular, it is shown that these functions are unimodal if δ2 – δ1 < 4 ln 2 and bimodal otherwise. The locations and values of the maxima are

  12. 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

  13. Summary: special waste form lysimeters - arid program

    International Nuclear Information System (INIS)

    Skaggs, R.L.; Walter, M.B.

    1987-01-01

    The purpose of the Special Waste Form Lysimeters - Arid Program is to determine the performance of solidified commercial low-level waste forms using a field-scale lysimeter facility constructed for measuring the release and migration of radionuclides from the waste forms. The performance of these waste forms, as measured by radionuclide concentrations in lysimeter effluent, will be compared to that predicted by laboratory characterization of the waste forms. Waste forms being tested include nuclear power reactor waste streams that have been solidified in cement, Dow polymer, and bitumen. To conduct the field leaching experiments a lysimeter facility was built to measure leachate under actual environmental conditions. Field-scale samples of waste were buried in lysimeters equipped to measure water balance components, effluent radionuclide concentrations, and to a limited extent, radionuclide concentrations in lysimeter soil samples. The waste forms are being characterized by standard laboratory leach tests to obtain estimates of radionuclide release. These estimates will be compared to leach rates observed in the field. Adsorption studies are being conducted to determine the amount of contaminant available for transport after the release. Theoretical solubility calculations will also be performed to investigate whether common solid phases could be controlling radionuclide release. 4 references, 8 figures, 1 table

  14. Certain aspects of leaching kinetics of solidified ''radioactive wastes'' - Laboratory studies

    International Nuclear Information System (INIS)

    Amarantos, S.G.; Petropoulos, J.H.

    1981-01-01

    A laboratory study of the kinetics of leaching of Cs + and Sr ++ incorporated as sulphates in cement (Portland type) or base asphalt with distilled water as leachant is reported. The comparison of different leaching methods (including stagnant or stirred leachant with periodic renewal and a new method of ''continuous leachant renewal'') and the study of the effect of temperature were among the particular objectives of this work. The results for the cemented salts indicate that the new method of ''continuous leachant renewal'' is more efficient than the others. Is has been also found that the other leaching methods, including the widely used stagnant, periodically renewed leachant, can be seriously affected in certain circumstances by factors which have hitherto been left uncontrolled. An observed sharp diminution in the elution rate of Sr ++ from cement samples appears to be attributable to the action of atmospheric CO 2 . It has been observed that temperature affects the acceleration in the earlier and later stages of cement-Cs 2 SO 4 leaching. (T.A.)

  15. 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

  16. Characterization of voic volume VOC concentration in vented TRU waste drums. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Liekhus, K.J.

    1994-12-01

    A test program has been conducted at the Idaho National Engineering Laboratory to demonstrate that the concentration of volatile organic compounds within the innermost layer of confinement in a vented waste drum can be estimated using a model incorporating diffusion and permeation transport principles and limited waste drum sampling data. This final report summarizes the experimental measurements and model predictions for transuranic waste drums containing solidified sludges and solid waste.

  17. Item Banking with Embedded Standards

    Science.gov (United States)

    MacCann, Robert G.; Stanley, Gordon

    2009-01-01

    An item banking method that does not use Item Response Theory (IRT) is described. This method provides a comparable grading system across schools that would be suitable for low-stakes testing. It uses the Angoff standard-setting method to obtain item ratings that are stored with each item. An example of such a grading system is given, showing how…

  18. 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

  19. SHIPPING OF RADIOACTIVE ITEMS

    CERN Multimedia

    TIS/RP Group

    2001-01-01

    The TIS-RP group informs users that shipping of small radioactive items is normally guaranteed within 24 hours from the time the material is handed in at the TIS-RP service. This time is imposed by the necessary procedures (identification of the radionuclides, determination of dose rate, preparation of the package and related paperwork). Large and massive objects require a longer procedure and will therefore take longer.

  20. Radioactive waste in Federal Germany

    International Nuclear Information System (INIS)

    Brennecke, P.; Schumacher, J.; Warnecke, E.

    1988-01-01

    The Physikalisch-Technische Bundesanstalt (PTB) is responsible for the long-term storage and disposal of radioactive waste according to the Federal Atomic Energy Act. On behalf of the Federal Minister of the Environment, Nature Conservation and Nuclear Safety, since 1985, the PTB has been carrying out annual inquiries into the amounts of radioactive waste produced in the Federal Republic of Germany. Within the scope of this inquiry performed for the preceding year, the amounts of unconditioned and conditioned waste are compiled on a producer- and plant-specific basis. On the basis of the inquiry for 1986 and of data presented to the PTB by the waste producers, future amounts of radioactive waste have been estimated up to the year 2000. The result of this forecast is presented. In the Federal Republic of Germany two sites are under consideration for disposal of radioactive waste. In the abandoned Konrad iron mine in Salzgitter-Bleckenstedt it is intended to dispose of such radioactive waste which has a negligible thermal influence upon the host rock. The Gorleben salt dome is being investigated for its suitability for the disposal of all kinds of solid and solidified radioactive wastes, especially of heat-generating waste. Comparing the estimated amount of radioactive wastes with the capacity of both repositories it may be concluded that the Konrad and Gorleben repositories will provide sufficient capacity to ensure the disposal of all kinds of radioactive waste on a long-term basis in the Federal Republic of Germany. 1 fig., 2 tabs

  1. Treatment and disposal of radioactive wastes. [Important new developments

    Energy Technology Data Exchange (ETDEWEB)

    Krause, H; Starch, M

    1977-04-01

    Out of an abundance of contributions on this subject published during the last 2 years, only the most important new developments are dealt with here. All over the world known methods have been improved, new ones developed especially in the field of solidifying highly active liquid wastes. For general information, surveys and conference proceedings are listed as well.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  6. 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)

  7. Method of processing decontaminating liquid waste

    International Nuclear Information System (INIS)

    Kusaka, Ken-ichi

    1989-01-01

    When decontaminating liquid wastes are processed by ion exchange resins, radioactive nuclides, metals, decontaminating agents in the liquid wastes are captured in the ion exchange resins. When the exchange resins are oxidatively deomposed, most of the ingredients are decomposed into water and gaseous carbonic acid and discharged, while sulfur ingredient in the resins is converted into sulfuric acid. In this case, even less oxidizable ingredients in the decontaminating agent made easily decomposable by oxidative decomposition together with the resins. The radioactive nuclides and a great amount of iron dissolved upon decontamination in the liquid wastes are dissolved in sulfuric acid formed. When the sulfuric acid wastes are nuetralized with sodium hydroxide, since they are formed into sodium sulfate, which is most popular as wastes from nuclear facilities, they can be condensated and solidified by existent waste processing systms to thereby facilitate the waste processing. (K.M.)

  8. Design and construction of the defense waste processing facility project at the Savannah River Plant

    International Nuclear Information System (INIS)

    Baxter, R.G.

    1986-01-01

    The Du Pont Company is building for the Department of Energy a facility to vitrify high-level radioactive waste at the Savannah River Plant (SRP) near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes by immobilizing the waste in Processing Facility (DWPF) will solidify existing and future radioactives wastes by immobilizing the waste in borosilicate glass contained in stainless steel canisters. The canisters will be sealed, decontaminated and stored, prior to emplacement in a federal repository. At the present time, engineering and design is 90% complete, construction is 25% complete, and radioactive processing in the $870 million facility is expected to begin by late 1989. This paper describes the SRP waste characteristics, the DWPF processing, building and equipment features, and construction progress of the facility

  9. 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

  10. 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.

  11. Characterizing cemented TRU waste for RCRA hazardous constituents

    International Nuclear Information System (INIS)

    Yeamans, D.R.; Betts, S.E.; Bodenstein, S.A.

    1996-01-01

    Los Alamos National Laboratory (LANL) has characterized drums of solidified transuranic (TRU) waste from four major waste streams. The data will help the State of New Mexico determine whether or not to issue a no-migration variance of the Waste Isolation Pilot Plant (WIPP) so that WIPP can receive and dispose of waste. The need to characterize TRU waste stored at LANL is driven by two additional factors: (1) the LANL RCRA Waste Analysis Plan for EPA compliant safe storage of hazardous waste; (2) the WIPP Waste Acceptance Criteria (WAC) The LANL characterization program includes headspace gas analysis, radioassay and radiography for all drums and solids sampling on a random selection of drums from each waste stream. Data are presented showing that the only identified non-metal RCRA hazardous component of the waste is methanol

  12. 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)

  13. Immobilization of organic liquid wastes

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1985-01-01

    This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a a FFTF fuels reprocessing line. An incineration system already on-hand was determined to be too costly to operate for the 100 to 400 gallons per year organic liquid. Organic test liquids were dispersed into an aqueous phosphate liquid using an emulsifier. A total of 109 gallons of potential and radioactive aqueous immiscible organic liquid wastes from Hanford 300 Area operations were solidified with portland cement and disposed of as solid waste during a 3-month test program with in-drum mixers. Waste packing efficiencies varied from 32 to 40% and included pump oils, mineral spirits, and TBP-NPH type solvents

  14. Plastic solidification of radioactive wastes

    International Nuclear Information System (INIS)

    Moriyama, Noboru

    1981-01-01

    Over 20 years have elapsed after the start of nuclear power development, and the nuclear power generation in Japan now exceeds the level of 10,000 MW. In order to meet the energy demands, the problem of the treatment and disposal of radioactive wastes produced in nuclear power stations must be solved. The purpose of the plastic solidification of such wastes is to immobilize the contained radionuclides, same as other solidification methods, to provide the first barrier against their move into the environment. The following matters are described: the nuclear power generation in Japan, the radioactive wastes from LWR plants, the position of plastic solidification, the status of plastic solidification in overseas countries and in Japan, the solidification process for radioactive wastes with polyethylene, and the properties of solidified products, and the leachability of radionuclides in asphalt solids. (J.P.N.)

  15. Method of processing liquid waste containing fission product

    International Nuclear Information System (INIS)

    Funabashi, Kiyomi; Kawamura, Fumio; Matsuda, Masami; Komori, Itaru; Miura, Eiichi.

    1988-01-01

    Purpose: To prepare solidification products of low surface dose by removing cesium which is main radioactive nuclides from re-processing plants. Method: Liquid wastes containing a great amount of fission products are generated accompanying the reprocessing for spent nuclear fuels. After pH adjustment, the liquid wastes are sent to a concentrator to concentrate the dissolved ingredients. The concentrated liquid wastes are pumped to an adsorption tower in which radioactive cesium contributing much to the surface dose is removed. Then, the liquid wastes are sent by way of a surge tank to a mixing tank, in which they are mixed under stirring with solidifying agents such as cements. Then, the mixture is filled in a drum-can and solidified. According to this invention, since radioactive cesium is removed before solidification, it is possible to prepare solidification products at low surface dose and facilitate the handling of the solidification products. (Horiuchi, T.)

  16. Survey of concrete waste forms

    International Nuclear Information System (INIS)

    Moore, J.G.

    1981-01-01

    The incorporation of radioactive waste in cement has been widely studied for many years. It has been routinely used at nuclear research and production sites for some types of nuclear waste for almost three decades and at power reactor plants for nearly two decades. Cement has many favorable characteristics that have contributed to its popularity. It is a readily available material and has not required complex and/or expensive equipment to solidify radioactive waste. The resulting solid products are noncombustible, strong, radiation resistant, and have reasonable chemical and thermal stability. As knowledge increased on the possible dangers from radioactive waste, requirements for waste fixation became more stringent. A brief survey of some of the research efforts used to extend and improve cementitious waste hosts to meet these requirements is given in this paper. Selected data are presented from the rather extensive study of the applicability of concrete as a waste form for Savannah River defense waste and the use of polymer impregnation to reduce the leachability and improve the durability of such waste forms. Hot-pressed concretes that were developed as prospective host solids for high-level wastes are described. Highlights are given from two decades of research on cementitious waste forms at Oak Ridge National Laboratory. The development of the hydrofracture process for the disposal of all locally generated radioactive waste led to a process for the disposal of I-129 and to the current research on the German in-situ solidification process for medium-level waste and the Oak Ridge FUETAP process for all classes of waste including commercial and defense high-level wastes. Finally, some of the more recent ORNL concepts are presented for the use of cement in the disposal of inorganic and biological sludges, waste inorganic salts, trash, and krypton

  17. Regulation of radioactive waste management

    International Nuclear Information System (INIS)

    2002-01-01

    This bulletin contains information about activities of the Nuclear Regulatory Authority of the Slovak Republic (UJD). In this leaflet the regulation of radioactive waste management of the UJD are presented. Radioactive waste (RAW) is the gaseous, liquid or solid material that contains or is contaminated with radionuclides at concentrations or activities greater than clearance levels and for which no use is foreseen. The classification of radioactive waste on the basis of type and activity level is: - transition waste; - short lived low and intermediate level waste (LlLW-SL); - long lived low and intermediate level waste (LlLW-LL); - high level waste. Waste management (in accordance with Act 130/98 Coll.) involves collection, sorting, treatment, conditioning, transport and disposal of radioactive waste originated by nuclear facilities and conditioning, transport to repository and disposal of other radioactive waste (originated during medical, research and industrial use of radioactive sources). The final goal of radioactive waste management is RAW isolation using a system of engineered and natural barriers to protect population and environment. Nuclear Regulatory Authority of the Slovak Republic regulates radioactive waste management in accordance with Act 130/98 Coll. Inspectors regularly inspect and evaluate how the requirements for nuclear safety at nuclear facilities are fulfilled. On the basis of safety documentation evaluation, UJD issued permission for operation of four radioactive waste management facilities. Nuclear facility 'Technologies for treatment and conditioning contains bituminization plants and Bohunice conditioning centre with sorting, fragmentation, evaporation, incineration, supercompaction and cementation. Final product is waste package (Fibre reinforced container with solidified waste) acceptable for near surface repository in Mochovce. Republic repository in Mochovce is built for disposal of short lived low and intermediate level waste. Next

  18. Gender-Based Differential Item Performance in Mathematics Achievement Items.

    Science.gov (United States)

    Doolittle, Allen E.; Cleary, T. Anne

    1987-01-01

    Eight randomly equivalent samples of high school seniors were each given a unique form of the ACT Assessment Mathematics Usage Test (ACTM). Signed measures of differential item performance (DIP) were obtained for each item in the eight ACTM forms. DIP estimates were analyzed and a significant item category effect was found. (Author/LMO)

  19. 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)

  20. Item-focussed Trees for the Identification of Items in Differential Item Functioning.

    Science.gov (United States)

    Tutz, Gerhard; Berger, Moritz

    2016-09-01

    A novel method for the identification of differential item functioning (DIF) by means of recursive partitioning techniques is proposed. We assume an extension of the Rasch model that allows for DIF being induced by an arbitrary number of covariates for each item. Recursive partitioning on the item level results in one tree for each item and leads to simultaneous selection of items and variables that induce DIF. For each item, it is possible to detect groups of subjects with different item difficulties, defined by combinations of characteristics that are not pre-specified. The way a DIF item is determined by covariates is visualized in a small tree and therefore easily accessible. An algorithm is proposed that is based on permutation tests. Various simulation studies, including the comparison with traditional approaches to identify items with DIF, show the applicability and the competitive performance of the method. Two applications illustrate the usefulness and the advantages of the new method.

  1. Draft of regulations for road transport of radioactive wastes

    International Nuclear Information System (INIS)

    Gese, J.; Zizka, B.

    1979-06-01

    A draft regulation is presented for the transport of solid and solidified radioactive wastes from nuclear power plants. The draft takes into consideration dosimetric, safety and fire-fighting directives, transport organization, anticipated amounts of radioactive wastes, characteristics of containers, maintenance of vehicles, and equipment of vehicles and personnel. The draft is based on the provisional regulations governing the transport on public roads issued in 1973, valid directives, decrees, acts and standards, and complies with 1973 IAEA requirements. (J.P.)

  2. Item validity vs. item discrimination index: a redundancy?

    Science.gov (United States)

    Panjaitan, R. L.; Irawati, R.; Sujana, A.; Hanifah, N.; Djuanda, D.

    2018-03-01

    In several literatures about evaluation and test analysis, it is common to find that there are calculations of item validity as well as item discrimination index (D) with different formula for each. Meanwhile, other resources said that item discrimination index could be obtained by calculating the correlation between the testee’s score in a particular item and the testee’s score on the overall test, which is actually the same concept as item validity. Some research reports, especially undergraduate theses tend to include both item validity and item discrimination index in the instrument analysis. It seems that these concepts might overlap for both reflect the test quality on measuring the examinees’ ability. In this paper, examples of some results of data processing on item validity and item discrimination index were compared. It would be discussed whether item validity and item discrimination index can be represented by one of them only or it should be better to present both calculations for simple test analysis, especially in undergraduate theses where test analyses were included.

  3. Thermomechanical effects of the salt rock on the solidified waste product during ultimate stoage of radioactive waste

    International Nuclear Information System (INIS)

    Schoen, R.

    1981-01-01

    The thermal stresses in the salt to be expected in the elastic case are very much reduced by the viscous behavior of the salt rock. The occurrence of tensile stresses may be prevented by reducing the differential temperatures by means of a decrease of the mould heat rate and/or the mechanical behavior of the glass as well as design measures. As far as the mechanical aspect is concerned thicker coverings have no positive effect on the stress in the glass. In the course of time the three principal stresses in the salt rock are matching. At the terminal point of the reference calculations these stresses amount to 12.5 MPa and 15 MPa in the horizontal and vertical direction respectively. (DG) [de

  4. 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.)

  5. Waste isolation facility description: bedded salt

    Energy Technology Data Exchange (ETDEWEB)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria. (LK)

  6. Waste isolation facility description: bedded salt

    International Nuclear Information System (INIS)

    1976-09-01

    The waste isolation facility is designed to receive and store three basic types of solidified wastes: high-level wastes, intermediate level high-gamma transuranic waste, and low-gamma transuranic wastes. The facility under consideration in this report is designed for bedded salt at a depth of approximately 1800 ft. The present design for the facility includes an area which would be used initially as a pilot facility to test the viability of the concept, and a larger facility which would constitute the final storage area. The total storage area in the pilot facility is planned to be 77 acres and in the fuel facility 1601 acres. Other areas for shaft operations and access would raise the overall size of the total facility to slightly less than 2,000 acres. The following subjects are discussed in detail: surface facilities, shaft design and characteristics, design and construction of the underground waste isolation facility, ventilation systems, and design requirements and criteria

  7. Significance of chemotoxic admixtures in radioactive wastes

    International Nuclear Information System (INIS)

    Merz, E.R.

    1989-01-01

    The double hazard potential of mixed wastes is characterized by several criteria: radioactivity on the one hand, and chemical toxicity, in flammability, corrosiveness as well as chemical reactivity on the other. The author argues that mixed wastes assigned for ultimate disposal should therefore be thoroughly detoxified, inertized, or mineralized, prior to conditioning and packaging. Strategies and techniques are presented which ensure the elimination of hazardous organic chemicals and minimizing waste volumes to be disposed of. Advantage can be taken of mixing mineralized filter dusts, arising in the combustion of hazardous chemical wastes with low-activity inertized radioactive wastes as a solidifying reagent. Simultaneous geological disposal of such mixed waste is feasible without any drawbacks

  8. Summary of Waste Calcination at INTEC

    Energy Technology Data Exchange (ETDEWEB)

    O' Brien, Barry Henry; Newby, Bill Joe

    2000-10-01

    Fluidized-bed calcination at the Idaho Nuclear Technologies and Engineering Center (INTEC, formally called the Idaho Chemical Processing Plant) has been used to solidify acidic metal nitrate fuel reprocessing and incidental wastes wastes since 1961. A summary of waste calcination in full-scale and pilot plant calciners has been compiled for future reference. It contains feed compositions and operating conditions for all the processing campaigns for the original Waste Calcining Facility (WCF), the New Waste Calcining Facility (NWCF) started up in 1982, and numerous small scale pilot plant tests for various feed types. This summary provides a historical record of calcination at INTEC, and will be useful for evaluating calcinability of future wastes.

  9. Method of processing radioactive liquid waste

    International Nuclear Information System (INIS)

    Hasegawa, Akira; Kuribayashi, Hiroshi; Soda, Kenzo; Mihara, Shigeru.

    1988-01-01

    Purpose: To obtain satisfactory plastic solidification products rapidly and smoothly by adding oxidizers to radioactive liquid wastes. Method: Sulfuric acid, etc. are added to radioactive liquid wastes to adjust the pH value of the liquid wastes to less than 3.0. Then, ferrous sulfates are added such that the iron concentration in the liquid wastes is 100 mg/l. Then, after adjusting pH suitably to the drying powderization by adding alkali such as hydroxide, the liquid wastes are dried and powderized. The resultant powder is subjected to plastic solidification by using polymerizable liquid unsaturated polyester resins as the solidifying agent. The thus obtained solidification products are stable in view of the physical property such as strength or water proofness, as well as stable operation is possible even for those radioactive liquid wastes in which the content ingredients are unknown. (Takahashi, M.)

  10. Research and development on radioactive waste management and storage: Third annual progress report (1982) of the European Community programme 1980-1984

    International Nuclear Information System (INIS)

    Anon.

    1984-01-01

    This book examines the European Community's program for nuclear waste management and storage. Topics considered include the characterization of conditioned low and medium activity waste forms, conditioning of high activity solid waste, treatment and conditioning processes for low and medium activity liquid waste, processing of alpha-contaminated waste, testing and evaluation of solidified high activity waste forms, immobilization and storage of gaseous waste, shallow land burial of solid low activity waste, storage and disposal in geological formations, and the performance and safety evaluation of radioactive waste disposed in geological formations

  11. 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.

  12. 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.

  13. 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%.

  14. 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%.

  15. Liquid secondary waste: Waste form formulation and qualification

    Energy Technology Data Exchange (ETDEWEB)

    Cozzi, A. D. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Dixon, K. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Hill, K. A. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); Nichols, R. L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL)

    2017-07-31

    The Hanford Site Effluent Treatment Facility (ETF) currently treats aqueous waste streams generated during site cleanup activities. When the Hanford Tank Waste Treatment and Immobilization Plant (WTP) begins operations, including Direct Feed Low Activity Waste (DFLAW) vitrification, a liquid secondary waste (LSW) stream from the WTP will need to be treated. The volume of effluent for treatment at the ETF will increase significantly. The powdered salt waste form produced by the ETF will be replaced by a stabilized solidified waste form for disposal in Hanford’s Integrated Disposal Facility (IDF). Washington River Protection Solutions is implementing a Secondary Liquid Waste Immobilization Technology Development Plan to address the technology needs for a waste form and solidification process to treat the increased volume of waste planned for disposal at the IDF. Waste form testing to support this plan is composed of work in the near term to provide data as input to a performance assessment (PA) for Hanford’s IDF. In 2015, three Hanford Liquid Secondary Waste simulants were developed based on existing and projected waste streams. Using these waste simulants, fourteen mixes of Hanford Liquid Secondary Waste were prepared and tested varying the waste simulant, the water-to-dry materials ratio, and the dry materials blend composition.1 In FY16, testing was performed using a simulant of the EMF process condensate blended with the caustic scrubber—from the Low Activity Waste (LAW) melter—, processed through the ETF. The initial EMF-16 simulant will be based on modeling efforts performed to determine the mass balance of the ETF for the DFLAW.2 The compressive strength of all of the mixes exceeded the target of 3.4 MPa (500 psi) to meet the requirements identified as potential IDF Waste Acceptance Criteria in Table 1 of the Secondary Liquid Waste Immobilization Technology Development Plan.3 The hydraulic properties of the waste forms tested (hydraulic conductivity

  16. 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)

  17. 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.

  18. 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.

  19. 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

  20. Treatment of radioactive metallic waste by the electro-slag melting method

    International Nuclear Information System (INIS)

    Ochiai, Atsuhiro; Nagura, Kanetake; Noura, Tsuyoshi

    1983-01-01

    The applicability of the electro-slag melting method for treating plutonuim contaminated metallic waste was studied. A 100kg test furnace was built and simulated metallic waste was melted and solidified in this furnace. Waste volume was reduced to 1/25 with a decontamination factor of 25 and the slag and the copper mold are repeatedly usable. The process is expected to be employed in the project of PWTF (Plutonium contaminated Wate Treatment Facilities). (author)

  1. Thermal Predictions of the Cooling of Waste Glass Canisters

    Energy Technology Data Exchange (ETDEWEB)

    Donna Post Guillen

    2014-11-01

    Radioactive liquid waste from five decades of weapons production is slated for vitrification at the Hanford site. The waste will be mixed with glass forming additives and heated to a high temperature, then poured into canisters within a pour cave where the glass will cool and solidify into a stable waste form for disposal. Computer simulations were performed to predict the heat rejected from the canisters and the temperatures within the glass during cooling. Four different waste glass compositions with different thermophysical properties were evaluated. Canister centerline temperatures and the total amount of heat transfer from the canisters to the surrounding air are reported.

  2. 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

  3. 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

  4. 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.

  5. 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.

  6. Properties and solidification of decontamination wastes

    International Nuclear Information System (INIS)

    Davis, M.S.; Piciulo, P.L.; Bowerman, B.S.; Adams, J.W.; Milian, L.

    1983-01-01

    LWRs will require one or more chemical decontaminations to achieve their designed lifetimes. Primary system decontamination is designed to lower radiation fields in areas where plant maintenance personnel must work. Chemical decontamination methods are either hard (concentrated chemicals, approximately 5 to 25 weight percent) or soft (dilute chemicals less than 1 percent by weight). These methods may have different chemical reagents, some tailor-made to the crud composition and many methods are and will be proprietary. One factor common to most commercially available processes is the presence of organic acids and chelates. These types of organic reagents are known to enhance the migration of radionuclides after disposal in a shallow land burial site. The NRC sponsors two programs at Brookhaven National Laboratory that are concerned with the management of decontamination wastes which will be generated by the full system decontamination of LWRs. These two programs focus on potential methods for degrading or converting decontamination wastes to more acceptable forms prior to disposal and the impact of disposing of solidified decontamination wastes. The results of the solidification of simulated decontamination resin wastes will be presented. Recent results on combustion of simulated decontamintion wastes will be described and procedures for evaluating the release of decontamination reagents from solidified wastes will be summarized

  7. Problems with the factor analysis of items: Solutions based on item response theory and item parcelling

    Directory of Open Access Journals (Sweden)

    Gideon P. De Bruin

    2004-10-01

    Full Text Available The factor analysis of items often produces spurious results in the sense that unidimensional scales appear multidimensional. This may be ascribed to failure in meeting the assumptions of linearity and normality on which factor analysis is based. Item response theory is explicitly designed for the modelling of the non-linear relations between ordinal variables and provides a strong alternative to the factor analysis of items. Items may also be combined in parcels that are more likely to satisfy the assumptions of factor analysis than do the items. The use of the Rasch rating scale model and the factor analysis of parcels is illustrated with data obtained with the Locus of Control Inventory. The results of these analyses are compared with the results obtained through the factor analysis of items. It is shown that the Rasch rating scale model and the factoring of parcels produce superior results to the factor analysis of items. Recommendations for the analysis of scales are made. Opsomming Die faktorontleding van items lewer dikwels misleidende resultate op, veral in die opsig dat eendimensionele skale as meerdimensioneel voorkom. Hierdie resultate kan dikwels daaraan toegeskryf word dat daar nie aan die aannames van lineariteit en normaliteit waarop faktorontleding berus, voldoen word nie. Itemresponsteorie, wat eksplisiet vir die modellering van die nie-liniêre verbande tussen ordinale items ontwerp is, bied ’n aantreklike alternatief vir die faktorontleding van items. Items kan ook in pakkies gegroepeer word wat meer waarskynlik aan die aannames van faktorontleding voldoen as individuele items. Die gebruik van die Rasch beoordelingskaalmodel en die faktorontleding van pakkies word aan die hand van data wat met die Lokus van Beheervraelys verkry is, gedemonstreer. Die resultate van hierdie ontledings word vergelyk met die resultate wat deur ‘n faktorontleding van die individuele items verkry is. Die resultate dui daarop dat die Rasch

  8. Method of processing radioactive wastes

    International Nuclear Information System (INIS)

    Funabashi, Kiyomi; Sugimoto, Yoshikazu; Kikuchi, Makoto; Yusa, Hideo.

    1979-01-01

    Purpose: To obtain solidified radioactive wastes at high packing density by packing radioactive waste pellets in a container and then packing and curing a thermosetting resin therein. Method: Radioactive liquid wastes are dried into power and subjected to compression molding. The pellets thus obtained are supplied in a predetermined amount from the hopper to the inside of a drum can. Then, thermosetting plastic and a curing agent are filled in the drum can. Gas between the pellets is completely expelled by the intrusion of the thermosetting resin and the curing agent among the pellets. Thereafter, the drum can is heated by a heater and curing is effected. After the curing, the drum can is sealed. (Kawakami, Y.)

  9. 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

  10. 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

  11. Method for solidification of radioactive iodine-containing solid wastes

    International Nuclear Information System (INIS)

    Ozawa, Yoshihiro; Funabashi, Kiyomi; Uetake, Naoto.

    1987-01-01

    Purpose: To process radioactive iodine containing solid wastes as non-leaching solidified wastes with no risk of iodine release. Method: It has been known for the thermal stability of CuI, PbI 2 or adsorbents containing the same that they do not release iodine in an inert gas atmosphere or in a reducing atmosphere at a temperature lower than 480 deg C. In view of the above, adsorbents containing iodine in the chemical form of CuI or PbI 2 , or CuI or powdery PbI 2 per se are sealed and solidified into low melting glass at a temperature of lower than 480 deg C at which no iodine release occurs in a non-oxidative atmosphere. Since the products are vitrified wastes, they scarcely show leaching property and are excellent in durability and stability. (Takahashi, M.)

  12. Solidification of ion exchange resin wastes in hydraulic cement

    International Nuclear Information System (INIS)

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

    1982-01-01

    Work has been conducted to investigate the solidification of ion exchange resin wastes with portland cements. These efforts have been directed toward the development of acceptable formulations for the solidification of ion exchange resin wastes and the characterization of the resultant waste forms. This paper describes formulation development work and defines acceptable formulations in terms of ternary phase compositional diagrams. The effects of cement type, resin type, resin loading, waste/cement ratio and water/cement ratio are described. The leachability of unsolidified and solidified resin waste forms and its relationship to full-scale waste form behavior is discussed. Gamma irradiation was found to improve waste form integrity, apparently as a result of increased resin crosslinking. Modifications to improve waste form integrity are described. 3 tables

  13. Processing method for cleaning water waste from cement kneader

    International Nuclear Information System (INIS)

    Soda, Kenzo; Fujita, Hisao; Nakajima, Tadashi.

    1990-01-01

    The present invention concerns a method of processing cleaning water wastes from a cement kneader in a case of processing liquid wastes containing radioactive wastes or deleterious materials such as heavy metals by means of cement solidification. Cleaning waste wastes from the kneader are sent to a cleaning water waste tank, in which gentle stirring is applied near the bottom and sludges are retained so as not to be coagulated. Sludges retained at the bottom of the cleaning water waste tank are sent after elapse of a predetermined time and then kneaded with cements. Thus, since the sludges in the cleaning water are solidified with cement, inhomogenous solidification products consisting only of cleaning sludges with low strength are not formed. The resultant solidification product is homogenous and the compression strength thereof reaches such a level as capable of satisfying marine disposal standards required for the solidification products of radioactive wastes. (I.N.)

  14. ITEM LEVEL DIAGNOSTICS AND MODEL - DATA FIT IN ITEM ...

    African Journals Online (AJOL)

    Global Journal

    Item response theory (IRT) is a framework for modeling and analyzing item response ... data. Though, there is an argument that the evaluation of fit in IRT modeling has been ... National Council on Measurement in Education ... model data fit should be based on three types of ... prediction should be assessed through the.

  15. Item Response Data Analysis Using Stata Item Response Theory Package

    Science.gov (United States)

    Yang, Ji Seung; Zheng, Xiaying

    2018-01-01

    The purpose of this article is to introduce and review the capability and performance of the Stata item response theory (IRT) package that is available from Stata v.14, 2015. Using a simulated data set and a publicly available item response data set extracted from Programme of International Student Assessment, we review the IRT package from…

  16. MIMIC Methods for Assessing Differential Item Functioning in Polytomous Items

    Science.gov (United States)

    Wang, Wen-Chung; Shih, Ching-Lin

    2010-01-01

    Three multiple indicators-multiple causes (MIMIC) methods, namely, the standard MIMIC method (M-ST), the MIMIC method with scale purification (M-SP), and the MIMIC method with a pure anchor (M-PA), were developed to assess differential item functioning (DIF) in polytomous items. In a series of simulations, it appeared that all three methods…

  17. Formulation development for PREPP concreted waste forms

    International Nuclear Information System (INIS)

    Neilson, R.M. Jr.; Welch, J.M.

    1984-05-01

    Analysis of variance and logistic regression techniques have been used to develop models describing the effects of formulation variables and their interactions on compressive strength, solidification, free-standing water, and workability of hydraulic cement grouts incorporating simulated Process Experimental Pilot Plant (PREPP) wastes. These models provide the basis for specifications of grout formulations to solidify these wastes. The experimental test matrix, formulation preparation, and test methods employed are described. The development of analytical models for formulation behavior and the conclusions drawn regarding appropriate formulation variable ranges are discussed. 13 references, 9 figures, 15 tables

  18. 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

  19. Prediction of waste glass melt rates

    International Nuclear Information System (INIS)

    Lee, L.

    1987-01-01

    Under contract to the Department of Energy, the Du Pont Company has begun construction of a Defense Waste Processing Facility to immobilize radioactive wastes now stored as liquids at the Department of Energy's Savannah River Plant. The immobilization process solidifies waste sludge by vitrification into a leach-resistant borosilicate glass. Development of this process has been the responsibility of the Savannah River Laboratory. As part of the development, a simple model was developed to predict the melt rates for the waste glass melter. This model is based on an energy balance for the cold cap and gives very good agreement with melt rate data obtained from experimental campaigns in smaller scale waste glass melters

  20. Method for treating waste containing stainless steel

    International Nuclear Information System (INIS)

    Kujawa, S.T.; Battleson, D.M.; Rademacher, E.L. Jr.; Cashell, P.V.; Filius, K.D.; Flannery, P.A.; Whitworth, C.G.

    1999-01-01

    A centrifugal plasma arc furnace is used to vitrify contaminated soils and other waste materials. An assessment of the characteristics of the waste is performed prior to introducing the waste into the furnace. Based on the assessment, a predetermined amount of iron is added to each batch of waste. The waste is melted in an oxidizing atmosphere into a slag. The added iron is oxidized into Fe 3 O 4 . Time of exposure to oxygen is controlled so that the iron does not oxidize into Fe 2 O 3 . Slag in the furnace remains relatively non-viscous and consequently it pours out of the furnace readily. Cooled and solidified slag produced by the furnace is very resistant to groundwater leaching. The slag can be safely buried in the earth without fear of contaminating groundwater. 3 figs

  1. Cementation of wastes with boric acid; Cimentacao de rejeitos contendo acido borico

    Energy Technology Data Exchange (ETDEWEB)

    Tello, Cledola C.O.; Haucz, Maria Judite A. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Alves, Lilian J.L.; Oliveira, Arno H. [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2000-07-01

    In nuclear power plants (PWR) are generated wastes, such as concentrate, which comes from the evaporation of liquid radioactive wastes, and spent resins. Both have boron in their composition. The cementation process is one of the options to solidify these wastes, but the boron has a negative effect on the setting of the cement mixture. In this paper are presented the experiments that are being carried out in order to overcome this problem and also to improve the efficiency of the process. Simulated wastes were cemented using additives (clays, admixtures etc.). In the process and product is being evaluated the effect of the amount, type and addition order of the materials. The mixtures were selected in accordance with their workability and incorporated waste. The solidified products are monolithic without free water with a good mechanical resistance. (author)

  2. 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)

  3. 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

  4. Heat transfer in vitrified radioactive waste

    International Nuclear Information System (INIS)

    Palancar, M.C.; Luis, M.A.; Luis, P.; Aragon, J.M.; Montero, M.A.

    1987-01-01

    An experimental method for measuring the thermal conductivity and convection coefficient of borosilicate glass cylinders, containing a simulated high level radioactive waste, is described. A simulation of the thermal behaviour of matrices of solidified waste during the cooling in air, water and a geological repository has been done. The experimental values of the thermal conductivity are ranging from 0.267 to 0.591 w/m K, for matrices with simulated waste contents of 10 to 40% (the waste is simulated by no radioactive isotopes). The convection coefficient for air/cylinders under the operating conditions used is 116 w/m 2 K. The simulated operation of cooling in air shows that about 1-2 days are enough to cool a solidified waste cylinder 0.6m diameter from 900 to 400 0 C. The cooling under water from 400 to near 80 0 C is faster than in air, but sharp temperature gradients within the matrices could be expected. The simulation of geological repositories lead to some criteria of arranging the matrices for avoiding undesirable high temperature points. (author) 1 fig

  5. Solidification method of radioactive wastes

    Energy Technology Data Exchange (ETDEWEB)

    Baba, Tsutomu; Chino, Koichi; Sasahira, Akira; Ikeda, Takashi

    1992-07-24

    Metal solidification material can completely seal radioactive wastes and it has high sealing effect even if a trace amount of evaporation should be caused. In addition, the solidification operation can be conducted safely by using a metal having a melting point of lower than that of the decomposition temperature of the radioactive wastes. Further, the radioactive wastes having a possibility of evaporation and scattering along with oxidation can be solidified in a stable form by putting the solidification system under an inert gas atmosphere. Then in the present invention, a metal is selected as a solidification material for radioactive wastes, and a metal, for example, lead or tin having a melting point of lower than that of the decomposition temperature of the wastes is used in order to prevent the release of the wastes during the solidification operation. Radioactive wastes which are unstable in air and scatter easily, for example, Ru or the like can be converted into a stable solidification product by conducting the solidification processing under an inert gas atmosphere. (T.M.).

  6. Selecting Items for Criterion-Referenced Tests.

    Science.gov (United States)

    Mellenbergh, Gideon J.; van der Linden, Wim J.

    1982-01-01

    Three item selection methods for criterion-referenced tests are examined: the classical theory of item difficulty and item-test correlation; the latent trait theory of item characteristic curves; and a decision-theoretic approach for optimal item selection. Item contribution to the standardized expected utility of mastery testing is discussed. (CM)

  7. 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

  8. 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

  9. The influence of interfacial energies and gravitational levels on the directionally solidified structures in hypermonotectic alloys

    Science.gov (United States)

    Andrews, J. B.; Curreri, P. A.; Sandlin, A. C.

    1988-01-01

    Various Cu-Pb-Al alloys were directionally solidified under 1-g conditions and alternating high-g/low-g conditions (achieved using NSAS's KC-135 aircraft) as a means of studying the influence of interfacial energies and gravitational levels on the resulting microstructures. Directional solidification of low Al content alloys was found to result in samples with coarser more irregular microstructures than in alloys with high Al contents under all the gravity conditions considered. Structures are correlated with interfacial energies, growth rates, and gravitational levels.

  10. 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.

  11. Secondary Waste Form Down-Selection Data Package—Fluidized Bed Steam Reforming Waste Form

    Energy Technology Data Exchange (ETDEWEB)

    Qafoku, Nikolla; Westsik, Joseph H.; Strachan, Denis M.; Valenta, Michelle M.; Pires, Richard P.

    2011-09-12

    The Hanford Site in southeast Washington State has 56 million gallons of radioactive and chemically hazardous wastes stored in 177 underground tanks (ORP 2010). The U.S. Department of Energy (DOE), Office of River Protection (ORP), through its contractors, is constructing the Hanford Tank Waste Treatment and Immobilization Plant (WTP) to convert the radioactive and hazardous wastes into stable glass waste forms for disposal. Within the WTP, the pretreatment facility will receive the retrieved waste from the tank farms and separate it into two treated process streams. These waste streams will be vitrified, and the resulting waste canisters will be sent to offsite (high-level waste [HLW]) and onsite (immobilized low-activity waste [ILAW]) repositories. As part of the pretreatment and ILAW processing, liquid secondary wastes will be generated that will be transferred to the Effluent Treatment Facility (ETF) on the Hanford Site for further treatment. These liquid secondary wastes will be converted to stable solid waste forms that will be disposed of in the Integrated Disposal Facility (IDF). To support the selection of a waste form for the liquid secondary wastes from WTP, Washington River Protection Solutions (WRPS) has initiated secondary waste form testing work at Pacific Northwest National Laboratory (PNNL). In anticipation of a down-selection process for a waste form for the Solidification Treatment Unit to be added to the ETF, PNNL is developing data packages to support that down-selection. The objective of the data packages is to identify, evaluate, and summarize the existing information on the four waste forms being considered for stabilizing and solidifying the liquid secondary wastes. At the Hanford Site, the FBSR process is being evaluated as a supplemental technology for treating and immobilizing Hanford LAW radioactive tank waste and for treating secondary wastes from the WTP pretreatment and LAW vitrification processes.

  12. Assessing item fit for unidimensional item response theory models using residuals from estimated item response functions.

    Science.gov (United States)

    Haberman, Shelby J; Sinharay, Sandip; Chon, Kyong Hee

    2013-07-01

    Residual analysis (e.g. Hambleton & Swaminathan, Item response theory: principles and applications, Kluwer Academic, Boston, 1985; Hambleton, Swaminathan, & Rogers, Fundamentals of item response theory, Sage, Newbury Park, 1991) is a popular method to assess fit of item response theory (IRT) models. We suggest a form of residual analysis that may be applied to assess item fit for unidimensional IRT models. The residual analysis consists of a comparison of the maximum-likelihood estimate of the item characteristic curve with an alternative ratio estimate of the item characteristic curve. The large sample distribution of the residual is proved to be standardized normal when the IRT model fits the data. We compare the performance of our suggested residual to the standardized residual of Hambleton et al. (Fundamentals of item response theory, Sage, Newbury Park, 1991) in a detailed simulation study. We then calculate our suggested residuals using data from an operational test. The residuals appear to be useful in assessing the item fit for unidimensional IRT models.

  13. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-07-01

    The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the low-level liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Hanford Site Maps, road evaluation for the grout treatment facility, Department of Ecology certificate of non-designation for centralia fly ash, double-shell tank waste compositional modeling, laboratory analysis reports for double-shell tank waste, stored in tanks 241-AN-103, 241-AN-106, and 241-AW-101, grout vault heat transfer results for M-106 grout formulation, test results for extraction procedure toxicity testing, test results for toxicity testing of double-shell tank grout, pilot-scale grout production test with a simulated low-level waste, characterization of simulated low-level waste grout produced in a pilot-scale test, description of the procedure for sampling nonaging waste storage tanks, description of laboratory procedures, grout campaign waste composition verification, variability in properties of grouted phosphate/sulfate N-reactor waste, engineering drawings, description of operating procedures, equipment list--transportable grout equipment, grout treatment facility--tank integrity assessment plan, long-term effects of waste solutions on concrete and reinforcing steel, vendor information, grout disposal facilities construction quality assurance plan, and flexible membrane liner/waste compatibility test results

  14. Studies Involving Immobilization Of Hazardous Wastes In Cement-ilmenite Matrix

    International Nuclear Information System (INIS)

    El-Dakrory, A.M.; Sayed, M.S.; Adham, K.

    1999-01-01

    Ilmenite was added to Ordinary Portland Cement to Modify the characteristic properties of the matrix as density, compressive strength and thermal stability . Coal tar and radiocesium were solidified as hazardous waste in cement-ilmenite matrix. The physical properties as density, sitting times and porosity were studied. The mechanical properties as compressive strength values and the chemical properties as leaching were measured

  15. Verifying generator waste certification: NTS waste characterization QA requirements

    International Nuclear Information System (INIS)

    Williams, R.E.; Brich, R.F.

    1988-01-01

    Waste management activities managed by the US Department of Energy (DOE) at the Nevada Test Site (NTS) include the disposal of low-level wastes (LLW) and mixed waste (MW), waste which is both radioactive and hazardous. A majority of the packaged LLW is received from offsite DOE generators. Interim status for receipt of MW at the NTS Area 5 Radioactive Waste Management Site (RWMS) was received from the state of Nevada in 1987. The RWMS Mixed Waste Management Facility (MWMF) is expected to be operational in 1988 for approved DOE MW generators. The Nevada Test Site Defense Waste Acceptance Criteria and Certification Requirements (NVO-185, Revision 5) delineates waste acceptance criteria for waste disposal at the NTS. Regulation of the hazardous component of mixed waste requires the implementation of US Environmental Protection Agency (EPA) requirements pursuant to the Resource Conservation and Recovery Act (RCRA). Waste generators must implement a waste certification program to provide assurance that the disposal site waste acceptance criteria are met. The DOE/Nevada Operations Office (NV) developed guidance for generator waste certification program plans. Periodic technical audits are conducted by DOE/NV to assess performance of the waste certification programs. The audit scope is patterned from the waste certification program plan guidance as it integrates and provides a common format for the applicable criteria. The criteria focus on items and activities critical to processing, characterizing, packaging, certifying, and shipping waste

  16. 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

  17. Stabilization and Solidification of Nitric Acid Effluent Waste at Y-12

    Energy Technology Data Exchange (ETDEWEB)

    Singh, Dileep [Argonne National Lab. (ANL), Argonne, IL (United States); Lorenzo-Martin, Cinta [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-12-16

    Consolidated Nuclear Security, LLC (CNS) at the Y-12 plant is investigating approaches for the treatment (stabilization and solidification) of a nitric acid waste effluent that contains uranium. Because the pH of the waste stream is 1-2, it is a difficult waste stream to treat and stabilize by a standard cement-based process. Alternative waste forms are being considered. In this regard, Ceramicrete technology, developed at Argonne National Laboratory, is being explored as an option to solidify and stabilize the nitric acid effluent wastes.

  18. Studies for geologic storage of radioactive waste in the southeast

    International Nuclear Information System (INIS)

    Marine, I.W.

    1977-01-01

    The National Waste Terminal Storage (NWTS) program was initiated to conduct the research necessary to select a site for a geologic repository for the storage of high-level, solidified radioactive waste from commercial power reactors. The program also includes the design and construction of the facility and its operation once completed. As part of this program, the Savannah River Laboratory is conducting geological research that is particularly relevant to potential repository sites in the Southeast, but is also of generic applicability. This paper describes the National Waste Terminal Storage program as well as the research program at the Savannah River Laboratory

  19. Studies for geologic storage of radioactive waste in the southeast

    International Nuclear Information System (INIS)

    Marine, I.W.

    1978-01-01

    The National Waste Terminal Storage (NWTS) program was initiated to conduct the research necessary to select a site for a geologic repository for the storage of high-level, solidified radioactive waste from commercial power reactors. The program also includes the design and construction of the facility and its operation once completed. As part of this program, the Savannah River Laboratory is conducting geological research that is particularly relevant to potential repository sites in the southeast, but is also of generic applicability. This paper describes the National Waste Terminal Storage program as well as the research program at the Savannah River Laboratory. 31 figures

  20. Defense waste processing facility project at the Savannah River Plant

    International Nuclear Information System (INIS)

    Baxter, R.G.; Maher, R.; Mellen, J.B.; Shafranek, L.F.; Stevens, W.R. III.

    1984-01-01

    The Du Pont Company is building for the Department of Energy a facility to vitrify high-level waste at the Savannah River Plant near Aiken, South Carolina. The Defense Waste Processing Facility (DWPF) will solidify existing and future radioactive wastes produced by defense activities at the site. At the present time engineering and design are 45% complete, the site has been cleared, and startup is expected in 1989. This paper will describe project status as well as features of the design. 9 figures

  1. Gas formation in drum waste packages of Paks NPP

    International Nuclear Information System (INIS)

    Molnar, M.; Palcsu, L.; Svingor, E.; Szanto, Z.; Futo, I.; Ormai, P.

    2000-01-01

    Gas composition measurements have been carried out by mass spectrometry analysis of samples taken from the headspace of ten drum waste packages generated and temporarily stored at Paks NPP. Four drums contained compacted solid waste, three drums were filled with grouted (solidified) sludge and three drums contained solid waste without compaction. The drums have been equipped with a special gas outlet system to make repeated sampling possible. Based on the first measurements significant differences in the gas composition and the rate of gas generation among the drums were found. (author)

  2. Treatment of liquid waste containing alpha nuclides by adsorption

    Energy Technology Data Exchange (ETDEWEB)

    Jishu, Zeng; Xiguang, Su; Dejing, Xia; Sianhua, Fan [China Inst. of Atomic Energy, Beijing (China). Radiochemistry Dept.

    1997-02-01

    In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10{sup 3} Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs.

  3. Treatment of liquid waste containing alpha nuclides by adsorption

    International Nuclear Information System (INIS)

    Zeng Jishu; Su Xiguang; Xia Dejing; Fan Sianhua

    1997-01-01

    In this paper, experimental investigations on the removal of actinides from a decontaminating waste stream by using adsorption technique following the cementation of a resultant absorbent sludge are described. One kind of apatites was selected as an actinide absorbent from a number of indigenous materials by batch equilibrium tests. The influence of contact time, temperature, particle size and pH variables on the adsorption of actinides is given. The removal of total alpha activity is higher tan 97% by absorbent precipitation process when the absorbent addition percentage of the liquid waste is more than 3.25 wt%, making alpha-activity level of the primary waste stream below 3.7 x 10 3 Bq/L, which can meet the acceptance requirements of the Low Level Radwaste Treatment Plant. The studies on the cementation of the absorbent sludge included the selection of cements used for solidification, formulation and characterization of the selected cemented waste forms. The results obtained have shown that both 525 type Portland cement and 325 type Portland pozzolana cement were compatible with the absorbent sludge. The selected cemented waste forms meet the requirements of the Chinese National Standard (GB 14569.1-93): Characteristic Requirements for Solidified Waste of Low and Intermediate Level Radioactive Waste - Cement Solidified Waste. (author). 9 refs, 3 figs, 14 tabs

  4. 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.

  5. 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

  6. Chapter 7. Radioactive wastes

    International Nuclear Information System (INIS)

    2000-01-01

    The inspection and assessment activities of Nuclear Regulatory Authority of the Slovak Republic (UJD) focused on minimization of activity and the quantity of produced radioactive waste (RAW), and on increasing safety of waste management. The general scheme of rad-waste management in the Slovak Republic is presented. The radioactive wastes produced during the operation of NPP V-1, NPP V-2 and NPP Mochovce in 1999 are listed.Liquid RAW was treated and conditioned into a solid form at the nuclear facility Technology for treatment and conditioning of RAW. In 1999 combustible solid waste was treated at the nuclear facility Incinerator of VUJE Trnava. Produced liquid and solid RAW are stored at designed equipment at individual nuclear installations (in case of NPP V-1, NPP V-2 Bohunice and NPP Mochovce in compliance with the Regulation No. 67/1987 Coll. law).The status of free capacity of these storages as of 31.121999 is presented. Storage solidified product built the SE-VYZ was fully filled at the end of 1999. In 1999 there was a significant improvement in the process of radioactive waste management by: (A) issuing approval for commissioning the National Repository for RAW, (B) issuing approval for commissioning the Treatment and Conditioning Center for RAW, (C) having the application for approval to transport conditioned RAW to the National repository Mochovce in the final stage of evaluation. At the beginning of 2000 it is realistic to expect that RAW conditioned in the Conditioning center of RAW will start to be disposed at the National repository of RAW in Mochovce

  7. Software for automated tracking of open items at NRC

    International Nuclear Information System (INIS)

    DeWispelare, A.R.; Mackin, P.C.; Johnson, R.L.

    1995-01-01

    The Open Item Tracking System (OITS) was developed in response to the Nuclear Regulatory Commission (NRC) need for a reliable, easy to use automated database system, to track all open (awaiting resolution) items related to regulatory, institutional, and technical uncertainties for the Department of Energy's (DOE's) high-level waste (HLW) disposal program. The OITS system was integrated with the Regulatory Program Database (RPD) Version 1.1, resulting in the RPD/OITS Version 2.0 system. RPD/OITS is a network bases system with client server architecture and a graphical user interface. This paper outlines the system and results of its implementation

  8. Container for processing and disposing radioactive wastes and industrial wastes

    International Nuclear Information System (INIS)

    Araki, Kunio; Kasahara, Yuko; Kasai, Noboru; Sudo, Giichi; Ishizaki, Kanjiro.

    1978-01-01

    Purpose: To improve the performance of containers for radioactive wastes for ocean disposal and on-land disposal such as impact strength, chemical resistance, fire resistance, corrosion resistance, water impermeability and the like. Constitution: Steel fiber-reinforced concrete previously molded in a shape of a container is impregnated with polymerizable impregnating agent selected from the group consisting of a polymerizable monomer, liquid mixture of a polymerizable monomer and an oligomer, a polymer solution, a copolymer solution and the liquid mixture thereof. Then, the polymerizable impregnating agent is polymerized to solidify in the concrete by way of heat-polymerization or radiation-induced polymerization to form a waste container. The container thus obtained can be improved with the impact resistance and wear resistance and further improved with salt water resistance, acid resistance, corrosion resistance and solidity by the impregnation of the polymer, as well as can effectively be prevented from leaching out of radioactive substances. (Furukawa, Y.)

  9. 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

  10. 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

  11. 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

  12. 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.

  13. 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.

  14. 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.

  15. 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.

  16. 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)

  17. 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.

  18. 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.

  19. 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.

  20. Construction of solid waste form test facility

    International Nuclear Information System (INIS)

    Park, Hyun Whee; Lee, Kang Moo; Koo, Jun Mo; Jung, In Ha; Lee, Jong Ryeul; Kim, Sung Whan; Bae, Sang Min; Cho, Kang Whon; Sung, Suk Jong

    1989-02-01

    The Solid Waste Form Test Facility (SWFTF) is now construction at DAEDUCK in Korea. In SWFTF, the characteristics of solidified waste products as radiological homogeneity, mechanical and thermal property, water resistance and lechability will be tested and evaluated to meet conditions for long-term storage or final disposal of wastes. The construction of solid waste form test facility has been started with finishing its design of a building and equipments in Sep. 1984, and now building construction is completed. Radioactive gas treatment system, extinguishers, cooling and heating system for the facility, electrical equipments, Master/Slave manipulator, power manipulator, lead glass and C.C.T.V. has also been installed. SWFTF will be established in the beginning of 1990's. At this report, radiation shielding door, nondestructive test of the wall, instrumentation system for the utility supply system and cell lighting system are described. (Author)

  1. The immobilization of organic liquid wastes

    International Nuclear Information System (INIS)

    Greenhalgh, W.O.

    1986-01-01

    This report describes a portland cement immobilization process for the disposal treatment of radioactive organic liquid wastes which would be generated in a FFTF fuels reprocessing line. An incineration system already on-hand was determined to be too costly to operate for the 100 to 400 gallons per year organic liquid. Organic test liquids were dispersed into an aqueous phosphate liquid using an emulsifier. A total of 109 gallons of potential and radioactive aqueous immiscible organic liquid wastes from Hanford 300 Area operations were solidified with portland cement and disposed of as solid waste during a 3 month test program with in-drum mixers. Waste packing efficiencies varied from 32 to 40% and included pump oils, mineral spirits, and TBP-NPH type solvents

  2. Multibarrier waste forms. Part III: Process considerations

    International Nuclear Information System (INIS)

    Lokken, R.O.

    1979-10-01

    The multibarrier concept for the solidification and storage of radioactive waste utilizes up to three barriers to isolate radionuclides from the environment: a solidified waste inner core, an impervious coating, and a metal matrix. The coating and metal matrix give the composite waste form enhanced inertness with improvements in thermal stability, mechanical strength, and leach resistance. Preliminary process flow rates and material costs were evaluated for four multibarrier waste forms with the process complexity increasing thusly: glass marbles, uncoated supercalcine, glass-coated supercalcine, and PyC/Al 2 O 3 -coated supercalcine. This report discusses the process variables and their effect on optimization of product quality, processing simplicity, and material cost. 11 figures, 2 tables

  3. Processing method for miscellaneous radioactive solid waste

    International Nuclear Information System (INIS)

    Matsuda, Masami; Komori, Itaru; Nishi, Takashi.

    1995-01-01

    Miscellaneous solid wastes are subjected to heat treatment at a temperature not lower than a carbonizing temperature of organic materials in the wastes and not higher than the melting temperature of inorganic materials in the wastes, for example, not lower than 200degC but not higher than 660degC, and then resultant miscellaneous solid wastes are solidified using a water hardening solidification material. With such procedures, the organic materials in the miscellaneous solids are decomposed into gases. Therefore, solid materials excellent in long term stability can be formed. In addition, since the heat treatment is conducted at a relatively low temperature such as not higher than 660degC, the generation amount of off gases is reduced to simplify an off gas processing system, and since molten materials are not formed, handing is facilitated. (T.M.)

  4. 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.

  5. MODELING SOLIDIFICATION-INDUCED STRESSES IN CERAMIC WASTE FORMS CONTAINING NUCLEAR WASTES

    International Nuclear Information System (INIS)

    Solbrig, Charles W.; Bateman, Kenneth J.

    2010-01-01

    The goal of this work is to produce a ceramic waste form (CWF) that permanently occludes radioactive waste. This is accomplished by absorbing radioactive salts into zeolite, mixing with glass frit, heating to a molten state 915 C to form a sodalite glass matrix, and solidifying for long-term storage. Less long term leaching is expected if the solidifying cooling rate doesn't cause cracking. In addition to thermal stress, this paper proposes that a stress is formed during solidification which is very large for fast cooling rates during solidification and can cause severe cracking. A solidifying glass or ceramic cylinder forms a dome on the cylinder top end. The temperature distribution at the time of solidification causes the stress and the dome. The dome height, ''the length deficit,'' produces an axial stress when the solid returns to room temperature with the inherent outer region in compression, the inner in tension. Large tensions will cause cracking of the specimen. The temperature deficit, derived by dividing the length deficit by the coefficient of thermal expansion, allows solidification stress theory to be extended to the circumferential stress. This paper derives the solidification stress theory, gives examples, explains how to induce beneficial stresses, and compares theory to experimental data.

  6. Method of processing waste sodium

    International Nuclear Information System (INIS)

    Shimoyashiki, Shigehiro; Takahashi, Kazuo.

    1982-01-01

    Purpose: To enable safety store of waste sodium in the form of intermetallic compounds. Method: Waste sodium used in a reactor is mixed with molten metal under an inert gas atmosphere and resulted intermetallic compounds are stored in a closely sealed container to enable quasi-permanent safety store as inert compound. Used waste sodium particularly, waste sodium in the primary system containing radioactive substances is charged in a waste sodium melting tank having a heater on the side, the tank is evacuated by a vacuum pump and then sealed with gaseous argon supplied from a gaseous argon tank, and waste sodium is melted under heating. The temperature and the amount of the liquid are measured by a thermometer and a level meter respectively. While on the other hand, molten metal such as Sn, Pb and Zn having melting point above 300 0 C are charged in a metal melting tank and heated by a heater. The molten sodium and the molten metals are charged into a mixing tank and agitated to mix by an induction type agitator. Sodium vapors in the tank are collected by traps. The air in the tank is replaced with gaseous argon. The molten mixture is closely sealed in a drum can and cooled to solidify for safety storage. (Seki, T.)

  7. Technology for commercial radioactive waste management

    International Nuclear Information System (INIS)

    1979-05-01

    A general analysis of transportation requirements for postfission radioactive wastes that are produced from the commercial light water reactor (LWR) fuel cycle and that are assumed to require Federal custody for storage or disposal is given. Possible radioactive wastes for which transportation requirements are described include: spent fuel, solidified high-level waste, fuel residues (cladding wastes), plutonium, and non-high-level transuranic (TRU) wastes. Transportation is described for wastes generated in three fuel cycle options: once-through fuel cycle, uranium recycle only, and recycle of uranium and plutonium. The geologic considerations essential for repository selection, the nature of geologic formations that are potential repository media, the thermal criteria for waste placement in geologic repositories, and conceptual repositories in four different geologic media are described. The media are salt deposits, granite, shale, and basalt. Possible alternatives for managing retired facilities and procedures for decommissioning are reviewed. A qualitative comparison is made of wastes generated by the uranium fuel cycle and the thorium fuel cycle. This study presents data characterizing wastes from prebreeder light water breeder reactors using thorium and slightly enriched uranium-235. The prebreeder LWBRs are essentially LWRs using thorium. The operation of HTGR and LWBR cycles are conceptually designed, and wastes produced in these cycles are compared for potential differences

  8. System for disposing of radioactive waste

    International Nuclear Information System (INIS)

    Gablin, K.A.; Hansen, L.J.

    1979-01-01

    A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container. 30 claims

  9. System for disposing of radioactive waste

    International Nuclear Information System (INIS)

    Gablin, K.A.; Hansen, L.J.

    1977-01-01

    A system is described for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container

  10. Solidification of ion exchange resin wastes

    International Nuclear Information System (INIS)

    1982-08-01

    Solidification media investigated included portland type I, portland type III and high alumina cements, a proprietary gypsum-based polymer modified cement, and a vinyl ester-styrene thermosetting plastic. Samples formulated with hydraulic cement were analyzed to investigate the effects of resin type, resin loading, waste-to-cement ratio, and water-to-cement ratio. The solidification of cation resin wastes with portland cement was characterized by excessive swelling and cracking of waste forms, both after curing and during immersion testing. Mixed bed resin waste formulations were limited by their cation component. Additives to improve the mechanical properties of portland cement-ion exchange resin waste forms were evaluated. High alumina cement formulations dislayed a resistance to deterioration of mechanical integrity during immersion testing, thus providing a significant advantage over portland cements for the solidification of resin wastes. Properties of cement-ion exchange resin waste forms were examined. An experiment was conducted to study the leachability of 137 Cs, 85 Sr, and 60 Co from resins modified in portland type III and high alumina cements. The cumulative 137 Cs fraction release was at least an order of magnitude greater than that of either 85 Sr or 60 Co. Release rates of 137 Cs in high alumina cement were greater than those in portland III cement by a factor of two.Compressive strength and leach testing were conducted for resin wastes solidified with polymer-modified gypsum based cement. 137 Cs, 85 Sr, and 60 Co fraction releases were about one, two and three orders of magnitude higher, respectively, than in equivalent portland type III cement formulations. As much as 28.6 wt % dry ion exchange resin was successfully solidified using vinyl ester-styrene compared with a maximum of 25 wt % in both portland and gypsum-based cement

  11. The effect of concentration on the structure and crystallinity of a cementitious waste form for caustic wastes

    International Nuclear Information System (INIS)

    Chung, Chul-Woo; Turo, Laura A.; Ryan, Joseph V.; Johnson, Bradley R.; McCloy, John S.

    2013-01-01

    Highlights: ► Cast Stone: Portland cement, fly ash, blast furnace slag, and simulated nuclear waste. ► Caustic secondary waste from the off-gas of a vitrification process was targeted. ► Crystallinity, micro- and mesostructure, and engineering properties characterized. ► Waste concentration varied from 0 to 2.5 M, but caused minimal changes. ► Cast Stone shows good compositional versatility as a secondary waste form. -- Abstract: Cement-based waste forms have long been considered economical technologies for disposal of various types of waste. A solidified cementitious waste form, Cast Stone, has been identified to immobilize the radioactive secondary waste from vitrification processes. In this work, Cast Stone was considered for a Na-based caustic liquid waste, and its physical properties were analyzed as a function of liquid waste loading up to 2 M Na. Differences in crystallinity (phase composition), microstructure, mesostructure (pore size distribution and surface area), and macrostructure (density and compressive strength) were investigated using various analytical techniques, in order to assess the suitability of Cast Stone as a chemically durable waste. It was found that the concentration of secondary waste simulant (caustic waste) had little effect on the relevant engineering properties of Cast Stone, showing that Cast Stone could be an effective and tolerant waste form for a wide range of concentrations of high sodium waste

  12. Thermal treatment of toxic metals of industrial hazardous wastes with fly ash and clay

    Energy Technology Data Exchange (ETDEWEB)

    Singh, I.B. [Regional Research Laboratory, Council of Scientific and Industrial Research, Hoshangabad Road, Bhopal 462026 (India)]. E-mail: ibsingh58@yahoo.com; Chaturvedi, K. [Regional Research Laboratory, Council of Scientific and Industrial Research, Hoshangabad Road, Bhopal 462026 (India); Morchhale, R.K. [Regional Research Laboratory, Council of Scientific and Industrial Research, Hoshangabad Road, Bhopal 462026 (India); Yegneswaran, A.H. [Regional Research Laboratory, Council of Scientific and Industrial Research, Hoshangabad Road, Bhopal 462026 (India)

    2007-03-06

    Waste generated from galvanizing and metal finishing processes is considered to be a hazardous due to the presence of toxic metals like Pb, Cu, Cr, Zn, etc. Thermal treatment of such types of wastes in the presence of clay and fly ash can immobilizes their toxic metals to a maximum level. After treatment solidified mass can be utilized in construction or disposed off through land fillings without susceptibility of re-mobilization of toxic metals. In the present investigation locally available clay and fly ash of particular thermal power plant were used as additives for thermal treatment of both of the wastes in their different proportions at 850, 900 and 950 deg. C. Observed results indicated that heating temperature to be a key factor in the immobilization of toxic metals of the waste. It was noticed that the leachability of metals of the waste reduces to a negligible level after heating at 950 deg. C. Thermally treated solidified specimen of 10% waste and remaining clay have shown comparatively a higher compressive strength than clay fired bricks used in building construction. Though, thermally heated specimens made of galvanizing waste have shown much better strength than specimen made of metal finishing waste. The lechability of toxic metals like Cr, Cu, Pb and Zn became far below from their regulatory threshold after heating at 950 deg. C. Addition of fly ash did not show any improvement either in engineering property or in leachability of metals from the solidified mass. X-ray diffraction (XRD) analysis of the solidified product confirmed the presence of mixed phases of oxides of metals.

  13. Bituminization of low- and medium-level radioactive wastes

    International Nuclear Information System (INIS)

    Lefillatre, G.

    1976-01-01

    French operations are presented concerning mainly: the bituminization of radioactive wastes produced in light water reactors; the direct bituminization of liquid effluents without concentration; the experiments carried out for 18 months on the land burial of concentrates solidified by bitumen into blocks of 100 liters. The knowledge acquired in France is exposed: in Valduc and in Saclay with facilities equipped with thin film evaporators and in Marcoule with the conditioning of trilaurylamine and tributyl-phosphate. At last the Cadarache bituminization [fr

  14. 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

  15. 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)

  16. Nordic study on reactor waste. Technical part 1 and 2

    International Nuclear Information System (INIS)

    1981-08-01

    An important part of the Nordic studies on system- and safety analysis of the management of low and medium level radioactive waste from nuclear power plants, is the safety analysis of a Reference System. This reference system was established within the study and is described in this Technical Part 1. The reference system covers waste management Schemes that are potential possibilities in either one of the four participating Nordic countries. The reference system is based on: a power reactor system consisting of 6 BWR's of 500 MWe each, operated simultaneously over the same 30 year period, and deep bed granular ion exchange resin wastes from the Reactor Water Clean-Up System (RWCS and powdered ion exchange resin from the Spent Fuel Pool Cleanup System (SFPCS)). Both waste types are supposed to be solidified by mixing with cement and bitumen. Two basic types of containers are considered. Standard 200 liter steel drums and specially made cubicreinforced concrete moulds with a net volume of 1 m 3 . The Nordic study assumes temporary storage of the solidified waste for a maximum of 50 years before the waste is transferred to the disposal site. Transportation of the waste from the storage facilitiy to the disposal site will be by road or sea. Three different disposal facilities are considered: Shallow land burial, near surface concrete bunker, and rock cavern with about 30 m granite cover. (EG)

  17. Liquid Secondary Waste Grout Formulation and Waste Form Qualification

    Energy Technology Data Exchange (ETDEWEB)

    Um, Wooyong [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Williams, B. D. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Snyder, Michelle M. V. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Wang, Guohui [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2016-05-23

    This report describes the results from liquid secondary waste (LSW) grout formulation and waste form qualification tests performed at Pacific Northwest National Laboratory (PNNL) for Washington River Protection Solutions (WRPS) to evaluate new formulations for preparing a grout waste form with high-sulfate secondary waste simulants and the release of key constituents from these grout monoliths. Specific objectives of the LSW grout formulation and waste form qualification tests described in this report focused on five activities: 1.preparing new formulations for the LSW grout waste form with high-sulfate LSW simulants and solid characterization of the cured LSW grout waste form; 2.conducting the U.S. Environmental Protection Agency (EPA) Method 1313 leach test (EPA 2012) on the grout prepared with the new formulations, which solidify sulfate-rich Hanford Tank Waste Treatment and Immobilization Plant (WTP) off-gas condensate secondary waste simulant, using deionized water (DIW); 3.conducting the EPA Method 1315 leach tests (EPA 2013) on the grout monoliths made with the new dry blend formulations and three LSW simulants (242-A evaporator condensate, Environmental Restoration Disposal Facility (ERDF) leachate, and WTP off-gas condensate) using two leachants, DIW and simulated Hanford Integrated Disposal Facility (IDF) Site vadose zone pore water (VZPW); 4.estimating the 99Tc desorption Kd (distribution coefficient) values for 99Tc transport in oxidizing conditions to support the IDF performance assessment (PA); 5.estimating the solubility of 99Tc(IV)-bearing solid phases for 99Tc transport in reducing conditions to support the IDF PA.

  18. Radiological protection criteria risk assessments for waste disposal options

    International Nuclear Information System (INIS)

    Hill, M.D.

    1982-01-01

    Radiological protection criteria for waste disposal options are currently being developed at the National Radiological Protection Board (NRPB), and, in parallel, methodologies to be used in assessing the radiological impact of these options are being evolved. The criteria and methodologies under development are intended to apply to all solid radioactive wastes, including the high-level waste arising from reprocessing of spent nuclear fuel (because this waste will be solidified prior to disposal) and gaseous or liquid wastes which have been converted to solid form. It is envisaged that the same criteria will be applied to all solid waste disposal options, including shallow land burial, emplacement on the ocean bed (sea dumping), geological disposal on land and sub-seabed disposal

  19. Loviisa starts low-level operating waste disposal in 1997

    International Nuclear Information System (INIS)

    Snellman, J.

    1996-01-01

    At an early stage Imatran Voima Oy (IVO) decided to construct a waste repository for Loviisa NPP. The suitability of the power plant site for final disposal of low- and intermediate- level operating waste was studied. In the site report in 1982 the plant site was found to be geologically suitable and economically feasible for construction. The necessary preparations started in 1992. The repository will be constructed in three phases. The first phase will cover the transport tunnel, construction of one maintenance waste tunnel and the excavation of another maintenance waste tunnel together with a hall for solidified wastes. This phase will be finished by the end of 1996. During the second phase in the beginning of next century the remaining already excavated rooms will be furnished. Finally in the third phase the repository will be extended for the decommissioning waste somewhere around years 2020-2025. (3 figs., 1 tab.)

  20. 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

  1. Leaching behavior of phosphate-bonded ceramic waste forms

    International Nuclear Information System (INIS)

    Singh, D.; Wagh, A.S.; Jeong, S.Y.; Dorf, M.

    1996-04-01

    Over the last few years, Argonne National Laboratory has been developing room-temperature-setting chemically bonded phosphate ceramics for solidifying and stabilizing low-level mixed wastes. This technology is crucial for stabilizing waste streams that contain volatile species and off-gas secondary waste streams generated by high-temperature treatment of such wastes. We have developed a magnesium phosphate ceramic to treat mixed wastes such as ash, salts, and cement sludges. Waste forms of surrogate waste streams were fabricated by acid-base reactions between the mixtures of magnesium oxide powders and the wastes, and phosphoric acid or acid phosphate solutions. Dense and hard ceramic waste forms are produced in this process. The principal advantage of this technology is that the contaminants are immobilized by both chemical stabilization and subsequent microencapsulation of the reaction products. This paper reports the results of durability studies conducted on waste forms made with ash waste streams spiked with hazardous and radioactive surrogates. Standard leaching tests such as ANS 16.1 and TCLP were conducted on the final waste forms. Fates of the contaminants in the final waste forms were established by electron microscopy. In addition, stability of the waste forms in aqueous environments was evaluated with long-term water-immersion tests

  2. 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

  3. 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

  4. 48 CFR 852.214-72 - Alternate item(s).

    Science.gov (United States)

    2010-10-01

    ... AND FORMS SOLICITATION PROVISIONS AND CONTRACT CLAUSES Texts of Provisions and Clauses 852.214-72... 2008) Bids on []* will be given equal consideration along with bids on []** and any such bids received... [].** * Contracting officer will insert an alternate item that is considered acceptable. ** Contracting officer will...

  5. Utilization of crushed radioactive concrete for mortar to fill waste container void space

    International Nuclear Information System (INIS)

    Ishikura, Takeshi; Ohnishi, Kazuhiko; Oguri, Daiichiro; Ueki, Hiroyuki

    2004-01-01

    Minimizing the volume of radioactive waste generated during dismantling of nuclear power plants is a matter of great importance. In Japan waste forms buried in a shallow burial disposal facility as low level radioactive waste must be solidified by cement or other materials with adequate strength and must provide no harmful opening. The authors have developed an improved method to minimize radioactive waste volume by utilizing radioactive concrete for fine aggregate for mortars to fill void space in waste containers. Tests were performed with pre-placed concrete waste and with filling mortar using recycled fine aggregate produced from concrete. It was estimated that the improved method substantially increases the waste fill ratio in waste containers, thereby decreasing the total volume of disposal waste. (author)

  6. Radioactive wastes management in fiscal year 1983 in the fuel reprocessing plant

    International Nuclear Information System (INIS)

    1985-01-01

    In the nuclear fuel reprocessing plant of Power Reactor and Nuclear Fuel Development Corporation, the releases of radioactive gaseous and liquid wastes are so managed not to exceed the respective objective release levels. Of the radioactive liquid wastes, the high level concentrated wastes are stored in tanks and the low level wastes are stored in tanks or asphalt solidified. For radioactive solid wastes, high level solid wastes are stored in casks, low level solid wastes and asphalt solids in drums etc. The releases of radioactive gaseous and liquid wastes in the fiscal year 1983 were below the objective release levels. The radioactive wastes management in the fuel reprocessing plant in fiscal year 1983 is given in tables, the released quantities, the stored quantities, etc. (Mori, K.)

  7. Characterization of radioactive waste forms. Volume 1

    International Nuclear Information System (INIS)

    Brodersen, K.; Nilsson, K.

    1989-01-01

    This document is the second yearbook for Task 3 of the European Communities 1985-89 programme of research on radioactive waste management and disposal carried out by public organizations and private firms in the Community through costsharing contracts with the Commission of the European Communities. The report, in two volumes, describes progress made in 1987 within the field of Task 3: Testing and evaluation of conditioned waste and engineered barriers. The first volume of the report covers Item 3.1 Characterization of low and medium-level radioactive waste forms and Item 3.5 Development of test methods for quality assurance. The second volume covers Item 3.2: High-level and alpha waste characterization and Item 3.3: Other engineered barriers. Item 3.4 on the round robin study will be treated in a separate report

  8. Characterization of radioactive waste forms. Volume 2

    International Nuclear Information System (INIS)

    Smith, D.L.; Green, T.H.

    1989-01-01

    This document is the second yearbook for Task 3 of the European Communities 1985-89 programme of research on radioactive waste management and disposal carried out by public organizations and private firms in the Community through cost-sharing contracts with the Commission of the European Communities. The report, in two volumes, describes progress made in 1987 within the field of Task 3: Testing and evaluation of conditioned waste and engineered barriers. The first volume of the report covers Item 3.1 Characterization of low and medium level radioactive waste forms and Item 3.5. Development of test methods for quality assurance. The second volume covers Item 3.2: High-level and alpha waste characterization and Item 3.3: Other engineered barriers. Item 3.4 on the round robin study will be treated in a separate report

  9. 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.

  10. 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.

  11. 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.

  12. 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

  13. 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.

  14. 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

  15. 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.

  16. 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.

  17. 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)

  18. 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.

  19. 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

  20. 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...

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. Management of Radioactive Wastes in Developing Countries

    International Nuclear Information System (INIS)

    Abdel Ghani, A.H.

    1999-01-01

    The management of radioactive wastes is one area of increasing interest especially in developing countries having more and more activities in the application of radioisotopes in medicine, research and industry. For a better understanding of radioactive waste management in developing countries this work will discuss the following items:Classification of countries with respect to waste management programs. Principal Radionuclides used in medicine, biological research and others and the range of radioactivity commonly used. Estimation of radioactive waste volumes and activities. Management of liquid wastes Collection. Treatment. Management of small volumes of organic liquid waste. Collection Treatment. Packaging and storage of radioactive wastes

  6. Solidification of hazardous and mixed radioactive waste at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Boehmer, A.M.; Larsen, M.M.

    1986-01-01

    EG and G Idaho has initiated a program to develop treatment options for the hazardous and mixed wastes generated at the Idaho National Engineering Laboratory (INEL). This program includes development of solidification methods for some of these wastes. Testing has shown that toxic wastes can be successfully solidified using cement, cement-silicate, or ENVIROSTONE binders to produce nontoxic stable waste forms for safe, long term disposal. This paper presents the results of the solidification development program conducted at the INEL by EG and G Idaho

  7. 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

  8. Solidification of hazardous and mixed radioactive waste at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Boehmer, A.M.; Larsen, M.M.

    1986-03-01

    EG and G Idaho has initiated a program to develop treatment options for the hazardous and mixed wastes generated at the Idaho National Engineering Laboratory (INEL). This program includes development of solidification methods for some of these wastes. Testing has shown that toxic wastes can be successfully solidified using cement, cement-silicate, or ENVIROSTONE binders to produce nontoxic stable waste forms for safe, long term disposal. This paper presents the results of the solidification development program conducted at the INEL by EG and G Idaho

  9. Hazardous and mixed waste solidification development conducted at the Idaho National Engineering Laboratory

    International Nuclear Information System (INIS)

    Boehmer, A.M.; Larsen, M.M.

    1986-04-01

    EG and G Idaho, Inc., has initiated a program to develop safe, efficient, cost-effective solidification treatment methods for the disposal of some of the hazardous and mixed wastes generated at the Idaho National Engineering Laboratory (INEL). Testing has shown that Extraction Procedure (EP) toxic wastes can be successfully solidified using cement, cement-silicate, or ENVIROSTONE binders to produce nontoxic stable waste forms for safe, long-term disposal as general or low-level waste, depending upon the radioactivity. The results of the solidification development program are presented in this report

  10. Modelling sequentially scored item responses

    NARCIS (Netherlands)

    Akkermans, W.

    2000-01-01

    The sequential model can be used to describe the variable resulting from a sequential scoring process. In this paper two more item response models are investigated with respect to their suitability for sequential scoring: the partial credit model and the graded response model. The investigation is

  11. Item level diagnostics and model - data fit in item response theory ...

    African Journals Online (AJOL)

    Item response theory (IRT) is a framework for modeling and analyzing item response data. Item-level modeling gives IRT advantages over classical test theory. The fit of an item score pattern to an item response theory (IRT) models is a necessary condition that must be assessed for further use of item and models that best fit ...

  12. Regional waste treatment with monolith disposal for low-level radioactive waste

    International Nuclear Information System (INIS)

    Forsberg, C.W.

    1983-01-01

    An alternative system is proposed for the disposal of low-level radioactive waste. This system, called REgional Treatment with MOnolith Disposal (RETMOD), is based on integrating three commercial technologies: automated package warehousing, whole-barrel rotary kiln incineration, and cement-based grouts for radioactive waste disposal. In the simplified flowsheet, all the sludges, liquids, resins, and combustible wastes are transported to regional facilities where they are incinerated. The ash is then mixed with special cement-based grouts, and the resulting mixture is poured into trenches to form large waste-cement monoliths. Wastes that do not require treatment, such as damaged and discarded equipment, are prepositioned in the trenches with the waste-cement mixture poured on top. The RETMOD system may provide higher safety margins by conversion of wastes into a solidified low-leach form, creation of low-surface area waste-cement monoliths, and centralization of waste processing into a few specialized facilities. Institutional problems would be simplified by placing total responsibility for safe disposal on the disposal site operator. Lower costs may be realized through reduced handling costs, the economics of scale, simplified operations, and less restrictive waste packaging requirements

  13. 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

  14. Solidification as low cost technology prior to land filling of industrial hazardous waste sludge.

    Science.gov (United States)

    El-Sebaie, O; Ahmed, M; Ramadan, M

    2000-01-01

    The aim of this study is to stabilize and solidify two different treated industrial hazardous waste sludges, which were selected from factories situated close to Alexandria. They were selected to ensure their safe transportation and landfill disposal by reducing their potential leaching of hazardous elements, which represent significant threat to the environment, especially the quality of underground water. The selected waste sludges have been characterized. Ordinary Portland Cement (OPC), Cement Kiln Dust (CKD) from Alexandria Portland Cement Company, and Calcium Sulphate as a by-product from the dye industry were used as potential solidification additives to treat the selected treated waste sludges from tanning and dyes industry. Waste sludges as well as the solidified wastes have been leach-tested, using the General Acid Neutralization Capacity (GANC) procedure. Concentration of concerning metals in the leachates was determined to assess changes in the mobility of major contaminants. The treated tannery waste sludge has an acid neutralization capacity much higher than that of the treated dyes waste sludge. Experiment results demonstrated the industrial waste sludge solidification mix designs, and presented the reduction of contaminant leaching from two types of waste sludges. The main advantages of solidification are that it is simple and low cost processing which includes readily available low cost solidification additives that will convert industrial hazardous waste sludges into inert materials.

  15. Method of processing radioactive metal wastes

    International Nuclear Information System (INIS)

    Inoue, Yoichi; Kitagawa, Kazuo; Tsuzura, Katsuhiko.

    1980-01-01

    Purpose: To enable long and safety storage for radioactive metal wastes such as used fuel cans after the procession or used pipe, instruments and the likes polluted with various radioactive substances, by compacting them to solidify. Method: Metal wastes such as used fuel cans, which have been cut shorter and reprocessed, are pressed into generally hexagonal blocks. The block is charged in a capsule of a hexagonal cross section made of non-gas permeable materials such as soft steels, stainless steels and the likes. Then, the capsule is subjected to static hydraulic hot pressing as it is or after deaeration and sealing. While various combinations are possible for temperature, pressure and time as the conditions for the static hydraulic hot pressing, dense block with no residual gas pores can be obtained, for example, under the conditions of 900 0 C, 1000 Kg/cm 2 and one hour where the wastes are composed of zircaloy. (Kawakami, Y.)

  16. Fate of nuclear waste site remains unclear

    International Nuclear Information System (INIS)

    Anderson, E.V.

    1980-01-01

    The only commercial nuclear fuel reprocessing plant in the U.S., located in West Valley, N.Y., has been shut down since 1972, and no efforts have yet been made to clean up the site. The site contains a spent-fuel pool, high level liquid waste storage tanks, and two radioactive waste burial grounds. Nuclear Fuel Services, Inc., has been leasing the site from the New York State Energy RandD Authority. Federal litigation may ensue, prompted by NRC and DOE, if the company refuses to decontaminate the area when its lease expires at the end of 1980. DOE has developed a plan to solidify the liquid wastes at the facility but needs additional legislation and funding to implement the scheme

  17. Grout treatment facility dangerous waste permit application

    International Nuclear Information System (INIS)

    1992-07-01

    The Grout Treatment Facility (GTF) will provide permanent disposal for approximately 43 Mgal of low-level radioactive liquid waste currently being stored in underground tanks on the Hanford Site. The first step in permanent disposal is accomplished by solidifying the liquid waste with cementitious dry materials. The resulting grout is cast within underground vaults. This report on the GTF contains information on the following: Geologic data, hydrologic data, groundwater monitoring program, information, detection monitoring program, groundwater characterization drawings, building emergency plan--grout treatment facility, response action plan for grout treatment facility, Hanford Facility contingency plan, training course descriptions, overview of the Hanford Facility Grout Performance, assessment, bland use and zoning map, waste minimization plan, cover design engineering report, and clay liners (ADMIXTURES) in semiarid environments

  18. Nordic study on reactor waste

    International Nuclear Information System (INIS)

    1981-08-01

    In 1981, 14 nuclear power reactors are in operation and 2 under construction in the Nordic countries. So far, the reactor waste originating from day-to-day operation of these plants has been stored in solidified form at the reactor sites. Within a few years a satisfactory disposal procedure needs to be established. While the main R and D effects in the waste field have earlier been devoted to the question of irradiated fuel and waste from reprocessing, there is therefore now an increased interest in reactor waste with its much lower radioactivity but somewhat larger volumes. Since 1977, efforts have been made in a joint Nordic study to examine which facts need to be known in order to perform a comprehensive safety assessment of a reactor waste management system. In the present study a Reference system related to the waste generated over 30 years from six 500 MW-reactors is examined. The dominating radionuclides during storage and transportation accident scenarios are Cs-134, Cs-137 and Co-60. For most of the release scenarios from repositories Cs-137 and Sr-90 are dominating. Some scenarios are, however, dominated by the very longlived nuclides I-129 and C-14. A closer examination of the concentration in the waste of these nuclides and of their leaching properties indicates that their small - but significant - influence, as calculated, is probably grossly overestimated. The mechanical stability obtained in routine solidification processes of reactor waste products in conjunction with the outer container (steel drum, transport container, etc.) turns out to be sufficient. Difficulties were encountered in applying ICRP methodology and available dose calculation methods to calculation of population doses due to small activity releases, and effects extending into the far future. (EG)

  19. Treatment and conditioning of radioactive waste solution by natural clay minerals

    International Nuclear Information System (INIS)

    El-Dessouky, M.I.; El-Massry, E.H.; Khalifa, S.M.; Aly, H.F.

    1999-01-01

    Natural inorganic exchangers. Was used to remove caesium, cobalt and europium using zinc sulfate as coagulant also different clay minerals. These calys include, feldrspare, aswanly, bentionite, hematite, mud, calcite, basalt, magnetite, kaoline sand stone, limonite and sand. The factros affecting the removal process namely PH, particle size, temperature and weight of the clay have been studied. Highest removal for Cs-137, Co-60 and Eu-152 and 154 was achived by asswanly and bentonite. Sand stone is more effective than the other clays. Removal of Cs-137 from low level waste solution is in the order the sequence, aswanly (85.5%)> bentonite (82.2%)> sandstone (65.4%). Solidified cement products have been evaluated to determine optimum conditions of mixing most sludges contained clays by testing mechanical strength and leaching rates of the waste products. The solidified waste forms were found more acceptable for handing, storage and ultimate disposal

  20. Psychometric Consequences of Subpopulation Item Parameter Drift

    Science.gov (United States)

    Huggins-Manley, Anne Corinne

    2017-01-01

    This study defines subpopulation item parameter drift (SIPD) as a change in item parameters over time that is dependent on subpopulations of examinees, and hypothesizes that the presence of SIPD in anchor items is associated with bias and/or lack of invariance in three psychometric outcomes. Results show that SIPD in anchor items is associated…

  1. Generalizability theory and item response theory

    NARCIS (Netherlands)

    Glas, Cornelis A.W.; Eggen, T.J.H.M.; Veldkamp, B.P.

    2012-01-01

    Item response theory is usually applied to items with a selected-response format, such as multiple choice items, whereas generalizability theory is usually applied to constructed-response tasks assessed by raters. However, in many situations, raters may use rating scales consisting of items with a

  2. Radioactive waste management from nuclear facilities

    International Nuclear Information System (INIS)

    2005-06-01

    This report has been published as a NSA (Nuclear Systems Association, Japan) commentary series, No. 13, and documents the present status on management of radioactive wastes produced from nuclear facilities in Japan and other countries as well. Risks for radiation accidents coming from radioactive waste disposal and storage together with risks for reactor accidents from nuclear power plants are now causing public anxiety. This commentary concerns among all high-level radioactive waste management from nuclear fuel cycle facilities, with including radioactive wastes from research institutes or hospitals. Also included is wastes produced from reactor decommissioning. For low-level radioactive wastes, the wastes is reduced in volume, solidified, and removed to the sites of storage depending on their radioactivities. For high-level radioactive wastes, some ten thousand years must be necessary before the radioactivity decays to the natural level and protection against seismic or volcanic activities, and terrorist attacks is unavoidable for final disposals. This inevitably results in underground disposal at least 300 m below the ground. Various proposals for the disposal and management for this and their evaluation techniques are described in the present document. (S. Ohno)

  3. Waste removal sequencing using ProdMod

    International Nuclear Information System (INIS)

    Paul, P.K.; Gregory, M.V.; Davis, N.R.; Brooke, J.N.

    1996-01-01

    The Savannah River Site (SRS) is starting to solidify its accumulated high-level radioactive waste into borosilicate glass in stainless steel canisters for eventual permanent storage. The in-tank precipitation process (ITP) and extended sludge processing (ESP) are two key operations in the waste processing complex. The supernate and dissolved salt from the waste storage tanks are transferred to the ITP process tank where the solution is decontaminated in batch processes. Soluble radioactive cesium is precipitated with sodium tetraphenylborate and strontium, uranium, and plutonium are adsorbed on monosodium titanate. The precipitate and adsorbent solids, which now contain the radionuclides, are concentrated using crossflow filters. The concentrated solids are sent to the high-level waste vitrification process. The decontaminated salt solution is sent to the low-level waste solidification process to form cement grout. In parallel with the precipitate operations, insoluble sludges that settled originally to the bottom of the waste tanks are reslurried and sent to ESP to undergo washing to reduce soluble salt content and aluminum dissolution, if required. In the vitrification process in the Defense Waste Processing Facility (DWPF), the concentrated precipitate from the ITP is mixed with the washed sludge from ESP and glass frit in proportion to form a stable borosilicate glass. A novel and fast-running Production Planning Model (ProdMod) has been developed to simulate the waste processing operation. This paper describes the application of ProdMod in sequencing the ITP batches and scheduling the ESP batches

  4. Spanish program on disposal of radioactive wastes

    International Nuclear Information System (INIS)

    Lopez Perez, B.; Ramos Salvador, L.; Martines Martinez, A.

    1977-01-01

    The Spanish Energetic Program assumes an installed nuclear electrical power of 23.000 MWe by the year 1985. Therefore, Spain is making an effort in the managment of radioactive wastes, that can be synthesized in the following points: 1.- Make-up and review of the regulation on the management of radioactive wastes. 2.- Development of the processes and equipment for the treatment of solid, liquid and gaseous wastes from the CNEN ''Juan Vigon'', as well as those from the Nuclear Center of Soria. Solidification studies of RAA wastes arisen from the reprocessing. 3.- Evaluation of radioactive waste treatment systems of the new installed nuclear power plants. Assistance to the nuclear and radioactive facilities operators. 4.- Increase the storage capacity of the pilot repository for solid radioactive wastes of categories 1 and 2 IAEA, located in Sierra Albarrana. Studies of adequate geological formation for storage of solid wastes of IAEA categories 3 and 4. 5.- Studies about long term surface storage systems for solidified RAA wastes arisen from the reprocessing [es

  5. Treatment of radioactive organics liquid wastes

    International Nuclear Information System (INIS)

    Morales Galarce, Tania

    1999-01-01

    Because of the danger that radioactive wastes can pose to society and to the environment a viable treatment alternative must be developed to prepare these wastes for final disposal. The waste studied in this work is a liquid organic waste contaminated with the radioisotope tritium. This must be treated and then changed into solid form in a 200 liter container. This study defined an optimum formulation that immobilizes the liquid waste. The organic waste is first submitted to an absorption treatment, with Celite absorbent, which had the best physical characteristics from the point of view of radioactive waste management. Then this was solidified by forming a cement mortar, using a highly resistant local cement, Polpaico 400. Various mixes were tested, with different water/cement, waste/absorbent and absorbed waste/cement ratios, until a mixture that met the quality control requirements was achieved. The optimum mixture obtained has a water/cement ratio of 0.35 (p/p) that is the amount of water needed to make the mixture workable, and minimum water for hydrating the cement; a waste/absorbent ration of 0.5 (v/v), where the organic liquid is totally absorbed, and is incorporated in the solid's crystalline network; and an absorbed waste/cement ratio of 0.8 (p/p), which represents the minimum amount of cement needed to obtain a solid product with the required mechanical resistance. The mixture's components join together with no problem, to produce a good workable mixture. It takes about 10 hours for the mixture to harden. After 14 days, the resulting solid product has a resistance to compression of 52 Kgf/cm2. The formulation contains 22.9% immobilized organic waste, 46.5% cement, 14.3% Celite and 16.3% water. Organic liquid waste can be treated and a solid product obtained, that meets the qualitative and quantitative parameters required for its disposal. (CW)

  6. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    International Nuclear Information System (INIS)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-01-01

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE's mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies

  7. Test plan for immobilization of salt-containing surrogate mixed wastes using polyester resins

    Energy Technology Data Exchange (ETDEWEB)

    Biyani, R.K.; Douglas, J.C.; Hendrickson, D.W.

    1997-07-07

    Past operations at many Department of Energy (DOE) sites have resulted in the generation of several waste streams with high salt content. These wastes contain listed and characteristic hazardous constituents and are radioactive. The salts contained in the wastes are primarily chloride, sulfate, nitrate, metal oxides, and hydroxides. DOE has placed these types of wastes under the purview of the Mixed Waste Focus Area (MWFA). The MWFA has been tasked with developing and facilitating the implementation of technologies to treat these wastes in support of customer needs and requirements. The MWFA has developed a Technology Development Requirements Document (TDRD), which specifies performance requirements for technology owners and developers to use as a framework in developing effective waste treatment solutions. This project will demonstrate the use of polyester resins in encapsulating and solidifying DOE`s mixed wastes containing salts, as an alternative to conventional and other emerging immobilization technologies.

  8. Generalizability theory and item response theory

    OpenAIRE

    Glas, Cornelis A.W.; Eggen, T.J.H.M.; Veldkamp, B.P.

    2012-01-01

    Item response theory is usually applied to items with a selected-response format, such as multiple choice items, whereas generalizability theory is usually applied to constructed-response tasks assessed by raters. However, in many situations, raters may use rating scales consisting of items with a selected-response format. This chapter presents a short overview of how item response theory and generalizability theory were integrated to model such assessments. Further, the precision of the esti...

  9. Method of processing solidification product of radioactive waste

    International Nuclear Information System (INIS)

    Daime, Fumiyoshi.

    1988-01-01

    Purpose: To improve the long-time stability of solidification products by providing solidification products with liquid tightness, gas tightness, abrasion resistance, etc., of the products in the course of the solidification for the treatment of radioactive wastes. Method: The surface of solidification products prepared by mixing solidifying agents with powder or pellets is entirely covered with high molecular polymer such as epoxy resin. The epoxy resin has excellent properties such as radiation-resistance, heat resistance, water proofness and chemical resistance, as well as have satisfactory mechanical properties. This can completely isolate the solidification products of radioactive wastes from the surrounding atmosphere. (Yoshino, Y.)

  10. 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

  11. Yucca Mountain Site Characterization Project Waste Package Plan

    International Nuclear Information System (INIS)

    Harrison-Giesler, D.J.; Jardine, L.J.

    1991-02-01

    The goal of the US Department of Energy's (DOE) Yucca Mountain Site Characterization Project (YMP) waste package program is to develop, confirm the effectiveness of, and document a design for a waste package and associated engineered barrier system (EBS) for spent nuclear fuel and solidified high-level nuclear waste (HLW) that meets the applicable regulatory requirements for a geologic repository. The Waste Package Plan describes the waste package program and establishes the technical approach against which overall progress can be measured. It provides guidance for execution and describes the essential elements of the program, including the objectives, technical plan, and management approach. The plan covers the time period up to the submission of a repository license application to the US Nuclear Regulatory Commission (NRC). 1 fig

  12. Solid waste processing experience at Susquehanna Steam Electric Station

    International Nuclear Information System (INIS)

    Phillips, J.W.; Granus, M.W.

    1984-01-01

    This paper reviews the first year's operation at the Susquehanna Steam Electric Station (SSES) with respect to the Westinghouse Hittman Nuclear Incorporated (Hittman) mobile solidification system and the dry activated waste generation, handling and processing. Experiences pertinent to the mobile solidification system are reviewed with emphasis on the integration of the system into the plant, problems associated with unexpected waste properties and the myriad of operating procedures that had to be prepared. The processing history for 1983 is reviewed in terms of the volume of waste, including solidified wastes, dewatered wastes an DAW. Factors that must be considered in evaluating processing alternatives, i.e., dewatering vs. solidification; steel liners vs. HICs, are discussed. Actions taken by Hittman and SSES to maximize the processing economics are also discussed. Finally, recommendations are provided to the utility considering implementing mobile solification services to ensure a smooth and timely integration of services into the plant

  13. 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

  14. Continuous using of the scaling factors for radionuclide evaluation in the packaged solid wastes originated from the Japanese Nuclear Power Plants since 2003

    International Nuclear Information System (INIS)

    2005-03-01

    The amounts and concentration of the nuclides in the waste packages are estimated by measuring some key nuclides, mostly gamma emitters, from outside of the packages and by applying the scaling factor method (using the relationship between some easy to measure key nuclides and the other difficult to measure nuclides). The solid wastes are classified into two kinds of packages: homogeneous solid wastes made from concentrated liquid wastes and spent fuels solidified with cement asphalt, or plastics and heterogeneous solid wastes made of cutting metals, compacted or fused filters solidified with mortars. Japan Nuclear Energy Safety Organization (JNES) established in 2005 is in charge of the confirmation of the inside contents with radionuclide information and compliance with formalities for safety maintenance and control. (S. Ohno)

  15. EXAMPLE OF A RISK-BASED DISPOSAL APPROVAL: SOLIDIFICATION OF HANFORD SITE TRANSURANIC (TRU) WASTE

    International Nuclear Information System (INIS)

    PRIGNANO AL

    2007-01-01

    The Hanford Site requested, and the U.S. Environmental Protection Agency (EPA) Region 10 approved, a Toxic Substances Control Act of 1976 (TSCA) risk-based disposal approval (RBDA) for solidifying approximately four cubic meters of waste from a specific area of one of the K East Basin: the North Loadout Pit (NLOP). The NLOP waste is a highly radioactive sludge that contained polychlorinated biphenyls (PCBs) regulated under TSCA. The prescribed disposal method for liquid PCB waste under TSCA regulations is either thermal treatment or decontamination. Due to the radioactive nature of the waste, however, neither thermal treatment nor decontamination was a viable option. As a result, the proposed treatment consisted of solidifying the material to comply with waste acceptance criteria at the Waste Isolation Pilot Plant (WPP) in Carlsbad, New Mexico, or possibly the Environmental Restoration Disposal Facility at the Hanford Site, depending on the resulting transuranic (TRU) content of the stabilized waste. The RBDA evaluated environmental risks associated with potential airborne PCBs. In addition, the RBDA made use of waste management controls already in place at the treatment unit. The treatment unit, the T Plant Complex, is a Resource Conservation and Recovery Act of 1976 (RCRA)-permitted facility used for storing and treating radioactive waste. The EPA found that the proposed activities did not pose an unreasonable risk to human health or the environment. Treatment took place from October 26,2005 to June 9,2006, and 332 208-liter (55-gallon) containers of solidified waste were produced. All treated drums assayed to date are TRU and will be disposed at WIPP

  16. 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

  17. 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

  18. Teoria da Resposta ao Item Teoria de la respuesta al item Item response theory

    Directory of Open Access Journals (Sweden)

    Eutalia Aparecida Candido de Araujo

    2009-12-01

    Full Text Available A preocupação com medidas de traços psicológicos é antiga, sendo que muitos estudos e propostas de métodos foram desenvolvidos no sentido de alcançar este objetivo. Entre os trabalhos propostos, destaca-se a Teoria da Resposta ao Item (TRI que, a princípio, veio completar limitações da Teoria Clássica de Medidas, empregada em larga escala até hoje na medida de traços psicológicos. O ponto principal da TRI é que ela leva em consideração o item particularmente, sem relevar os escores totais; portanto, as conclusões não dependem apenas do teste ou questionário, mas de cada item que o compõe. Este artigo propõe-se a apresentar esta Teoria que revolucionou a teoria de medidas.La preocupación con las medidas de los rasgos psicológicos es antigua y muchos estudios y propuestas de métodos fueron desarrollados para lograr este objetivo. Entre estas propuestas de trabajo se incluye la Teoría de la Respuesta al Ítem (TRI que, en principio, vino a completar las limitaciones de la Teoría Clásica de los Tests, ampliamente utilizada hasta hoy en la medida de los rasgos psicológicos. El punto principal de la TRI es que se tiene en cuenta el punto concreto, sin relevar las puntuaciones totales; por lo tanto, los resultados no sólo dependen de la prueba o cuestionario, sino que de cada ítem que lo compone. En este artículo se propone presentar la Teoría que revolucionó la teoría de medidas.The concern with measures of psychological traits is old and many studies and proposals of methods were developed to achieve this goal. Among these proposed methods highlights the Item Response Theory (IRT that, in principle, came to complete limitations of the Classical Test Theory, which is widely used until nowadays in the measurement of psychological traits. The main point of IRT is that it takes into account the item in particular, not relieving the total scores; therefore, the findings do not only depend on the test or questionnaire

  19. Development of thermal conditioning technology for alpha-contaminated wastes: a study on leaching characteristics and long-term safety assessment of simulated waste forms

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Yong Chil [Yonsei University, Seoul (Korea); Lee, Sang Hoon; Yoo, Jong Ik; Choi, Yong Cheol [Yonsei University, Seoul (Korea)

    2001-04-01

    Radioactive wastes should be stabilized for safe management during several hundred years. To assess stability of solidified waste forms, mechanical properties and chemical durability of the waste forms should be analyzed. Chemical durability is one of the most important factors in the assessment of waste forms, which could be examined by leaching tests. Various methods in leaching test are suggested by different organizations, but a formal test method in Korea is not ready yet. Therefore, the leaching test method applicable to various constituents is necessary for the safe management of radioactive wastes In this study, leaching behavior and characteristics of components such as solidification materials, heavy metals and radioactive nuclids were analyzed for cement waste form and glassy waste form. 58 refs., 25 figs., 8 tabs. (Author)

  20. 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

  1. Process Waste Assessment - Paint Shop

    International Nuclear Information System (INIS)

    Phillips, N.M.

    1993-06-01

    This Process Waste Assessment was conducted to evaluate hazardous wastes generated in the Paint Shop, Building 913, Room 130. Special attention is given to waste streams generated by the spray painting process because it requires a number of steps for preparing, priming, and painting an object. Also, the spray paint booth covers the largest area in R-130. The largest and most costly waste stream to dispose of is open-quote Paint Shop wasteclose quotes -- a combination of paint cans, rags, sticks, filters, and paper containers. These items are compacted in 55-gallon drums and disposed of as solid hazardous waste. Recommendations are made for minimizing waste in the Paint Shop. Paint Shop personnel are very aware of the need to minimize hazardous wastes and are continuously looking for opportunities to do so

  2. Centralized collection of radioactive wastes

    International Nuclear Information System (INIS)

    1985-06-01

    The standard based upon TGL-190-921/02 applies to solid wastes of the category A1 and the radiation protection groups S1 and S2. The following items are specified: (1) requirements concerning the form and properties of the waste (permitted composition, unpermitted components, type of packaging, maximum weight per package/container), (2) technical conditions for connecting technical means of collection (lifting devices, traffic connections) with customer, and (3) tasks in handing/taking over the waste in relation to waste type (controls, operation of facilities, decontamination, transport documents)

  3. Centralized collection of radioactive wastes

    International Nuclear Information System (INIS)

    1985-06-01

    The standard based upon TGL-190-921/03 applies to solid wastes of the category A2 and the radiation protection groups S3, S4 and S5. The following items are specified: (1) requirements concerning the form and properties of the waste (permitted composition, unpermitted components, type of packaging, maximum weight per package/container), (2) technical conditions for connecting technical means of collection (lifting devices, traffic connections) with customer, and (3) tasks in handing/taking over the waste in relation to waste type (controls, operation of facilities, decontamination, transport documents)

  4. 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.)

  5. In-situ thermoelectric stabilization of radioactive wastes

    International Nuclear Information System (INIS)

    Brouns, R.A.; Timmerman, C.L.

    1982-02-01

    A new process for stabilizing buried radioactive wastes without exhumation is being developed by Pacific Northwest Laboratory (PNL). The process, known as in situ vitrification, converts waste and contaminated soil to a durable glass and crystalline material by passing an electric current between electrodes placed in the ground. Joule heating created by the flowing current has generated temperatures over 1700 0 C which cause the soil to melt and dissolve or encapsulate the wastes. Engineering-scale tests conducted in the laboratory have melted approximately 45 kgs (30 liters) of soil at a time by this technique. Encouraging results from these engineering-scale tests led to the design and construction of a pilot-scale field test unit which has solidified approximately 9000 kg of simulated contaminated soil per test. Test results and evaluations to date have been very promising. No detectable migration of hazardous species into uncontaminated soil has been found, and volatilization during melting has been very low. Leach studies have found the vitrified soil to be a highly durable waste form similar to pyrex glass. Electrical power costs to solidify a disposal site have been calculated at less than $70 per cubic meter ($2/ft 3 ) of waste. Future activities include both radioactive and nonradioactive pilot and large-scale tests

  6. 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

  7. Characterisation of bitumenised waste in SFR 1

    International Nuclear Information System (INIS)

    Pettersson, Michael; Elert, M.

    2001-06-01

    The waste deposited in the Final Repository for Radioactive Operational Waste, SFR, consists in part of waste solidified in bitumen. Bitumen is considered to have favourable chemical and physical properties to act as a fixation material for radioactive waste. However, during interim storage and subsequent disposal bitumen's properties may change. This may influence the stability of the bitumen matrix to retain radionuclides. This report discusses different processes affecting the long-term performance of bitumenised waste, and an evaluation of these properties in waste deposited in SFR 1 is made. The possible effect of a bitumen barrier on the release rate of radionuclides from SFR 1 is assessed. Based on leaching experiments reviewed in this study, it could take some thousand years, possibly more, to release all radionuclides in a 200-litre drum. The results are, however, extrapolated from experiments performed during a short period of time. Long- term deteriorating effects and the effect of a low temperature on the bitumen matrix are not very well documented. The literature focuses principally on bitumenised evaporator concentrate, but the bitumenised waste deposited in SFR 1 consists mainly of ion exchange resins. There are indications that the non-radioactive waste products usually investigated overestimate bitumen's ability to retain waste. Radiolytic effects has been estimated in this work to be negligible for waste categories F.17, F.20 and B.20 deposited in SFR 1, but for categories B.05, B.06 and F.18 the possibility of increased water uptake rate due to radiolysis can not be excluded. A more reasonable assumption is that bitumen will act as an effective barrier for radionuclide release during a time span from some hundreds to thousand of years. Generally, the majority of the inventory of radionuclides in SFR 1 is not solidified in bitumen. By taking the bitumen barrier into account in the modelling of release of radio- nuclides from SFR 1, the total

  8. Characterisation of bitumenised waste in SFR 1

    Energy Technology Data Exchange (ETDEWEB)

    Pettersson, Michael; Elert, M. [Kemakta Konsult AB, Stockholm (Sweden)

    2001-06-01

    The waste deposited in the Final Repository for Radioactive Operational Waste, SFR, consists in part of waste solidified in bitumen. Bitumen is considered to have favourable chemical and physical properties to act as a fixation material for radioactive waste. However, during interim storage and subsequent disposal bitumen's properties may change. This may influence the stability of the bitumen matrix to retain radionuclides. This report discusses different processes affecting the long-term performance of bitumenised waste, and an evaluation of these properties in waste deposited in SFR 1 is made. The possible effect of a bitumen barrier on the release rate of radionuclides from SFR 1 is assessed. Based on leaching experiments reviewed in this study, it could take some thousand years, possibly more, to release all radionuclides in a 200-litre drum. The results are, however, extrapolated from experiments performed during a short period of time. Long- term deteriorating effects and the effect of a low temperature on the bitumen matrix are not very well documented. The literature focuses principally on bitumenised evaporator concentrate, but the bitumenised waste deposited in SFR 1 consists mainly of ion exchange resins. There are indications that the non-radioactive waste products usually investigated overestimate bitumen's ability to retain waste. Radiolytic effects has been estimated in this work to be negligible for waste categories F.17, F.20 and B.20 deposited in SFR 1, but for categories B.05, B.06 and F.18 the possibility of increased water uptake rate due to radiolysis can not be excluded. A more reasonable assumption is that bitumen will act as an effective barrier for radionuclide release during a time span from some hundreds to thousand of years. Generally, the majority of the inventory of radionuclides in SFR 1 is not solidified in bitumen. By taking the bitumen barrier into account in the modelling of release of radio- nuclides from SFR 1, the

  9. Sharing the cost of redundant items

    DEFF Research Database (Denmark)

    Hougaard, Jens Leth; Moulin, Hervé

    2014-01-01

    We ask how to share the cost of finitely many public goods (items) among users with different needs: some smaller subsets of items are enough to serve the needs of each user, yet the cost of all items must be covered, even if this entails inefficiently paying for redundant items. Typical examples...... are network connectivity problems when an existing (possibly inefficient) network must be maintained. We axiomatize a family cost ratios based on simple liability indices, one for each agent and for each item, measuring the relative worth of this item across agents, and generating cost allocation rules...... additive in costs....

  10. 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.

  11. 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

  12. 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)

  13. 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.

  14. 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%

  15. 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.

  16. 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.

  17. 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)

  18. 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.

  19. 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.

  20. 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.