Corrosion-Resistant Container for Molten-Material Processing

Science.gov (United States)

Stern, Theodore G.; McNaul, Eric

2010-01-01

In a carbothermal process, gaseous methane is passed over molten regolith, which is heated past its melting point to a temperature in excess of 1,625 C. At this temperature, materials in contact with the molten regolith (or regolith simulant) corrode and lose their structural properties. As a result, fabricating a crucible to hold the molten material and providing a method of contact heating have been problematic. Alternative containment approaches use a large crucible and limit the heat zone of the material being processed, which is inefficient because of volume and mass constraints. Alternative heating approaches use non-contact heating, such as by laser or concentrated solar energy, which can be inefficient in transferring heat and thus require higher power heat sources to accomplish processing. The innovation is a combination of materials, with a substrate material having high structural strength and stiffness and high-temperature capability, and a coating material with a high corrosion resistance and high-temperature capability. The material developed is a molybdenum substrate with an iridium coating. Creating the containment crucible or heater jacket using this material combination requires only that the molybdenum, which is easily processed by conventional methods such as milling, electric discharge machining, or forming and brazing, be fabricated into an appropriate shape, and that the iridium coating be applied to any surfaces that may come in contact with the corrosive molten material. In one engineering application, the molybdenum was fashioned into a container for a heat pipe. Since only the end of the heat pipe is used to heat the regolith, the container has a narrowing end with a nipple in which the heat pipe is snugly fit, and the external area of this nipple, which contacts the regolith to transfer heat into it, is coated with iridium. At the time of this reporting, no single material has been found that can perform the functions of this combination

Characterization of plutonium-containing materials and storage canisters

International Nuclear Information System (INIS)

Mason, R.E.

1997-01-01

Throughout the weapons complex, plutonium materials are stored in various containers. Some plutonium has been stored for 20 yr or more. The physical and chemical properties of the plutonium material and the containers that hold it are often not well characterized. The U.S. Department of Energy (DOE) 3013 standard sets criteria to which stored material must conform. The 3013 standard regulates materials that hold 50% or greater plutonium, and the other 50% is not specified and is usually unknown. The Materials Identification and Surveillance project is tasked to characterize representative materials and begin to characterize the other 50% and to show that materials can be brought into 3013 criteria conformance through thermal treatments

Transport of bundles and equipment which contain radioactive material

International Nuclear Information System (INIS)

1987-01-01

This norm settles down: 1) The requirements that should be completed in relation to safety precautions and protection against ionizing radiations during the transport radioactive material and/or equipment containing it, in order to avoid risks to the collective and the environment. 2) The basic information on procedures that will be completed in the event of happening accidents during the transport or the transit storage of radioactive material and/or equipment that contain it. 3) The measures of security and physical protection during the transport of radioactive material and/or equipment containing it. This norm is applied: 1) To all the ways of transport (by air, by ground and by ship, fluvial and marine) of radioactive material and/or equipment that contain it. 2) To all natural or legal, public or private person, devoted to install, produce, trade, market, import or export radioactive materials and/or equipment containing it, and that needs to transport them as main or secondary activity [es

Candidate container materials for Yucca Mountain waste package designs

International Nuclear Information System (INIS)

McCright, R.D.; Halsey, W.G.; Gdowski, G.E.; Clarke, W.L.

1991-09-01

Materials considered as candidates for fabricating nuclear waste containers are reviewed in the context of the Conceptual Design phase of a potential repository located at Yucca Mountain. A selection criteria has been written for evaluation of candidate materials for the next phase -- Advanced Conceptual Design. The selection criteria is based on the conceptual design of a thin-walled container fabricated from a single metal or alloy; the criteria consider the performance requirements on the container and the service environment in which the containers will be emplaced. A long list of candidate materials is evaluated against the criteria, and a short list of materials is proposed for advanced characterization in the next design phase

Disposal containers for radioactive waste materials and separation systems for radioactive waste materials

International Nuclear Information System (INIS)

Rubin, L.S.

1986-01-01

A separation system for dewatering radioactive waste materials includes a disposal container, drive structure for receiving the container, and means for releasably attaching the container to the drive structure. The separation structure disposed in the container adjacent the inner surface of the side wall structure retains solids while allowing passage of liquids. The inlet port structure in the container top wall is normally closed by first valve structure that is centrifugally actuated to open the inlet port and the discharge port structure at the container periphery receives liquid that passes through the separation structure and is normally closed by a second valve structure that is centrifugally actuated to open the discharge ports. The container also includes a coupling structure for releasable engagement with the centrifugal drive structure. The centrifugal force produced when the container is driven in rotation by the drive structure opens the valve structures, and radioactive waste material introduced into the container through the open inlet port is dewatered, and the waste is compacted. The ports are automatically closed by the valves when the container drum is not subjected to centrifugal force such that containment effectiveness is enhanced and exposure of personnel to radioactive materials is minimized. (author)

Risk ranking of LANL nuclear material storage containers for repackaging prioritization.

Science.gov (United States)

Smith, Paul H; Jordan, Hans; Hoffman, Jenifer A; Eller, P Gary; Balkey, Simon

2007-05-01

Safe handling and storage of nuclear material at U.S. Department of Energy facilities relies on the use of robust containers to prevent container breaches and subsequent worker contamination and uptake. The U.S. Department of Energy has no uniform requirements for packaging and storage of nuclear materials other than those declared excess and packaged to DOE-STD-3013-2000. This report describes a methodology for prioritizing a large inventory of nuclear material containers so that the highest risk containers are repackaged first. The methodology utilizes expert judgment to assign respirable fractions and reactivity factors to accountable levels of nuclear material at Los Alamos National Laboratory. A relative risk factor is assigned to each nuclear material container based on a calculated dose to a worker due to a failed container barrier and a calculated probability of container failure based on material reactivity and container age. This risk-based methodology is being applied at LANL to repackage the highest risk materials first and, thus, accelerate the reduction of risk to nuclear material handlers.

Storage and transport containers for radioactive medical materials

International Nuclear Information System (INIS)

Suthanthiran, K.

1989-01-01

This patent describes a storage and transport container for small-diameter ribbon-like lengths of material including radioactive substances for use in medical treatments, comprising: an exterior shell for radiation shielding metal having top and bottom members of radiation shielding metal integral therewith; radiation shielding metal extending downward from the top of the container and forming a central cavity, the central cavity being separate from the exterior shell material of the container and extending downwardly a distance less than the height of the container; a plurality of small diameter carrier tubes located within the interior of the container and having one end of each tube opening through one side of the container and the other end of such tube opening through the opposite lateral side of the container with the central portion of each tube passing under the central cavity; and a plug of radiation shielding metal removably located in the top the central cavity for shielding the radiation from radiation sources located within the container

Assessment of materials for nuclear fuel immobilization containers

Energy Technology Data Exchange (ETDEWEB)

Nuttall, K; Urbanic, V F

1981-09-01

A wide range of engineering metals and alloys has been assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The expected range of service conditions in the disposal vault are discussed, as well as the material properties required for this application. An important requirement is that the container last at least 500 years without being breached. The assessment is treated in two parts. Part I concentrates on the physical and mechanical metallurgy, with special reference to strength, weldability, potential embrittlement mechanisms and some economic aspects. Part II discusses possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditions in the vault. Localized corrosion and delayed fracture processes are identified as being most likely to limit container lifetime. Hence an essential requirement is that such processes either be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further consideration as possible container materials: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the enviromental conditions in the vault. 42 figures, 31 tables.

Asbestos-Containing Materials (ACM) and Demolition

Science.gov (United States)

There are specific federal regulatory requirements that require the identification of asbestos-containing materials (ACM) in many of the residential buildings that are being demolished or renovated by a municipality.

Buried waste containment system materials. Final Report

International Nuclear Information System (INIS)

Weidner, J.R.; Shaw, P.G.

1997-10-01

This report describes the results of a test program to validate the application of a latex-modified cement formulation for use with the Buried Waste Containment System (BWCS) process during a proof of principle (POP) demonstration. The test program included three objectives. One objective was to validate the barrier material mix formulation to be used with the BWCS equipment. A basic mix formula for initial trials was supplied by the cement and latex vendors. The suitability of the material for BWCS application was verified by laboratory testing at the Idaho National Engineering and Environmental Laboratory (INEEL). A second objective was to determine if the POP BWCS material emplacement process adversely affected the barrier material properties. This objective was met by measuring and comparing properties of material prepared in the INEEL Materials Testing Laboratory (MTL) with identical properties of material produced by the BWCS field tests. These measurements included hydraulic conductivity to determine if the material met the US Environmental Protection Agency (EPA) requirements for barriers used for hazardous waste sites, petrographic analysis to allow an assessment of barrier material separation and segregation during emplacement, and a set of mechanical property tests typical of concrete characterization. The third objective was to measure the hydraulic properties of barrier material containing a stop-start joint to determine if such a feature would meet the EPA requirements for hazardous waste site barriers

Mineralogical conversion of asbestos containing materials

International Nuclear Information System (INIS)

Pulsford, S.K.; Foltz, A.D.; Ek, R.B.

1996-01-01

The principal objective of the Technical Task Plan (TTP) is to demonstrate a thermal-chemical mineralogical asbestos conversion unit at the Hanford Site, which converts non-radiological asbestos containing materials (ACMs) into an asbestos-free material. The permanent thermal-chemical mineralogical conversion of ACMs to a non-toxic, non-hazardous, potentially marketable end product should not only significantly reduce the waste stream volumes but terminate the open-quotes cradle to graveclose quotes ownership liabilities

Radioactive material dry-storage facility and radioactive material containing method

International Nuclear Information System (INIS)

Kanai, Hidetoshi; Kumagaya, Naomi; Ganda, Takao.

1997-01-01

The present invention provides a radioactive material dry storage facility which can unify the cooling efficiency of a containing tube and lower the pressure loss in a storage chamber. Namely, a cylindrical body surrounds a first containing tube situated on the side of an air discharge portion among a plurality of containing tubes and forms an annular channel extending axially between the cylindrical body and the first containing tube. An air flow channel partitioning member is disposed below a second containing tube situated closer to an air charging portion than the first containing tube. A first air flow channel is formed below the air channel partitioning member extending from the air charging portion to the annular channel. The second air channel is formed above the air channel partitioning member and extends from the air charging portion to the air discharge portion by way of a portion between the second containing tubes and the portion between the cylindrical body and the first containing tube. Then, low temperature air can be led from the air charging portion to the periphery of the first containing tube. The effect of cooling the first containing tube can be enhanced. The difference between the cooling efficiency between the second containing tube and the first containing tube is decreased. (I.S.)

Treating agent for urea containing radioactive materials

International Nuclear Information System (INIS)

Ogawa, Hiroshi; Maki, Kentaro.

1973-01-01

Object: To add a coagulant into urea containing radioactive material to precipitate and remove the radioactive material in the urea. Structure: Iodosalt is added into urea and next, a mixed reagent in which silver ion or silver acetic ion and iron hydroxide precipitation or ferrite ion coexist is added therein. The urea is treated to have a sufficient alkaline, after which it is introduced into a basket type centrifuge formed with a filter layer in combination of an upper glass fiber layer and a lower active carbon layer. The treating agent can uniformly remove radioactive ion and radioactive chelate within urea containing inorganic salt and various metabolites. (Nakamura, S.)

Nondestructive Examination Of Plutonium-Bearing Material Containers

International Nuclear Information System (INIS)

Yerger, L.; Mcclard, J.; Traver, L.; Grim, T.

2010-01-01

The first nondestructive examination (NDE) of 3013-type containers as part of the Department of Energy's (DOE's) Integrated Surveillance Program (ISP) was performed in February, 2005. Since that date 280 NDE surveillances on 255 containers have been conducted. These containers were packaged with plutonium-bearing materials at multiple DOE sites. The NDE surveillances were conducted at Hanford, Lawrence Livermore National Laboratory (LLNL), and Savannah River Site (SRS). These NDEs consisted of visual inspection, mass verification, radiological surveys, prompt gamma analysis, and radiography. The primary purpose of performing NDE surveillances is to determine if there has been a significant pressure buildup inside the inner 3013 container. This is done by measuring the lid deflection of the inner 3013 container using radiography images. These lid deflection measurements are converted to pressure measurements to determine if a container has a pressure of a 100 psig or greater. Making this determination is required by Surveillance and Monitoring Plan (S and MP). All 3013 containers are designed to withstand at least 699 psig as specified by DOE-STD-3013. To date, all containers evaluated have pressures under 50 psig. In addition, the radiography is useful in evaluating the contents of the 3013 container as well as determining the condition of the walls of the inner 3013 container and the convenience containers. The radiography has shown no signs of degradation of any container, but has revealed two packaging anomalies. Quantitative pressure measurements based on lid deflections, which give more information than the 'less than or greater than 100 psig' (pass/fail) data are also available for many containers. Statistical analyses of the pass/fail data combined with analysis of the quantitative data show that it is extremely unlikely that any container in the population of 3013 containers considered in this study (e.g., containers packaged according to the DOE-STD-3013

Reuse of Material Containing Natural Radionuclides - 12444

Energy Technology Data Exchange (ETDEWEB)

Metlyaev, E.G.; Novikova, N.J. [Burnasyan Federal Medical Biophysical Centre, Moscow (Russian Federation)

2012-07-01

Disposal of and use of wastes containing natural radioactive material (NORM) or technologically enhanced natural radioactive material (TENORM) with excessive natural background as a building material is very important in the supervision body activity. At the present time, the residents of Octyabrsky village are under resettlement. This village is located just near the Priargunsky mining and chemical combine (Ltd. 'PPGHO'), one of the oldest uranium mines in our country. The vacated wooden houses in the village are demolished and partly used as a building material. To address the issue of potential radiation hazard of the wooden beams originating from demolition of houses in Octyabrsky village, the contents of the natural radionuclides (K-40, Th-232, Ra-226, U- 238) are being determined in samples of the wooden beams of houses. The NORM contents in the wooden house samples are higher, on average, than their content in the reference sample of the fresh wood shavings, but the range of values is rather large. According to the classification of waste containing the natural radionuclides, its evaluation is based on the effective specific activity. At the effective specific activity lower 1.5 kBq/kg and gamma dose rate lower 70 μR/h, the material is not considered as waste and can be used in building by 1 - 3 classes depending upon A{sub eff} value. At 1.5 kBq/kg < A{sub eff} ≤ 4 kBq/kg (4 class), the wooden beams might be used for the purpose of the industrial building, if sum of ratios between the radionuclide specific activity and its specific activity of minimum significance is lower than unit. The material classified as the waste containing the natural radionuclides has A{sub eff} higher 1.5 kBq /kg, and its usage for the purpose of house-building and road construction is forbidden. As for the ash classification and its future usage, such usage is unreasonable, because, according to the provided material, more than 50% of ash samples are considered as

Storage vessel for containing radiation contaminated material

International Nuclear Information System (INIS)

Ogawa, Kazuya.

1995-01-01

A container pipe and an outer pipe are coaxially assembled integrally in a state where securing spacers are disposed between the container pipe and the outer pipe, and an annular flow channel is formed around the container pipe. Radiation contaminated material-containing body (glass solidified package) is contained in the container pipe. The container pipe and the outer pipe in an integrated state are suspended from a ceiling plug of a cell chamber of a storage vessel, and supporting devices are assembled between the pipes and a support structure. A shear/lug mechanism is used for the supporting devices. The combination of the shear/lug allows radial and vertical movement but restrict horizontal movement of the outer tube. The supporting devices are assembled while visually recognizing the state of the shear/lug mechanism between the outer pipe and the support mechanism. Accordingly, operationability upon assembling the container pipe and the outer pipe is improved. (I.N.)

Self-closing shielded container for use with radioactive materials

Science.gov (United States)

Smith, J.E.

A container for storage of radioactive material comprises a container body and a closure member. The closure member is coupled to the container body to enable the closure body to move automatically from a first position (e.g., closed) to a second position (open).

Guidelines for Assessment and Abatement of Asbestos-Containing Materials in Buildings.

Science.gov (United States)

Pielert, James H.; Mathey, Robert G.

This report presents guidelines, based on available information, for the assessment and abatement of asbestos-containing materials in buildings. Section 1 provides background information on the history and use of asbestos-containing products in buildings, the characteristics of asbestos fibers, products and materials containing asbestos, and…

An assessment of materials for nuclear fuel immobilization containers

International Nuclear Information System (INIS)

Nuttall, K.; Urbanic, V.F.

1981-09-01

A wide range of engineering metals and alloys was assessed for their suitability as container materials for irradiated nuclear fuel intended for permanent disposal in a deep, underground hard-rock vault. The container must last at least 500 years without being breached. Materials were assessed for their physical and mechanical metallurgy, weldability, potential embrittlement mechanisms, and economics. A study of the possible mechanisms of metallic corrosion for the various engineering alloys and the expected range of environmental conditons in the vault showed that localized corrosion and delayed fracture processes are the most likely to limit container lifetime. Thus such processes either must be absent or proceed at an insignificant rate. Three groups of alloys are recommended for further study: AISI 300 series austenitic stainless steels, high nickel-base alloys and very dilute titanium-base alloys. Specific alloys from each group are indicated as having the optimum combination of required properties, including cost. For container designs where the outer container shell does not independently support the service loads, copper should also be considered. The final material selection will depend primarily on the environmental conditions in the vault

Surveillance of sealed containers with plutonium oxide materials (ms163)

Science.gov (United States)

Worl, Laura; Berg, John; Ford, Doris; Martinez, Max; McFarlan, Jim; Morris, John; Padilla, Dennis; Rau, Karen; Smith, Coleman; Veirs, Kirk; Hill, Dallas; Prenger, Coyne

2000-07-01

DOE is embarking upon a program to store large quantities of plutonium-bearing materials for up to fifty years. Materials destined for long-term storage include metals and oxides that are stabilized and packaged according to the DOE storage standard, where the packaging consists of two nested, welded, stainless steel containers. We have designed instrumented storage containers that mimic the inner storage can specified in the 3013 standard at both full- and small-scale capacities (2.4 liter and 0.005 liter, respectively), Figures 1 and 2. The containers are designed to maintain the volume to material mass ratio while allowing the gas composition and pressure to be monitored over time.

Double container system for the transport and storage of radioactive materials

International Nuclear Information System (INIS)

Popp, F.W.; Pontani, B.; Ernst, E.

1987-01-01

The double container system consists of an inner storage container made of steel for the gastight inclusion of the radioactive material to be stored and an outer shielding container which ensures the necessary shielding and mechanical safety in handling and transport. A neutron moderator layer of material containing hydrogen, preferably polyethylene, is present in the annular gap between the outer shielding container and the inner storage container. In order to achieve good shielding with simultaneous very good heat conduction from the inside to the outside, the moderator layer consists of individual polyethylene rings stacked above one another. There is an H profile ring made of heat conducting metal material between each two polyethylene rings. The legs of the H profile ring surround the sides of the two polyethylene rings for fixing it. (orig.) [de

SULPHUR DIOXIDE LEACHING OF URANIUM CONTAINING MATERIAL

Science.gov (United States)

Thunaes, A.; Rabbits, F.T.; Hester, K.D.; Smith, H.W.

1958-12-01

A process is described for extracting uranlum from uranium containing material, such as a low grade pitchblende ore, or mill taillngs, where at least part of the uraniunn is in the +4 oxidation state. After comminuting and magnetically removing any entrained lron particles the general material is made up as an aqueous slurry containing added ferric and manganese salts and treated with sulfur dioxide and aeration to an extent sufficient to form a proportion of oxysulfur acids to give a pH of about 1 to 2 but insufficient to cause excessive removal of the sulfur dioxide gas. After separating from the solids, the leach solution is adjusted to a pH of about 1.25, then treated with metallic iron in the presence of a precipitant such as a soluble phosphate, arsonate, or fluoride.

49 CFR 176.76 - Transport vehicles, freight containers, and portable tanks containing hazardous materials.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Transport vehicles, freight containers, and... TRANSPORTATION HAZARDOUS MATERIALS REGULATIONS CARRIAGE BY VESSEL General Handling and Stowage § 176.76 Transport... paragraphs (b) through (f) of this section, hazardous materials authorized to be transported by vessel may be...

Corrosion aspects of steel radioactive waste containers in cementitious materials

International Nuclear Information System (INIS)

Smart, Nick

2012-01-01

Nick Smart from Serco, UK, gave an overview of the effects of cementitious materials on the corrosion of steel during storage and disposal of various low- and intermediate-level radioactive wastes. Steel containers are often used as an overpack for the containment of radioactive wastes and are routinely stored in an open atmosphere. Since this is an aerobic and typically humid environment, the steel containers can start to corrode whilst in storage. Steel containers often come into contact with cementitious materials (e.g. grout encapsulants, backfill). An extensive account of different steel container designs and of steel corrosion mechanisms was provided. Steel corrosion rates under conditions buffered by cementitious materials have been evaluated experimentally. The main conclusion was that the cementitious environment generally facilitates the passivation of steel materials. Several general and localised corrosion mechanisms need to be considered when evaluating the performance of steel containers in cementitious environments, and environmental thresholds can be defined and used with this aim. In addition, the consequences of the generation of gaseous hydrogen by the corrosion of carbon steel under anoxic conditions must be taken into account. Discussion of the paper included: Is crevice corrosion really significant in cementitious systems? Crevice corrosion is unlikely in the cementitious backfill considered because it will tend to neutralise any acidic conditions in the crevice. What is the role of microbially-induced corrosion (MIC) in cementitious systems? Microbes are likely to be present in a disposal facility but their effect on corrosion is uncertain

Application of electron irradiation to food containers and packaging materials

International Nuclear Information System (INIS)

Ueno, Koji

2010-01-01

Problems caused by microbial contamination and hazardous chemicals have attracted much attention in the food industry. The number of systems such as hygienic management systems and Hazard Analysis Critical Control Point (HACCP) systems adopted in the manufacturing process is increasing. As manufacturing process control has become stricter, stricter control is also required for microbial control for containers and packaging materials (from disinfection to sterilization). Since safe and reliable methods for sterilizing food containers and packaging materials that leave no residue are required, electron beam sterilization used for medical equipment has attracted attention from the food industry. This paper describes an electron irradiation facility, methods for applying electron beams to food containers and packaging materials, and products irradiated with electron beams. (author)

System for indicating the level of material in a container

International Nuclear Information System (INIS)

Erb, T.L.

1980-01-01

In a radiation detecting system for controlling the level of material in a container, the first counter accumulates pulses generated by a geiger tube at a rate related to the level of material and a second counter accumulates clock pulses. A race condition is established between a NAND circuit indicating that the first counter has reached a predetermined total, and a NAND circuit indicating that the second counter has reached a second predetermined total representing a fixed counting interval. The first NAND circuit to respond to its predetermined total actuates a circuit to reset both counters and, if indicative of the material level being below a predetermined minimum, actuates an alarm or operates a control circuit to add material to the container. In the example shown, an additional NAND circuit responds to a different count in the first counter which count in the same time interval corresponds to a higher level, and when material is being added to the container, the race condition is between two NAND circuits. The effect of this is to provide a hysteresis effect preventing the circuit from 'hunting' around one level of material. (author)

Anthropogenic materials and products containing natural radionuclides. Pt. 1a. Radiation properties of raw materials and waste materials. A literature study

International Nuclear Information System (INIS)

Reichelt, A.; Roehrer, J.; Lehmann, K.H.

1995-12-01

Cased on the literature study, the publication presents relevant data on raw materials and wastes containing natural radionuclides. The study is part 1a of the project on ''Anthropogenic materials and waste materials containing natural radionuclides''. Part 1 of the project gives data and information on about 100 different materials and wastes or products for household or industrial applications which contain significant amounts of natural radioactivity. In addition, part 1 presents for some of these materials information on their applications, consumption, radioactivity and resulting radiation doses. The raw materials and waste materials on the list in part 1 are characterised in this 1a report. Wherever appropriate, two or more materials are dealt with in one chapter, as e.g. felspar and felspar sands (pegmatite), talcum, and soapstone. The wastes are dealt with in the chapters discussing the relevant raw materials. The information given is as derived from the literature and does not include comments or evaluation by the authors of this report. Whenever the literature study did not yield information on radiological aspects of a material on the list, an appropriate notice is given. (Orig./DG) [de

Preparation and multi-properties determination of radium-containing rocklike material

Science.gov (United States)

Hong, Changshou; Li, Xiangyang; Zhao, Guoyan; Jiang, Fuliang; Li, Ming; Zhang, Shuai; Wang, Hong; Liu, Kaixuan

2018-02-01

The radium-containing rocklike material were fabricated using distilled water, ordinary Portland cement and additives mixed aggregates and admixtures according to certain proportion. The physico-mechanical properties as well as radioactive properties of the prepared rocklike material were measured. Moreover, the properties of typical granite sample were also investigated. It is found on one hand, similarities exist in physical and mechanical properties between the rocklike material and the granite sample, this confirms the validity of the proposed method; on the other hand, the rocklike material generally performs more remarkable radioactive properties compared with the granite sample, while radon diffusive properties in both materials are essentially matching. This study will provide a novel way to prepare reliable radium-containing samples for radon study of underground uranium mine.

Materials performance in off-gas systems containing iodine

International Nuclear Information System (INIS)

Beavers, J.A.; Berry, W.E.; Griess, J.C.

1981-11-01

During the reprocessing of spent reactor fuel elements, iodine is released to gas streams from which it is ultimately removed by conversion to nonvolatile iodic acid. Under some conditions iodine can produce severe corrosion in off-gas lines; in this study these conditions were established. Iron- and nickel-based alloys containing more than 6% molybdenum, such as Hastelloy G (7%), Inconel 625 (9%), and Hastelloy C-276 (16%), as well as titanium and zirconium, remained free of attack under all conditions tested. When the other materials, notably the austenitic stainless steels, were exposed to gas streams containing even only low concentrations of iodine and water vapors at 25 and 40 0 C, a highly corrosive, brownish-green liquid formed on their surfaces. In the complete absence of water vapor, the iodine-containing liquid did not form and all materials remained unaffected. The liquid that formed had a low pH (usually 2 inhibited attack

On material and energy sources of formation of fuel-containing materials during Chernobyl NPP UNIT 4 accident

Directory of Open Access Journals (Sweden)

O. V. Mikhailov

2016-12-01

Full Text Available Results of detailed analysis of material substance of lava-like fuel-containing materials sources (FCM and clusters with high uranium concentration were presented. Material and energy balance are aggregated in a process model for optimal composition of sacrificial materials and FCM. Quantitative estimate is given for spent nuclear fuel’ afterheat in a number of other heat energy sources in reactor vault. Conclusion was made that upon condition of 50 % heat loss, remained amount of “useful” heat would be sufficient for proceeding of blast furnace version of fuel-containing materials.

Glucose from a cellulose-containing raw material

Energy Technology Data Exchange (ETDEWEB)

Feniksova, R V

1977-01-01

Glucose was produced by mixing the raw material with an enzyme and water, enzymic hydrolysis of raw material, and separation of the hydrolyzate. To increase the yield of glucose, the hydrolysis took place in 2 stages: at the 1st stage, the hydrolysis was continued until the mixture of glucose and oligosaccharides was obtained under the influence of an active enzyme (1 to 20 units/g of cellulose present in the raw material); at the 2nd stage, cellulase immobilized on an insoluble carrier (by covalent bonds) was used for the hydrolysis of oligosaccharides. The hydromodulus of the suspension of cellulose-containing raw material was maintained 5 to 30. A carrier nonhydrolyzable by cellulase, for instance a homogenious macroporous Aerosil gel, a porous glass, or porous ceramics, was used.

Researches of real observation geometry in monitoring fuel-containing materials' subcriticality

International Nuclear Information System (INIS)

Vysotskij, E.D.; Shevchenko, V.G.; Shevchenko, M.V.

2002-01-01

The effectiveness of fuel-containing materials monitoring is discussed in the part related to the feasibilities of researches and realization of optimal geometry (detectors - source) of survey of neutron activity dynamics in nuclearly hazardous areas with clusters of fuel-containing materials concentrated in the premises 305/2

Container for storage of environmental incompatible materials

International Nuclear Information System (INIS)

Ruggenthaler, P.T.

1984-01-01

The container consists of a cuboid chamber, closed to five sides, just as the cover made of concrete. Iron mountings for use with lifting gears are coupled with the armouring of the container. The cover is made in such a way that mountings are hidden by the recesses at its borders. Therefore it is possible to stick these boxes. Concrete employed for is enriched with sealing materials of synthetics, the box is painted too. Sensors on the outside ensure telemetering of closeness of the boxes. (J.K.) [de

Ensuring the 50 year life of a fissile material container

International Nuclear Information System (INIS)

Glass, R.E.; Towne, T.L.

1997-12-01

Sandia was presented with an opportunity in 1993 to design containers for the long term storage and transport of fissile material. This program was undertaken at the direction of the US Department of Energy and in cooperation with Lawrence Livermore National Laboratory and Los Alamos National Laboratory which were tasked with developing the internal fixturing for the contents. The hardware is being supplied by Allied Signal Federal Manufacturing and Technologies, and the packaging will occur at Mason and Hangar Corporation's Pantex Plant. The unique challenge was to design a container that could be sealed with the fissile material contents; and, anytime during the next 50 years, the container could be transported with only the need for the pre-shipment leak test. This required not only a rigorous design capable of meeting the long term storage and transportation requirements, but also resulted in development of a surveillance program to ensure that the container continues to perform as designed over the 50-year life. This paper addresses the design of the container, the testing that was undertaken to demonstrate compliance with US radioactive materials transport regulations, and the surveillance program that has been initiated to ensure the 50-year performance

Determination of Fire Enviroment in Stacked Cargo Containers with Radioactive Materials Packages

Energy Technology Data Exchange (ETDEWEB)

Arviso, M.; Bobbe, J.G.; Dukart, R.D.; Koski, J.A.

1999-05-01

Results from a Fire Test with a three-by-three stack of standard 6 m long International Standards Organization shipping containers containing combustible fuels and empty radioactive materials packages are reported and discussed. The stack is intended to simulate fire conditions that could occur during on-deck stowage on container cargo ships. The fire is initated by locating the container stack adjacent to a 9.8 x 6 m pool fire. Temperatures of both cargoes (empty and simulated radioactive materials packages) and containers are recorded and reported. Observations on the duration, intensity and spread of the fire are discussed. Based on the results, models for simulation of fire exposure of radioactive materials packages in such fires are suggested.

Immobilization of INEL low-level radioactive wastes in ceramic containment materials

International Nuclear Information System (INIS)

Seymour, W.C.; Kelsey, P.V.

1978-11-01

INEL low-level radioactive wastes have an overall chemical composition that lends itself to self-containment in a ceramic-based material. Fewer chemical additives would be needed to process the wastes than to process high-level wastes or use a mixture containment method. The resulting forms of waste material could include a basalt-type glass or glass ceramic and a ceramic-type brick. Expected leach resistance is discussed in relationshp to data found in the literature for these materials and appears encouraging. An overview of possible processing steps for the ceramic materials is presented

A Review of Removable Surface Contamination on Radioactive Materials Transportation Containers

Energy Technology Data Exchange (ETDEWEB)

Kennedy, Jr, W. E.; Watson, E. C.; Murphy, D. W.; Harrer, B. J.; Harty, R.; Aldrich, J. M.

1981-05-01

This report contains the results of a study sponsored by the U.S. Nuclear Regulatory Commission (NRC) of removable surface contamination on radioactive materials transportation containers. The purpose of the study is to provide information to the NRC during their review of existing regulations. Data was obtained from both industry and literature on three major topics: 1) radiation doses, 2) economic costs, and 3) contamination frequencies. Containers for four categories of radioactive materials are considered including radiopharmaceuticals, industrial sources, nuclear fuel cycle materials, and low-level radioactive waste. Assumptions made in this study use current information to obtain realistic yet conservative estimates of radiation dose and economic costs. Collective and individual radiation doses are presented for each container category on a per container basis. Total doses, to workers and the public, are also presented for spent fuel cask and low-level waste drum decontamination. Estimates of the additional economic costs incurred by lowering current limits by factors of 10 and 100 are presented. Current contamination levels for each category of container are estimated from the data collected. The information contained in this report is designed to be useful to the NRC in preparing their recommendations for new regulations.

19 CFR 10.540 - Packing materials and containers for shipment.

Science.gov (United States)

2010-04-01

...-Singapore Free Trade Agreement Rules of Origin § 10.540 Packing materials and containers for shipment. (a... the United States. Accordingly, in applying either the build-down or build-up method for determining... shipping container which it purchased from Company B in Singapore. The shipping container is originating...

Container materials for isolation of radioactive waste in salt

International Nuclear Information System (INIS)

Streicher, M.A.; Andrews, A.

1987-10-01

The workshop reviewed the extensive data on the corrosion resistance of low-carbon steel in simulated salt repository environments, determined whether these data were sufficient to recommend low-carbon steel for fabrication of the container, and assessed the suitability of other materials under consideration in the SRP. The panelists determined the need for testing and research programs, recommended experimental approaches, and recommended materials based on existing technology. On the first day of the workshop, presentations were made on waste package requirements; the expected corrosion environment; degradation processes, including a review of data from corrosion tests on carbon steel; and rationales for container design and materials, modeling studies, and planned future work. The second day was devoted to a panel caucus, presentation of workshop findings, and open discussion. 76 refs., 2 figs., 3 tabs

Comparison of solute-binding properties of plastic materials used as pharmaceutical product containers.

Science.gov (United States)

Jenke, Dennis; Couch, Tom; Gillum, Amy

2010-01-01

Material/water equilibrium binding constants (E(b)) were determined for 11 organic solutes and 2 plastic materials commonly used in pharmaceutical product containers (plasticized polyvinyl chloride and polyolefin). In general, solute binding by the plasticized polyvinyl chloride material was greater, by nearly an order of magnitude, than the binding by the polyolefin (on an equal weight basis). The utilization of the binding constants to facilitate container compatibility assessments (e.g., drug loss by container binding) for drug-containing products is discussed.

19 CFR 10.461 - Retail packaging materials and containers.

Science.gov (United States)

2010-04-01

... Free Trade Agreement Rules of Origin § 10.461 Retail packaging materials and containers. Packaging... classification set out in General Note 26(n), HTSUS. If the good is subject to a regional value content... non-originating materials, as the case may be, in calculating the regional value content of the good...

Heat-processing method and facility for helium-containing metal material

International Nuclear Information System (INIS)

Kato, Takahiko; Kodama, Hideyo; Matsumoto, Toshimi; Aono, Yasuhisa; Nagata, Tetsuya; Hattori, Shigeo; Kaneda, Jun-ya; Ono, Shigeki.

1996-01-01

Electric current is supplied to an objective portion of a He-containing metal material to be applied with heat processing without causing melting, to decrease the He content of the portion. Subsequently, the defect portion of the tissues of the He-containing metal is modified by heating the portion with melting. Since electric current can be supplied to the metal material in a state where the metal material is heated and the temperature thereof is elevated, an effect of further reducing the He content can be obtained. Further, if the current supply and/or the heating relative to the metal material is performed in a vacuum or inert gas atmosphere, an effect of reducing the degradation of the surface of the objective portion to be supplied with electric current can be obtained. (T.M.)

Impact of container material on the development of Aedes aegypti larvae at different temperatures.

Science.gov (United States)

Kumar, Gaurav; Singh, R K; Pande, Veena; Dhiman, R C

2016-01-01

Aedes aegypti, the primary vector of dengue generally breeds in intradomestic and peridomestic containers made up of different materials, i.e. plastic, iron, rubber, earthen material etc. The material of container is likely to affect the temperature of water in container with variation in environmental temperature. The present study was aimed to determine the effect of different container materials on larval development of Ae. aegypti at different temperatures. Newly hatched I instar larvae (2-4 h old) were used in the study and experiments were conducted using three different containers made up of plastic, iron and earthen material. Three replicates for each type of container at 22, 26, 30, 34, 38, 40, and 42°C were placed in environmental chamber for the development of larvae. At temperatures >22°C, 50% pupation was completed in earthen pot within 4.3±0.6 to 6.3±0.6 days followed by plastic containers (5±0 to 8±0 days) and iron containers (6±0 to 9±0 days). Developmental time for 50% pupation in the three containers differed significantly (p containers (p containers resulted in significant variations in the developmental period of larvae. More than 35°C temperature of water was found inimical for pupal development. The results revealed the variation in temperature of water in different types of containers depending on the material of container, affecting duration of larval development. As the larval development was faster in earthen pot as compared to plastic and iron containers, community should be discouraged for storing the water in earthen pots. However, in view of containers of different materials used by the community in different temperature zones in the country, further studies are required for devising area-specific preventive measures for Aedes breeding.

Manufacturing method for radioactive material containing vessel

International Nuclear Information System (INIS)

Kamino, Yoshikazu; Nishioka, Eiji; Toyota, Michinori.

1997-01-01

A containing vessel for radioactive materials (for example, spent fuels) comprises an inner cylinder made of stainless steel having a space for containing radioactive materials at the inside and an outer cylinder made of stainless steel disposed at the outer side of the inner cylinder. Lead homogenization is applied to a space between the inner and the outer cylinders to deposit a lead layer. Then, molten lead heated to a predetermined temperature is cast into the space between the inner and the outer cylinders. A valve is opened to discharge the molten lead in the space from a molten lead discharge pipe, and heated molten lead is injected from a molten lead supply pipe. Then, the discharge of the molten lead and the injection of the molten lead are stopped, and the lead in the space is coagulated. With such procedures, gaps are not formed between the lead of the homogenized portion and the lead of cast portion even when the thickness of the inner and the outer cylinders is great. (I.N.)

Study of mechanoactivation of tungsten-molybdenum containing raw material in gas-jet mill

International Nuclear Information System (INIS)

Agnokov, T.Sh.; Gorobets, L.Zh.; Martynenko, V.P.; Fedorov, Yu.P.; Krakhmaleva, M.T.; Sokolova, L.A.

1988-01-01

Investigation is aimed at intensifying autoclave-soda leaching of tungsten-molybdenum-containing raw material. Connection of reactivity and physicochemical properties of crushed tungsten-molybdenum-containing products under different gas-jet crushing parameters is investigated. Optimal technological indices of hydrometallurgical reprocessing of tungsten-molybdenum-containing raw materials and products processed by gas-jet technique are given. The results obtained point out to perspectiveness of applying gas-jet technique of thermomechanical processing for intensifying and increasing the quality of tungsten- and molybdenum-containing raw materials and products of hydrometallurgical production

Use of the radiochemical method for identification of seized nuclear material - container

International Nuclear Information System (INIS)

Rosskopfova, O.; Matel, L.; Rajec, P.; Dobias, M.

2003-01-01

Analysis of the nuclear material of the seized illicit trafficking container was performed in the laboratory LARCHA. The container was professionally taken apart in the hot cell of HUMA-LAB APEKO. The compact core of container with mass of 36.00 kg was separated and analysed in the LARCHA laboratory. The results of alpha spectrometry proved that the core of container was made of depleted uranium. A suspicious container was seized, that was believed to be used for transportation of radioactive and nuclear materials. Container, according to Slovak Law, was transported to a Civil defence and police started with expertise analysis. On the basis of previous results, it was assumed that the core of the container was made of a nuclear material. The container was cut inside a hot cell to separate parts at HUMA-Lab Kosice and individual parts were prepared for analyses. From shape and type of the detained container it was assumed that the core of the container was made of depleted uranium and suspected cut pieces were sent to the accredited laboratory LARCHA Bratislava for isotope analysis. Samples from different parts of the container were analysed by using HPGe gamma. The results of gamma spectrometry proved the presence of Co-60, Cs-137 and U-235 with values 11.8 Bq, 5.6 Bq a 3.8 Bq respectively. On the basis of measured results it was calculated mass content (%) of uranium isotopes in samples. It was concluded that: - container is composed of different parts, which are made from steel, lead and uranium; - the wipe tests proved the presence of radionuclides of Cs-137 and Co-60; - the core of the container is made of depleted uranium; - chosen methods, which were used proved that are suitable for analysis and identification of nuclear materials. (authors)

New CSA guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances

International Nuclear Information System (INIS)

Rhodes, M.; Kwong, A.

2011-01-01

The Canadian Standards Association (CSA) guideline N292.5, Guideline for the exemption or clearance from regulatory control of materials that contain, or potentially contain, nuclear substances, was recently developed to address a need for guidance on approaches for clearance of materials from facilities licensed by the Canadian Nuclear Safety Commission (CNSC) consistent with Canadian and international recommendations. This guideline is also applicable to determining if an activity associated with materials that contain nuclear substances is exempt from requiring a CNSC licence. The guideline summarizes the regulatory requirements associated with the exemption and clearance of materials and provides a graded approach to designing a survey based on the risk of residual contamination being present. (author)

Preliminary selection criteria for the Yucca Mountain Project waste package container material

International Nuclear Information System (INIS)

Halsey, W.G.

1991-01-01

The Department of Energy's Yucca Mountain Project (YMP) is evaluating a site at Yucca Mountain in Nevada for construction of a geologic repository for the storage of high-level nuclear waste. Lawrence Livermore National Laboratory's (LLNL) Nuclear Waste Management Project (NWMP) has the responsibility for design, testing, and performance analysis of the waste packages. The design is performed in an iterative manner in three sequential phases (conceptual design, advanced conceptual design, and license application design). An important input to the start of the advanced conceptual design is the selection of the material for the waste containers. The container material is referred to as the 'metal barrier' portion of the waste package, and is the responsibility of the Metal Barrier Selection and Testing task at LLNL. The selection will consist of several steps. First, preliminary, material-independent selection criteria will be established based on the performance goals for the container. Second, a variety of engineering materials will be evaluated against these criteria in a screening process to identify candidate materials. Third, information will be obtained on the performance of the candidate materials, and final selection criteria and quantitative weighting factors will be established based on the waste package design requirements. Finally, the candidate materials will be ranked against these criteria to determine whether they meet the mandated performance requirements, and to provide a comparative score to choose the material for advanced conceptual design activities. This document sets forth the preliminary container material selection criteria to be used in screening candidate materials. 5 refs

Hearth furnace for distilling powdered materials containing hydrocarbons

Energy Technology Data Exchange (ETDEWEB)

1937-06-21

The present invention has for its object a hearth furnace particularly intended for the distillation of powdered material containing hydrocarbons. It consists of a fixed circular hearth above which are placed the moving scrapers intended to displace the material from the center toward the periphery. The material is poured by a central vertical pipe at the base of which is arranged a rotating ring for distributing the material on the hearth; this ring, which is fastened on the vertical axis of the drive, carries the radial arms to which are attached the scrapers arranged obliquely by the support on the arms and intended to displace the material on the hearth. The hearth is heated from below by means of forced circulation of gases produced in a fire-box and maintained at a convenient temperature by mixing with a part of the cold gases.

Accumulation of organic compounds leached from plastic materials used in biopharmaceutical process containers.

Science.gov (United States)

Jenke, Dennis R; Zietlow, David; Garber, Mary Jo; Sadain, Salma; Reiber, Duane; Terbush, William

2007-01-01

Plastic materials are widely used in medical items, such as solution containers, transfusion sets, transfer tubing, and devices. An emerging trend in the biotechnology industry is the utilization of plastic containers to prepare, transport, and store an assortment of solutions including buffers, media, and in-process and finished product. The direct contact of such containers with the product at one or more points in its lifetime raises the possibility that container leachables may accumulate in the finished product. The interaction between several commercially available container materials and numerous model test solutions (representative of buffers and media used in biopharmaceutical applications) was investigated. This paper summarizes the identification of leachables associated with the container materials and documents the levels to which targeted leachables accumulate in the test solutions under defined storage conditions.

An annotated history of container candidate material selection

International Nuclear Information System (INIS)

McCright, R.D.

1988-07-01

This paper documents events in the Nevada Nuclear Waste Storage Investigations (NNWSI) Project that have influenced the selection of metals and alloys proposed for fabrication of waste package containers for permanent disposal of high-level nuclear waste in a repository at Yucca Mountain, Nevada. The time period from 1981 to 1988 is covered in this annotated history. The history traces the candidate materials that have been considered at different stages of site characterization planning activities. At present, six candidate materials are considered and described in the 1988 Consultation Draft of the NNWSI Site Characterization Plan (SCP). The six materials are grouped into two alloy families, copper-base materials and iron to nickel-base materials with an austenitic structure. The three austenitic candidates resulted from a 1983 survey of a longer list of candidate materials; the other three candidates resulted from a special request from DOE in 1984 to evaluate copper and copper-base alloys. 24 refs., 2 tabs

Modification of clay-based waste containment materials

International Nuclear Information System (INIS)

Adu-Wusu, K.; Whang, J.M.; McDevitt, M.F.

1997-01-01

Bentonite clays are used extensively for waste containment barriers to help impede the flow of water in the subsurface because of their low permeability characteristics. However, they do little to prevent diffusion of contaminants, which is the major transport mechanism at low water flows. A more effective way of minimizing contaminant migration in the subsurface is to modify the bentonite clay with highly sorptive materials. Batch sorption studies were conducted to evaluate the sorptive capabilities of organo-clays and humic- and iron-based materials. These materials proved to be effective sorbents for the organic contaminants 1,2,4-trichlorobenzene, nitrobenzene, and aniline in water, humic acid, and methanol solution media. The sorption capacities were several orders of magnitude greater than that of unmodified bentonite clay. Modeling results indicate that with small amounts of these materials used as additives in clay barriers, contaminant flux through walls could be kept very small for 100 years or more. The cost of such levels of additives can be small compared to overall construction costs

Endurance test report of rubber sealing materials for the containment vessel

International Nuclear Information System (INIS)

Yamamoto, R.; Watanabe, K.; Hanashima, K.

2015-01-01

In the event of a nuclear power plant accident such as a core meltdown and a cooling system failure, the containment contains radioactive materials released from the reactor pressure vessel to reduce the activity of the radioactive materials and the effects of radiation in the vicinity of the plant. Since high sealing performance and high pressure resistance are required of the containment, a silicone or EPDM rubber gasket with high heat and radiation resistance is used for the sealing of the sealing boundary of the containment. In recent years, it has been shown that a large amount of steam is released into the containment in the case of a severe accident. Consequently, radiation resistance at high temperature as well as steam resistance is required of the rubber gasket placed at the sealing boundary. However, the steam resistance of silicone rubber is not necessarily as good as that of EPDM rubber. Therefore, it is necessary to evaluate the sealing characteristics of rubber gaskets in such a degrading environment in a severe accident. O. Kato et al. [1] conducted a study on the degradation status of rubber gaskets and their application limits at high temperature. However, few studies have evaluated rubber gaskets in high-temperature radiation and steam environments. In this study, we degraded silicone rubber and EPDM rubber used for the containment in the high-temperature radiation and steam environments expected to occur in a severe accident and evaluated the useful life of the rubber as a sealing material by estimating the change in its performance as a sealing material from the change in permanent compressive strain in the rubber. (author)

Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

Science.gov (United States)

Batyaev, V. F.; Skliarov, S. V.

2018-01-01

The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW). The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration), meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g) confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

Detection of tiny amounts of fissile materials in large-sized containers with radioactive waste

Directory of Open Access Journals (Sweden)

Batyaev V.F.

2018-01-01

Full Text Available The paper is devoted to non-destructive control of tiny amounts of fissile materials in large-sized containers filled with radioactive waste (RAW. The aim of this work is to model an active neutron interrogation facility for detection of fissile ma-terials inside NZK type containers with RAW and determine the minimal detectable mass of U-235 as a function of various param-eters: matrix type, nonuniformity of container filling, neutron gen-erator parameters (flux, pulse frequency, pulse duration, meas-urement time. As a result the dependence of minimal detectable mass on fissile materials location inside container is shown. Nonu-niformity of the thermal neutron flux inside a container is the main reason of the space-heterogeneity of minimal detectable mass in-side a large-sized container. Our experiments with tiny amounts of uranium-235 (<1 g confirm the detection of fissile materials in NZK containers by using active neutron interrogation technique.

19 CFR 10.602 - Packing materials and containers for shipment.

Science.gov (United States)

2010-04-01

...-Central America-United States Free Trade Agreement Rules of Origin § 10.602 Packing materials and... regional value content calculation. Packing materials and containers for shipment, as defined in § 10.593(m) of this subpart, are to be disregarded in determining the regional value content of a good imported...

Corrosion of candidate container materials by Yucca Mountain bacteria

International Nuclear Information System (INIS)

Horn, J; Jones, D; Lian, T; Martin, S; Rivera, A

1999-01-01

Several candidate container materials have been studied in modified Yucca Mountain (YM) ground water in the presence or absence of YM bacteria. YM bacteria increased corrosion rates by 5-6 fold in UNS G10200 carbon steel, and nearly 100-fold in UNS NO4400 Ni-Cu alloy. YM bacteria caused microbiologically influenced corrosion (MIC) through de-alloying or Ni-depletion of Ni-Cu alloy as evidenced by scanning electronic microscopy (SEM) and inductively coupled plasma spectroscopy (ICP) analysis. MIC rates of more corrosion-resistant alloys such as UNS NO6022 Ni-Cr- MO-W alloy, UN's NO6625 Ni-Cr-Mo alloy, and UNS S30400 stainless steel were measured below 0.05 umyr, however YM bacteria affected depletion of Cr and Fe relative to Ni in these materials. The chemical change on the metal surface caused by depletion was characterized in anodic polarization behavior. The anodic polarization behavior of depleted Ni-based alloys was similar to that of pure Ni. Key words: MIC, container materials, YM bacteria, de-alloying, Ni-depletion, Cr-depletion, polarization resistance, anodic polarization,

Packaging material and flexible medical tubing containing thermally exfoliated graphite oxide

Science.gov (United States)

Prud'homme, Robert K. (Inventor); Aksay, Ilhan A. (Inventor)

2011-01-01

A packaging material or flexible medical tubing containing a modified graphite oxide material, which is a thermally exfoliated graphite oxide with a surface area of from about 300 m.sup.2/g to 2600 m.sup.2/g.

Methods of pretreating comminuted cellulosic material with carbonate-containing solutions

Energy Technology Data Exchange (ETDEWEB)

Francis, Raymond

2012-11-06

Methods of pretreating comminuted cellulosic material with an acidic solution and then a carbonate-containing solution to produce a pretreated cellulosic material are provided. The pretreated material may then be further treated in a pulping process, for example, a soda-anthraquinone pulping process, to produce a cellulose pulp. The pretreatment solutions may be extracted from the pretreated cellulose material and selectively re-used, for example, with acid or alkali addition, for the pretreatment solutions. The resulting cellulose pulp is characterized by having reduced lignin content and increased yield compared to prior art treatment processes.

Development of Latent Heat Storage Phase Change Material Containing Plaster

Directory of Open Access Journals (Sweden)

Diana BAJARE

2016-05-01

Full Text Available This paper reviews the development of latent heat storage Phase Change Material (PCM containing plaster as in passive application. Due to the phase change, these materials can store higher amounts of thermal energy than traditional building materials and can be used to add thermal inertia to lightweight constructions. It was shown that the use of PCMs have advantages stabilizing the room temperature variations during summer days, provided sufficient night ventilation is allowed. Another advantage of PCM usage is stabilized indoor temperature on the heating season. The goal of this study is to develop cement and lime based plaster containing microencapsulated PCM. The plaster is expected to be used for passive indoor applications and enhance the thermal properties of building envelope. The plaster was investigated under Scanning Electron Microscope and the mechanical, physical and thermal properties of created plaster samples were determined.

Leaching assessment of road materials containing primary lead and zinc slags.

Science.gov (United States)

Barna, R; Moszkowicz, P; Gervais, C

2004-01-01

Characterisation of the leaching behaviour of waste-containing materials is a crucial step in the environmental assessment for reuse scenarios. In our research we applied the multi-step European methodology ENV 12-920 to the leaching assessment of road materials containing metallurgical slag. A Zn slag from an imperial smelting furnace (ISF) and a Pb slag from a lead blast furnace (LBF) are investigated. The two slags contain up to 11.2 wt% of lead and 3.5 wt% of zinc and were introduced as a partial substitute for sand in two road materials, namely sand-cement and sand-bitumen. At the laboratory scale, a leaching assessment was performed first through batch equilibrium leaching tests. Second, the release rate of the contaminants was evaluated using saturated leaching tests on monolithic material. Third, laboratory tests were conducted on monolithic samples under intermittent wetting conditions. Pilot-scale tests were conducted for field testing of intermittent wetting conditions. The results show that the release of Pb and Zn from the materials in a saturated scenario was controlled by the pH of the leachates. For the intermittent wetting conditions, an additional factor, blocking of the pores by precipitation during the drying phase is proposed. Pilot-scale leaching behaviour only partially matched with the laboratory-scale test results: new mass transfer mechanisms and adapted laboratory leaching tests are discussed.

Direct vitrification of plutonium-containing materials (PCM`s) with the glass material oxidation and dissolution system (GMODS)

Energy Technology Data Exchange (ETDEWEB)

Forsberg, C.W. Beahm, E.C.; Parker, G.W.; Rudolph, J.C.; Haas, P.A.; Malling, G.F.; Elam, K.; Ott, L.

1995-10-30

The end of the cold war has resulted in excess PCMs from nuclear weapons and associated production facilities. Consequently, the US government has undertaken studies to determine how best to manage and dispose of this excess material. The issues include (a) ensurance of domestic health, environment, and safety in handling, storage, and disposition, (b) international arms control agreements with Russia and other countries, and (c) economics. One major set of options is to convert the PCMs into glass for storage or disposal. The chemically inert characteristics of glasses make them a desirable chemical form for storage or disposal of radioactive materials. A glass may contain only plutonium, or it may contain plutonium along with other radioactive materials and nonradioactive materials. GMODS is a new process for the direct conversion of PCMs (i.e., plutonium metal, scrap, and residues) to glass. The plutonium content of these materials varies from a fraction of a percent to pure plutonium. GMODS has the capability to also convert other metals, ceramics, and amorphous solids to glass, destroy organics, and convert chloride-containing materials into a low-chloride glass and a secondary clean chloride salt strewn. This report is the initial study of GMODS for vitrification of PCMs as input to ongoing studies of plutonium management options. Several tasks were completed: initial analysis of process thermodynamics, initial flowsheet analysis, identification of equipment options, proof-of-principle experiments, and identification of uncertainties.

Photometric determination of yttrium in zirconium-containing materials

International Nuclear Information System (INIS)

Barbina, T.M.; Polezhaev, Yu.M.

1984-01-01

Comparative evaluation of the effect of different ways of eliminating the zirconium interfering effect on the results of yttrium photometric determination with arsenazo 2 in artificial mixtures of Y 2 O 3 and ZrO 2 , containing 5 and 10 mol.% Y 2 O 3 , has been carried out. The effect of Zr is eliminated by means of its precipitation by ammonium solution in the form of hydroxide and using camouflaging with 25% sulfosalicylic acid. Both ways do not provide a correct enough result. The use of non-reagent thermohydrolytic Zr precipitation during the analysis of zirconium-containing materials permits to obtain correct and well-reproducible results

16 CFR 1145.2 - Paint (and other similar surface-coating materials) containing lead; toys, children's articles...

Science.gov (United States)

2010-01-01

... materials) containing lead; toys, children's articles, and articles of furniture bearing such paint (or... materials) containing lead; toys, children's articles, and articles of furniture bearing such paint (or...) Paint and other similar surface-coating materials containing lead and toys, children's articles, and...

Automated nuclear material recovery and decontamination of large steel dynamic experiment containers

International Nuclear Information System (INIS)

Dennison, D.K.; Gallant, D.A.; Nelson, D.C.; Stovall, L.A.; Wedman, D.E.

1999-01-01

A key mission of the Los Alamos National Laboratory (LANL) is to reduce the global nuclear danger through stockpile stewardship efforts that ensure the safety and reliability of nuclear weapons. In support of this mission LANL performs dynamic experiments on special nuclear materials (SNM) within large steel containers. Once these experiments are complete, these containers must be processed to recover residual SNM and to decontaminate the containers to below low level waste (LLW) disposal limits which are much less restrictive for disposal purposes than transuranic (TRU) waste limits. The purpose of this paper is to describe automation efforts being developed by LANL for improving the efficiency, increasing worker safety, and reducing worker exposure during the material cleanout and recovery activities performed on these containers

Container materials in environments of corroded spent nuclear fuel

Science.gov (United States)

Huang, F. H.

1996-07-01

Efforts to remove corroded uranium metal fuel from the K Basins wet storage to long-term dry storage are underway. The multi-canister overpack (MCO) is used to load spent nuclear fuel for vacuum drying, staging, and hot conditioning; it will be used for interim dry storage until final disposition options are developed. Drying and conditioning of the corroded fuel will minimize the possibility of gas pressurization and runaway oxidation. During all phases of operations the MCO is subjected to radiation, temperature and pressure excursions, hydrogen, potential pyrophoric hazard, and corrosive environments. Material selection for the MCO applications is clearly vital for safe and efficient long-term interim storage. Austenitic stainless steels (SS) such as 304L SS or 316L SS appear to be suitable for the MCO. Of the two, Type 304L SS is recommended because it possesses good resistance to chemical corrosion, hydrogen embrittlement, and radiation-induced corrosive species. In addition, the material has adequate strength and ductility to withstand pressure and impact loading so that the containment boundary of the container is maintained under accident conditions without releasing radioactive materials.

To the problem of structural materials serviceability in nitrogen-hydrogen-containing environments

International Nuclear Information System (INIS)

Bichuya, A.L.

1982-01-01

The analysis of the factors which affect high-temperature serviceability of structural materials in nitrogen-hydrogen-containing environments, in particular in ammonia, has been carried out on the basis of the published and own experimental data. It is shown that the observed reduction of serviceability of structural materials, under the effect of high temperatures and nitrogen-hydrogen-containing environments, can occur as a result of corrosion failure connected with nitriding, and also hydrogen embrittlement appearing as a result of the penetration of hydrogen formed during adsorbed gaseous phase dissociation on the metal being deformed. The suggested scheme of high-temperature metal fracture under the effect of nitrogen-hydrogen-containing environments, that in contrast to the previous ones includes the factor of hydrogen ebrittlement, allows to give a real estimation of structional materials serviceability under product service conditions

POLYMER COMPOSITES MODIFIED BY WASTE MATERIALS CONTAINING WOOD FIBRES

Directory of Open Access Journals (Sweden)

Bernardeta Dębska

2016-11-01

Full Text Available In recent years, the idea of sustainable development has become one of the most important require-ments of civilization. Development of sustainable construction involves the need for the introduction of innovative technologies and solutions that will combine beneficial economic effects with taking care of the health and comfort of users, reducing the negative impact of the materials on the environment. Composites obtained from the use of waste materials are part of these assumptions. These include modified epoxy mortar containing waste wood fibres, described in this article. The modification consists in the substitution of sand by crushed waste boards, previously used as underlays for panels, in quantities of 0%, 10%, 20%, 35% and 50% by weight, respectively. Composites containing up to 20% of the modifier which were characterized by low water absorption, and good mechanical properties, also retained them after the process of cyclic freezing and thawing.

Corrosion behaviour of container materials for geological disposal of high level radioactive waste

International Nuclear Information System (INIS)

Accary, A.

1985-01-01

The disposal of high level radioactive waste in geological formations, based on the multibarrier concept, may include the use of a container as one of the engineered barriers. In this report the requirements imposed on this container and the possible degradation processes are reviewed. Further on an overview is given of the research being carried out by various research centres in the European Community on the assessment of the corrosion behaviour of candidate container materials. The results obtained on a number of materials under various testing conditions are summarized and evaluated. As a result, three promising materials have been selected for a detailed joint testing programme. It concerns two highly corrosion resistant alloys, resp. Ti-Pd (0.2 Pd%) and Hastelloy C4 and one consumable material namely a low carbon steel. Finally the possibilities of modelling the corrosion phenomena are discussed

Radiation exposure by man-modified materials containing natural radionuclides

Energy Technology Data Exchange (ETDEWEB)

Becker, D.E. [Technical Inspection Agency of Bavaria, Munich (Germany); Eder, E. [Government of Bavaria, Ministry for State Development and Environmental Affairs Development, Munich (Germany); Reichelt, A. [Technical Inspection Agency of Bavaria, Munich (Germany)

1992-07-01

More than one hundred materials, containing natural radioactive nuclides, are being investigated due to radiation exposure to people. This paper deals with thoriated gas mantles and shows that the radiation exposure by inhalation of radionuclides released while burning and exchange is not negligible. (author)

Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

International Nuclear Information System (INIS)

Ko, A Reum; Lee, Jeong Ik; Yoon, Ho Joon

2016-01-01

Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases

Preliminary Study of Applying Phase Change Materials (PCM) for Containment Passive Cooling

Energy Technology Data Exchange (ETDEWEB)

Ko, A Reum; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [KUSTAR, Abu Dhabi (United Arab Emirates)

2016-05-15

Most of Pressurized Water Reactor (PWR) containments use fan cooler systems and containment spray systems. However, the importance of passive safety system has increased after the Fukushima accident. As the main passive safety system, Passive Containment Cooling System (PCCS), which utilizes natural phenomena to remove the heat released from the reactor, is suggested in the advanced pressurized water reactor (APWR). To increase the efficiency of passive cooling, additional passive containment cooling method using Phase Change Material (PCM) is suggested in this paper. For containment using PCMs, there are many advantages. Phase Change Material (PCM) is proposed as an additional passive containment cooling method to increase the efficiency of passive cooling in this paper. To apply proper PCMs to containment, commercially available PCMs were screened while reviewing thermophysical properties data and suggested selection criteria. A sensitivity study was also carried out to identify the effect of potential installation location of PCM using the CAP code. The pressure of containment in most cases showed slightly higher than that of the initial case. For the temperature of steam and water and humidity, similar results with the initial case were showed in most cases.

Radioactive material transporting container

International Nuclear Information System (INIS)

Watabe, Yukio.

1990-01-01

As a supporting member of a sealing container for containing spent fuels, etc., a straight pipe or a cylinder has been used. However, upon dropping test, the supporting member is buckled toward the central axis of a transporting container and a shock absorber is crushed in the axial direction to prevent its pushing force to the outer side, which may possibly hinder normal shock moderating function. Then, at least more than one-half of the supporting member is protruded radially to the outer side of the sealing container beyond the fixed portion with the sealed container, so that the member has a portion extended in the radial outside of the transporting container with an angle greater than the angle formed between a line connecting the outer circumference at the bottom of an outer cylinder with the gravitational center of the transporting container and the central axis of the transporting container. As a result, buckling of the supporting member toward the central axis of the transporting container upon dropping test can be prevented and the deformation of the shock absorber is neither not prevented to exhibit normal shock absorbing effect. This can improve the reliability and reduce the amount of shock absorbers. (N.H.)

Containers for short-term storage of nuclear materials at the Los Alamos plutonium facility

International Nuclear Information System (INIS)

Hagan, R.; Gladson, J.

1997-01-01

The Los Alamos Plutonium Facility for the past 18 yr has stored nuclear samples for archiving and in support of nuclear materials research and processing programs. In the past several years, a small number of storage containers have been found in a deteriorated condition. A failed plutonium container can cause personnel contamination exposure and expensive physical area decontamination. Containers are stored in a physically secure radiation area vault, making close inspection costly in the form of personnel radiation exposure and work time. A moderate number of these containers are used in support of plutonium processing and must withstand daily handling abuse. A 2-yr evaluation of failed containers and those that have shown no deterioration has been conducted. Based on that study, a program was established to formalize our packing methods and materials and standardize the size and shape of containers that are used for short-term use. A standardized set of containers was designed, evaluated, tested, and procured for use in the facility. This paper reviews our vault storage problems, shows some failed containers, and presents our planned solutions to provide safe and secure containment of nuclear materials

MULCHES AND OTHER COVER MATERIALS TO REDUCE WEED GROWTH IN CONTAINER-GROWN NURSERY STOCK.

Science.gov (United States)

Rys, F; Van Wesemael, D; Van Haecke, D; Mechant, E; Gobin, B

2014-01-01

Due to the recent EU-wide implementation of Integrated Pest Management (IPM), alternative methods to reduce weed growth in container-grown nursery stock are needed to cut back the use of herbicides. Covering the upper layer of the substrate is known as a potential method to prevent or reduce weed growth in plant containers. As a high variety of mulches and other cover materials are on the market, however, it is no longer clear for growers which cover material is most efficient for use in containers. Therefore, we examined the effect on weed growth of different mulches and other cover materials, including Pinus maritima, P. sylvestris, Bio-Top Basic, Bio-Top Excellent, coco chips fine, hemp fibres, straw pellets, coco disk 180LD and jute disk. Cover materials were applied immediately after repotting of Ligustrum ovalifolium or planting of Fagus sylvatica. At regular times, both weed growth and side effects (e.g., plant growth, water status of the substrate, occurrence of mushrooms, foraging of birds, complete cover of the substrate and fixation) were assessed. All examined mulches or other cover materials were able to reduce weed growth on the containers during the whole growing season. Weed suppression was even better than that of a chemical treated control. Although all materials showed some side effects, the impact on plant growth is most important to the grower and depends not only on material characteristics (e.g., biodegradation, nutrient leaching and N-immobilisation) but also on container size and climatic conditions. In conclusion, mulches and other cover materials can be a valuable tool within IPM to lower herbicide use. To enable a deliberate choice of which cover material is best used in a specific situation more research is needed on lifespan and stability as well as on economic characteristics of the materials.

Radioactive material-containing vessel and method of manufacturing the same

International Nuclear Information System (INIS)

Kanazawa, Hiroshi; Wada, Katsuyoshi; Ota, Shigeo; Nishioka, Eiji; Okuno, Michinori.

1995-01-01

In a vessel for containing radioactive materials having an outer wall with a structure of interposing a lead layer, as a shielding material between inner and outer cylinders made of steel plates, the inner cylinder and the lead layer are in close contact by way of a thin layer of a lead/tin type soldering material and to such an extent that the boundary layer is not detected by supersonic inspection. In addition, flux is coated to the steel plate, which forms the inner cylinder, on the surface being in contact with the lead layer, then a thin layer of the soldering material such as lead or tin is formed, to cast the lead between the inner and the outer cylinders. Then, since the inner cylinder and the lead layer are thermally joined tightly, heat generated at the inside can effectively be released to the outside, so that it is effective as a high-performance cask for transporting a large amount of radioactive materials such as spent nuclear fuels having high temperature afterheat. In addition, a containing vessel with good contact between the inner cylinder and the lead can be manufactured at a low cost only applying a simple primer treatment on the surface of the inner cylinder in addition to an existent lead casting method. (N.H.)

Experiments on container materials for Swiss high-level waste disposal projects. Part 2

International Nuclear Information System (INIS)

Simpson, J.P.

1984-12-01

The present concept for final disposal of high-level waste in Switzerland consists of a repository at a depth of 1000 to 1500 m in the crystalline bedrock of northern Switzerland. The waste will be placed in a container which is required to function as a high integrity barrier for at least 1000 years. This report is the second of a set of two dealing with the evaluation of potential materials for such containers. Four materials were identified for further evaluation in the first of these reports; they were cast steel, nodular cast iron, copper and Ti-Code 12. It was concluded that some testing was needed, in particular with respect to corrosion, in order to confirm these materials as candidate container materials. The experimental programme included: 1) corrosion tests on copper under gamma radiation; 2) immersion corrosion tests on the four candidate materials including welded specimens; 3) corrosion testing of the four materials in saturated bentonite; 4) constant strain rate testing of Ti-Code 12 and copper at 80 degrees C; 5) the behaviour of copper, Ti-Code 12 and Zircaloy-2 when immersed in liquid lead; 6) corrosion potential and galvanic current measurements on several material pairs. The standard test medium was natural mineral water from the Bad Saeckingen source. This water has a total dissolved solids content of approx. 3200 mg/l, about 1600 mg/l as chloride. The oxygen level was defined as 0.1 μg/g. In certain cases this medium was modified in order to test under more severe conditions. The results of the corrosion tests confirm in general the evaluation in the first part of the report. All of the materials are suitable for high-level waste containers: cast steel, nodular cast iron and copper as single layer containers, and Ti-Code 12 as an outer corrosion resistant layer. Copper could also be used under an outer steel layer, where it could arrest local penetration

Preliminary sensitivity analyses of corrosion models for BWIP [Basalt Waste Isolation Project] container materials

International Nuclear Information System (INIS)

Anantatmula, R.P.

1984-01-01

A preliminary sensitivity analysis was performed for the corrosion models developed for Basalt Waste Isolation Project container materials. The models describe corrosion behavior of the candidate container materials (low carbon steel and Fe9Cr1Mo), in various environments that are expected in the vicinity of the waste package, by separate equations. The present sensitivity analysis yields an uncertainty in total uniform corrosion on the basis of assumed uncertainties in the parameters comprising the corrosion equations. Based on the sample scenario and the preliminary corrosion models, the uncertainty in total uniform corrosion of low carbon steel and Fe9Cr1Mo for the 1000 yr containment period are 20% and 15%, respectively. For containment periods ≥ 1000 yr, the uncertainty in corrosion during the post-closure aqueous periods controls the uncertainty in total uniform corrosion for both low carbon steel and Fe9Cr1Mo. The key parameters controlling the corrosion behavior of candidate container materials are temperature, radiation, groundwater species, etc. Tests are planned in the Basalt Waste Isolation Project containment materials test program to determine in detail the sensitivity of corrosion to these parameters. We also plan to expand the sensitivity analysis to include sensitivity coefficients and other parameters in future studies. 6 refs., 3 figs., 9 tabs

19 CFR 10.539 - Retail packaging materials and containers.

Science.gov (United States)

2010-04-01

...-Singapore Free Trade Agreement Rules of Origin § 10.539 Retail packaging materials and containers. Packaging... requirement. The United States importer of good C decides to use the build-down method, RVC=((AV−VNM)/AV... content requirement. In applying this method, the non-originating blister packages are taken into account...

Stowing of packages containing radioactive materials on conveyances

International Nuclear Information System (INIS)

Draulans, J.; Lafontaine, I.; Chevalier, G.; Gilles, P.; Jolys, J.C.; Pouard, M.

1986-04-01

The Commission of the European Communities has financed some research work carried out jointly by the ''Commissariat a l'Energie Atomique'' and the belgian company ''TRANSNUBEL'', in the field of stowing containers for radioactive materials on trucks. 2 reference type accidents are selected: . a front-end collision against a rigid barrier at an impact speed of 50 km/h . a side-on collision of an impacting vehicle at a speed of 25-35 km/h against a truck loaded with a container. A mathematical model has been developed by means of the CEA Trico code to compute a frontal impact in which the container (1.3 t weight), is stowed by means of 4 tie-down members, each for a nominal load of 2 t. Results indicate the stowing being insufficient and the attachment points too weak to keep the container on the platform. Real tests have been performed to verify these results. Tie-down members and chocks have been defined on the basis of static- and dynamic tests for being used in 8 crash tests. Different containers (low- and high center of gravity) and different ways of stowing have been tried out. An attempt is made to work a code of good practice for stowing, by means of tie-down members and chocks, packages on a truck platform. 19 refs

Potential containment materials for liquid-lead and lead-bismuth eutectic spallation neutron source

International Nuclear Information System (INIS)

Park, J.J.; Butt, D.P.; Beard, C.A.

1997-11-01

Lead (Pb) and lead-bismuth eutectic (44Pb-56Bi) have been the two primary candidate liquid-metal target materials for the production of spallation neutrons. Selection of a container material for the liquid-metal target will greatly affect the lifetime and safety of the target subsystem. For the lead target, niobium-1 (wt%) zirconium (Nb-1Zr) is a candidate containment material for liquid lead, but its poor oxidation resistance has been a major concern. The oxidation rate of Nb-1Zr was studied based on the calculations of thickness loss due to oxidation. According to these calculations, it appeared that uncoated Nb-1Zr may be used for a one-year operation at 900 C at P O 2 = 1 x 10 -6 torr, but the same material may not be used in argon with 5-ppm oxygen. Coating technologies to reduce the oxidation of Nb-1Zr are reviewed, as are other candidate refractory metals such as molybdenum, tantalum, and tungsten. For the Pb-Bi target, three candidate containment materials are suggested based on a literature survey of the materials compatibility and proton irradiation tests: Croloy 2-1/4, modified 9Cr-1Mo, and 12Cr-1Mo (HT-9) steel. These materials seem to be used only if the lead-bismuth is thoroughly deoxidized and treated with zirconium and magnesium

Ultrasonic Fingerprinting of Structural Materials: Spent Nuclear Fuel Containers Case-Study

Science.gov (United States)

Sednev, D.; Lider, A.; Demyanuk, D.; Kroening, M.; Salchak, Y.

Nowadays, NDT is mainly focused on safety purposes, but it seems possible to apply those methods to provide national and IAEA safeguards. The containment of spent fuel in storage casks could be dramatically improved in case of development of so-called "smart" spent fuel storage and transfer casks. Such casks would have tamper indicating and monitoring/tracking features integrated directly into the cask design. The microstructure of the containers material as well as of the dedicated weld seam is applied to the lid and the cask body and provides a unique fingerprint of the full container, which can be reproducibly scanned by using an appropriate technique. The echo-sounder technique, which is the most commonly used method for material inspection, was chosen for this project. The main measuring parameter is acoustic noise, reflected from material's artefacts. The purpose is to obtain structural fingerprinting. Reference measurement and additional measurement results were compared. Obtained results have verified the appliance of structural fingerprint and the chosen control method. The successful authentication demonstrates the levels of the feature points' compliance exceeding the given threshold which differs considerably from the percentage of the concurrent points during authentication from other points. Since reproduction or doubling of the proposed unique identification characteristics is impossible at the current state science and technology, application of this technique is considered to identify the interference into the nuclear materials displacement with high accuracy.

49 CFR 176.170 - Transport of Class 1 (explosive) materials in freight containers.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Transport of Class 1 (explosive) materials in... REGULATIONS CARRIAGE BY VESSEL Detailed Requirements for Class 1 (Explosive) Materials Cargo Transport Units and Shipborne Barges § 176.170 Transport of Class 1 (explosive) materials in freight containers. (a...

Waste handling and REACH : Recycling of materials containing SVHCs: daily practice challenges

NARCIS (Netherlands)

Janssen MPM; van Broekhuizen FA; MSP; M&V

2017-01-01

To achieve a circular economy it is essential to recycle substances, materials and products created by that economy. Recycling, however, becomes more difficult when said materials and products contain substances that are so hazardous that their use is restricted. This is the case with any substance

Development of a container for the transportation and storage of plutonium bearing materials

International Nuclear Information System (INIS)

Ammerman, D.; Geinitz, R.; Thorp, D.; Rivera, M.

1998-03-01

There is a large backlog of plutonium contaminated materials at the Rocky Flats Environmental Technology Site near Denver, Colorado, USA. The clean-up of this site requires this material to be packaged in such a way as to allow for efficient transportation to other sites or to a permanent geologic repository. Prior to off-site shipment of the material, it may be stored on-site for a period of time. For this reason, it is desirable to have a container capable of meeting the requirements for storage as well as the requirements for transportation. Most of the off-site transportation is envisioned to take place using the TRUPACT-II Type B package, with the Waste Isolation Pilot Plant (WIPP) as the destination. Prior to the development of this new container, the TRUPACT-II had a limit of 325 FGE (fissile gram equivalents) of plutonium due to criticality control concerns. Because of the relatively high plutonium content in the material to be transported, transporting 325 FGE per TRUPACT-II is uneconomical. Thus, the purpose of the new containers is to provide criticality control to increase the allowed TRUPACT-II payload and to provide a safe method for on-site storage prior to transport. This paper will describe the analysis and testing used to demonstrate that the Pipe Overpack Container provides safe on-site storage of plutonium bearing materials in unhardened buildings and provides criticality control during transportation within the TRUPACT-II. Analyses included worst-case criticality analyses, analyses of fork-lift time impacts, and analyses of roof structure collapse onto the container. Testing included dynamic crush tests, bare pipe impact tests, a 30-minute totally engulfing pool-fire test, and multiple package impact tests in end-on and side-on orientations

Experiments on container materials for Swiss high-level waste disposal projects. Part IV

International Nuclear Information System (INIS)

Simpson, J.P.

1989-12-01

One concept for final disposal of high-level waste in switzerland consists of a repository at a depth of 1000 to 1500 m in the crystalline bedrock of Northern Switzerland. The waste will be placed in a container which will be required to function as a high integrity barrier for at least 100 years. This report is the fourth and last in the current series dealing with the evaluation of potential materials for such containers. Four materials were identified for further evaluation in the first of these reports: cast steel, nodular cast iron, copper and Ti-Code 12. This report deals with the problem of demonstrating that cast steel containers will not fail by stress corrosion cracking and with the problem of hydrogen produced by the reduction of water. The experimental results on pre-cracked specimens revealed no susceptibility of cast steel to stress corrosion cracking under model repository conditions. No crack growth was detected on compact DCB specimens exposed in aerobic and anaerobic groundwaters at 80 and 140 o C for 16-24 months. Cast steel remains a candidate material for high-level waste containers. As expected from thermodynamic considerations no hydrogen could be detected from copper immersed in model groundwaters at 50 o C. Hydrogen is evolved from corroding steel under anaerobic conditions. Hydrogen evolution due to corrosion of iron or steel in waste repositories has to be considered in any safety analysis; the amounts produced can be significant. Evidence todate suggests that both cast steel and copper are suitable container materials. Because the corrosion behaviour of both materials is sensitive to service conditions, in particular length of the aerobic phase, groundwater chemistry and temperature, further testing should be undertaken when a specific site has been identified. (author) 9 tabs., 11 figs., 25 refs

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Gdowski, G.E.; Bullen, D.B.

1988-08-01

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials [CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)], which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

Energy Technology Data Exchange (ETDEWEB)

Gdowski, G.E.; Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

1988-08-01

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of containers for disposal of high-level radioactive waste. This waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr. During the first 50 yr after emplacement, they must be retrievable from the disposal site. Shortly after the containers are emplaced in the repository, they will be exposed to high temperatures and high gamma radiation fields from the decay of the high-level waste. This volume surveys the available data on oxidation and corrosion of the iron- to nickel-based austenitic materials (Types 304L and 316L stainless steels and Alloy 825) and the copper-based alloy materials (CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni)), which are the present candidates for fabrication of the containers. Studies that provided a large amount of data are highlighted, and those areas in which little data exists are identified. Examples of successful applications of these materials are given. On the basis of resistance to oxidation and general corrosion, the austenitic materials are ranked as follows: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is as follows: CDA 715 and CDA 613 (both best), and CDA 102 (worst). 110 refs., 30 figs., 13 tabs.

Containers for the transport of radioactive materials

International Nuclear Information System (INIS)

Bochard, C.

1975-01-01

The container for heat evolving radioactive materials has a metallic outer casing formed with outwardly projecting heat dissipating or cooling members, such as pins or fins, while each of its ends is formed with a flat flange which extends radially beyond the outer ends of the cooling members. A perforated wall extends between the flanges to define with same and with the periphery of the outer casing an annular space within which the cooling members are enclosed. This perforated wall is adapted to support a flexible covering sleeve the ends of which are clamped by inflatable seals between the periphery of the flanges and outer rings removably secured to the latter. Spraying means are provided within the aforesaid space to permit of projecting an uncontaminated liquid on the cooling members to cool the container before and/or while the latter is immersed in a loading and unloading pond with the sleeve mounted in position. The lower flange is provided with liquid collecting and evacuating means and compressed air may be injected into the said space to force the collected liquid outwardly. (auth)

Calculations of radiation damage in target, container and window materials for spallation neutron sources

International Nuclear Information System (INIS)

Wechsler, M.S.; Mansur, L.K.

1996-01-01

Radiation damage in target, container, and window materials for spallation neutron sources is am important factor in the design of target stations for accelerator-driver transmutation technologies. Calculations are described that use the LAHET and SPECTER codes to obtain displacement and helium production rates in tungsten, 316 stainless steel, and Inconel 718, which are major target, container, and window materials, respectively. Results are compared for the three materials, based on neutron spectra for NSNS and ATW spallation neutron sources, where the neutron fluxes are normalized to give the same flux of neutrons of all energies

Container material for the disposal of highly radioactive wastes: corrosion chemistry aspects

International Nuclear Information System (INIS)

Grauer, R.

1984-08-01

Prior to disposal in crystalline formations it is planned to enclose vitrified highly radioactive waste from nuclear power plants in metallic containers ensuring their isolation from the groundwater for at least 1,000 years. Appropriate metals can be either thermodynamically stable in the repository environment (such as copper), passive materials with very low corrosion rates (titanium, nickel alloys), or metals such as cast iron or unalloyed cast steels which, although they corrode, can be used in sections thick enough to allow for this corrosion. The first part of the report presents the essentials of corrosion science in order to enable even a non-specialist to follow the considerations and arguments necessary to choose the material and design the container against corrosion. Following this, the principles of the long-term extrapolation of corrosion behaviour are discussed. The second part summarizes and comments upon the literature search carried out to identify published results relevant to corrosion in a repository environment. Results of archeaological studies are included wherever possible. Not only the general corrosion behaviour but also localized corrosion and stress corrosion cracking are considered, and the influence of hydrogen on the material behaviour is discussed. Taking the corrosion behaviour as criterion, the author suggests the use either of copper or of cast iron or steel as an appropriate container material. The report concludes with proposals for further studies. (Auth.)

Standard practice for sampling special nuclear materials in multi-container lots

CERN Document Server

American Society for Testing and Materials. Philadelphia

1987-01-01

1.1 This practice provides an aid in designing a sampling and analysis plan for the purpose of minimizing random error in the measurement of the amount of nuclear material in a lot consisting of several containers. The problem addressed is the selection of the number of containers to be sampled, the number of samples to be taken from each sampled container, and the number of aliquot analyses to be performed on each sample. 1.2 This practice provides examples for application as well as the necessary development for understanding the statistics involved. The uniqueness of most situations does not allow presentation of step-by-step procedures for designing sampling plans. It is recommended that a statistician experienced in materials sampling be consulted when developing such plans. 1.3 The values stated in SI units are to be regarded as the standard. 1.4 This standard does not purport to address all of the safety problems, if any, associated with its use. It is the responsibility of the user of this standar...

Dose and radon measurements inside houses containing ash as building material

International Nuclear Information System (INIS)

Bodnar, R.; Lendvai, Z.; Somlai, J.; Nemeth, C.

1996-01-01

Radon concentration and external dose have been measured in dwellings that contain by-products of coal burning for building materials. The concentrations of 40 K, 232 Th, 238 U and 226 Ra have been determined in the materials. The date are analyzed according to indices frequently used for decision of utilizing the by-products. The observed daily fluctuation of the radon concentration in dwellings might exceed a factor of 5. (author)

Selection criteria for container materials at the proposed Yucca Mountain high level nuclear waste repository

International Nuclear Information System (INIS)

Halsey, W.G.

1989-11-01

A geological repository has been proposed for the permanent disposal of the nation's high level nuclear waste at Yucca Mountain in the Nevada desert. The containers for this waste must remain intact for the unprecedented service lifetime of 1000 years. A combination of engineering, regulatory, and licensing requirements complicate the container material selection. In parallel to gathering information regarding the Yucca Mountain service environment and material performance data, a set of selection criteria have been established which compare candidate materials to the performance requirements, and allow a quantitative comparison of candidates. These criteria assign relative weighting to varied topic areas such as mechanical properties, corrosion resistance, fabricability, and cost. Considering the long service life of the waste containers, it is not surprising that the corrosion behavior of the material is a dominant factor. 7 refs

19 CFR 10.462 - Packing materials and containers for shipment.

Science.gov (United States)

2010-04-01

... Free Trade Agreement Rules of Origin § 10.462 Packing materials and containers for shipment. (a... disregarded in determining the regional value content of a good imported into the United States. Accordingly, in applying either the build-down or build-up method for determining the regional value content of...

Performance of Cement Containing Laterite as Supplementary Cementing Material

Directory of Open Access Journals (Sweden)

Abbas Bukhari, Z. S.

2013-03-01

Full Text Available The utilization of different industrial waste, by-products or other materials such as ground granulated blast furnace slag, silica fume, fly ash, limestone, and kiln dust, etc. as supplemen- tary cementing materials has received considerable attention in recent years. A study has been conducted to look into the performance of laterite as Supplementary Cementing Materials (SCM. The study focuses on compressive strength performance of blended cement containing different percentage of laterite. The cement is replaced accordingly with percentage of 2 %, 5 %, 7 % and 10 % by weight. In addition, the effect of use of three chemically different laterites have been studied on physical performance of cement as in setting time, Le-Chatlier expansion, loss on ignition, insoluble residue, free lime and specifically compressive strength of cement cubes tested at the age of 3, 7, and 28 days. The results show that the strength of cement blended with laterite as SCM is enhanced. Key words: Portland cement, supplementary cementing materials (SCM, laterite, compressive strength KUI – 6/2013 Received January 4, 2012 Accepted February 11, 2013

Method for contamination control and barrier apparatus with filter for containing waste materials that include dangerous particulate matter

Science.gov (United States)

Pinson, Paul A.

1998-01-01

A container for hazardous waste materials that includes air or other gas carrying dangerous particulate matter has incorporated in barrier material, preferably in the form of a flexible sheet, one or more filters for the dangerous particulate matter sealably attached to such barrier material. The filter is preferably a HEPA type filter and is preferably chemically bonded to the barrier materials. The filter or filters are preferably flexibly bonded to the barrier material marginally and peripherally of the filter or marginally and peripherally of air or other gas outlet openings in the barrier material, which may be a plastic bag. The filter may be provided with a backing panel of barrier material having an opening or openings for the passage of air or other gas into the filter or filters. Such backing panel is bonded marginally and peripherally thereof to the barrier material or to both it and the filter or filters. A coupling or couplings for deflating and inflating the container may be incorporated. Confining a hazardous waste material in such a container, rapidly deflating the container and disposing of the container, constitutes one aspect of the method of the invention. The chemical bonding procedure for producing the container constitutes another aspect of the method of the invention.

Method for contamination control and barrier apparatus with filter for containing waste materials that include dangerous particulate matter

International Nuclear Information System (INIS)

Pinson, P.A.

1998-01-01

A container for hazardous waste materials that includes air or other gas carrying dangerous particulate matter has incorporated barrier material, preferably in the form of a flexible sheet, and one or more filters for the dangerous particulate matter sealably attached to such barrier material. The filter is preferably a HEPA type filter and is preferably chemically bonded to the barrier materials. The filter or filters are preferably flexibly bonded to the barrier material marginally and peripherally of the filter or marginally and peripherally of air or other gas outlet openings in the barrier material, which may be a plastic bag. The filter may be provided with a backing panel of barrier material having an opening or openings for the passage of air or other gas into the filter or filters. Such backing panel is bonded marginally and peripherally thereof to the barrier material or to both it and the filter or filters. A coupling or couplings for deflating and inflating the container may be incorporated. Confining a hazardous waste material in such a container, rapidly deflating the container and disposing of the container, constitutes one aspect of the method of the invention. The chemical bonding procedure for producing the container constitutes another aspect of the method of the invention. 3 figs

Corrosion of container materials for disposal of high-level radioactive wastes

Energy Technology Data Exchange (ETDEWEB)

Chun, K.S.; Park, H.S.; Yeon, J.W.; Ha, Y.K. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-01-01

In the corrosion aspect of container for the deep geological disposal of high-level radioactive waste, disposal concepts and the related container materials, which have been developed by advanced countries, have been reviewed. The disposal circumstances could be divided into the saturated and the unsaturated zones. The candidate materials in the countries, which consider the disposal in the unsaturated zone, are the corrosion resistant materials such as supper alloys and stainless steels, but those in the saturated zone is cupper, one of the corrosion allowable materials. By the results of the pitting corrosion test of sensitized stainless steels (such as 304, 304L, 316 and 316L), pitting potential is decreased with the degree of sensitization and the pitting corrosion resistance of 316L is higher than others. And so, the long-term corrosion experiment with 316L stainless steel specimens, sebsitized and non-sensitized, under the compacted bentonite and synthetic granitic groundwater has been being carried out. The results from the experiment for 12 months indicate that no evidence of pitting corrosion of the specimens has been observed but the crevice corrosion has occurred on the sensitized specimens even for 3 months. (author). 33 refs., 19 figs., 10 tabs.

Processing Uranium-Bearing Materials Containing Coal and Loam

Energy Technology Data Exchange (ETDEWEB)

Civin, V; Prochazka, J [Research and Development Laboratory No. 3 of the Uranium Industry, Prague, Czechoslovakia (Czech Republic)

1967-06-15

Among the ores which are classified as low-grade in the CSSR are mixtures of coal and bentonitic loam of tertiary origin, containing approximately 0.1% U and with a moisture content at times well above 20-30%. The uranium is held mainly by the carbonaceous component. Conventional processing of these materials presents various difficulties which are not easily overcome. During leaching the pulp thickens and frequently becomes pasty, due to the presence of montmorillonites. Further complications arise from the high sorption capacity of the materials (again primarily due to montmorillonites) and poor sedimentation of the viscous pulps. In addition, the materials are highly refractory to the leaching agents. The paper presents experience gained in solving the problems of processing these ores. The following basic routes were explored: (1) separation of the carbonaceous and loamy components: The organic component appears to be the main activity carrier. Processing the concentrated material upon separation of the inactive or less active loam may not only remove the thixotropic behaviour but also substantially reduce the cost of the ore treatment; (2) 'liquifying' the pulps or preventing the thickening of the pulp by addition of suitable agents; (3) joint acid or carbonate processing of the materials in question with current ore types; (4) removal or suppression of thixotropic behaviour by thermal pretreatment of the material; and (5) application of the 'acid cure' method. The first method appears to be the most effective, but it presents considerable difficulties due to the extreme dispersion of the carbonaceous phase and further research is being carried out. Methods 2 and 3 proved to be unacceptable. Method 4, which includes roasting at 300-400{sup o}C, is now being operated on an industrial scale. The final method has also shown definite advantages for particular deposits of high montmorillonite content material. (author)

12 CFR 503.2 - Exemptions of records containing investigatory material compiled for law enforcement purposes.

Science.gov (United States)

2010-01-01

... 12 Banks and Banking 5 2010-01-01 2010-01-01 false Exemptions of records containing investigatory material compiled for law enforcement purposes. 503.2 Section 503.2 Banks and Banking OFFICE OF THRIFT SUPERVISION, DEPARTMENT OF THE TREASURY PRIVACY ACT § 503.2 Exemptions of records containing investigatory material compiled for law enforcement...

Numerical Simulation on Dense Packing of Granular Materials by Container Oscillation

OpenAIRE

Jun Liu; Dongxu You

2013-01-01

The packing of granular materials is a basic and important problem in geomechanics. An approach, which generates dense packing of spheres confined in cylindrical and cuboidal containers in three steps, is introduced in this work. A loose packing structure is first generated by means of a reference lattice method. Then a dense packing structure is obtained in a container by simulating dropping of particles under gravitational forces. Furthermore, a scheme that makes the bottom boundary fluctua...

Design and construction of thermal desorption measurement system for tritium contained materials

International Nuclear Information System (INIS)

Hara, M.; Hatano, Y.; Calderoni, P.; Shimada, M.

2014-01-01

The dual-mode thermal desorption analysis system was designed and built in Idaho National Laboratory (INL) to examine the evolution of the hydrogen isotope gas from materials. The system is equipped with a mass spectrometer for stable hydrogen isotopes and an ionization chamber for tritium components. The performance of the system built was tested with using tritium contained materials. The evolution of tritiated gas species from contaminated materials was measured successfully by using the system. (author)

Shipping container for a bottle of radioactive material

International Nuclear Information System (INIS)

Soldan, D.W.

1975-01-01

A shipping container for a bottle of radioactive material comprises a can having a cavity therein for receiving the bottle, a screw cap for the can and the cavity, and an annular bottle retainer extending downwardly in the cavity from its upper end having an outwardly extending flange at its upper end clamped between the cap and the upper end of the cavity and an inwardly extending flange at its lower end receiving the neck of the bottle. The cap carries a sponge to absorb spillage from the bottle. (U.S.)

Alkaline degradation of organic materials contained in TRU wastes under repository conditions

International Nuclear Information System (INIS)

Otsuka, Yoshiki; Banba, Tsunetaka

2007-09-01

Alkaline degradation tests for 9 organic materials were conducted under the conditions of TRU waste disposal: anaerobic alkaline conditions. The tests were carried out at 90degC for 91 days. The sample materials for the tests were selected from the standpoint of constituent organic materials of TRU wastes. It has been found that cellulose and plastic solidified products are degraded relatively easily and that rubbers are difficult to degrade. It could be presumed that the alkaline degradation of organic materials occurs starting from the functional group in the material. Therefore, the degree of degradation difficulty is expected to be dependent on the kinds of functional group contained in the organic material. (author)

Fluorine uptake into human enamel around fluoride-containing dental materials during cariogenic pH cycling

Energy Technology Data Exchange (ETDEWEB)

Komatsu, H. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan)], E-mail: kom@den.hokudai.ac.jp; Yamamoto, H. [Graduate School of Dentistry, Osaka University, 1-8 Yamada-Oka, Suita 565-0871 (Japan); Nomachi, M. [Graduate School of Science, Osaka University, 1-1 Machikaneyama, Toyonaka 560-0043 (Japan); Yasuda, K. [The Wakasa wan Energy Research Center, 64-52-1 Hase, Tsuruga 914-0192 (Japan); Matsuda, Y.; Kinugawa, M.; Kijimura, T.; Sano, H. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan); Satou, T.; Oikawa, S.; Kamiya, T. [Advanced Radiation Technology, TARRI, JAEA, 1233 Watanuki-machi, Takasaki 370-1292 (Japan)

2009-06-15

Using PIGE (Proton Induced Gamma Emission) technique at TARRI (Takasaki Advanced Radiation Research Institute), Japan, we measured fluorine (F) uptake into the tooth enamel around two fluoride-containing materials during caries progression using pH cycling. Class V cavities in extracted human teeth were drilled and filled with fluoride-containing materials (i.e. 'Fuji IX' (FN) and 'UniFil flow with MEGA bond' (UF)) and a non-fluoride-containing material (i.e. 'SOLARE with MEGA bond' (SO)). Three 120 {mu}m longitudinal sections including the filling material were obtained from each tooth. In order to simulate daily acid attack occurring in the oral cavity, the pH cycling (pH 6.8-4.5) was carried out for 1, 3 and 5 weeks, separately. After pH cycling, the caries progression in all specimens was observed using transverse microradiography (TMR). The F and calcium distributions of the specimens were evaluated using PIGE and PIXE techniques. The F distribution of the specimens clearly showed the F uptake from FN into enamel adjacent to the filling material, while the F uptakes from UF and SO were not detected. For UF, the MEGA bond (non-fluoride-containing) between the tooth and UniFil flow interfered with the F absorption into the tooth. For FN, the amount of F uptake into the subsurface enamel increased during pH cycling. The amount of F uptake in 5-week pH cycling had significantly higher value compared to those in 1- and 3-week pH cycling. For UF and SO, there were no significant differences between the different durations of pH cycling. Among fluoride-containing materials, there were some differences in the F uptake with increased pH cycling, which could possibly lead to obtaining difference in clinical performance. The data obtained using PIGE and PIXE techniques were useful in understanding the benefit of fluorine by means of fluoride-containing material for preventing caries.

Fluorine uptake into human enamel around fluoride-containing dental materials during cariogenic pH cycling

International Nuclear Information System (INIS)

Komatsu, H.; Yamamoto, H.; Nomachi, M.; Yasuda, K.; Matsuda, Y.; Kinugawa, M.; Kijimura, T.; Sano, H.; Satou, T.; Oikawa, S.; Kamiya, T.

2009-01-01

Using PIGE (Proton Induced Gamma Emission) technique at TARRI (Takasaki Advanced Radiation Research Institute), Japan, we measured fluorine (F) uptake into the tooth enamel around two fluoride-containing materials during caries progression using pH cycling. Class V cavities in extracted human teeth were drilled and filled with fluoride-containing materials (i.e. 'Fuji IX' (FN) and 'UniFil flow with MEGA bond' (UF)) and a non-fluoride-containing material (i.e. 'SOLARE with MEGA bond' (SO)). Three 120 μm longitudinal sections including the filling material were obtained from each tooth. In order to simulate daily acid attack occurring in the oral cavity, the pH cycling (pH 6.8-4.5) was carried out for 1, 3 and 5 weeks, separately. After pH cycling, the caries progression in all specimens was observed using transverse microradiography (TMR). The F and calcium distributions of the specimens were evaluated using PIGE and PIXE techniques. The F distribution of the specimens clearly showed the F uptake from FN into enamel adjacent to the filling material, while the F uptakes from UF and SO were not detected. For UF, the MEGA bond (non-fluoride-containing) between the tooth and UniFil flow interfered with the F absorption into the tooth. For FN, the amount of F uptake into the subsurface enamel increased during pH cycling. The amount of F uptake in 5-week pH cycling had significantly higher value compared to those in 1- and 3-week pH cycling. For UF and SO, there were no significant differences between the different durations of pH cycling. Among fluoride-containing materials, there were some differences in the F uptake with increased pH cycling, which could possibly lead to obtaining difference in clinical performance. The data obtained using PIGE and PIXE techniques were useful in understanding the benefit of fluorine by means of fluoride-containing material for preventing caries.

A review of a radioactive material shipping container including design, testing, upgrading compliance program and shipping logistics

International Nuclear Information System (INIS)

Celovsky, A.; Lesco, R.; Gale, B.; Sypes, J.

2003-01-01

Ten years ago Atomic Energy of Canada developed a Type B(U)-85 shipping container for the global transport of highly radioactive materials. This paper reviews the development of the container, including a summary of the design requirements, a review of the selected materials and key design elements, and the results of the major qualification tests (drop testing, fire test, leak tightness testing, and shielding integrity tests). As a result of the testing, improvements to the structural, thermal and containment design were made. Such improvements, and reasons thereof, are noted. Also provided is a summary of the additional analysis work required to upgrade the package from a Type B(U) to a Type B(F), i.e. essentially upgrading the container to include fissile radioisotopes to the authorized radioactive contents list. Having a certified shipping container is only one aspect governing the global shipments of radioactive material. By necessity the shipment of radioactive material is a highly regulated environment. This paper also explores the experiences with other key aspects of radioactive shipments, including the service procedures used to maintain the container certification, the associated compliance program for radioactive material shipments, and the shipping logistics involved in the transport. (author)

Stress corrosion cracking of steam generator tubing materials in lead containing solution

International Nuclear Information System (INIS)

Kim, H.P.; Hwang, S.S.; Kim, J.S.; Hong, J.H.

2007-01-01

Stress corrosion cracking (SCC) in lead (Pb) containing environments has been one of key issues in the nuclear power industry since Pb had been identified as a cause of the SCC of steam generator (SG) tubing materials in some power plants. To mitigate or prevent degradation of SG tubing materials, a mechanistic understanding of SCC in Pb containing environment is needed, along with an understanding of the source and transport behaviors of Pb species in the secondary circuit. In this work, SCC behaviors of Alloy 600 in Pb containing environments were studied. Influences of microstructures of Alloy 600 and the inhibitive additives were investigated using the C-ring and the slow strain rate tests in caustic solution and demineralized water at 315 o C. Microstructures of Alloy 600 were varied by heat treatment at different temperatures. The additives examined were nickel boride (NiB) and cerium boride (CeB 6 ). The surface films were analyzed using Auger Electron Spectroscopy (AES) and Energy Dispersive X-ray Spectroscopy (EDS). The SCC mode varied with microstructure. Effectiveness of the additives in Pb containing environments is discussed. (author)

Gas-handling system for studies of tritium-containing materials

International Nuclear Information System (INIS)

Carstens, D.H.W.

1975-01-01

A gas handling system for preparation and study of tritium containing compounds and materials is described. The system at any one time can handle amounts of DT gas up to about 3 moles and has provisions for purification, storage, and measurement of the gas. Experimental conditions covering the ranges 20 to 800 0 C and 0.1 Pa to 137 MPa (10 -2 torr to 20,000 psi) can be maintained. (auth)

49 CFR 837.3 - Published reports, material contained in the public accident investigation dockets, and accident...

Science.gov (United States)

2010-10-01

... 49 Transportation 7 2010-10-01 2010-10-01 false Published reports, material contained in the... OF RECORDS IN LEGAL PROCEEDINGS § 837.3 Published reports, material contained in the public accident... submitted, in writing, to the Public Inquiries Branch. Demands for specific published reports and studies...

Process of preparing ethanol by continuous fermentation of polysaccharide-containing materials

Energy Technology Data Exchange (ETDEWEB)

Ehnstroem, L.K.J.

1981-04-16

The invention concerns a process of preparing ethanol by continuous fermentation of polysaccharide - containing raw materials. Fermentation, hereby, occurs in one or several fermentors while dividing one stream of the fermentation liquid into a yeast-concentrate stream and a yeast-free stream and, if neccessary, a sludge stream. The yeast-concentrate stream is re-fed into the fermentor and at least part of the yeast-free stream is directed into a simple evaporator corresponding to one or several distilling stages where it is separated partially in an ethanol-enriched initial vapour stream supplying a facility to produce the desired ethanol quality, and partially in a liquid initial bottom stream re-fed at least in part into the fermentor. The characteristic feature of this new process is that a raw-material stream is fed into a closed circuit containing the fermentor and the evaporator, and that, in the evaporator, the raw-material stream is hydrolysed to a fermentable state. This hydrolysis is carried out most favourably by enzymes - preferably a gluco-amylase - at a temperature ranging from 35/sup 0/C to 75/sup 0/C.

Kinetics and Efficiency of H2O2 Activation by Iron-Containing Minerals and Aquifer Materials

Science.gov (United States)

Pham, Anh Le-Tuan; Doyle, Fiona M.; Sedlak, David L.

2014-01-01

To gain insight into factors that control H2O2 persistence and ˙OH yield in H2O2-based in situ chemical oxidation systems, the decomposition of H2O2 and transformation of phenol were investigated in the presence of iron-containing minerals and aquifer materials. Under conditions expected during remediation of soil and groundwater, the stoichiometric efficiency, defined as the amount of phenol transformed per mole of H2O2 decomposed, varied from 0.005 to 0.28%. Among the iron-containing minerals, iron oxides were 2 to 10 times less efficient in transforming phenol than iron-containing clays and synthetic iron-containing catalysts. In both iron-containing mineral and aquifer materials systems, the stoichiometric efficiency was inversely correlated with the rate of H2O2 decomposition. In aquifer materials systems, the stoichiometric efficiency was also inversely correlated with the Mn content, consistent with the fact that the decomposition of H2O2 on manganese oxides does not produce ˙OH. Removal of iron and manganese oxide coatings from the surface of aquifer materials by extraction with citrate-bicarbonate-dithionite slowed the rate of H2O2 decomposition on aquifer materials and increased the stoichiometric efficiency. In addition, the presence of 2 mM of dissolved SiO2 slowed the rate of H2O2 decomposition on aquifer materials by over 80% without affecting the stoichiometric efficiency. PMID:23047055

Piezoelectric ceramic material, containing PbNb2O6, K2Nb2O6

International Nuclear Information System (INIS)

Fesenko, E.G.; Filip'ev, V.S.; Razumovskaya, O.N.; Cherner, Ya.E.; Rudkovskaya, L.M.; Zav'yalov, V.P.; Molchanova, R.A.; Kryshtop, V.G.; Panich, A.E.; Servuli, V.A.

1984-01-01

A new piezoelectric ceramic material including PbNb 2 O 6 , K 2 Nb 2 O 6 is prepared. Above the new material contains Nb 2 O 5 . The invention relates to piezotechnique. The principal advantage of this material for acoustic converters is high anisotropy of piezoelectric properties as well as high Curie temperature (T C =539-553 deg C). The composition containing 93.96 mole% PbNb 2 O 6 ; 2.48 mole% K 2 Nb 2 O 6 and 3.56 mole% Nb 2 O 5 has optimum content of parameters

Closure for casks containing radioactive materials

International Nuclear Information System (INIS)

Hall, G.V.B.; Mallory, C.W.

1990-01-01

This patent describes an improved closure for covering and sealing an opening in a single cask for containing radioactive material, wherein the opening is characterized by a ledge. It comprises: an inner lid receivable within the opening and having a gasket means that is seatable over the ledge; an outer lid which is likewise receivable into the opening and securable therearound when the outer lid is rotated relative to the opening. The inner lid remaining stationary relative to the cask opening when the outer lid is rotated and having no torque applied thereto by the outer lid when the outer lid is rotated, and bolt means threadedly mounted through the outer lid for applying a compressive force between the inner and outer lids after the outer lid has been secured to the opening in order to depress the gasket means of the inner lid into sealing engagement with the ledge while avoiding the application of torsion between the gasket means and the ledge

Recommended welding criteria for use in the fabrication of shipping containers for radioactive materials

International Nuclear Information System (INIS)

Monroe, R.E.; Woo, H.H.; Sears, R.G.

1984-03-01

Welding and related operations are evaluated to assess the controls required to prevent weld-related failure of shipping containers used for transportation of radioactive materials. The report includes (1) recommended criteria for controlling welding as applied to shipping containers and (2) a discussion of modifications of the recommended industry Codes as applied to shipping containers. 13 references, 2 tables

Preparation and use of polymeric materials containing hydrophobic anions and plasticizers for separation of cesium and strontium

International Nuclear Information System (INIS)

Abney, K.D.; Kinkead, S.A.; Mason, C.F.V.; Rais, J.

1997-01-01

Preparation and use is described for polymeric materials containing hydrophobic anions and plasticizers for extraction of cesium and strontium. The use of polymeric materials containing plasticizers which are solvents for hydrophobic anions such as derivatives of cobalt dicarbollide or tetraphenylborate which are capable of extracting cesium and strontium ions from aqueous solutions in contact with the polymeric materials, is described. The polymeric material may also include a synergistic agent for a given ion like polyethylene glycol or a crown ether, for removal of radioactive isotopes of cesium and strontium from solutions of diverse composition and, in particular, for solutions containing large excess of sodium nitrate

Inherent N,O-containing carbon frameworks as electrode materials for high-performance supercapacitors.

Science.gov (United States)

Hu, Fangyuan; Wang, Jinyan; Hu, Shui; Li, Linfei; Wang, Gang; Qiu, Jieshan; Jian, Xigao

2016-09-15

N,O-Containing micropore-dominated materials have been developed successfully via temperature-dependent cross-linking of 4,4'-(dioxo-diphenyl-2,3,6,7-tetraazaanthracenediyl)dibenzonitrile (DPDN) monomers. By employing a molecular engineering strategy, we have designed and synthesized a series of porous heteroatom-containing carbon frameworks (PHCFs), in which nitrogen and oxygen heteroatoms are distributed homogeneously throughout the whole framework at the atomic level, which can ensure the stability of its electrical properties. The as-made PHCFs@550 exhibits a high specific capacitance of 378 F g -1 , with an excellent long cycling life, including excellent cycling stability (capacitance retention of ca. 120% over 20 000 cycles). Moreover, the successful preparation of PHCFs provides new insights for the fabrication of nitrogen and oxygen-containing electrode materials from readily available components via a facile route.

Contained scanning electron microscope facility for examining radioactive materials

International Nuclear Information System (INIS)

Hsu, C.W.

1986-03-01

At the Savannah River Laboratory (SRL) radioactive solids are characterized with a scanning electron microscope (SEM) contained in a glove box. The system includes a research-grade Cambridge S-250 SEM, a Tracor Northern TN-5500 x-ray and image analyzer, and a Microspec wavelength-dispersive x-ray analyzer. The containment facility has a glove box train for mounting and coating samples, and for housing the SEM column, x-ray detectors, and vacuum pumps. The control consoles of the instruments are located outside the glove boxes. This facility has been actively used since October 1983 for high alpha-activity materials such as plutonium metal and plutonium oxide powders. Radioactive defense waste glasses and contaminated equipment have also been examined. During this period the facility had no safety-related incidents, and personnel radiation exposures were maintained at less than 100 mrems

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Bullen, D.B.; Gdowski, G.E.

1988-08-01

Three copper-based alloys and three iron- to nickel-based austenitic alloys are being considered as possible materials for fabrication of high-level radioactive-waste disposal containers. The waste will include spent fuel assemblies from reactors as well as high-level waste in borosilicate glass and will be sent to the prospective site at Yucca Mountain, Nevada, for disposal. The copper-based alloy materials are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). The austenitic materials are Types 304L and 316L stainless steels and Alloy 825. The waste-package containers must maintain substantially complete containment for at least 300 yr and perhaps as long as 1000 yr, and they must be retrievable from the disposal site during the first 50 yr after emplacement. The containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. This volume surveys the available data on the phase stability of both groups of candidate alloys. The austenitic alloys are reviewed in terms of the physical metallurgy of the iron-chromium-nickel system, martensite transformations, carbide formation, and intermetallic-phase precipitation. The copper-based alloys are reviewed in terms of their phase equilibria and the possibility of precipitation of the minor alloying constituents. For the austenitic materials, the ranking based on phase stability is: Alloy 825 (best), Type 316L stainless steel, and then Type 304L stainless steel (worst). For the copper-based materials, the ranking is: CDA 102 (oxygen-free copper) (best), and then both CDA 715 and CDA 613. 75 refs., 24 figs., 6 tabs

Fabricating Composite-Material Structures Containing SMA Ribbons

Science.gov (United States)

Turner, Travis L.; Cano, Roberto J.; Lach, Cynthia L.

2003-01-01

An improved method of designing and fabricating laminated composite-material (matrix/fiber) structures containing embedded shape-memory-alloy (SMA) actuators has been devised. Structures made by this method have repeatable, predictable properties, and fabrication processes can readily be automated. Such structures, denoted as shape-memory-alloy hybrid composite (SMAHC) structures, have been investigated for their potential to satisfy requirements to control the shapes or thermoelastic responses of themselves or of other structures into which they might be incorporated, or to control noise and vibrations. Much of the prior work on SMAHC structures has involved the use SMA wires embedded within matrices or within sleeves through parent structures. The disadvantages of using SMA wires as the embedded actuators include (1) complexity of fabrication procedures because of the relatively large numbers of actuators usually needed; (2) sensitivity to actuator/ matrix interface flaws because voids can be of significant size, relative to wires; (3) relatively high rates of breakage of actuators during curing of matrix materials because of sensitivity to stress concentrations at mechanical restraints; and (4) difficulty of achieving desirable overall volume fractions of SMA wires when trying to optimize the integration of the wires by placing them in selected layers only.

Liquid filter for liquids containing radioactive materials

International Nuclear Information System (INIS)

Rohleder, N.; Schwarz, F.

1986-01-01

A device for filtering radioactive liquids loaded with solids is described, which has a pressure-resistant housing with a lid and an incomer for the turbid liquid and a collecting space and drain for the filtrate at the bottom of the housing. A filter cartridge is present in this housing. Such a filtering device must be suitable for use in nuclear plants, must be easy to replace by remote control and must minimise the carrying over of radioactive particles. This problem should be solved by the filter cartridge consisting of a large number of horizontal filter plates stacked above one another, which carry a deep layer filter material acting in the sub-micron range. The turbid liquid runs into the centre of the stack of filter plates via a vertical central duct. The intermediate spaces between the filter places are connected to this central duct via the layer of filter material. The filter plates are sealed against one another on the outer circumference and have radial drain openings for the filtrate on the outside. The central duct is sealed at the lower end by a plate. When the filter cartridge is replaced, the radioactive waste in the filter cartridge remains safely enclosed and can be conditioned in suitable containers. (orig.) [de

Corrosion of container and infrastructure materials under clay repository conditions

International Nuclear Information System (INIS)

Debruyn, W.; Dresselaers, J.; Vermeiren, P.; Kelchtermans, J.; Tas, H.

1991-01-01

With regard to the disposal of high-level radioactive waste, it was recommended in a IAEA Technical Committee meeting to perform tests in realistic environments corresponding with normal and accidental conditions, to qualify and apply corrosion monitoring techniques for corrosion evaluation under real repository conditions and to develop corrosion and near-field evolution models. The actual Belgian experimental programme for the qualification of a container for long-term HLW storage in clay formations complies with these recommendations. The emphasis in the programme is indeed on in situ corrosion testing and monitoring and on in situ control of the near-field chemistry. Initial field experiments were performed in a near-surface clay quarry at Terhaegen. Based on a broad laboratory material screening programme and in agreement with the Commission of the European Communities, three reference materials were chosen for extensive in situ overpack testing. Ti/0.2 Pd and Hastelloy C-4 were chosen as reference corrosion resistant materials and a low-carbon steel as corrosion allowance reference material. This report summarizes progress made in the material qualification programme since the CEC contract of 1983-84. 57 Figs.; 15 Tabs.; 18 Refs

Conversion of plutonium-containing materials into borosilicate glass using the glass material oxidation and dissolution system

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.; Parker, G.W.

1996-01-01

The end of the cold war has resulted in excess plutonium-containing materials (PCMs) in multiple chemical forms. Major problems are associated with the long-term management of these materials: safeguards and nonproliferation issues; health, environment, and safety concerns; waste management requirements; and high storage costs. These issues can be addressed by conversion of the PCMs to glass: however, conventional glass processes require oxide-like feed materials. Conversion of PCMs to oxide-like materials followed by vitrification is a complex and expensive process. A new vitrification process has been invented, the Glass Material Oxidation and Dissolution System (GMODS) to allow direct conversion of PCMs to glass. GMODS directly converts metals, ceramics, and amorphous solids to glass; oxidizes organics with the residue converted to glass; and converts chlorides to borosilicate glass and a secondary sodium chloride stream. Laboratory work has demonstrated the conversion of cerium (a plutonium surrogate), uranium (a plutonium surrogate), Zircaloy, stainless steel, multiple oxides, and other materials to glass. Equipment options have been identified for processing rates between 1 and 100,000 t/y. Significant work, including a pilot plant, is required to develop GMODS for applications at an industrial scale

Modeling of the solution interaction properties of plastic materials used in pharmaceutical product container systems.

Science.gov (United States)

Jenke, Dennis; Couch, Tom; Gillum, Amy; Sadain, Salma

2009-01-01

Material/water equilibrium binding constants (Eb) were determined for 14 organic solutes and 17 plastic raw materials that could be used in pharmaceutical product container systems. Correlations between the measured binding constants and the organic solute's octanol/water and hexane/water partition coefficients were obtained. In general, while the materials examined exhibited a wide range of binding characteristics, the tested materials by and large fell within two broad classes: (1) those that were octanol-like in their binding characteristics, and (2) those that were hexane-like. Materials of the same class (e.g., polypropylenes) generally had binding models that were very similar. Rank ordering of the materials in terms of their magnitude of drug binding (least binding to most binding) was as follows: polypropylene < polyethylene < polyamide < styrene-ethylene-butylene-styrene < copolyester ether elastomer approximately equal to amine-terminated poly fatty acid amide polymer. The utilization of the developed models to estimate drug loss via sorption by the container is discussed.

Martensitic/ferritic steels as container materials for liquid mercury target of ESS

International Nuclear Information System (INIS)

Dai, Y.

1996-01-01

In the previous report, the suitability of steels as the ESS liquid mercury target container material was discussed on the basis of the existing database on conventional austenitic and martensitic/ferritic steels, especially on their representatives, solution annealed 316 stainless steel (SA 316) and Sandvik HT-9 martensitic steel (HT-9). Compared to solution annealed austenitic stainless steels, martensitic/ferritic steels have superior properties in terms of strength, thermal conductivity, thermal expansion, mercury corrosion resistance, void swelling and irradiation creep resistance. The main limitation for conventional martensitic/ferritic steels (CMFS) is embrittlement after low temperature (≤380 degrees C) irradiation. The ductile-brittle transition temperature (DBTT) can increase as much as 250 to 300 degrees C and the upper-shelf energy (USE), at the same time, reduce more than 50%. This makes the application temperature range of CMFS is likely between 300 degrees C to 500 degrees C. For the present target design concept, the temperature at the container will be likely controlled in a temperature range between 180 degrees C to 330 degrees C. Hence, CMFS seem to be difficult to apply. However, solution annealed austenitic stainless steels are also difficult to apply as the maximum stress level at the container will be higher than the design stress. The solution to the problem is very likely to use advanced low-activation martensitic/ferritic steels (LAMS) developed by the fusion materials community though the present database on the materials is still very limited

Thermodynamics of a natural draught-cooled container deposit for radioactive materials

International Nuclear Information System (INIS)

Schoenfeld, R.; Dorst, H.J.

1982-01-01

This article interprets a calculation model for the determination of heat dissipation and temperature distribution of cooling air within an intermediate container deposit cooled with air by natural draught for self-heating materials. The mathematical modelling of the physical processes in the flowing medium 'air' takes place stationarily and one-dimensionelly in flow direction. The calculation model is based on known procedures for the convective heat transmission, for the description of channel flow, and for the determination of pressure losses. It provides the thermodynamic characteristic data which are required for an engineering and safety design of a container deposit. (orig.) [de

10 CFR 32.74 - Manufacture and distribution of sources or devices containing byproduct material for medical use.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Manufacture and distribution of sources or devices... SPECIFIC DOMESTIC LICENSES TO MANUFACTURE OR TRANSFER CERTAIN ITEMS CONTAINING BYPRODUCT MATERIAL Generally Licensed Items § 32.74 Manufacture and distribution of sources or devices containing byproduct material for...

Detecting nuclear materials smuggling: performance evaluation of container inspection policies.

Science.gov (United States)

Gaukler, Gary M; Li, Chenhua; Ding, Yu; Chirayath, Sunil S

2012-03-01

In recent years, the United States, along with many other countries, has significantly increased its detection and defense mechanisms against terrorist attacks. A potential attack with a nuclear weapon, using nuclear materials smuggled into the country, has been identified as a particularly grave threat. The system for detecting illicit nuclear materials that is currently in place at U.S. ports of entry relies heavily on passive radiation detectors and a risk-scoring approach using the automated targeting system (ATS). In this article we analyze this existing inspection system and demonstrate its performance for several smuggling scenarios. We provide evidence that the current inspection system is inherently incapable of reliably detecting sophisticated smuggling attempts that use small quantities of well-shielded nuclear material. To counter the weaknesses of the current ATS-based inspection system, we propose two new inspection systems: the hardness control system (HCS) and the hybrid inspection system (HYB). The HCS uses radiography information to classify incoming containers based on their cargo content into "hard" or "soft" containers, which then go through different inspection treatment. The HYB combines the radiography information with the intelligence information from the ATS. We compare and contrast the relative performance of these two new inspection systems with the existing ATS-based system. Our studies indicate that the HCS and HYB policies outperform the ATS-based policy for a wide range of realistic smuggling scenarios. We also examine the impact of changes in adversary behavior on the new inspection systems and find that they effectively preclude strategic gaming behavior of the adversary. © 2011 Society for Risk Analysis.

Application of the ASME code in designing containment vessels for packages used to transport radioactive materials

International Nuclear Information System (INIS)

Raske, D.T.; Wang, Z.

1992-01-01

The primary concern governing the design of shipping packages containing radioactive materials is public safety during transport. When these shipments are within the regulatory jurisdiction of the US Department of Energy, the recommended design criterion for the primary containment vessel is either Section III or Section VIII, Division 1, of the ASME Boiler and Pressure Vessel Code, depending on the activity of the contents. The objective of this paper is to discuss the design of a prototypic containment vessel representative of a packaging for the transport of high-level radioactive material

Possibilities of using new technology materials in constructing the radioactive waste containers The paper will consider using the latest technologies in material science for building

International Nuclear Information System (INIS)

Itu, Razvan Bogdan

2008-01-01

The paper will consider using the latest technologies in materials science for building the radioactive waste containers. A new amorphous steel has been discovered by the scientists from the University of Virginia, a material three times stronger then conventional steel and non-magnetic. Scientists shown that this steel, DARVA - Glass 101, has superior anticorrosive proprieties. The paper will also consider using Para-Aramides in protecting the radioactive waste containers. Chemical and physical properties of these materials shown a great tensile strength and the inter-chain bonds make these materials extremely strong. (author)

Evaluation of copper, aluminum bronze, and copper-nickel container material for the Yucca mountain project

International Nuclear Information System (INIS)

Kass, J.

1990-01-01

Copper, 70 percent aluminum bronze, and 70/30 copper-nickel were evaluated as potential waste-packaging materials as part of the Yucca Mountain Project. The proposed waste repository site is under a desert mountain in southern Nevada. The expected temperatures at the container surface are higher than at other sites, about 250C at the beginning of the containment period; they could fall below the boiling point of water during this period, but will be exposed to very little water, probably less than 5 l/a. Initial gamma flux will be 10 4 rad/h, and no significant hydrostatic or lithostatic pressure is expected. Packages will contain PWR or BWR fuel, or processed-glass waste. Three copper alloys are being considered for containers: oxygen-free copper (CDA 102); 7 percent aluminum bronze (CDA 613); and 70/30 copper-nickel (CDA 715). Phase separation due to prolonged thermal exposure could be a problem for the two alloys, causing embrittlement. The reduction of internal oxides present in pure copper by hydrogen could cause mechanical degradation. Corrosion and oxidation rates measured for the three materials in well water with and without gamma irradiation at flux rates about ten times higher than those expected were all quite small. The corrosion/oxidation rates for CDA715 show a marked increase under irradiation, but are still acceptable. In the presence of ammonia and other nitrogen-bearing species stress corrosion cracking (SCC) is a concern. Welded U-bend specimens of all three materials have been tested for up to 10000 h in highly irradiated environments, showing no SCC. There was some alloy segregation in the Al bronze specimens. The investigators believe that corrosion and mechanical properties will not present problems for these materials at this site. Further work is needed in the areas of weld inspection, welding techniques, embrittlement of weld metal, the effects of dropping the containers during emplacement, and stress corrosion cracking. Other materials

ERG review of waste package container materials selection and corrosion

International Nuclear Information System (INIS)

Moak, D.P.; Perrin, J.S.

1986-07-01

The Engineering Review Group (ERG) was established by the Office of Nuclear Waste Isolation (ONWI) to help evaluate engineering-related issues in the US Department of Energy's nuclear waste repository program. The October 1984 meeting of the ERG reviewed the waste package container materials selection and corrosion. This report documents the ERG's comments and recommendations on these subjects and the ONWI response to the specific points raised by the ERG

Process for surface treatment of zirconium-containing cladding materials for fuel element or other components for nuclear reactors

International Nuclear Information System (INIS)

Videm, K.G.; Lunde, L.R.; Kooyman, H.H.

1975-01-01

A process for the surface treatment of zirconium-base cladding materials for fuel elements or other components for nuclear reactors is described. The treatment includes pickling the cladding material in a fluoride-containing bath, and then applying a protective coating through oxidation to the pickled cladding material. The fluoride-containing contaminants which remain on the surface of the cladding material during pickling are removed or rendered harmless by anodic oxidation

Improved process for producing a fermentation product from a lignocellulose-containing material

DEFF Research Database (Denmark)

2017-01-01

The present invention relates to the production of hydrolyzates from a lignocellulose-containing material, and to fermentation of the hydrolyzates. More specifically, the present invention relates to the detoxification of phenolic inhibitors and toxins formed during the processing of lignocellulose...

A process for producing a fermentation product from a lignocellulose-containing material

DEFF Research Database (Denmark)

2016-01-01

The present invention relates to the production of hydrolyzates from a lignocellulose-containing material, and to fermentation of the hydrolyzates. More specifically, the present invention relates to the detoxification of phenolic inhibitors and toxins formed during the processing of lignocellulose...

Direct containment heating experiments in Zion Nuclear Power Plant geometry using prototypic materials

International Nuclear Information System (INIS)

Binder, J.L.; McUmber, L.M.; Spencer, B.W.

1993-01-01

Direct Containment Heating (DCH) experiments have been completed which utilize prototypic core materials. The experiments reported on here are a continuation of the Integral Effects Testing (IET) DCH program. The experiments incorporated a 1/40 scale model of the Zion Nuclear Power Plant containment structures. The model included representations of the primary system volume, RPV lower head, cavity and instrument tunnel, and the lower containment structures. The experiments were steam driven. Iron-alumina thermite with chromium was used as a core melt stimulant in the earlier IET experiments. These earlier IET experiments at Argonne National Laboratory (ANL) and Sandia National Laboratories (SNL) provided useful data on the effect of scale on DCH phenomena; however, a significant question concerns the potential experiment distortions introduced by the use of non-prototypic iron/alumina thermite. Therefore, further testing with prototypic materials has been carried out at ANL. Three tests have been completed, DCH-U1A, U1B and U2. DCH-U1A and U1B employed an inerted containment atmosphere and are counterpart to the IET-1RR test with iron/alumina thermite. DCH-U2 employed nominally the same atmosphere composition of its counterpart iron/alumina test, IET-6. All tests, with prototypic material, have produced lower peak containment pressure rises; 45, 111 and 185 kPa in U1A, U1B and U2, compared to 150 and 250 kPa IET-1RR and 6. Hydrogen production, due to metal-steam reactions, was 33% larger in U1B and U2 compared to IET-1RR and IET-6. The pressurization efficiency was consistently lower for the corium tests compared to the IET tests

Safety analysis report: packages. Argonne National Laboratory SLSF test train shipping container, P-1 shipment. Fissile material. Final report

International Nuclear Information System (INIS)

Meyer, C.A.

1975-06-01

The package is used to ship an instrumented test fuel bundle (test train) containing fissile material. The package assembly is Argonne National Laboratory (ANL) Model R1010-0032. The shipment is fissile class III. The packaging consists of an outer carbon steel container into which an inner container is placed; the inner container is separated from the outer container by urethane foam cushioning material. The test train is supported in the inner container by a series of transverse supports spaced along the length of the test train. Both the inner and outer containers are closed with bolted covers. The covers do not seal the containers in a leaktight manner. The gross weight of the shipment is about 8350 lb. The unirradiated fissile material content is less than 3 kg of UO 2 of up to 93.2 percent enrichment. This is a Type A quantity (transport group III and less than 3 curies) of radioactive material which does not require shielding, cooling or heating, or neutron absorption or moderation functions in its packaging. The maximum exterior dimensions of the container are 37 ft 11 in. long, 24 1 / 2 in. wide, and 19 3 / 4 in. high

Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

Science.gov (United States)

Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

2009-01-27

A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

Proportioning equipment for vibration filling and compacting of grain materials in pipe containers, especially of fuel elements

International Nuclear Information System (INIS)

Pinkas, V.; Filip, Z.; Beranek, J.

1981-01-01

The equipment consists of a base plate to which are attached the fastening collar fo the pipe container and the guide column with the height-adjustable support. The filling pipe is fixed to the support. The proportioning equipment prevents particles of grain material from segregation, thus allowing to achieve homogeneity of the material in the whole volume to be compacted. It also allows determining the height of the column of material in the pipe container without destructive effects on the stacked material. The equipment is designed for the manufacture of shortened fuel elements. (J.B.)

Material characteristics and construction methods for a typical research reactor concrete containment in Iran

International Nuclear Information System (INIS)

Ebrahimia, Mahsa; Suha, Kune Y.; Eghbalic, Rahman; Jahan, Farzaneh Asadi malek

2012-01-01

Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran

Material characteristics and construction methods for a typical research reactor concrete containment in Iran

Energy Technology Data Exchange (ETDEWEB)

Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

2012-10-15

Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

Radiation levels on empty cylinders containing heel material

Energy Technology Data Exchange (ETDEWEB)

Shockley, C.W. [Martin Marietta Energy Systems, Inc., Paducah, KY (United States)

1991-12-31

Empty UF{sub 6} cylinders containing heel material were found to emit radiation levels in excess of 200 mr/hr, the maximum amount stated in ORO-651. The radiation levels were as high as 335 mr/hr for thick wall (48X and 48Y) cylinders and 1050 mr/hr for thin wall (48G and 48H) cylinders. The high readings were found only on the bottom of the cylinders. These radiation levels exceeded the maximum levels established in DOT 49 CFR, Part 173.441 for shipment of cylinders. Holding periods of four weeks for thick-wall cylinders and ten weeks for thin-wall cylinders were established to allow the radiation levels to decay prior to shipment.

Spectral studies on the interaction of acetylacetone with aluminum-containing MCM-41 mesoporous materials

International Nuclear Information System (INIS)

Zanjanchi, M.A.; Vaziri, M.

2008-01-01

Diffuse reflectance spectroscopy (DRS) was used to study the interaction of acetylacetone (acac) with the mesoporous aluminum-containing MCM-41 materials. A room temperature synthesis method was used for preparation of purely siliceous MCM-41 and for aluminum-containing MCM-41 materials. Samples with Si/Al ratios of 50, 20, 10 and 5 were synthesized. The synthesized mesoporous materials possess highly ordered structure and high surface area as evidenced from X-ray diffraction and nitrogen physisorption measurements, respectively. The treatment of the as-synthesized aluminum-containing MCM-41 samples with acac shows a distinct band at ∼290 nm. This band is assigned to six coordinated aluminum atoms in the structure which is produced by diffusion of acac molecules through surfactant micelles and their interaction with aluminum atoms. The 290-nm band disappears upon several successive washing of the sample with ethanol. The treatment of the calcined aluminum-containing MCM-41 sample with acac produces the same 290-nm band where its intensity increases with the aluminum content of the sample. The intensity of this band is reduced upon successive ethanol washing, but remains nearly constant after three times washing. This irremovable aluminum species can be assigned to framework aluminum. The measured acidity for our aluminum-containing MCM-41 samples correlates linearly with the intensity of 290-nm band for the ethanol treated samples. This supports the idea that the Bronsted acidity in aluminum-modified MCM-41 samples is a function of the amount of tetrahedral framework aluminum in the structure

Contained x-ray diffraction goniometer for examination of radioactive materials

International Nuclear Information System (INIS)

Smith, P.K.; Osgood, B.C.; Blaser, D.E.; Howell, R.E.; Stuhler, H.; Stauver, J.

1987-11-01

Radioactive materials are being characterized for chemical form and certain physical properties with an x-ray diffraction goniometer customized for containment in a shielded alpha glovebox. A Siemens D500 goniometer was customized by Siemens to locate the associated electronics and x-ray generator outside the glovebox to minimize corrosion and facilitate maintenance. A graphite monochromator is used with a shielded scintillation detector to separate diffracted x-radiation from nuclear radiation. The diffraction system is computer automated for data acquisition and reduction. The facility is designed to handle primarily alpha- and beta-emitting samples with moderate neutron and gamma radiation. Samples containing plutonium, enriched uranium, and other transuranic elements are analyzed in support of site nuclear operations and development programs on nuclear waste, chemical separations, reactor fuels, and product forms

Development of assessment methods for transport and storage containers with higher content of metallic recycling material

International Nuclear Information System (INIS)

Zencker, U.; Qiao Linan; Droste, B.

2004-01-01

The mechanical behaviour of transport and storage containers made of ductile cast iron melted with higher content of metallic recycling material from decommissioning and dismantling of nuclear installations is investigated. With drop tests of cubic container-like models, the influence of different real targets on the stresses in the cask body and the fracture behaviour is examined. A test stand foundation is suggested, which can be manufactured simply and improves the reproducibility of the test results strongly. The test objects are partially equipped with artificial cracklike defects. Dynamic fracture mechanics analyses of these defects were performed by means of finite element calculations to uncover safety margins. Numerous test results show depending on the requirements that containers for final disposal can be built by means of a ductile cast iron with fracture toughness more than half under the lower bound value for the licensed material qualities yet. The application limits of the material are determined also by the opportunities of the safety assessment methods. This project supports the application of brittle fracture safe transport and storage packages for radioactive materials as recommended in App. VI of the Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (IAEA No. TS-G-1.1)

Development of assessment methods for transport and storage containers with higher content of metallic recycling material

Energy Technology Data Exchange (ETDEWEB)

Zencker, U.; Qiao Linan; Droste, B. [Bundesanstalt fuer Materialforschung und -pruefung (BAM), Berlin (Germany)

2004-07-01

The mechanical behaviour of transport and storage containers made of ductile cast iron melted with higher content of metallic recycling material from decommissioning and dismantling of nuclear installations is investigated. With drop tests of cubic container-like models, the influence of different real targets on the stresses in the cask body and the fracture behaviour is examined. A test stand foundation is suggested, which can be manufactured simply and improves the reproducibility of the test results strongly. The test objects are partially equipped with artificial cracklike defects. Dynamic fracture mechanics analyses of these defects were performed by means of finite element calculations to uncover safety margins. Numerous test results show depending on the requirements that containers for final disposal can be built by means of a ductile cast iron with fracture toughness more than half under the lower bound value for the licensed material qualities yet. The application limits of the material are determined also by the opportunities of the safety assessment methods. This project supports the application of brittle fracture safe transport and storage packages for radioactive materials as recommended in App. VI of the Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material (IAEA No. TS-G-1.1).

Micromechanical Analysis of Crack Closure Mechanism for Intelligent Material Containing TiNi Fibers

Science.gov (United States)

Araki, Shigetoshi; Ono, Hiroyuki; Saito, Kenji

In our previous study, the micromechanical modeling of an intelligent material containing TiNi fibers was performed and the stress intensity factor KI at the tip of the crack in the material was expressed in terms of the magnitude of the shape memory shrinkage of the fibers and the thermal expansion strain in the material. In this study, the value of KI at the tip of the crack in the TiNi/epoxy material is calculated numerically by using analytical expressions obtained in our first report. As a result, we find that the KI value decreases with increasing shrink strain of the fibers, and this tendency agrees with that of the experimental result obtained by Shimamoto etal.(Trans. Jpn. Soc. Mech. Eng., Vol. 65, No. 634 (1999), pp. 1282-1286). Moreover, there exists an optimal value of the shrink strain of the fibers to make the KI value zero. The change in KI with temperature during the heating process from the reference temperature to the inverse austenitic finishing temperature of TiNi fiber is also consistent with the experimental result. These results can be explained by the changes in the shrink strain, the thermal expansion strain, and the elastic moduli of TiNi fiber with temperature. These results may be useful in designing intelligent materials containing TiNi fibers from the viewpoint of crack closure.

A formula for determination of swelling characteristics of buffer material containing bentonite and evaluation of self-sealing performance

International Nuclear Information System (INIS)

Komine, Hideo; Ogata, Nobuhide

1998-01-01

High level radioactive waste disposal facility is planned to construct in a rock board deeper than some hundreds meter at underground, it is necessary to develop a material to refill the gap between waste container and its circumferential rock board. For this material it is required to have high water sealing and swellability, and bentonite is expected to use such application. Production of soil buffer materials containing bentonite is difficult to obtain required dry density by solidifying due to in situ roll-pressing, so it is, at present, thought to be an effective method to carry a block produced in a factory to a pit for disposal to settle. When supposing to produce and settle such buffer material, a gap forming between the buffer material and circumferential rock board or waste container has a large possibility to make a water path when remaining without treating it. Therefore, the buffer material is required to have a function to fill gap portion by swelling deformation and to proof water. In this study, in order to evaluated self-sealing of bentonite, due to a theoretical examination and a laboratory experiment on swelling behavior of soil materials containing bentonite, a swelling evaluation equation of the buffer materials was proposed. And, an application example for outlined design of an actual high level radioactive waste disposal facility was introduced. (G.K.)

Nupack, the new ASME code for radioactive material transportation packaging containments

International Nuclear Information System (INIS)

Turula, P.

1998-01-01

The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as Nupack, has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used for the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper

Luminescent materials based on Tb, Eu-containing layered double hydroxides

International Nuclear Information System (INIS)

Zhuravleva, N.G.; Eliseev, A.A.; Lukashin, A.V.; Kinast, U.; Tret'yakov, Yu.D.

2004-01-01

Luminescent materials on the basis of magnesium-aluminium layered double hydroxides with intercalated anionic complexes of terbium and europium picolinates were synthesized. Relying on data of spectroscopy, elementary and X-ray phase analyses, the change in the rare earth complex structure and metal/ligand ratio, depending on the hydroxide layer charge, determined by Mg/Al ratio in the double hydroxide, were ascertained. The values of quantum yields of luminescence for terbium-containing samples amounted to 30-50% [ru

TEMPERATURE PREDICTION IN 3013 CONTAINERS IN K AREA MATERIAL STORAGE (KAMS) FACILITY USING REGRESSION METHODS

International Nuclear Information System (INIS)

Gupta, N

2008-01-01

3013 containers are designed in accordance with the DOE-STD-3013-2004. These containers are qualified to store plutonium (Pu) bearing materials such as PuO2 for 50 years. DOT shipping packages such as the 9975 are used to store the 3013 containers in the K-Area Material Storage (KAMS) facility at Savannah River Site (SRS). DOE-STD-3013-2004 requires that a comprehensive surveillance program be set up to ensure that the 3013 container design parameters are not violated during the long term storage. To ensure structural integrity of the 3013 containers, thermal analyses using finite element models were performed to predict the contents and component temperatures for different but well defined parameters such as storage ambient temperature, PuO 2 density, fill heights, weights, and thermal loading. Interpolation is normally used to calculate temperatures if the actual parameter values are different from the analyzed values. A statistical analysis technique using regression methods is proposed to develop simple polynomial relations to predict temperatures for the actual parameter values found in the containers. The analysis shows that regression analysis is a powerful tool to develop simple relations to assess component temperatures

Modeling property evolution of container materials used in nuclear waste storage

Science.gov (United States)

Li, Dongsheng; Garmestani, Hamid; Khaleel, Moe; Sun, Xin

2010-03-01

Container materials under irradiation for a long time will raise high energy in the structure to generate critical structural damage. This study investigated what kind of mesoscale microstructure will be more resistant to radiation damage. Mechanical properties evolution during irradiation was modeled using statistical continuum mechanics. Preliminary results also showed how to achieve the desired microstructure with higher resistance to radiation.

Polymers Containing Diphenylvinyl-Substituted Indole Rings as Charge-Transporting Materials for OLEDs

Science.gov (United States)

Grigalevicius, S.; Zostautiene, R.; Sipaviciute, D.; Stulpinaite, B.; Volyniuk, D.; Grazulevicius, J. V.; Liu, L.; Xie, Z.; Zhang, B.

2016-02-01

Monomers and polymers containing electronically isolated diphenylvinyl-substituted indole rings were synthesized and characterized by nuclear magnetic resonance (NMR) and mass spectroscopies as well as by gel permeation chromatography. The polymers represent amorphous materials with glass transition temperatures of 91-109°C and thermal decomposition starting above 307°C. Electron photoemission spectra of thin films of the synthesized polymers revealed ionization potentials of 5.54-5.58 eV. The synthesized polymers were tested as hole-transporting materials in simple electroluminescent organic light-emitting diode (OLED) devices with tris(quinolin-8-olato)aluminium (Alq3) as an emitter as well as an electron-transporting layer. A green OLED device containing a hole-transporting layer of poly[1-(2,3-epithiopropyl)-2-methyl-3-(2,2-diphenylvinyl)índole] exhibited the best overall performance with a driving voltage of 4.0 V, maximum photometric efficiency of 2.8 cd/A and maximum brightness of about 4200 cd/m2.

Analysis of wallboard containing a phase change material

Science.gov (United States)

Tomlinson, J. J.; Heberle, D. P.

Phase change materials (PCMs) used on the interior of buildings hold the promise for improved thermal performance by reducing the energy requirements for space conditioning and by improving thermal comfort by reducing temperature swings inside the building. Efforts are underway to develop a gypsum wallboard containing a hydrocarbon PCM. With a phase change temperature in the room temperature range, the PCM wallboard adds substantially to the thermal mass of the building while serving the same architectural function as conventional wallboard. To determine the thermal and economic performance of this PCM wallboard, the Transient Systems Simulation Program (TRNSYS) was modified to accommodate walls that are covered with PCM plasterboard, and to apportion the direct beam solar radiation to interior surfaces of a building. The modified code was used to simulate the performance of conventional and direct-gain passive solar residential-sized buildings with and without PCM wallboard. Space heating energy savings were determined as a function of PCM wallboard characteristics. Thermal comfort improvements in buildings containing the PCM were qualified in terms of energy savings. The report concludes with a present worth economic analysis of these energy savings and arrives at system costs and economic payback based on current costs of PCMs under study for the wallboard application.

Waste container and method for containing waste

International Nuclear Information System (INIS)

Ono, Akira; Matsushita, Mitsuhiro; Doi, Makoto; Nakatani, Seiichi.

1990-01-01

In a waste container, water-proof membranes and rare earth element layers are formed on the inner surface of a steel plate concrete container in which steel plates are embedded. Further, rear earth element detectors are disposed each from the inner side of the steel plate concrete container by way of a pressure pipe to the outer side of the container. As a method for actually containing wastes, when a plurality of vessels in which wastes are fixed are collectively enhoused to the waste container, cussioning materials are attached to the inner surface of the container and wastes fixing containers are stacked successively in a plurality of rows in a bag made of elastic materials. Subsequently, fixing materials are filled and tightly sealed in the waste container. When the waste container thus constituted is buried underground, even if it should be deformed to cause intrusion of rain water to the inside of the container, the rare earth elements in the container dissolved in the rain water can be detected by the detectors, the containers are exchanged before the rain water intruding to the inner side is leached to the surrounding ground, to previously prevent the leakage of radioactive nuclides. (K.M.)

Gamma motes for detection of radioactive materials in shipping containers

International Nuclear Information System (INIS)

Harold McHugh; William Quam; Stephan Weeks; Brendan Sever

2007-01-01

Shipping containers can be effectively monitored for radiological materials using gamma (and neutron) motes in distributed mesh networks. The mote platform is ideal for collecting data for integration into operational management systems required for efficiently and transparently monitoring international trade. Significant reductions in size and power requirements have been achieved for room-temperature cadmium zinc telluride (CZT) gamma detectors. Miniaturization of radio modules and microcontroller units are paving the way for low-power, deeply-embedded, wireless sensor distributed mesh networks

Tunable multichannel filter in photonic crystal heterostructure containing permeability-negative materials

International Nuclear Information System (INIS)

Hu Xiaoyong; Liu Zheng; Gong Qihuang

2008-01-01

A tunable multichannel filter is demonstrated theoretically based on a one-dimensional photonic crystal heterostructure containing permeability-negative material. The filtering properties of the photonic crystal filter, including the channel number and frequency, can be tuned by adjusting the structure parameters or by a pump laser. The angular response of the photonic crystal filter and the influences of the losses on the filtering properties are also analyzed

Tunable multichannel filter in photonic crystal heterostructure containing permeability-negative materials

Energy Technology Data Exchange (ETDEWEB)

Hu Xiaoyong [State Key Laboratory for Mesoscopic Physics, Department of Physics, Peking University, Beijing 100871 (China)], E-mail: xiaoyonghu@pku.edu.cn; Liu Zheng [State Key Laboratory for Mesoscopic Physics, Department of Physics, Peking University, Beijing 100871 (China); Gong Qihuang [State Key Laboratory for Mesoscopic Physics, Department of Physics, Peking University, Beijing 100871 (China)], E-mail: qhgong@pku.edu.cn

2008-01-14

A tunable multichannel filter is demonstrated theoretically based on a one-dimensional photonic crystal heterostructure containing permeability-negative material. The filtering properties of the photonic crystal filter, including the channel number and frequency, can be tuned by adjusting the structure parameters or by a pump laser. The angular response of the photonic crystal filter and the influences of the losses on the filtering properties are also analyzed.

Los Alamos National Laboratory new generation standard nuclear material storage container - the SAVY4000 design

International Nuclear Information System (INIS)

Stone, Timothy Amos

2010-01-01

Incidents involving release of nuclear materials stored in containers of convenience such as food pack cans, slip lid taped cans, paint cans, etc. has resulted in defense board concerns over the lack of prescriptive performance requirements for interim storage of nuclear materials. Los Alamos National Laboratory (LANL) has shared in these incidents and in response proactively moved into developing a performance based standard involving storage of nuclear material (RD003). This RD003 requirements document has sense been updated to reflect requirements as identified with recently issued DOE M 441.1-1 'Nuclear Material Packaging Manual'. The new packaging manual was issued at the encouragement of the Defense Nuclear Facilities Safety Board with a clear directive for protecting the worker from exposure due to loss of containment of stored materials. The Manual specifies a detailed and all inclusive approach to achieve a high level of protection; from package design and performance requirements, design life determinations of limited life components, authorized contents evaluations, and surveillance/maintenance to ensure in use package integrity over time. Materials in scope involve those stored outside an approved engineered-contamination barrier that would result in a worker exposure of in excess of 5 rem Committed Effective Does Equivalent (CEDE). Key aspects of meeting the challenge as developed around the SAVY-3000 vented storage container design will be discussed. Design performance and acceptance criteria against the manual, bounding conditions as established that the user must ensure are met to authorize contents in the package (based upon the activity of heat-source plutonium (90% Pu-238) oxide, which bounds the requirements for weapons-grade plutonium oxide), interface as a safety class system within the facility under the LANL plutonium facility DSA, design life determinations for limited life components, and a sense of design specific surveillance program

40 CFR 1612.3 - Published reports and material contained in the public incident investigation dockets.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 32 2010-07-01 2010-07-01 false Published reports and material... Published reports and material contained in the public incident investigation dockets. (a) Demands for published investigation reports should be directed to the Office of Congressional and Public Affairs, U.S...

An examination of source material requirements contained in 10 CFR Part 40

International Nuclear Information System (INIS)

Nussbaumer, D.; Smith, D.A.; Wiblin, C.

1992-10-01

This report identifies issues for consideration for rule-making to update the requirements for source material in 10 CFR Part 40 and examines options for resolving these issues. The contemplated rulemaking is intended to update 10 CFR Part 40 to reflect current radiation protection principles and regulatory practices. It is expected that such an update would make requirements for the control of source material more comparable to those pertaining to byproduct material contained in 10 CFR Part 30. The newer biological data and dose calculation methodology reflected in revised 10 CFR Part 20 will be used in analyses of potential regulatory amendments. This report presents historical background information and discussion on the various issues identified and makes preliminary recommendations concerning needed regulatory changes and approaches to rulemaking

Transport containers for radioactive material

International Nuclear Information System (INIS)

Bibby, D.

1978-01-01

A transport container for transporting irradiated nuclear fuel is described comprising a steel flask with detachable cover and having external heat exchange fins. The flask contains a solid annular shield comprised of discrete bodies of Pb or Fe bonded together by a solid matrix, for attenuating gamma rays and neutron emission. This may comprise lead shot bonded together by concrete or polyethylene, or alternatively iron shot bonded by concrete. (UK)

Comparative analysis of non-destructive methods to control fissile materials in large-size containers

Directory of Open Access Journals (Sweden)

Batyaev V.F.

2017-01-01

Full Text Available The analysis of various non-destructive methods to control fissile materials (FM in large-size containers filled with radioactive waste (RAW has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one.

ENGINEERING CONTROL PRACTICES FOR REDUCING EMISSIONS DURING DRILLING OF ASBESTOS-CONTAINING FLOORING MATERIALS

Science.gov (United States)

This report describes the implementation and testing of control measures to reduce airborne asbestos generated by the drilling of asbestos-containing flooring materials, an OSHA Class III asbestos maintenance activity. Bosch 11224 and 11222 rotary drills were fitted with shrouds ...

76 FR 41178 - Pesticides; Policies Concerning Products Containing Nanoscale Materials; Opportunity for Public...

Science.gov (United States)

2011-07-13

... Pesticides; Policies Concerning Products Containing Nanoscale Materials; Opportunity for Public Comment; Extension of Comment Period AGENCY: Environmental Protection Agency (EPA). ACTION: Proposed policy statement; extension of comment period. SUMMARY: EPA issued a proposed policy statement in the Federal Register of June...

Numerical Simulation on Dense Packing of Granular Materials by Container Oscillation

Directory of Open Access Journals (Sweden)

Jun Liu

2013-01-01

Full Text Available The packing of granular materials is a basic and important problem in geomechanics. An approach, which generates dense packing of spheres confined in cylindrical and cuboidal containers in three steps, is introduced in this work. A loose packing structure is first generated by means of a reference lattice method. Then a dense packing structure is obtained in a container by simulating dropping of particles under gravitational forces. Furthermore, a scheme that makes the bottom boundary fluctuate up and down was applied to obtain more denser packing. The discrete element method (DEM was employed to simulate the interactions between particle-particle and particle-boundary during the particles' motions. Finally, two cases were presented to indicate the validity of the method proposed in this work.

A method to calculate flux distribution in reactor systems containing materials with grain structure

International Nuclear Information System (INIS)

Stepanek, J.

1980-01-01

A method is proposed to compute the neutron flux spatial distribution in slab, spherical or cylindrical systems containing zones with close grain structure of material. Several different types of equally distributed particles embedded in the matrix material are allowed in one or more zones. The multi-energy group structure of the flux is considered. The collision probability method is used to compute the fluxes in the grains and in an ''effective'' part of the matrix material. Then the overall structure of the flux distribution in the zones with homogenized materials is determined using the DPN ''surface flux'' method. Both computations are connected using the balance equation during the outer iterations. The proposed method is written in the code SURCU-DH. Two testcases are computed and discussed. One testcase is the computation of the eigenvalue in simplified slab geometry of an LWR container of one zone with boral grains equally distributed in an aluminium matrix. The second is the computation of the eigenvalue in spherical geometry of the HTR pebble-bed cell with spherical particles embedded in a graphite matrix. The results are compared to those obtained by repeated use of the WIMS Code. (author)

Reactor container

International Nuclear Information System (INIS)

Ichiki, Tadaharu; Saba, Kazuhisa.

1979-01-01

Purpose: To improve the earthquake resistance as well as reduce the size of a container for a nuclear reactor with no adverse effects on the decrease of impact shock to the container and shortening of construction step. Constitution: Reinforcing profile steel materials are welded longitudinally and transversely to the inner surface of a container, and inner steel plates are secured to the above profile steel materials while keeping a gap between the materials and the container. Reactor shielding wall planted to the base concrete of the container is mounted to the pressure vessel, and main steam pipeways secured by the transverse beams and led to the outside of container is connected. This can improve the rigidity earthquake strength and the safetiness against the increase in the inside pressure upon failures of the container. (Yoshino, Y.)

Evaluation of iron-base materials for waste package containers in a salt repository

International Nuclear Information System (INIS)

Westerman, R.E.; Nelson, J.L.; Kuhn, W.L.; Basham, S.G.; Moak, D.A.; Pitman, S.G.

1983-11-01

Design studies for high-level nuclear waste packages for salt repositories have identified low-carbon steel as a candidate material for containers. Among the requirements are strength, corrosion resistance, and fabricability. The studies of the corrosion resistance and structural stability of iron-base materials (particularly low-carbon steel) are treated in this paper. The materials have been exposed in brines that are characteristic of the potential sites for salt repositories. The effects of temperature, radiation level, oxygen level and other parameters are under investigation. The initial development of corrosion models for these environments is presented with discussion of the key mechanisms under consideration. 6 references, 5 figures

Preliminary analysis of the creep behaviour of nuclear fuel-waste container materials

International Nuclear Information System (INIS)

Dutton, R.; Leitch, B.W.; Crosthwaite, J.L.; Kasprick, G.R.

1996-12-01

In the Canadian Nuclear Fuel Waste Management Program, it is proposed that nuclear fuel waste be placed in a durable container and disposed of in a deep underground vault. Consideration of various disposal-container designs has identified either titanium or copper as the material suitable for constructing the container shell. As part of the R and D program to examine the structural integrity of the container, creep tests are being conducted on commercially pure titanium and oxygen-free copper. This report presents the preliminary data obtained. It also describes the evaluation of various constitutive equations to represent the creep curves, thus providing the basis for extrapolation of the creep behaviour over the design lifetime of the container. In this regard, a specific focus is placed on equations derived from the 0-Projection Concept. Recognizing that the container lifetime will be determined by the onset of tertiary creep leading to creep rupture, we present the results of the metallographic examination of creep damage. This shows that the tertiary stage in titanium is associated with the formation of transgranular cavities within the region of localized necking of the creep specimens. In contrast, creep damage in copper is in the form of intergranular cavities uniformly distributed throughout the gauge length. These results are analyzed within the context of the extant literature, and their implications for future container design are discussed. (author)

41 CFR 101-42.1101 - Federal supply classification (FSC) groups and classes which contain hazardous materials.

Science.gov (United States)

2010-07-01

..., building paper, and thermal insulation materials Asbestos cloth which has loose fibers or particles that... shipment. 5965 Headsets, handsets, microphones, and speakers Items containing magnetic material. 5970... hazardous chemicals, solvents. 6625 Electrical and electronic properties measuring and testing instruments...

Nupack, the new Asme code for radioactive material transportation packaging containments

International Nuclear Information System (INIS)

Turula, P.

1998-01-01

The American Society of Mechanical Engineers (ASME) has added a new division to the nuclear construction section of its Boiler and Pressure Vessel Code (B and PVC). This Division, commonly referred to as 'Nupack', has been written to provide a consistent set of technical requirements for containment vessels of transportation packagings for high-level radioactive materials. This paper provides an introduction to Nupack, discusses some of its technical provisions, and describes how it can be used the design and construction of packaging components. Nupack's general provisions and design requirements are emphasized, while treatment of materials, fabrication and inspection is left for another paper. Participation in the Nupack development work described in this paper was supported by the U.S. Department of Energy. (authors)

Evaluation of calculational and material models for concrete containment structures

International Nuclear Information System (INIS)

Dunham, R.S.; Rashid, Y.R.; Yuan, K.A.

1984-01-01

A computer code utilizing an appropriate finite element, material and constitutive model has been under development as a part of a comprehensive effort by the Electric Power Research Institute (EPRI) to develop and validate a realistic methodology for the ultimate load analysis of concrete containment structures. A preliminary evaluation of the reinforced and prestressed concrete modeling capabilities recently implemented in the ABAQUS-EPGEN code has been completed. This effort focuses on using a state-of-the-art calculational model to predict the behavior of large-scale reinforced concrete slabs tested under uniaxial and biaxial tension to simulate the wall of a typical concrete containment structure under internal pressure. This paper gives comparisons between calculations and experimental measurements for a uniaxially-loaded specimen. The calculated strains compare well with the measured strains in the reinforcing steel; however, the calculations gave diffused cracking patterns that do not agree with the discrete cracking observed in the experiments. Recommendations for improvement of the calculational models are given. (orig.)

Comparative analysis of non-destructive methods to control fissile materials in large-size containers

Science.gov (United States)

Batyaev, V. F.; Sklyarov, S. V.

2017-09-01

The analysis of various non-destructive methods to control fissile materials (FM) in large-size containers filled with radioactive waste (RAW) has been carried out. The difficulty of applying passive gamma-neutron monitoring FM in large containers filled with concreted RAW is shown. Selection of an active non-destructive assay technique depends on the container contents; and in case of a concrete or iron matrix with very low activity and low activity RAW the neutron radiation method appears to be more preferable as compared with the photonuclear one. Note to the reader: the pdf file has been changed on September 22, 2017.

Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

International Nuclear Information System (INIS)

Kuznetsov, Andrey; Evsenin, Alexey; Osetrov, Oleg; Vakhtin, Dmitry; Gorshkov, Igor

2009-01-01

Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types--based on BGO, NaI and LaBr 3 crystals is presented.

Device for Detection of Explosives, Nuclear and Other Hazardous Materials in Luggage and Cargo Containers

Science.gov (United States)

Kuznetsov, Andrey; Evsenin, Alexey; Gorshkov, Igor; Osetrov, Oleg; Vakhtin, Dmitry

2009-12-01

Device for detection of explosives, radioactive and heavily shielded nuclear materials in luggage and cargo containers based on Nanosecond Neutron Analysis/Associated Particles Technique (NNA/APT) is under construction. Detection module consists of a small neutron generator with built-in position-sensitive detector of associated alpha-particles, and several scintillator-based gamma-ray detectors. Explosives and other hazardous chemicals are detected by analyzing secondary high-energy gamma-rays from reactions of fast neutrons with materials inside a container. The same gamma-ray detectors are used to detect unshielded radioactive and nuclear materials. An array of several neutron detectors is used to detect fast neutrons from induced fission of nuclear materials. Coincidence and timing analysis allows one to discriminate between fission neutrons and scattered probing neutrons. Mathematical modeling by MCNP5 and MCNP-PoliMi codes was used to estimate the sensitivity of the device and its optimal configuration. Comparison of the features of three gamma detector types—based on BGO, NaI and LaBr3 crystals is presented.

Hydrolysis of cellulose-containing materials by cellulase of the Trichoderma lignorum OM 534 fungus

Energy Technology Data Exchange (ETDEWEB)

Romanov, S L; Lobanok, A G

1977-01-01

Of the cellulose containing materials, hydrocellulose was most easily degraded while lignocellulose was hardest to break down with cellulase from T. lignorum grown on lactose or cellulose. Grinding and heat treatment (at 200/sup 0/) of lignocellulose enhanced its enzymic degradability. Hydrolysis was highest by cellulase from lactose-cultured Trichoderma. The hydrolysis products contained glucose, galactose, xylose, and mannose. Filtrates from T. lignorum grown on a lignocellulose were enzymically active after purification.

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Bullen, D.B.; Gdowski, G.E.; Weiss, H.

1988-06-01

Three copper-based alloys, CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni), are being considered along with three austenitic candidates as possible materials for fabrication of containers for disposal of high-level radioactive waste. The waste will include spent fuel assemblies from reactors as well as high-level reprocessing wastes in borosilicate glass and will be sent to the prospective repository at Yucca Mountain, Nevada, for disposal. The containers must maintain mechanical integrity for 50 yr after emplacement to allow for retrieval of waste during the preclosure phase of repository operation. Containment is required to be substantially complete for up to 300 to 1000 yr. During the early period, the containers will be exposed to high temperatures and high gamma radiation fields from the decay of high-level waste. The final closure joint will be critical to the integrity of the containers. This volume surveys the available data on the metallurgy of the copper-based candidate alloys and the welding techniques employed to join these materials. The focus of this volume is on the methods applicable to remote-handling procedures in a hot-cell environment with limited possibility of postweld heat treatment. The three copper-based candidates are ranked on the basis of the various closure techniques. On the basis of considerations regarding welding, the following ranking is proposed for the copper-based alloys: CDA 715 (best) > CDA 102 > CDA 613 (worst). 49 refs., 15 figs., 1 tab

Dry containment of radioactive materials

International Nuclear Information System (INIS)

Williams, C.E.

1980-01-01

A cask for the dry containment of radioactive fuel elements is described. The cask has a cover which contains valved drain and purge passageways. These passageways are sealed by after purge cover seals which are clamped over them and to the outer surface of the cover. The cover seals are tested by providing them with a pair of concentric ring seal elements squeezed between the cover seal and the outer surface of the cover and by forcing a gas under pressure into the annular region between the seal element

Nuclear power plant containment metallic pressure boundary materials and plans for collecting and presenting their properties

International Nuclear Information System (INIS)

Oland, C.B.

1995-04-01

A program is being conducted at the Oak Ridge National Laboratory (ORNL to assist the Nuclear Regulatory Commission (NRC)) in their assessment of the effects of degradation (primarily corrosion) on the structural capacity and leaktight integrity of metal containments and steel liners of reinforced concrete structures in nuclear power plants. One of the program objectives is to characterize and quantify manifestations of corrosion on the properties of steels used to construct containment pressure boundary components. This report describes a plan for use in collecting and presenting data and information on ferrous alloys permitted for use in construction of pressure retaining components in concrete and metal containments. Discussions about various degradation mechanisms that could potentially affect the mechanical properties of these materials are also included. Conclusions and recommendations presented in this report will be used to guide the collection of data and information that will be used to prepare a material properties data base for containment steels

Glycerol Salicylate-based Pulp-Capping Material Containing Portland Cement.

Science.gov (United States)

Portella, Fernando Freitas; Collares, Fabrício Mezzomo; Santos, Paula Dapper; Sartori, Cláudia; Wegner, Everton; Leitune, Vicente Castelo Branco; Samuel, Susana Maria Werner

2015-01-01

The purpose of this study was to evaluate the water sorption, solubility, pH and ability to diffuse into dentin of a glycerol salicylate-based, pulp-capping cement in comparison to a conventional calcium hydroxide-based pulp capping material (Hydcal). An experimental cement was developed containing 60% glycerol salicylate resin, 10% methyl salicylate, 25% calcium hydroxide and 5% Portland cement. Water sorption and solubility were determined based on mass changes in the samples before and after the immersion in distilled water for 7 days. Material discs were stored in distilled water for 24 h, 7 days and 28 days, and a digital pHmeter was used to measure the pH of water. The cement's ability to diffuse into bovine dentin was assessed by Raman spectroscopy. The glycerol salicylate-based cement presented higher water sorption and lower solubility than Hydcal. The pH of water used to store the samples increased for both cements, reaching 12.59 ± 0.06 and 12.54 ± 0.05 after 7 days, for Hydcal and glycerol salicylate-based cements, respectively. Both cements were able to turn alkaline the medium at 24 h and sustain its alkalinity after 28 days. Hydcal exhibited an intense diffusion into dentin up to 40 µm deep, and the glycerol salicylate-based cement penetrated 20 µm. The experimental glycerol salicylate-based cement presents good sorption, solubility, ability to alkalize the surrounding tissues and diffusion into dentin to be used as pulp capping material.

21 CFR 1308.35 - Exemption of certain cannabis plant material, and products made therefrom, that contain...

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 9 2010-04-01 2010-04-01 false Exemption of certain cannabis plant material, and... ENFORCEMENT ADMINISTRATION, DEPARTMENT OF JUSTICE SCHEDULES OF CONTROLLED SUBSTANCES Exempt Cannabis Plant... cannabis plant material, and products made therefrom, that contain tetrahydrocannabinols. (a) Any processed...

Anthropogenic materials and products containing natural radionuclides. Pt. 1. Survey of the major exposure pathways

International Nuclear Information System (INIS)

Becker, D.E.; Reichelt, A.

1991-06-01

Knowledge of the possible exposure pathways permits to perform an overall assessment of the radiation doses and qualities affecting the population, as well as their inter-relations: A catalogue was established of products, raw materials and waste materials containing natural radioactivity that are processed, produced or dumped in Bavaria and that contribute above negligible level to the radiation exposure of the population and to occupational radiation doses. A literature study rounds up the information on anthropogenic sources containing natural radioactivity and thus representing a radiation source generally to be considered for assessments. Some of these sources are discussed in more detail, indicating their radiological significance for the population and the environment in Bavaria. (Orig./DG) [de

Feasibility assessment of copper-base waste package container materials in a tuff repository

International Nuclear Information System (INIS)

Acton, C.F.; McCright, R.D.

1986-01-01

This report discussed progress made during the second year of a two-year study on the feasibility of using copper or a copper-base alloy as a container material for a waste package in a potential repository in tuff rock at the Yucca Mountain site in Nevada. Corrosion testing in potentially corrosive irradiated environments received emphasis during the feasibility study. Results of experiments to evaluate the effect of a radiation field on the uniform corrosion rate of the copper-base materials in repository-relevant aqueous environments are given as well as results of an electrochemical study of the copper-base materials in normal and concentrated J-13 water. Results of tests on the irradiation of J-13 water and on the subsequent formation of hydrogen peroxide are given. A theoretical study was initiated to predict the long-term corrosion behavior of copper in the repository. Tests were conducted to determine whether copper would adversely affect release rates of radionuclides to the environment because of degradation of the Zircaloy cladding. A manufacturing survey to determine the feasibility of producing copper containers utilizing existing equipment and processes was completed. The cost and availability of copper was also evaluated and predicted to the year 2000. Results of this feasibility assessment are summarized

Processing method and processing device for liquid waste containing surface active agent and radioactive material

International Nuclear Information System (INIS)

Nishi, Takashi; Matsuda, Masami; Baba, Tsutomu; Yoshikawa, Ryozo; Yukita, Atsushi.

1998-01-01

Washing liquid wastes containing surface active agents and radioactive materials are sent to a deaerating vessel. Ozone is blown into the deaerating vessel. The washing liquid wastes dissolved with ozone are introduced to a UV ray irradiation vessel. UV rays are irradiated to the washing liquid wastes, and hydroxy radicals generated by photodecomposition of dissolved ozone oxidatively decompose surface active agents contained in the washing liquid wastes. The washing liquid wastes discharged from the UV ray irradiation vessel are sent to an activated carbon mixing vessel and mixed with powdery activated carbon. The surface active agents not decomposed in the UV ray irradiation vessel are adsorbed to the activated carbon. Then, the activated carbon and washing liquid wastes are separated by an activated carbon separating/drying device. Radioactive materials (iron oxide and the like) contained in the washing liquid wastes are mostly granular, and they are separated and removed from the washing liquid wastes in the activated carbon separating/drying device. (I.N.)

Method to prepare essentially organic waste liquids containing radioactive or toxic materials

International Nuclear Information System (INIS)

Baehr, W.; Drobnik, S.H.; Hild, W.; Kroebel, R.; Meyer, A.; Naumann, G.

1976-01-01

Waste solutions occuring in nuclear technology containing radioactive or toxic materials can be solidified by mixing with a polymerisable mixture with subsequent polymerization. An improvement of this method, especially for liquids in which the radioactive components are present as organic compounds is achieved by adding a mixture of at least one monomeric vinyl compound, at least one polyvinyl compound and appropriate catalysts and by polymerizing at temperatures between 15 and 150 0 C. Should the waste liquid contain mineral acid, this is first neutralized by the addition of CaO or MgO. In processing oils or soaps, the addition of swelling agent for polystyrol resins is advantageous. 16 examples illustrate the invention. (UWI) [de

Physicochemical and bioactive properties of innovative resin-based materials containing functional halloysite-nanotubes fillers.

Science.gov (United States)

Degrazia, Felipe Weidenbach; Leitune, Vicente Castelo Branco; Takimi, Antonio Shigueaki; Collares, Fabrício Mezzomo; Sauro, Salvatore

2016-09-01

This study aimed to assess the degree of conversion, microhardness, solvent degradation, contact angle, surface free energy and bioactivity (e.g., mineral precipitation) of experimental resin-based materials containing, pure or triclosan-encapsulated, aluminosilicate-(halloysite) nanotubes. An experimental resin blend was prepared using bis-GMA/TEGDMA, 75/25wt% (control). Halloysite nanotubes (HNT) doped with or without triclosan (TCN) were first analyzed using transmission electron microscopy (TEM). HNT or HNT/TCN fillers were incorporated into the resin blend at different concentrations (5, 10, and 20wt%). Seven experimental resins were created and the degree of conversion, microhardness, solvent degradation and contact angle were assessed. Bioactive mineral precipitation induced by the experimental resins was evaluated through Raman spectroscopy and SEM-EDX. TEM showed a clear presence of TCN particles inside the tubular lumen and along the outer surfaces of the halloysite nanotubes. The degree of conversion, surface free energy, microhardness, and mineral deposition of polymers increased with higher amount of HNTs. Conversely, the higher the amount (20wt%) of TCN-loaded HNTs the lower the microhardness of the experimental resins. The incorporation of pure or TCN-loaded aluminosilicate-(halloysite) nanotubes into resin-based materials increase the bioactivity of such experimental restorative materials and promotes mineral deposition. Therefore, innovative resin-based materials containing functional halloysite-nanotube fillers may represent a valuable alternative for therapeutic minimally invasive treatments. Copyright © 2016 The Academy of Dental Materials. Published by Elsevier Ltd. All rights reserved.

Estimates of fire environments in ship holds containing radioactive material packages

International Nuclear Information System (INIS)

Koski, J.A.; Cole, J.K.; Hohnstreiter, G.F.; Wix, S.D.

1995-01-01

Fire environments that occur on cargo ships differ significantly from the fire environments found in land transport. Cargo ships typically carry a large amount of flammable fuel for propulsion and shipboard power, and may transport large quantities of flammable cargo. As a result, sea mode transport accident records contain instances of long lasting and intense fires. Since Irradiated Nuclear Fuel (INF) casks are not carried on tankers with large flammable cargoes, most of these dramatic, long burning fires are not relevant threats, and transport studies must concentrate on those fires that are most likely to occur. By regulation, INF casks must be separated from flammable cargoes by a fire-resistant, liquid-tight partition. This makes a fire in an adjacent ship hold the most likely fire threat. The large size of a cargo ship relative to any spent nuclear fuel casks on board, however, may permit a severe, long lasting fire to occur with little or no thermal impact on the casks. Although some flammable materials such as shipping boxes or container floors may exist in the same hold with the cask, the amount of fuel available may not provide a significant threat to the massive transport casks used for radioactive materials. This shipboard fire situation differs significantly from the regulatory conditions specified in 10 CFR 71 for a fully engulfing pool fire. To learn more about the differences, a series of simple thermal analyses has been completed to estimate cask behavior in likely marine and land thermal accident situations. While the calculations are based on several conservative assumptions, and are only preliminary, they illustrate that casks are likely to heat much more slowly in shipboard hold fires than in an open pool fire. The calculations also reinforce the basic regulatory concept that for radioactive materials, the shipping cask, not the ship, is the primary protection barrier to consider

Absorbent material for type a radioactive materials packaging containing liquids

International Nuclear Information System (INIS)

Saunders, G.A.

1989-11-01

The application of absorbent materials to the packaging and transport of liquid radioactive materials in Type A packages has not been reported in the literature. However, a significant body of research exists on absorbent materials for personal hygiene products such as diapers. Absorption capacity is dependent on both the absorbent material and the liquid being absorbed. Theoretical principles for capillary absorption in both the horizontal and the vertical plane indicate that small contact angle between the absorbent fibre and the liquid, and a small inter-fibre pore size are important. Some fluid parameters such as viscosity affect the rate of absorption but not the final absorption capacity. There appears to be little comparability between results obtained for the same absorbent and fluid using different test procedures. Test samples of materials from several classes of potential absorbents have been evaluated in this study, and shown to have a wide range of absorbent capacities. Foams, natural fibres, artificial fibres and granular materials are all potentially useful absorbents, with capacities ranging from as little as 0.86 to as much as 40.6 grams of distilled water per gram of absorbent. Two experimental procedures for evaluating the absorbent capacity of these materials have been detailed in this report, and found suitable for evaluating granular, fibrous or foam materials. Compression of the absorbent material reduces its capacity, but parameters such as relative humidity, pH, temperature, and viscosity appear to have little significant influence on capacity. When the materials were loaded to 50% of their one-minute absorbency, subsequent loss of the absorbed liquid was generally minimal. All of the absorbent materials rapidly lost their absorbed water through evaporation within twenty-four hours in still air at 21 degrees C and 50% relative humidity

Device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor

International Nuclear Information System (INIS)

Jaeger, W.

1984-01-01

This invention concerns a device for manufacturing methane or synthetic gas from materials containing carbon using a nuclear reactor, where part of the carbon is gasified with hydration and the remaining carbon is converted to synthetic gas by adding steam. This synthetic gas consists mainly of H 2 , CO, CO 2 and CH 4 and can be converted to methane in so-called methanising using a nickel catalyst. The hydrogen gasifier is situated in the first of two helium circuits of a high temperature reactor, and the splitting furnace is situated in the second helium circuit, where part of the methane produced is split into hydrogen at high temperature, which is used for the hydrating splitting of another part of the material containing carbon. (orig./RB) [de

Anthropogenic materials and products containing natural radionuclides. Pt. 2. Examination of radiation doses resulting from occupational exposure

International Nuclear Information System (INIS)

Reichelt, A.; Lehmann, K.H.

1993-11-01

The radiation doses are determined on the basis of dosimetric scanning of the materials and products and measurement of the ambient dose rates and inhaled doses at the place of work. For all places and conditions exmined, the average annual effective dose (ICRP) is of the order of 20mSv/annum. The substances and products examined are phosphate fertilizers. thoriated tungsten electrodes, or glass gas hoods, respectively, dental material containing uranium, and dental ceramics containing zirconium sands. The report also gives information on the occupational exposure in drinking-water conditioning plants. (Orig./DG) [de

Research and development activities at INE concerning corrosion of final repository container materials

International Nuclear Information System (INIS)

Kienzler, Bernhard

2017-01-01

The present work provides a historical overview of the research and development activities carried out at the (Nuclear) Research Center Karlsruhe (today KIT) since the beginning of the 1980s on the corrosion of materials which might be suitable for construction of containers for highly radioactive wastes. The report relates almost exclusively to the work performed by Dr. Emmanuel Smailos, who elaborated the corrosion of various materials at the Institute for Nuclear Waste Disposal (INE). The requirements for the containers and materials, which were subject to changes in time, are presented. The changes were strongly influenced by the changed perception of the use of nuclear energy. The selection of the materials under investigations, the boundary conditions for the corrosion experiments and the analytical methods are described. Results of the corrosion of the materials such as finegrained steel, Hastelloy C4, nodular cast iron, titanium-palladium and copper or copper-nickel alloys in typical salt solutions are summarized. The findings of special investigations, e.g. corrosion under irradiation or the influence of sulfide on the corrosion rates are shown. For construction of disposal canisters, experiments were conducted to determine the contact corrosion, the influence of the hydrogen embrittlement of Ti-Pd and fine-grained steels on the corrosion behavior as well as the corrosion behavior of welding and the influence of different welding processes with the resulting heat-affected zones on the corrosion behavior. The work was contributed to several European research programs and was well recognized in the USA. Investigations on the corrosion of steels in non-saline solutions and corrosion under interim storage conditions as well as under the expected conditions of the Konrad repository for low-level radioactive wastes are also described. In addition, the experiments on ceramic materials are presented and the results of the corrosion of Al 2 O 3 and ZrO 2 ceramics

Lining materials for waste disposal containment and waste storage facilities. (Latest citations from the NTIS bibliographic database). Published Search

International Nuclear Information System (INIS)

1993-11-01

The bibliography contains citations concerning the design characteristics, performance, and materials used to make liners for the waste disposal and storage industry. Liners made of concrete, polymeric materials, compacted clays, asphalt, and in-situ glass are discussed. The use of these liners to contain municipal wastes, hazardous waste liquids, and both low-level and high-level radioactive wastes is presented. Liner permeability, transport, stability, construction, and design are studied. Laboratory field measurements for specific wastes are included. (Contains a minimum of 213 citations and includes a subject term index and title list.)

Materials designed for containments

International Nuclear Information System (INIS)

Piehl, K.H.

1976-01-01

The present article points out that high-tensile fine-grained steels have been used successfully in the construction of reactor containments, spherical gasometers, and pressure vessels. It has been confirmed that their use requires safety measures concerning lay out and production. Viscosity properties of high-tensile, fine-grained steels can be improved significantly by means of electroslag remelting. The extent to which this improvement influences the heat-affected zone is being examined. (orig./RW) [de

Program to develop analytical tools for environmental and safety assessment of nuclear material shipping container systems

International Nuclear Information System (INIS)

Butler, T.A.

1978-11-01

This paper describes a program for developing analytical techniques to evaluate the response of nuclear material shipping containers to severe accidents. Both lumped-mass and finite element techniques are employed to predict shipping container and shipping container-carrier response to impact. The general impact problem is computationally expensive because of its nonlinear, three-dimensional nature. This expense is minimized by using approximate models to parametrically identify critical cases before more exact analyses are performed. The computer codes developed for solving the problem are being experimentally substantiated with test data from full-scale and scale-model container drop tests. 6 figures, 1 table

Electro-active polymers containing pendent 2,7-diarylfluorene fragments as materials for OLEDs

Science.gov (United States)

Krucaite, G.; Tavgeniene, D.; Peciulyte, L.; Buika, G.; Liu, L.; Zhang, B.; Xie, Z.; Grigalevicius, S.

2016-05-01

Poly[2-phenyl-7-(4-vinylphenyl)-9,9-diethylfluorene)], poly[2-(1-naphtyl)-7-(4-vinylphenyl)-9,9-diethylfluorene)] and poly[2-(4-biphenyl)-7-(4-vinylphenyl)-9,9-diethylfluorene)] were synthesized and characterized by NMR spectroscopy, elemental analysis and gel permeation chromatography. The derivatives represent materials of high thermal stability with initial thermal destruction temperatures from 390°C to 400 °C. The glass transition temperatures of the amorphous materials were 182 °C, 151 °C and 159 °C respectively. Hole-transporting properties of the polymeric materials were tested in the structures of organic light emitting diodes with Alq3 as the green emitter and electron transporting material. The device containing hole-transporting layers of polymer with 2-(4-biphenyl)-7-(4-vinylphenyl)-9,9-diethylfluorene moieties exhibited the best overall performance with turn on voltage of 3.6 V, a maximum photometric efficiency of 3.1 cd/A and maximum brightness of about 5300 cd/m2.

Discrimination of high-Z materials in concrete-filled containers using muon scattering tomography

Science.gov (United States)

Frazão, L.; Velthuis, J.; Thomay, C.; Steer, C.

2016-07-01

An analysis method of identifying materials using muon scattering tomography is presented, which uses previous knowledge of the position of high-Z objects inside a container and distinguishes them from similar materials. In particular, simulations were performed in order to distinguish a block of Uranium from blocks of Lead and Tungsten of the same size, inside a concrete-filled drum. The results show that, knowing the shape and position from previous analysis, it is possible to distinguish 5 × 5 × 5 cm3 blocks of these materials with about 4h of muon exposure, down to 2 × 2 × 2 cm3 blocks with 70h of data using multivariate analysis (MVA). MVA uses several variables, but it does not benefit the discrimination over a simpler method using only the scatter angles. This indicates that the majority of discrimination is provided by the angular information. Momentum information is shown to provide no benefits in material discrimination.

Compatibility of Space Nuclear Power Plant Materials in an Inert He/Xe Working Gas Containing Reactive Impurities

International Nuclear Information System (INIS)

MM Hall

2006-01-01

A major materials selection and qualification issue identified in the Space Materials Plan is the potential for creating materials compatibility problems by combining dissimilar reactor core, Brayton Unit and other power conversion plant materials in a recirculating, inert He/Xe gas loop containing reactive impurity gases. Reported here are results of equilibrium thermochemical analyses that address the compatibility of space nuclear power plant (SNPP) materials in high temperature impure He gas environments. These studies provide early information regarding the constraints that exist for SNPP materials selection and provide guidance for establishing test objectives and environments for SNPP materials qualification testing

Packet D: Fuel containing materials

International Nuclear Information System (INIS)

Tokarevskij, V.V.

1999-01-01

The tasks of the project 'D' are: increase of nuclear safety by fuel containing mass (FCM) characterisation, and development of a preliminary plan for FCM management which should be accomplished by FCM extraction

Fluorine uptake into human enamel around a fluoride-containing dental material during cariogenic pH cycling

International Nuclear Information System (INIS)

Komatsu, H.; Yamamoto, H.; Nomachi, M.; Yasuda, K.; Matsuda, Y.; Murata, Y.; Kijimura, T.; Sano, H.; Sakai, T.; Kamiya, T.

2007-01-01

Using PIGE (TIARA, JAPAN) technique, we measured fluorine (F) uptake into the tooth enamel around a fluoride-containing material during caries progression using pH cycling. Class I cavities in the buccal surfaces of 6 extracted human teeth were drilled and filled with fluoride-containing material; a glass ionomer cement (Fuji IX(GC)). Three 300 μm sections through the material were obtained from each tooth. Two of these specimens were utilized to measure the F distribution in enamel adjacent to the material. A 1.7 MeV proton beam accelerated by the TIARA single-ended accelerator was delivered to a micro-beam apparatus. The beam spot size was about 1 μm with a beam current of about 100 pA. A nuclear reaction, 19 F(p, αγ) 16 0, was used to measure the F concentration and the gamma-rays from this reaction were detected with a 4' NaI detector. X-rays induced by proton were detected with a Si(Li) detector to measure calcium concentration and the beam intensity was monitored with the X-ray yield from a copper foil for quantitative analysis. After measurement of F uptake, all specimens were polished to a thickness of 120 μm. In order to simulate daily acid challenges occurring in the oral cavity, the pH cycling (pH6.8-pH4.5) was carried out for 1, 3 and 5 weeks, separately. The duration that the solution remained below pH 5.5 was 37 min per cycle. The cycles were repeated 6 times per day with 2 h interval between cycles, and the specimens were kept in remineralizing solution for the rest of pH cycle. After pH cycling, F and calcium distribution of each specimen was evaluated using PIGE technique. The F distribution of the specimens before pH cycling clearly showed the F uptake from fluoride-containing material into enamel adjacent to the material. After pH cycling, the caries progression in all specimens was observed by the image of transverse microradiography (TMR). The depth of caries and mineral loss progressed with increasing the duration of pH cycling, although

Fluorine uptake into human enamel around a fluoride-containing dental material during cariogenic pH cycling

Energy Technology Data Exchange (ETDEWEB)

Komatsu, H. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan)]. E-mail: kom@den.hokudai.ac.jp; Yamamoto, H. [Graduate School of Dentistry, Osaka University, 1-8 Yamada-Oka, Suita 565-0871 (Japan); Nomachi, M. [Graduate School of Science, Osaka University, 1-1 Machikaneyama, Toyonaka 560-0043 (Japan); Yasuda, K. [Wakasa wan Energy Research Center, 64-52-1 Hase, Tsuruga 914-0192 (Japan); Matsuda, Y. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan); Murata, Y. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan); Kijimura, T. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan); Sano, H. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan); Sakai, T. [Takasaki Advanced Radiation Research Institute, JAEA, 1233 Watanuki-machi, Takasaki 370-1292 (Japan); Kamiya, T. [Takasaki Advanced Radiation Research Institute, JAEA, 1233 Watanuki-machi, Takasaki 370-1292 (Japan)

2007-07-15

Using PIGE (TIARA, JAPAN) technique, we measured fluorine (F) uptake into the tooth enamel around a fluoride-containing material during caries progression using pH cycling. Class I cavities in the buccal surfaces of 6 extracted human teeth were drilled and filled with fluoride-containing material; a glass ionomer cement (Fuji IX(GC)). Three 300 {mu}m sections through the material were obtained from each tooth. Two of these specimens were utilized to measure the F distribution in enamel adjacent to the material. A 1.7 MeV proton beam accelerated by the TIARA single-ended accelerator was delivered to a micro-beam apparatus. The beam spot size was about 1 {mu}m with a beam current of about 100 pA. A nuclear reaction, {sup 19}F(p, {alpha}{gamma}){sup 16}0, was used to measure the F concentration and the gamma-rays from this reaction were detected with a 4' NaI detector. X-rays induced by proton were detected with a Si(Li) detector to measure calcium concentration and the beam intensity was monitored with the X-ray yield from a copper foil for quantitative analysis. After measurement of F uptake, all specimens were polished to a thickness of 120 {mu}m. In order to simulate daily acid challenges occurring in the oral cavity, the pH cycling (pH6.8-pH4.5) was carried out for 1, 3 and 5 weeks, separately. The duration that the solution remained below pH 5.5 was 37 min per cycle. The cycles were repeated 6 times per day with 2 h interval between cycles, and the specimens were kept in remineralizing solution for the rest of pH cycle. After pH cycling, F and calcium distribution of each specimen was evaluated using PIGE technique. The F distribution of the specimens before pH cycling clearly showed the F uptake from fluoride-containing material into enamel adjacent to the material. After pH cycling, the caries progression in all specimens was observed by the image of transverse microradiography (TMR). The depth of caries and mineral loss progressed with increasing the

The interaction of iodine with organic material in containment

Energy Technology Data Exchange (ETDEWEB)

Wren, J C; Ball, J M; Glown, G A; Portmann, R; Sanipelli, G G [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-01

Organic impurities in containment water, originating from various painted structural surfaces and organic containment materials, could have a significant impact on iodine volatility following an accident. A research program has been designed to determine the impact of organic impurities on iodine volatility under accident conditions. The program consists of experimental, literature and modelling studies on the radiolysis or organic compounds in the aqueous phase, thermal and radiolytic formation and decomposition of organic iodides, dissolution of organic solvents from various painted surfaces into the aqueous phase, and iodine deposition on painted surfaces. The experimental studies consist of bench-scale `separate effects` tests as well as intermediate-scale `integrated effects` in the Radioiodine Test facility. The studies have shown that organic impurities will be found in containment water, arising from the dissolution of organic compounds from various surface paints and that these compounds can potentially have a significant impact on iodine volatility following an accident. The main impact of surface paints will occur through aqueous-phase reactions of the organic compounds that they release to the aqueous phase. Under the radiation conditions expected during an accident, these compounds will react to reduce the pH and dissolved oxygen concentration, consequently increasing the formation of I{sub 2} from I{sup -} that is present in the sump. It appears that the rates of these processes may be controlled by the dissolution kinetics of the organic compounds from the surface coatings. Moreover, the organic compounds may also react thermally and radiolytically with I{sub 2} to form organic iodides in the aqueous phase. Our studies have shown that the formation of organic iodides from soluble organics such as ketones, alcohols and phenols may have more impact on the total iodine volatility than the formation of CH{sub 3}I. (author) 13 figs., 2 tabs., 19 refs.

The interaction of iodine with organic material in containment

International Nuclear Information System (INIS)

Wren, J.C.; Ball, J.M.; Glown, G.A.; Portmann, R.; Sanipelli, G.G.

1996-01-01

Organic impurities in containment water, originating from various painted structural surfaces and organic containment materials, could have a significant impact on iodine volatility following an accident. A research program at the Whiteshell Laboratories of AECL has been designed to determine the impact of organic impurities on iodine volatility under accident conditions. The program consists of experimental, literature and modelling studies on the radiolysis or organic compounds in the aqueous phase, thermal and radiolytic formation and decomposition of organic iodides, dissolution of organic solvents from various painted surfaces into the aqueous phase, and iodine deposition on painted surfaces. The experimental studies consist of bench-scale 'separate effects' tests as well as intermediate-scale 'integrated effects' in the Radioiodine Test facility. The studies have shown that organic impurities will be found in containment water, arising from the dissolution of organic compounds from various surface paints and that these compounds can potentially have a significant impact on iodine volatility following an accident. The main impact of surface paints will occur through aqueous-phase reactions of the organic compounds that they release to the aqueous phase. Under the radiation conditions expected during an accident, these compounds will react to reduce the pH and dissolved oxygen concentration, consequently increasing the formation of I 2 from I - that is present in the sump. It appears that the rates of these processes may be controlled by the dissolution kinetics of the organic compounds from the surface coatings. Moreover, the organic compounds may also react thermally and radiolytically with I 2 to form organic iodides in the aqueous phase. Our studies have shown that the formation of organic iodides from soluble organics such as ketones, alcohols and phenols may have more impact on the total iodine volatility than the formation of CH 3 I. (author) 13 figs., 2

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Farmer, J.C.; McCright, R.D.; Kass, J.N.

1988-06-01

Three iron- to nickel-based austenitic alloys and three copper-based alloys are being considered as candidate materials for the fabrication of high-level radioactive-waste disposal containers. The austenitic alloys are Types 304L and 316L stainless steels and the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper), CDA 613 (Cu-7Al), and CDA 715 (Cu-30Ni). Waste in the forms of both spent fuel assemblies from reactors and borosilicate glass will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking; and transgranular stress corrosion cracking. Problems specific to welds, such as hot cracking, may also occur. A survey of the literature has been prepared as part of the process of selecting, from among the candidates, a material that is adequate for repository conditions. The modes of degradation are discussed in detail in the survey to determine which apply to the candidate alloys and the extent to which they may actually occur. The eight volumes of the survey are summarized in Sections 1 through 8 of this overview. The conclusions drawn from the survey are also given in this overview

Oxidation of a Silica-Containing Material in a Mach 0.3 Burner Rig

Science.gov (United States)

Nguyen, QuynhGiao N.; Cuy, Michael D.; Gray, Hugh R. (Technical Monitor)

2002-01-01

A primarily silica-containing material with traces of organic compounds, as well as aluminum and calcium additions, was exposed to a Mach 0.3 burner rig at atmospheric pressure using jet fuel. The sample was exposed for 5 continuous hours at 1370 C. Post exposure x-ray diffraction analyses indicate formation of cristobalite, quartz, NiO and Spinel (Al(Ni)CR2O4). The rig hardware is composed of a nickel-based superalloy with traces of Fe. These elements are indicated in the energy dispersive spectroscopy (EDS) results. This material was studied as a candidate for high temperature applications under an engine technology program.

Steam reforming as an alternative technique for treatment of oil sludge containing naturally occurring radioactive material

International Nuclear Information System (INIS)

Norasalwa Zakaria; Muhd Noor Muhd Yunus; Mohd Khairi Muhd Said; Mohamad Azman Che Mat Isa; Mohd Puad Abu

2004-01-01

Steam reforming treatment system is an innovative technology that holds a potential to treat mixed waste containing radioactive material. The system is utilizing the thermal heat of the superheated steam at 500 degree C to produce combustible gases and integrates it with ash melting at 1400 degree C for final destruction. In this system, liquids are evaporated, organics are converted into a hydrogen-rich gas, chlorinated compounds are converted in hydrochloric acid, and reactive chemicals in the waste containing radionuclide and heavy metals are converted into the stable product through ash melting dioxins and furans are not formed, but instead are destroyed in the reducing environment of the system. No secondary pollutants are produced from the system that requires subsequent treatment. The system is divided into three development stages, and currently the project is progressing at development stage 1. This project is an entailment of a concentrated effort to solve oil sludge containing radioactive material treatment issue. (Author)

Low temperature rheological properties of asphalt mixtures containing different recycled asphalt materials

Directory of Open Access Journals (Sweden)

Ki Hoon Moon

2017-01-01

Full Text Available Reclaimed Asphalt Pavement (RAP and Recycled Asphalt Shingles (RAS are valuable materials commonly reused in asphalt mixtures due to their economic and environmental benefits. However, the aged binder contained in these materials may negatively affect the low temperature performance of asphalt mixtures. In this paper, the effect of RAP and RAS on low temperature properties of asphalt mixtures is investigated through Bending Beam Rheometer (BBR tests and rheological modeling. First, a set of fourteen asphalt mixtures containing RAP and RAS is prepared and creep stiffness and m-value are experimentally measured. Then, thermal stress is calculated and graphically and statistically compared. The Huet model and the Shift-Homothety-Shift in time-Shift (SHStS transformation, developed at the École Nationale des Travaux Publics de l'État (ENTPE, are used to back calculate the asphalt binder creep stiffness from mixture experimental data. Finally, the model predictions are compared to the creep stiffness of the asphalt binders extracted from each mixture, and the results are analyzed and discussed. It is found that an addition of RAP and RAS beyond 15% and 3%, respectively, significantly change the low temperature properties of asphalt mixture. Differences between back-calculated results and experimental data suggest that blending between new and old binder occurs only partially. Based on the recent finding on diffusion studies, this effect may be associated to mixing and blending processes, to the effective contact between virgin and recycled materials and to the variation of the total virgin-recycled thickness of the binder film which may significantly influence the diffusion process. Keywords: Reclaimed Asphalt Pavement (RAP, Recycled Asphalt Shingles (RAS, Thermal stress, Statistical comparison, Back-calculation, Binder blending

Container material and design considerations for storage of low-level radioactive waste

International Nuclear Information System (INIS)

Temus, C.J.

1987-01-01

With the threat of increased burial site restrictions and increased surcharges; the ease with which waste is sent to the burial site has been reduced. For many generators of waste the only alternative after maximizing volume reduction efforts is to store the waste. Even after working through the difficult decision of deciding what type of storage facility to have, the decision of what type of container to store the waste in has to still be made. This paper explores the many parameters that affect not only the material selection but also the design. The proper selection of materials affect the ability of the container to survive the storage period. The material selection also directly affects the design and utilization of the storage facility. The impacts to the facility include the functional aspects as well as its operational cost and liability as related to such things as fire insurance and active environmental control systems. The advantages and disadvantages of many of the common systems such as carbon steel, various coatings, polyethylene, stainless steel, composites and concrete will be discussed and evaluated. Recognizing that the waste is to be disposed of in the future differentiates it from waste that is shipped directly to the disposal site. The stored waste has to have the capability to be handled not only once like the disposal site waste but potentially several times before ultimate disposal. This handling may be by several different systems both at the storage facility and the burial site. Some of these systems due to ALARA considerations are usually remote requiring various interfaces, while not interfering with handling, transportation or disposal operations

Fluorine uptake into the human enamel surface from fluoride-containing sealing materials during cariogenic pH cycling

Science.gov (United States)

Yasuhiro, Matsuda; Katsushi, Okuyama; Hiroko, Yamamoto; Hisanori, Komatsu; Masashi, Koka; Takahiro, Sato; Naoki, Hashimoto; Saiko, Oki; Chiharu, Kawamoto; Hidehiko, Sano

2015-04-01

To prevent the formation of caries and reduce dentin hypersensitivity, sealing materials, either with or without fluoride, are generally applied on the tooth in clinical practice. Application of fluoride-free sealing materials results in the formation of an acid-resistant layer on the tooth surface. On the other hand, fluoride-containing sealing materials might not only form an acid-resistant layer but could possibly further provide fluoride to enhance remineralization and reduce demineralization. In this study, the demineralization prevention ability and fluorine uptake rate in human enamel of fluoride-containing sealing materials ["MS coats F" (MSF)] and fluoride-free sealing materials ("hybrid coats 2" [HI]) were evaluated using an automatic pH cycling system. Each material was applied to the original tooth surface, the cut surfaces were covered with sticky wax, and the automatic pH-cycling system simulated daily acid changes (pH 6.8-4.5) occurring in the oral cavity for 4 weeks. Caries progression was analyzed using transverse microradiography (TMR) taken pre and post the 4 weeks of pH cycling. The fluorine and calcium distributions in the carious lesion in each specimen were evaluated using the proton-induced gamma emission (PIGE) and proton-induced X-ray (PIXE) techniques, respectively. TMR analysis showed that both MSF and HI had a caries-preventing effect after 4 weeks of pH cycling. PIGE/PIXE analysis demonstrated that only MSF resulted in fluoride uptake in the enamel surface. Therefore, MSF can help to form an acid-resistant layer and provide fluoride to the enamel surface. The presence of fluoride on the enamel surface suggested that MSF could prevent demineralization, even if the acid-resistant layer was removed, in clinical settings. The data obtained using the PIGE and PIXE techniques are useful for understanding the benefits of the use of a fluoride-containing sealing material for preventing caries.

Fluoride release and recharge abilities of contemporary fluoride-containing restorative materials and dental adhesives.

Science.gov (United States)

Dionysopoulos, Dimitrios; Koliniotou-Koumpia, Eugenia; Helvatzoglou-Antoniades, Maria; Kotsanos, Nikolaos

2013-01-01

The aim of this study was to evaluate the fluoride release of five fluoride-releasing restorative materials and three dental adhesives, before and after NaF solution treatment. Five restorative materials (Fuji IX GP, GC Corp.; Ketac N100, 3M ESPE; Dyract Extra, Dentsply; Beautifil II, Shofu Inc.; Wave, SDI) and three dental adhesives (Stae, SDI; Fluorobond II - Shofu Inc.; Prime & Bond NT, Dentsply) were investigated before and after NaF solution treatment. A fluoride ion-selective electrode was to measure fluoride concentrations. During the 86-day period before NaF solution treatment, Fuji IX GP released the highest amount of fluoride among the restorative materials while Prime & Bond NT was the highest among the dental adhesives. After NaF solution treatment, Fuji IX GP again ranked the highest in fluoride release among the restorative materials while Fluorobond II ranked the highest among dental adhesives. It was concluded that the compositions and setting mechanisms of fluoride-containing dental materials influenced their fluoride release and recharge abilities.

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D.; Gdowski, G.E.

1988-05-01

Three copper-based alloys --- CDA 102 (OFHC copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni) --- are being considered as possible materials for the fabrication of high-level radioactive-waste disposal containers. Waste will include fuel assemblies from reactors as well as borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada, for emplacement. The three copper-based alloys discussed here are being considered in addition to the iron- to nickel-based austenitic materials discussed in Volume 3. The decay of radionuclides will result in substantial heat generation and in fluxes of gamma radiation. In this environment, container materials may degrade by atmospheric oxidation, uniform aqueous phase corrosion, pitting, crevice corrosion, transgranular stress corrosion cracking (TGSCC) in tarnishing environments, or intergranular stress corrosion cracking (IGSCC) in nontarnishing environments. This report is a critical survey of available data on the stress corrosion cracking (SCC) of the three copper-based alloys. The requisite conditions for TGSCC and IGSCC include combinations of stress, oxygen, ammonia or nitrite, and water. Note that nitrite is generated by gamma radiolysis of moisture films in air but that ammonia is not. TGSCC has been observed in CDA 102 and CDA 613 exposed to moist ammonia-containing environments whereas SCC has not been documented for CDA 715 under similar conditions. SCC is also promoted in copper by nitrite ions. Furthermore, phosphorus-deoxidized copper is unusually susceptible to embrittlement in such environments. The presence of tin in CDA 613 prevents IGSCC. It is believed that tin segregates to grain boundaries, where it oxidizes very slowly, thereby inhibiting the oxidation of aluminum. 117 refs., 27 figs., 9 tabs

In vitro effects of alcohol-containing and alcohol-free mouthrinses on microhardness of some restorative materials.

Science.gov (United States)

Gürgan, S; Onen, A; Köprülü, H

1997-03-01

Daily application of mouthrinses has been recommended for the prevention and control of caries and periodontal disease. The aim of this study was to evaluate the effect of alcohol-containing and alcohol-free mouthrinses on the microhardness of three restorative materials. Materials tested included visible light cured (VLC) composite resin (Amelogen), VLC glass-ionomer cement (Fuji II LC) and a fissure sealant (Ultra Seal XT). Eighteen cylinders of each restorative were fabricated and initially stored in distilled water for 24 h. Six samples of the restoratives were stored for 12 hours to simulate a 2 min/day for 1 year exposure to mouthrinses in the following solutions: distilled water (control), alcohol-containing mouthrinse (Viadent) and alcohol-free mouthrinse (Rembrandt). At the end of the test period microhardness was measured with a Tukon microhardness tester. Kruskal-Wallis one-way analysis of variance was used to analyse the data. Both mouthrinses affected the hardness of the materials tested.

Simulation and Experimental Validation of Electromagnetic Signatures for Monitoring of Nuclear Material Storage Containers

International Nuclear Information System (INIS)

Aker, Pamela M.; Bunch, Kyle J.; Jones, Anthony M.

2013-01-01

Previous research at the Pacific Northwest National Laboratory (PNNL) has demonstrated that the low frequency electromagnetic (EM) response of a sealed metallic container interrogated with an encircling coil is a strong function of its contents and can be used to form a distinct signature which can confirm the presence of specific components without revealing hidden geometry or classified design information. Finite element simulations have recently been performed to further investigate this response for a variety of configurations composed of an encircling coil and a typical nuclear material storage container. Excellent agreement was obtained between simulated and measured impedance signatures of electrically conducting spheres placed inside an AT-400R nuclear container. Simulations were used to determine the effects of excitation frequency and the geometry of the encircling coil, nuclear container, and internal contents. The results show that it is possible to use electromagnetic models to evaluate the application of the EM signature technique to proposed versions of nuclear weapons containers which can accommodate restrictions imposed by international arms control and treaty verification legislation

Hierarchy concepts: classification and preparation strategies for zeolite containing materials with hierarchical porosity.

Science.gov (United States)

Schwieger, Wilhelm; Machoke, Albert Gonche; Weissenberger, Tobias; Inayat, Amer; Selvam, Thangaraj; Klumpp, Michael; Inayat, Alexandra

2016-06-13

'Hierarchy' is a property which can be attributed to a manifold of different immaterial systems, such as ideas, items and organisations or material ones like biological systems within living organisms or artificial, man-made constructions. The property 'hierarchy' is mainly characterised by a certain ordering of individual elements relative to each other, often in combination with a certain degree of branching. Especially mass-flow related systems in the natural environment feature special hierarchically branched patterns. This review is a survey into the world of hierarchical systems with special focus on hierarchically porous zeolite materials. A classification of hierarchical porosity is proposed based on the flow distribution pattern within the respective pore systems. In addition, this review might serve as a toolbox providing several synthetic and post-synthetic strategies to prepare zeolitic or zeolite containing material with tailored hierarchical porosity. Very often, such strategies with their underlying principles were developed for improving the performance of the final materials in different technical applications like adsorptive or catalytic processes. In the present review, besides on the hierarchically porous all-zeolite material, special focus is laid on the preparation of zeolitic composite materials with hierarchical porosity capable to face the demands of industrial application.

The adhesion characteristics of protective coating materials for the containment structure in nuclear power plants

International Nuclear Information System (INIS)

Lee, Sang-Kook; Shin, Jae-Chul

2003-01-01

Protective coating materials used in the containment structures should be durable for the designed 30 to 40 year lifetime of a nuclear power plant. At the present, these materials have not yet been developed. Therefore it is very important to keep the durability of the protective coating materials through persistent maintenance, and in order to achieve this, understanding the adhesion characteristics of the coating materials is of utmost importance. Therefore, this study attempts to find any methods for durability maintenance of these protective coating materials. To accomplish these aims, this study applied an experimental deterioration environment condition relevant to Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB), categorized as of Design Basis Accident (DBA), onto steel liner plate specimens covered with protective coating materials. Adhesion tests were performed on these deteriorated coating materials to characterize the physical properties and through these tests, the quantitative adhesion characteristics according to the history of deterioration environment were found

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D.; Gdowski, G.E.

1988-06-01

Three copper-based alloys, CDA 102 (oxygen-free, high-purity copper), CDA 613 (aluminum bronze), and CDA 715 (Cu-30Ni), are candidates for the fabrication of high-level radioactive-waste disposal containers. Waste will include spent fuel assemblies from reactors as well as borosilicate glass, and will be sent to the prospective repository site at Yucca Mountain in Nye County, Nevada. The decay of radionuclides will result in the generation of substantial heat and in fluxes of gamma radiation outside the containers. In this environment, container materials might degrade by atmospheric oxidation, general aqueous phase corrosion, localized corrosion (LC), and stress corrosion cracking (SCC). This volume is a critical survey of available data on pitting and crevice corrosion of the copper-based candidates. Pitting and crevice corrosion are two of the most common forms of LC of these materials. Data on the SCC of these alloys is surveyed in Volume 4. Pitting usually occurs in water that contains low concentrations of bicarbonate and chloride anions, such as water from Well J-13 at the Nevada Test Site. Consequently, this mode of degradation might occur in the repository environment. Though few quantitative data on LC were found, a tentative ranking based on pitting corrosion, local dealloying, crevice corrosion, and biofouling is presented. CDA 102 performs well in the categories of pitting corrosion, local dealloying, and biofouling, but susceptibility to crevice corrosion diminishes its attractiveness as a candidate. The cupronickel alloy, CDA 715, probably has the best overall resistance to such localized forms of attack. 123 refs., 11 figs., 3 tabs

Inelastic analysis acceptance criteria for radioactive material transportation containers

International Nuclear Information System (INIS)

Ammerman, D.J.; Ludwigsen, J.S.

1993-01-01

The design criteria currently used in the design of radioactive material (RAM) transportation containers are taken from the ASME Boiler and Pressure Vessel Code (ASME, 1992). These load-based criteria are ideally suited for pressure vessels where the loading is quasistatic and all stresses are in equilibrium with externally applied loads. For impact events, the use of load-based criteria is less supportable. Impact events tend to be energy controlled, and thus, energy-based acceptance criteria would appear to be more appropriate. Determination of an ideal design criteria depends on what behavior is desired. Currently there is not a design criteria for inelastic analysis for RAM nation packages that is accepted by the regulatory agencies. This lack of acceptance criteria is one of the major factors in limiting the use of inelastic analysis. In this paper inelastic analysis acceptance criteria based on stress and strain-energy density will be compared for two stainless steel test units subjected to impacts onto an unyielding target. Two different material models are considered for the inelastic analysis, a bilinear fit of the stress-strain curve and a power law hardening model that very closely follows the stress-strain curve. It is the purpose of this paper to stimulate discussion and research into the area of strain-energy density based inelastic analysis acceptance criteria

Design of an experiment to measure the fire exposure of radioactive materials packages aboard container cargo ships

International Nuclear Information System (INIS)

Koski, J.A.

1997-11-01

The test described in this paper is intended to measure the typical accident environment for a radioactive materials package in a fire aboard a container cargo ship. A stack of nine used standard cargo containers will be variously loaded with empty packages, simulated packages and combustible cargo and placed over a large hydrocarbon pool fire of one hour duration. Both internal and external fire container fire environments typical of on-deck stowage will be measured as well as the potential for container to container fire spread. With the use of the inverse heat conduction calculations, the local heat transfer to the simulated packages can be estimated from thermocouple data. Data recorded will also provide information on fire durations in each container, fire intensity and container to container fire spread characteristics

Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

Energy Technology Data Exchange (ETDEWEB)

Vinson, D.W.; Nutt, W.M.; Bullen, D.B. [Iowa State Univ. of Science and Technology, Ames, IA (United States)

1995-06-01

Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27.

Survey of the degradation modes of candidate materials for high-level radioactive waste disposal containers

International Nuclear Information System (INIS)

Vinson, D.W.; Nutt, W.M.; Bullen, D.B.

1995-06-01

Oxidation and atmospheric corrosion data suggest that addition of Cr provides the greatest improvement in oxidation resistance. Cr-bearing cast irons are resistant to chloride environments and solutions containing strongly oxidizing constituents. Weathering steels, including high content and at least 0.04% Cu, appear to provide adequate resistance to oxidation under temperate conditions. However, data from long-term, high-temperature oxidation studies on weathering steels were not available. From the literature, it appears that the low alloy steels, plain carbon steels, cast steels, and cast irons con-ode at similar rates in an aqueous environment. Alloys containing more than 12% Cr or 36% Ni corrode at a lower rate than plain carbon steels, but pitting may be worse. Short term tests indicate that an alloy of 9Cr-1Mo may result in increased corrosion resistance, however long term data are not available. Austenitic cast irons show the best corrosion resistance. A ranking of total corrosion performance of the materials from most corrosion resistant to least corrosion resistant is: Austenitic Cast Iron; 12% Cr = 36% Ni = 9Cr-1Mo; Carbon Steel = Low Alloy Steels; and Cast Iron. Since the materials to be employed in the Advanced Conceptual Design (ACD) waste package are considered to be corrosion allowance materials, the austenitic cast irons, high Cr steels, high Ni steels and the high Cr-Mo steels should not be considered as candidates for the outer containment barrier. Based upon the oxidation and corrosion data available for carbon steels, low alloy steels, and cast irons, a suitable list of candidate materials for a corrosion allowance outer barrier for an ACD waste package could include, A516, 2.25%Cr -- 1%Mo Steel, and A27

Radiation Shielding Materials Containing Hydrogen, Boron, and Nitrogen: Systematic Computational and Experimental Study. Phase I

Science.gov (United States)

Thibeault, Sheila A.; Fay, Catharine C.; Lowther, Sharon E.; Earle, Kevin D.; Sauti, Godfrey; Kang, Jin Ho; Park, Cheol; McMullen, Amelia M.

2012-01-01

The key objectives of this study are to investigate, both computationally and experimentally, which forms, compositions, and layerings of hydrogen, boron, and nitrogen containing materials will offer the greatest shielding in the most structurally robust combination against galactic cosmic radiation (GCR), secondary neutrons, and solar energetic particles (SEP). The objectives and expected significance of this research are to develop a space radiation shielding materials system that has high efficacy for shielding radiation and that also has high strength for load bearing primary structures. Such a materials system does not yet exist. The boron nitride nanotube (BNNT) can theoretically be processed into structural BNNT and used for load bearing structures. Furthermore, the BNNT can be incorporated into high hydrogen polymers and the combination used as matrix reinforcement for structural composites. BNNT's molecular structure is attractive for hydrogen storage and hydrogenation. There are two methods or techniques for introducing hydrogen into BNNT: (1) hydrogen storage in BNNT, and (2) hydrogenation of BNNT (hydrogenated BNNT). In the hydrogen storage method, nanotubes are favored to store hydrogen over particles and sheets because they have much larger surface areas and higher hydrogen binding energy. The carbon nanotube (CNT) and BNNT have been studied as potentially outstanding hydrogen storage materials since 1997. Our study of hydrogen storage in BNNT - as a function of temperature, pressure, and hydrogen gas concentration - will be performed with a hydrogen storage chamber equipped with a hydrogen generator. The second method of introducing hydrogen into BNNT is hydrogenation of BNNT, where hydrogen is covalently bonded onto boron, nitrogen, or both. Hydrogenation of BN and BNNT has been studied theoretically. Hyper-hydrogenated BNNT has been theoretically predicted with hydrogen coverage up to 100% of the individual atoms. This is a higher hydrogen content

Modelling effective dielectric properties of materials containing diverse types of biological cells

International Nuclear Information System (INIS)

Huclova, Sonja; Froehlich, Juerg; Erni, Daniel

2010-01-01

An efficient and versatile numerical method for the generation of different realistically shaped biological cells is developed. This framework is used to calculate the dielectric spectra of materials containing specific types of biological cells. For the generation of the numerical models of the cells a flexible parametrization method based on the so-called superformula is applied including the option of obtaining non-axisymmetric shapes such as box-shaped cells and even shapes corresponding to echinocytes. The dielectric spectra of effective media containing various cell morphologies are calculated focusing on the dependence of the spectral features on the cell shape. The numerical method is validated by comparing a model of spherical inclusions at a low volume fraction with the analytical solution obtained by the Maxwell-Garnett mixing formula, resulting in good agreement. Our simulation data for different cell shapes suggest that around 1MHz the effective dielectric properties of different cell shapes at different volume fractions significantly deviate from the spherical case. The most pronounced change exhibits ε eff between 0.1 and 1 MHz with a deviation of up to 35% for a box-shaped cell and 15% for an echinocyte compared with the sphere at a volume fraction of 0.4. This hampers the unique interpretation of changes in cellular features measured by dielectric spectroscopy when simplified material models are used.

Direct conversion of plutonium-containing materials to borosilicate glass for storage or disposal

International Nuclear Information System (INIS)

Forsberg, C.W.; Beahm, E.C.

1995-01-01

A new process, the Glass Material Oxidation and Dissolution System (GMODS), has been invented for the direct conversion of plutonium metal, scrap, and residue into borosilicate glass. The glass should be acceptable for either the long-term storage or disposition of plutonium. Conversion of plutonium from complex chemical mixtures and variable geometries into homogeneous glass (1) simplifies safeguards and security; (2) creates a stable chemical form that meets health, safety, and environmental concerns; (3) provides an easy storage form; (4) may lower storage costs; and (5) allows for future disposition options. In the GMODS process, mixtures of metals, ceramics, organics, and amorphous solids containing plutonium are fed directly into a glass melter where they are directly converted to glass. Conventional glass melters can accept materials only in oxide form; thus, it is its ability to accept materials in multiple chemical forms that makes GMODS a unique glass making process. Initial proof-of-principle experiments have converted cerium (plutonium surrogate), uranium, stainless steel, aluminum, and other materials to glass. Significant technical uncertainties remain because of the early nature of process development

Comparison of DNA Damage and Apoptosis Induced By Silver Nanoparticle-containing Dressing Materials During Wound Healing.

Science.gov (United States)

Choi, Young Suk; Gwak, Heui-Chul; Park, Jae Keun; Lim, Ji Yun; Yeo, Eui Dong; Park, Eunseok; Kim, Junyong; Lee, Young Koo

2018-04-13

Silver nanoparticle (AgNP)-containing dressings are used worldwide for the treatment of wounds; however, many studies have indicated that AgNPs are toxic to humans and cause cell death, primarily via apoptosis. In this study, the investigators compare the apoptotic effects of various AgNP dressing materials, with the hypothesis that nanosilver would be less toxic than ionic silver. For the in vivo experiments, Sprague-Dawley (SD) and streptozotocin (STZ)-induced diabetic rats were treated with 5 dressing materials: Aquacel Ag (product A, silver ion; ConvaTec, Berkshire, UK), Acticoat (product B, AgNP; Smith & Nephew, Fort Worth, TX), Medifoam Silver (product C, silver ion; Genewel Science Co Ltd, Seongnam, South Korea), PolyMem Silver (product D, AgNP; Ferris Mfg Corp, Fort Worth, TX), and Vaseline-impregnated dressing gauze (control; Unilever, London, UK). All treatments were applied 3 times per week. After 14 days of treatment, the SD and STZ rats were euthanized, and wound samples were examined for apoptosis. The analysis included immunohistochemistry, terminal deoxynucleotidyl transferase dUTP nick end labeling (TUNEL) staining, Western blotting, and reverse transcription polymerase chain reaction for a semiquantitative evaluation of apoptosis. The AgNP-containing dressing materials were more cytotoxic than the silver dressings. Compared with the AgNP dressing materials, no significant levels of apoptotic factors were observed in the silver dressing-treated wounds. The TUNEL staining showed that product C-dressed wounds contained the most apoptotic cells, while some apoptotic cells were observed in product B-dressed wounds. Moreover, apoptotic gene expression was altered, including a decline in B-cell lymphoma-2 and activation of caspase-3. This was most evident in wounds treated with product C. Interestingly, apoptotic gene expression was not induced in product A-treated wounds. Finally, product D had a relatively lower silver concentration and was less toxic

EBER - development of evaluation methods for shipping and storage containers with an increased content of metallic residual materials

International Nuclear Information System (INIS)

Droste, B.; Voelzke, H.

1995-01-01

Containers which are manufactured using recycled metallic residual materials from decommissioned nuclear facilities must be designed in accordance with the requirements of transport, interim and ultimate storage of radioactive waste. When metallic residual materials are added to the melt in the manufacture of containers made of cast iron with nodular graphite (GGG 40), how the effect on and characteristics of materials are to be observed as well as the permissible limits. The influence on charcteristics of relevance to safety, such as fracture toughness and the types and magnitudes of defects in construction components are the most important factors. In BAM's 'EBER' project presented here, the studies concentrate on design-based safety against the mechanical effects of accidents. (orig./DG) [de

Adsorption and revaporisation studies on iodine oxide aerosols deposited on containment surface materials in LWR

International Nuclear Information System (INIS)

Tietze, S.; Foreman, M.R.StJ.; Ekberg, C.; Kaerkelae, T.; Auvinen, A.; Tapper, U.; Lamminmaeki, S.; Jokiniemi, J.

2012-12-01

During a hypothetical severe nuclear accident, the radiation field will be very high in the nuclear reactor containment building. As a result gaseous radiolysis products will be formed. Elemental iodine can react in the gaseous phase with ozone to form solid iodine oxide aerosol particles (iodine oxide). Within the AIAS (Adsorption of Iodine oxide Aerosols on Surfaces) project the interactions of iodine oxide (IOx) aerosols with common containment surface materials were investigated. Common surface materials in Swedish and Finnish LWRs are Teknopox Aqua V A paint films and metal surfaces such as Cu, Zn, Al and SS, as well as Pt and Pd surfaces from hydrogen recombiners. Non-radioactive and 131 I labelled iodine oxide aerosols were produced with the EXSI CONT facility from elemental iodine and ozone at VTT Technical Research Centre of Finland. The iodine oxide deposits were analysed with microscopic and spectroscopic measurement techniques to identify the kind of iodine oxide formed and if a chemical conversion on the different surface materials occurs. The revaporisation behaviour of the deposited iodine oxide aerosol particles from the different surface materials was studied under the influence of heat, humidity and gamma irradiation at Chalmers University of Technology, Sweden. Studies on the effects of humidity were performed using the FOMICAG facility, while heat and irradiation experiments were performed in a thermostated heating block and with a gammacell 22 having a dose rate of 14 kGy/h. The revaporisation losses were measured using a HPGe detector. The revaporisated 131 I species from the surfaces were chemically tested for elemental iodine formation. The parameter dominating the degradation of the produced iodine oxide aerosols was humidity. Cu and Zn surfaces were found to react with iodine from the iodine oxide aerosols to form iodides, while no metal iodides were detected for Al and SS samples. Most of the iodine oxide aerosols are assumed to be

Use of various types of carbon-containing raw materials to produce thermal energy

Directory of Open Access Journals (Sweden)

В. Б. Кусков

2016-08-01

Full Text Available Many types of carbon-containing organic compounds and all possible carbon-containing products or wastes in low demand can be used to produce thermal energy. A technology has been developed for producing highly flammable briquettes on the basis of bituminous coal. These briquettes have a special incendiary layer. It is easily ignites from low energy heat sources (e.g. matches, and then flame spreads to the rest of briquette. Use of coal slacks and paper wastes as carbon-containing components playing the role of binders provides an opportunity to get a fuel briquette easy in terms of production and plain in composition while at the same time dispose of coal and paper wastes. Such briquettes may also have a special incendiary layer. Technology for fuel briquettes production from wood and slate wastes employed no binding agents, as wood products acted as binders. Thus technologies have been developed to produce fuel briquettes from various carbon-containing materials in low demand. The briquettes are intended for household boilers, fireplaces, different ovens in order to cook food, heat residential and utility premises, cabins, etc.

Preparation and characterization of macrocapsules containing microencapsulated PCMs (phase change materials) for thermal energy storage

International Nuclear Information System (INIS)

Han, Pengju; Lu, Lixin; Qiu, Xiaolin; Tang, Yali; Wang, Jun

2015-01-01

This paper was aimed to prepare, characterize and determine the comprehensive evaluation of promising composite macrocapsules containing microencapsulated PCMs (phase change materials) with calcium alginate gels as the matrix material. Macrocapsules containing microcapsules were fabricated by piercing-solidifying incuber method. Two kinds of microcapsules with n-tetradecane as core material, UF (urea-formaldehyde) and PMMA (poly(methyl methacrylate)) respectively as shell materials were prepared initially. For application concerns, thermal durability and mechanical property of macrocapsules were investigated by TGA (thermal gravimetric analysis) and Texture Analyser for the first time, respectively. The results showed excellent thermal stability and the compressive resistance of macrocapsules was sufficient for common application. The morphology and chemical structure of the prepared microcapsules and macrocapsules were characterized by SEM (scanning electron microscopy) and FT-IR (fourier transform infrared) spectroscopy method. Phase change behaviors and thermal durability of microcapsules and macrocapsules were investigated by DSC (differential scanning calorimetry). In order to improve latent heat of composite microcapsules, the core-shell weight ratio of tetradecane/UF shell microcapsules was chosen as 5.5:1 which obtained the phase change enthalpy of 194.1 J g −1 determined by DSC. In conclusion, these properties make it a feasible composite in applications of textile, building and cold-chain transportation. - Highlights: • We improved the phase change enthalpy with a higher core-shell ratio. • Urea-formaldehyde was firstly used as a shell material in the composite. • Mechanical and thermal durability property of the macrocapsules was firstly investigated in our work.

Storage containers for radioactive material

International Nuclear Information System (INIS)

Cassidy, D.A.; Dates, L.R.; Groh, E.F.

1981-01-01

A radioactive material storage system is disclosed for use in the laboratory. This system is composed of the following: a flat base plate with a groove in one surface thereof and a hollow pedestal extending perpendicularly away from the other surface thereof; a sealing gasket in the groove, a cover having a filter therein and an outwardly extending flange which fits over the plate; the groove and the gasket, and a clamp for maintaining the cover and the plate are sealed together, whereby the plate and the cover and the clamp cooperate to provide a storage area for radioactive material readily accessible for use or inventory. Wall mounts are provided to prevent accidental formation of critical masses during storage

Lightweight flywheel containment

Science.gov (United States)

Smith, James R.

2004-06-29

A lightweight flywheel containment composed of a combination of layers of various material which absorb the energy of a flywheel structural failure. The various layers of material act as a vacuum barrier, momentum spreader, energy absorber, and reaction plate. The flywheel containment structure has been experimentally demonstrated to contain carbon fiber fragments with a velocity of 1,000 m/s and has an aerial density of less than 6.5 g/square centimeters. The flywheel containment, may for example, be composed of an inner high toughness structural layer, and energy absorbing layer, and an outer support layer. Optionally, a layer of impedance matching material may be utilized intermediate the flywheel rotor and the inner high toughness layer.

Inorganic material candidates to replace a metallic or non-metallic concrete containment liner

Energy Technology Data Exchange (ETDEWEB)

Seni, C [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Mills, R H [Toronto Univ., ON (Canada)

1994-12-31

Internal liners for concrete containments are generally organic or metals. They have to be able to inhibit radioactive leakage into the environment, but both types have shortcomings. The results of research to develop a better liner are published in this paper. The best material found was fibre-reinforced mortar. Of the fibres considered, steel, kevlar and glass were the best, each showing advantages and disadvantages. 1 ref., 9 tabs., 12 figs.

Inorganic material candidates to replace a metallic or non-metallic concrete containment liner

International Nuclear Information System (INIS)

Seni, C.; Mills, R.H.

1994-01-01

Internal liners for concrete containments are generally organic or metals. They have to be able to inhibit radioactive leakage into the environment, but both types have shortcomings. The results of research to develop a better liner are published in this paper. The best material found was fibre-reinforced mortar. Of the fibres considered, steel, kevlar and glass were the best, each showing advantages and disadvantages. 1 ref., 9 tabs., 12 figs

Manufacturing method for radioactive material containing vessel

International Nuclear Information System (INIS)

Nishioka, Hideharu; Matsushita, Kazuo; Toyota, Michinori.

1997-01-01

Lead homogenization is applied on the inner surface of a space formed between an inner cylinder and an outer cylinder, and a molten lead heated to about 400 to 500degC is cast into a space formed between the inner cylinder and the outer cylinder in a state where the inner and the outer cylinders are heated to from 200 to 300degC. The space formed between the inner cylinder and the outer cylinder is heated to and kept at 330degC or higher for at least 2minutes after the casting of the molten lead, and then it is cooled. Thus, lowering of density of the molten lead due to excess elevation of temperature or dropping of the lead at the homogenization portion by heating the inner and the outer cylinders to an excessively high temperature are not caused. In addition, formation of gaps in the boundary between the inner cylinder and the outer cylinder or between the lead of the homogenized portion and that of the cast portion due to the melting of the lead of the homogenized portion in the space is prevented reliably thereby capable of forming a satisfactory shielding member. Then, even when the thickness of the inner cylinder and the outer cylinder is large, radioactive material containing vessel excellent in heat releasing property and radiation shielding property can be manufactured. (N.H.)

ESR study of the radiolysis of cellobiose, cellulose-containing materials, and their mixtures with methyl methacrylate

International Nuclear Information System (INIS)

Kozlova, E.Y.; Shostenko, A.G.; Ermolaev, S.V.

1995-01-01

The ESR spectra of γ-irradiated cellobiose, paper waste, and cellulose extracted from paper waste and waste pulp sludge were analyzed. The kinetics of formation and decay of cellobiose radicals were investigated, and the radiation-chemical yields of the radicals formed in cellulose-containing materials were calculated. The ESR spectra of cellobiose irradiated in the presence of methyl methacrylate (MMA) were obtained. A probable mechanism of MMA grafting onto cellulose-containing matrices is considered

Characteristics of Airborne Particulates Containing Naturally Occurring Radioactive Materials in Monazite Industry

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Geon; Choi, Cheol Kyu; Park, Il; Kim, Min Jun; Go, A Ra; Ji, Seung Woo; Kim, Kwang Pyo [Kyunghee University, Yongin (Korea, Republic of); Koo, Bon Cheol [KINS, Daejeon (Korea, Republic of)

2016-05-15

The objective of this study was to characterize physicochemical properties of airborne particulates at a monazite pulverization industry. The properties included particulate size distribution, concentration, shape, density, and radioactivity concentration. Monazite is one of the minerals containing naturally occurring radioactive material (NORM). Therefore, external and internal exposure can be occurred to the workers in monazite industry. The major exposure pathway of the workers is internal exposure due to inhalation of airborne particulates. According to International Commission on Radiological Protection (ICRP), radiation dose due to inhaled particulates containing NORM depends on particulate properties. Therefore, ICRP recommended the internal dose assessment using measured physicochemical properties of the airborne particulates. In the absence of specific information, ICRP provided default reference values. In this study, we characterized physicochemical properties of airborne particulates at a monazite pulverization industry. The databases of particulate information can be used for accurate internal dose assessment of worker.

Fluorine uptake into the human enamel surface from fluoride-containing sealing materials during cariogenic pH cycling

Energy Technology Data Exchange (ETDEWEB)

Yasuhiro, Matsuda, E-mail: matsuda@den.hokudai.ac.jp [Department of Restorative Dentistry, Graduate School of Dental Medicine Hokkaido University (Japan); Katsushi, Okuyama [Department of Restorative Dentistry, Graduate School of Dental Medicine Hokkaido University (Japan); Hiroko, Yamamoto [Graduate School of Dentistry, Osaka University (Japan); Hisanori, Komatsu [Department of Restorative Dentistry, Graduate School of Dental Medicine Hokkaido University (Japan); Masashi, Koka; Takahiro, Sato [Takasaki Advanced Radiation Research Institute, JAEA (Japan); Naoki, Hashimoto; Saiko, Oki; Chiharu, Kawamoto; Hidehiko, Sano [Department of Restorative Dentistry, Graduate School of Dental Medicine Hokkaido University (Japan)

2015-04-01

To prevent the formation of caries and reduce dentin hypersensitivity, sealing materials, either with or without fluoride, are generally applied on the tooth in clinical practice. Application of fluoride-free sealing materials results in the formation of an acid-resistant layer on the tooth surface. On the other hand, fluoride-containing sealing materials might not only form an acid-resistant layer but could possibly further provide fluoride to enhance remineralization and reduce demineralization. In this study, the demineralization prevention ability and fluorine uptake rate in human enamel of fluoride-containing sealing materials [“MS coats F” (MSF)] and fluoride-free sealing materials (“hybrid coats 2” [HI]) were evaluated using an automatic pH cycling system. Each material was applied to the original tooth surface, the cut surfaces were covered with sticky wax, and the automatic pH-cycling system simulated daily acid changes (pH 6.8–4.5) occurring in the oral cavity for 4 weeks. Caries progression was analyzed using transverse microradiography (TMR) taken pre and post the 4 weeks of pH cycling. The fluorine and calcium distributions in the carious lesion in each specimen were evaluated using the proton-induced gamma emission (PIGE) and proton-induced X-ray (PIXE) techniques, respectively. TMR analysis showed that both MSF and HI had a caries-preventing effect after 4 weeks of pH cycling. PIGE/PIXE analysis demonstrated that only MSF resulted in fluoride uptake in the enamel surface. Therefore, MSF can help to form an acid-resistant layer and provide fluoride to the enamel surface. The presence of fluoride on the enamel surface suggested that MSF could prevent demineralization, even if the acid-resistant layer was removed, in clinical settings. The data obtained using the PIGE and PIXE techniques are useful for understanding the benefits of the use of a fluoride-containing sealing material for preventing caries.

Fluorine uptake into the human enamel surface from fluoride-containing sealing materials during cariogenic pH cycling

International Nuclear Information System (INIS)

Yasuhiro, Matsuda; Katsushi, Okuyama; Hiroko, Yamamoto; Hisanori, Komatsu; Masashi, Koka; Takahiro, Sato; Naoki, Hashimoto; Saiko, Oki; Chiharu, Kawamoto; Hidehiko, Sano

2015-01-01

To prevent the formation of caries and reduce dentin hypersensitivity, sealing materials, either with or without fluoride, are generally applied on the tooth in clinical practice. Application of fluoride-free sealing materials results in the formation of an acid-resistant layer on the tooth surface. On the other hand, fluoride-containing sealing materials might not only form an acid-resistant layer but could possibly further provide fluoride to enhance remineralization and reduce demineralization. In this study, the demineralization prevention ability and fluorine uptake rate in human enamel of fluoride-containing sealing materials [“MS coats F” (MSF)] and fluoride-free sealing materials (“hybrid coats 2” [HI]) were evaluated using an automatic pH cycling system. Each material was applied to the original tooth surface, the cut surfaces were covered with sticky wax, and the automatic pH-cycling system simulated daily acid changes (pH 6.8–4.5) occurring in the oral cavity for 4 weeks. Caries progression was analyzed using transverse microradiography (TMR) taken pre and post the 4 weeks of pH cycling. The fluorine and calcium distributions in the carious lesion in each specimen were evaluated using the proton-induced gamma emission (PIGE) and proton-induced X-ray (PIXE) techniques, respectively. TMR analysis showed that both MSF and HI had a caries-preventing effect after 4 weeks of pH cycling. PIGE/PIXE analysis demonstrated that only MSF resulted in fluoride uptake in the enamel surface. Therefore, MSF can help to form an acid-resistant layer and provide fluoride to the enamel surface. The presence of fluoride on the enamel surface suggested that MSF could prevent demineralization, even if the acid-resistant layer was removed, in clinical settings. The data obtained using the PIGE and PIXE techniques are useful for understanding the benefits of the use of a fluoride-containing sealing material for preventing caries

Standard Practice for Laboratory Screening of Metallic Containment Materials for Use With Liquids in Solar Heating and Cooling Systems

CERN Document Server

American Society for Testing and Materials. Philadelphia

1980-01-01

1.1 This practice covers several laboratory test procedures for evaluating corrosion performance of metallic containment materials under conditions similar to those that may occur in solar heating and cooling systems. All test results relate to the performance of the metallic containment material only as a part of a metal/fluid pair. Performance in these laboratory test procedures, taken by itself, does not necessarily constitute an adequate basis for acceptance or rejection of a particular metal/fluid pair in solar heating and cooling systems, either in general or in a particular design. This practice is not intended to preclude the use of other screening tests, particularly when those tests are designed to more closely simulate field service conditions. 1.2 This practice describes apparatus and procedures for several tests, any one or more of which may be used to evaluate the deterioration of the metallic containment material in a metal/fluid pair. The procedures are designed to permit simulation, heating...

Tension in the tie-down chains of a shipping container for hazardous material

International Nuclear Information System (INIS)

Lo, T.Y.

1995-01-01

Chains are frequently used to tie the shipping containers of hazardous material to a truck bed. The tie-down system is nonlinear when the container is subjected to a triaxial force during transit. It is nonlinear because chains cannot carry compressive force, and the base of the container may partially lift off from the truck bed. A method was developed to calculate the amount of tension in the chains. This methodology includes three assumptions: (1) No friction exists between the container and the truck bed; (2) The container and the truck bed are rigid; and (3) All chains are properly tightened (i.e., no slacks) during preparation for shipment. The methodology employs an iterative process of a linear tie-down system. This linear system is derived from the nonlinear system with two additional assumptions: (a) All chains can carry compression as well as tension; and (b) There is a point contact between the container and the truck bed. This linear system has a closed-form solution. After the first solution of the linear system is obtained, the unreasonable, or physically impossible, rotational degree of freedom of the container and the chains with compressive force are eliminated in the follow-up iterative calculations using the same linear system. Unreasonable rotation is detected when the results indicate a rotation of the container into the truck bed. Zero rotational value is assigned to this degree of freedom in the follow-up iterations. For chains under compression, the chain length is replaced with a fictitiously large value because a long chain carries essentially no load. This process usually needs only two or three iterations and can easily be carried out in a spread sheet using such programs as Microsoft Excel or LOTUS 123

Membrane and Films Based on Novel Crown-Containing Dyes as Promising Chemosensoring Materials

Directory of Open Access Journals (Sweden)

Sergei Yu. Zaitsev

2010-12-01

Full Text Available This paper discusses several works on supramolecular systems such as monolayer and multilayer, polymer films of various crown-containing dyes, surface-active monomers and polymers. Design, production and investigation of the membrane nanostructures based on crown ethers is a rapidly developing field at the “junction” of materials sciences and nanotechnology. These nanostructures can serve as convenient models for studying the self-organization and molecular recognition processes at interfaces that are typical for biomembranes. Based on the results obtained for such structures by absorption and fluorescence spectroscopy, atomic force and Brewster-angle microscopy, surface pressure and surface potential isotherm measurements, the possibility of developing micro- and nanomaterials possessing a set of specified properties (including chemosensor, photochromic and photorefractive materials is demonstrated.

Selection of candidate canister materials for high-level nuclear waste containment in a tuff repository

International Nuclear Information System (INIS)

McCright, R.D.; Weiss, H.; Juhas, M.C.; Logan, R.W.

1983-11-01

A repository located at Yucca Mountain at the Nevada Test Site is a potential site for permanent geological disposal of high-level nuclear waste. The repository can be located in a horizon in welded tuff, a volcanic rock, which is above the static water level at this site. The environmental conditions in this unsaturated zone are expected to be air and water vapor dominated for much of the containment period. Type 304L stainless steel is the reference material for fabricating canisters to contain the solid high-level wastes. Alternative stainless alloys are considered because of possible susceptibility of 304L to localized and stress forms of corrosion. For the reprocessed glass wastes, the canisters serve as the recipient for pouring the glass with the result that a sensitized microstructure may develop because of the times at elevated temperatures. Corrosion testing of the reference and alternative materials has begun in tuff-conditioned water and steam environments. 21 references, 8 figures, 8 tables

THE NEED FOR A NEW JOINING TECHNOLOGY FOR THE CLOSURE WELDING OF RADIOACTIVE MATERIALS CONTAINERS

International Nuclear Information System (INIS)

CANNELL GR; HILL BE; GRANT GJ

2008-01-01

One of the activities associated with cleanup throughout the Department of Energy (DOE) complex is packaging radioactive materials into storage containers. Much of this work will be performed in high-radiation environments requiring fully remote operations, for which existing, proven systems do not currently exist. These conditions demand a process that is capable of producing acceptable (defect-free) welds on a consistent basis; the need to perform weld repair, under fully-remote operations, can be extremely costly and time consuming. Current closure welding technology (fusion welding) is not well suited for this application and will present risk to cleanup cost and schedule. To address this risk, Fluor and the Pacific Northwest National Laboratory (PNNL), are proposing that a new and emerging joining technology, Friction Stir Welding (FSW), be considered for this work. FSW technology has been demonstrated in other industries (aerospace and marine) to produce near flaw-free welds on a consistent basis. FSW is judged capable of providing the needed performance for fully-remote closure welding of containers for radioactive materials for the following reasons: FSW is a solid-state process; material is not melted. As such, FSW does not produce the type of defects associated with fusion welding, e.g., solidification-induced porosity, cracking, distortion due to weld shrinkage, and residual stress. In addition, because FSW is a low-heat input process, material properties (mechanical, corrosion and environmental) are preserved and not degraded as can occur with 'high-heat' fusion welding processes. When compared to fusion processes, FSW produces extremely high weld quality. FSW is performed using machine-tool technology. The equipment is simple and robust and well-suited for high radiation, fully-remote operations compared to the relatively complex equipment associated with the fusion-welding processes. Additionally, for standard wall thicknesses of radioactive materials

A fracture mechanics safety concept to assess the impact behavior of ductile cast iron containers for shipping and storage of radioactive materials

International Nuclear Information System (INIS)

Voelzke, H.; Roedel, R.; Droste, B.

1994-01-01

Within the scope of the German licensing procedures for shipping and storage containers for radioactive materials made of ductile cast iron, BAM performs approval design tests including material tests to ensure the main safety goals of shielding, leaktightness and subcriticality under ''Type B accident conditions''. So far the safety assessment concept of BAM is based essentially on the experimental proof of container strength by prototype testing under most damaging test conditions in connection with complete approval design tests, and has been developed especially for cylindrical casks like CASTOR- and TN-design. In connection with the development of new container constructions such as ''cubic cast containers'', and the fast developments in the area of numerical calculation methods, there is a need for a more flexible safety concept especially considering fracture mechanics aspects.This paper presents the state of work at BAM for such an extended safety concept for ductile cast iron containers, based on a detailed brittle fracture safe design proof. The requirements on stress analysis (experimental or numerical), material properties, material qualification, quality assurance provisions and fracture mechanics safety assessment, including well defined and justified factors of safety, are described. ((orig.))

Effect of chloride concentration and pH on pitting corrosion of waste package container materials

International Nuclear Information System (INIS)

Roy, A.K.; Fleming, D.L.; Gordon, S.R.

1996-12-01

Electrochemical cyclic potentiodynamic polarization experiments were performed on several candidate waste package container materials to evaluate their susceptibility to pitting corrosion at 90 degrees C in aqueous environments relevant to the potential underground high-level nuclear waste repository. Results indicate that of all the materials tested, Alloy C-22 and Ti Grade-12 exhibited the maximum corrosion resistance, showing no pitting or observable corrosion in any environment tested. Efforts were also made to study the effect of chloride ion concentration and pH on the measured corrosion potential (Ecorr), critical pitting and protection potential values

Improving Performance of Cold-Chain Insulated Container with Phase Change Material: An Experimental Investigation

Directory of Open Access Journals (Sweden)

Li Huang

2017-12-01

Full Text Available The cold-chain transportation is an important means to ensure the drug and food safety. An cold-chain insulated container incorporating with Phase Change Material (PCM has been developed for a temperature-controlled transportation in the range of 2~8 °C. The container configuration and different preconditioning methods have been determined to realize a 72-h transportation under extremely high, extremely low, and alternating temperature conditions. The experimental results showed that the temperature-controlled time was extended from 1 h to more than 80 h and the internal temperature maintained at 4~5 °C by using a PCM with a melting/freezing point of 5 °C, while the container presented a subcooling effect in a range of −1~2 °C when using water as PCM. The experimental values of the temperature-controlled time agreed well with the theoretical values.

Study of improving the thermal response of a construction material containing a phase change material

Science.gov (United States)

Laaouatni, A.; Martaj, N.; Bennacer, R.; Elomari, M.; El Ganaoui, M.

2016-09-01

The use of phase change materials (PCMs) for improving the thermal comfort in buildings has become an attractive application. This solution contributes to increasing the thermal inertia of the building envelope and reducing power consumption. A building element filled with a PCM and equipped with ventilation tubes is proposed, both for increasing inertia and contributing to refreshing building envelope. A numerical simulation is conducted by the finite element method in COMSOL Multiphysics, which aims to test the thermal behaviour of the developed solution. An experimental study is carried out on a concrete block containing a PCM with ventilation tubes. The objective is to see the effect of PCM coupled with ventilation on increasing the inertia of the block. The results show the ability of this new solution to ensure an important thermal inertia of a building.

Hybrid silica luminescent materials based on lanthanide-containing lyotropic liquid crystal with polarized emission

Energy Technology Data Exchange (ETDEWEB)

Selivanova, N.M., E-mail: natsel@mail.ru [Kazan National Research Technological University, 68 Karl Marx Str., Kazan 420015 (Russian Federation); Vandyukov, A.E.; Gubaidullin, A.T. [A.E. Arbuzov Institute of Organic and Physical Chemistry of the Kazan Scientific Center of the Russian Academy of Sciences, 8 Acad. Arbuzov Str., Kazan 420088 (Russian Federation); Galyametdinov, Y.G. [Kazan National Research Technological University, 68 Karl Marx Str., Kazan 420015 (Russian Federation)

2014-11-14

This paper represents the template method for synthesis of hybrid silica films based on Ln-containing lyotropic liquid crystal and characterized by efficient luminescence. Luminescence films were prepared in situ by the sol–gel processes. Lyotropic liquid crystal (LLC) mesophases C{sub 12}H{sub 25}O(CH{sub 2}CH{sub 2}O){sub 10}H/Ln(NO{sub 3}){sub 3}·6H{sub 2}O/H{sub 2}O containing Ln (III) ions (Dy, Tb, Eu) were used as template. Polarized optical microscopy, X-ray powder diffraction, and FT-IR-spectroscopy were used for characterization of liquid crystal mesophases and hybrid films. The morphology of composite films was studied by the atomic force microscopy method (AFM). The optical properties of the resulting materials were evaluated. It was found that hybrid silica films demonstrate significant increase of their lifetime in comparison with an LLC system. New effects of linearly polarized emission revealed for Ln-containing hybrid silica films. Polarization in lanthanide-containing hybrid composites indicates that silica precursor causes orientation of emitting ions. - Highlights: • We suggest a new simple approach for creating luminescence hybrid silica films. • Ln-containing hybrid silica films demonstrate yellow, green and red emissions. • Tb(III)-containing hybrid film have a high lifetime. • We report effects of linearly polarized emission in hybrid film.

Adsorption and revaporisation studies on iodine oxide aerosols deposited on containment surface materials in LWR

Energy Technology Data Exchange (ETDEWEB)

Tietze, S.; Foreman, M.R.StJ.; Ekberg, C. [Chalmers Univ. of Technology, Goeteborg (Sweden); Kaerkelae, T.; Auvinen, A.; Tapper, U.; Lamminmaeki, S.; Jokiniemi, J. [VTT Technical Research Centre of Finland, Espoo (Finland)

2012-12-15

During a hypothetical severe nuclear accident, the radiation field will be very high in the nuclear reactor containment building. As a result gaseous radiolysis products will be formed. Elemental iodine can react in the gaseous phase with ozone to form solid iodine oxide aerosol particles (iodine oxide). Within the AIAS (Adsorption of Iodine oxide Aerosols on Surfaces) project the interactions of iodine oxide (IOx) aerosols with common containment surface materials were investigated. Common surface materials in Swedish and Finnish LWRs are Teknopox Aqua V A paint films and metal surfaces such as Cu, Zn, Al and SS, as well as Pt and Pd surfaces from hydrogen recombiners. Non-radioactive and {sup 131}I labelled iodine oxide aerosols were produced with the EXSI CONT facility from elemental iodine and ozone at VTT Technical Research Centre of Finland. The iodine oxide deposits were analysed with microscopic and spectroscopic measurement techniques to identify the kind of iodine oxide formed and if a chemical conversion on the different surface materials occurs. The revaporisation behaviour of the deposited iodine oxide aerosol particles from the different surface materials was studied under the influence of heat, humidity and gamma irradiation at Chalmers University of Technology, Sweden. Studies on the effects of humidity were performed using the FOMICAG facility, while heat and irradiation experiments were performed in a thermostated heating block and with a gammacell 22 having a dose rate of 14 kGy/h. The revaporisation losses were measured using a HPGe detector. The revaporisated {sup 131}I species from the surfaces were chemically tested for elemental iodine formation. The parameter dominating the degradation of the produced iodine oxide aerosols was humidity. Cu and Zn surfaces were found to react with iodine from the iodine oxide aerosols to form iodides, while no metal iodides were detected for Al and SS samples. Most of the iodine oxide aerosols are assumed to

40 CFR 61.155 - Standard for operations that convert asbestos-containing waste material into nonasbestos...

Science.gov (United States)

2010-07-01

... to the information requirements of § 61.07(b)(3), a (i) Description of waste feed handling and...) Disposed of as asbestos-containing waste material according to § 61.150, or (ii) Recycled as waste feed... waste feed to the process. (2) Collect and analyze monthly composite samples (one 200-gram (7-ounce...

Grip force and heart rate responses to manual carrying tasks: effects of material, weight, and base area of the container.

Science.gov (United States)

Lee, Tzu-Hsien; Tseng, Chia-Yun

2014-01-01

This study recruited 16 industrial workers to examine the effects of material, weight, and base area of container on reduction of grip force (ΔGF) and heart rate for a 100-m manual carrying task. This study examined 2 carrying materials (iron and water), 4 carrying weights (4.4, 8.9, 13.3, 17.8 kg), and 2 base areas of container (24 × 24 cm, 35 × 24 cm). This study showed that carrying water significantly increased ΔGF and heart rate as compared with carrying iron. Also, ΔGF and heart rate significantly increased with carrying weight and base area of container. The effects of base area of container on ΔGF and heart rate were greater in carrying water condition than in carrying iron condition. The maximum dynamic effect of water on ΔGF and heart rate occurred when water occupied ~60%-80% of full volume of the container.

Development of a testing method for asbestos fibers in treated materials of asbestos containing wastes by transmission electron microscopy

Energy Technology Data Exchange (ETDEWEB)

Yamamoto, Takashi, E-mail: tyama@nies.go.jp [Center for Material Cycles and Waste Management Research, National Institute for Environmental Studies, 16-2 Onogawa, Tsukuba, Ibaraki 305-8506 (Japan); Kida, Akiko [Faculty of Agriculture, Ehime University, 3-5-7 Tarumi, Matsuyama, Ehime 790-8566 (Japan); Noma, Yukio [Department of Environmental Science, Fukuoka Womens University, 1-1-1 Kasumigaoka, Higashiku, Fukuoka 813-8529 (Japan); Terazono, Atsushi [Center for Material Cycles and Waste Management Research, National Institute for Environmental Studies, 16-2 Onogawa, Tsukuba, Ibaraki 305-8506 (Japan); Sakai, Shin-ichi [Environmental Preservation Research Center, Kyoto University, Yoshidahonmachi, Sakyoku, Kyoto 606-8501 (Japan)

2014-02-15

Highlights: • A high sensitive and selective testing method for asbestos in treated materials of asbestos containing wastes was developed. • Asbestos can be determined at a limits are a few million fibers per gram and a few μg g{sup −1}. • High temperature melting treatment samples were determined by this method. Asbestos fiber concentration were below the quantitation limit in all samples, and total fiber concentrations were determined as 47–170 × 10{sup 6} g{sup −1}. - Abstract: Appropriate treatment of asbestos-containing wastes is a significant problem. In Japan, the inertization of asbestos-containing wastes based on new treatment processes approved by the Minister of the Environment is promoted. A highly sensitive method for testing asbestos fibers in inertized materials is required so that these processes can be approved. We developed a method in which fibers from milled treated materials are extracted in water by shaking, and are counted and identified by transmission electron microscopy. Evaluation of this method by using asbestos standards and simulated slag samples confirmed that the quantitation limits are a few million fibers per gram and a few μg/g in a sample of 50 mg per filter. We used this method to assay asbestos fibers in slag samples produced by high-temperature melting of asbestos-containing wastes. Fiber concentrations were below the quantitation limit in all samples, and total fiber concentrations were determined as 47–170 × 10{sup −6} f/g. Because the evaluation of treated materials by TEM is difficult owing to the limited amount of sample observable, this testing method should be used in conjunction with bulk analytical methods for sure evaluation of treated materials.

Ceramics for Molten Materials Containment, Transfer and Handling on the Lunar Surface

Science.gov (United States)

Standish, Evan; Stefanescu, Doru M.; Curreri, Peter A.

2009-01-01

As part of a project on Molten Materials Transfer and Handling on the Lunar Surface, molten materials containment samples of various ceramics were tested to determine their performance in contact with a melt of lunar regolith simulant. The test temperature was 1600 C with contact times ranging from 0 to 12 hours. Regolith simulant was pressed into cylinders with the approximate dimensions of 1.25 dia x 1.25cm height and then melted on ceramic substrates. The regolith-ceramic interface was examined after processing to determine the melt/ceramic interaction. It was found that the molten regolith wetted all oxide ceramics tested extremely well which resulted in chemical reaction between the materials in each case. Alumina substrates were identified which withstood contact at the operating temperature of a molten regolith electrolysis cell (1600 C) for eight hours with little interaction or deformation. This represents an improvement over alumina grades currently in use and will provide a lifetime adequate for electrolysis experiments lasting 24 hours or more. Two types of non-oxide ceramics were also tested. It was found that they interacted to a limited degree with the melt resulting in little corrosion. These ceramics, Sic and BN, were not wetted as well as the oxides by the melt, and so remain possible materials for molten regolith handling. Tests wing longer holding periods and larger volumes of regolith are necessary to determine the ultimate performance of the tested ceramics.

YIELD FORMING EFFECT OF APPLICATION OF COMPOSTS CONTAINING POLYMER MATERIALS ENRICHED IN BIOCOMPONENTS

Directory of Open Access Journals (Sweden)

Florian Gambuś

2014-01-01

Full Text Available In a pot experiment the impact of composts containing polymeric materials modified with biocomponents on the diversity of crops of oats and mustard was examined. The composts used in the study were produced in the laboratory from wheat and rape straw, and pea seed cleaning waste with 8-percent addition of chopped biopolymer materials (films which were prepared in the Central Mining Institute (GIG in Katowice. Three polymers differing in content of starch and density were selected for the composting. The pot experiment was conducted on three substrates: light and medium soil and on the sediment obtained after flotation of zinc and lead ores, coming from the landfill ZGH “Boleslaw” S.A. in Bukowno. The need for using such materials and substrates results from the conditions of processing some morphological fractions of municipal waste and from improving methods of reclamation. Yield enhancing effect of composts depends on the substrate on which the compost was used, cultivated plants and crop succession. Application of composts prepared with 8% of polymeric materials based on polyethylene, modified with starch as biocomponent, resulted in significantly lower yields in sandy (light soil in case of oats and, in some cases, in medium soil. Subsequent plant yield did not differ significantly between the objects fertilized with compost.

Search for hidden high-Z materials inside containers with the Muon Portal Project

International Nuclear Information System (INIS)

Rocca, P La; Bandieramonte, M; Blancato, A A; Bonanno, D; Indelicato, V; Presti, D Lo; Petta, C; Antonuccio, V; Becciani, U; Belluso, M; Billotta, S; Bonanno, G; Costa, A; Garozzo, S; Massimino, P; Belluomo, F; Fallica, G; Leonora, E; Longhitano, F; Longo, S

2014-01-01

The Muon Portal is a recently born project that plans to build a large area muon detector for a noninvasive inspection of shipping containers in the ports, searching for the presence of potential fissile (U, Pu) threats. The technique employed by the project is the well-known muon tomography, based on cosmic muon scattering from high-Z materials. The design and operational parameters of the muon portal under construction will be described in this paper, together with preliminary simulation and test results

Search for hidden high-Z materials inside containers with the Muon Portal Project

Science.gov (United States)

La Rocca, P.; Antonuccio, V.; Bandieramonte, M.; Becciani, U.; Belluomo, F.; Belluso, M.; Billotta, S.; Blancato, A. A.; Bonanno, D.; Bonanno, G.; Costa, A.; Fallica, G.; Garozzo, S.; Indelicato, V.; Leonora, E.; Longhitano, F.; Longo, S.; Lo Presti, D.; Massimino, P.; Petta, C.; Pistagna, C.; Pugliatti, C.; Puglisi, M.; Randazzo, N.; Riggi, F.; Riggi, S.; Romeo, G.; Russo, G. V.; Santagati, G.; Valvo, G.; Vitello, F.; Zaia, A.; Zappalà, G.

2014-01-01

The Muon Portal is a recently born project that plans to build a large area muon detector for a noninvasive inspection of shipping containers in the ports, searching for the presence of potential fissile (U, Pu) threats. The technique employed by the project is the well-known muon tomography, based on cosmic muon scattering from high-Z materials. The design and operational parameters of the muon portal under construction will be described in this paper, together with preliminary simulation and test results.

Ekor - unique material for transportation, containment and disposal of radioactive wastes

International Nuclear Information System (INIS)

Belyaev, S.T.; Shvetsov, I.K.; Perevozchikov, S.A.; Kalinichenko, B.S.; Polivanov, A.N.; Makarenko, I.A.; Minasyan, R.A.; Semenkova, N.Y.; Kozodaeva, M.M.; Kozodaeva, N.M.; Gulko, P.

1998-01-01

EKOR - a unique radiation-resistant silicon-organic foam-type elastomer is presented as a new material for transportation, containment, isolation and disposal of radioactive wastes. EKOR has been developed and full-scale tested by a group of Russian scientists from the Kurchatov Institute, in collaboration with specialists from Euro-Asian Physical Society (EAPS) (President - Prof. S.P. Kapitza) and other organisations. EAPS is a patent holder for EKOR. The sole and exclusive licensee of the patents is Eurotech, Ltd. a U.S. company, with rights to sub-license the patents world-wide. EKOR maintains structural stability - does not disintegrate and preserves its structured properties under radiation, including α, β and γ rays, with the absorbed dose 10 Grad, transforming finally into foam-ceramics with mechanical compression strength within interval 5-10 kg/cm 2 . Material does not inflame and does not burn in the open flame, keeping its initial form and dimensions. It is not toxic under the impact of flame. EKOR has excellent adhesion to concrete, metal, glass without the primer. EKOR has resistance to corrosion caused by acids, alkalis and organic solvents. (authors)

Simulations and imaging algorithm development for a cosmic ray muon tomography system for the detection of special nuclear material in transport containers

International Nuclear Information System (INIS)

Jewett, C.; Anghel, V.N.P.; Armitage, J.; Boudjemline, K.; Botte, J.; Bryman, D.; Bueno, J.; Charles, E.; Cousins, T.; Didsbury, R.; Erhardt, L.; Erlandson, A.; Gallant, G.; Jason, A.; Jonkmans, G.; Liu, Z.; McCall, M.; Noel, S.; Oakham, F.G.; Ong, D.; Stocki, T.; Thompson, M.; Waller, D.

2011-01-01

The Cosmic Ray Inspection and Passive Tomography (CRIPT) collaboration is developing a cosmic ray muon tomography system to identify Special Nuclear Materials (SNM) in cargo containers. In order to gauge the viability of the technique, and to determine the best detector type, GEANT4 was used to simulate the passage of cosmic ray muons through a cargo container. The scattering density estimation (SDE) algorithm was developed and tested with data from these simulations to determine how well it could reconstruct the interior of a container. The simulation results revealed the ability of cosmic ray muon tomography techniques to image spheres of lead-shielded Special Nuclear Materials (SNM), such as uranium or plutonium, in a cargo container, containing a cargo of granite slabs. (author)

Simulations and imaging algorithm development for a cosmic ray muon tomography system for the detection of special nuclear material in transport containers

Energy Technology Data Exchange (ETDEWEB)

Jewett, C.; Anghel, V.N.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Armitage, J.; Boudjemline, K.; Botte, J. [Carleton Univ., Dept. of Physics, Ottawa, Ontario (Canada); Bryman, D. [Advanced Applied Physics Solutions, Vancouver, British Columbia (Canada); Univ. of British Columbia, Vancouver, British Columbia (Canada); Bueno, J. [Advanced Applied Physics Solutions, Vancouver, British Columbia (Canada); Charles, E. [Canada Border Services Agency, Ottawa, Ontario (Canada); Cousins, T. [International Safety Research, Ottawa, Ontario (Canada); Didsbury, R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Erhardt, L. [Defence Research and Development Canada, Ottawa, Ontario (Canada); Erlandson, A. [Carleton Univ., Dept. of Physics, Ottawa, Ontario (Canada); Gallant, G. [Canada Border Services Agency, Ottawa, Ontario (Canada); Jason, A. [Los Alamos National Laboratory, Los Alamos (United States); Jonkmans, G. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Liu, Z. [Advanced Applied Physics Solutions, Vancouver, British Columbia (Canada); Univ. of British Columbia, Vancouver, British Columbia (Canada); McCall, M.; Noel, S. [International Safety Research, Ottawa, Ontario (Canada); Oakham, F.G. [Carleton Univ., Dept. of Physics, Ottawa, Ontario (Canada); TRIUMF, Vancouver, British Columbia, (Canada); Ong, D.; Stocki, T. [Health Canada, Ottawa, Ontario (Canada); Thompson, M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Waller, D. [Defence Research and Development Canada, Ottawa, Ontario (Canada)

2011-07-01

The Cosmic Ray Inspection and Passive Tomography (CRIPT) collaboration is developing a cosmic ray muon tomography system to identify Special Nuclear Materials (SNM) in cargo containers. In order to gauge the viability of the technique, and to determine the best detector type, GEANT4 was used to simulate the passage of cosmic ray muons through a cargo container. The scattering density estimation (SDE) algorithm was developed and tested with data from these simulations to determine how well it could reconstruct the interior of a container. The simulation results revealed the ability of cosmic ray muon tomography techniques to image spheres of lead-shielded Special Nuclear Materials (SNM), such as uranium or plutonium, in a cargo container, containing a cargo of granite slabs. (author)

Sodium montmorillonite/amine-containing drugs complexes: new insights on intercalated drugs arrangement into layered carrier material.

Directory of Open Access Journals (Sweden)

Murilo L Bello

Full Text Available Layered drug delivery carriers are current targets of nanotechnology studies since they are able to accommodate pharmacologically active substances and are effective at modulating drug release. Sodium montmorillonite (Na-MMT is a clay that has suitable properties for developing new pharmaceutical materials due to its high degree of surface area and high capacity for cation exchange. Therefore Na-MMT is a versatile material for the preparation of new drug delivery systems, especially for slow release of protonable drugs. Herein, we describe the intercalation of several amine-containing drugs with Na-MMT so we can derive a better understanding of how these drugs molecules interact with and distribute throughout the Na-MMT interlayer space. Therefore, for this purpose nine sodium montmorillonite/amine-containing drugs complexes (Na-MMT/drug were prepared and characterized. In addition, the physicochemical properties of the drugs molecules in combination with different experimental conditions were assessed to determine how these factors influenced experimental outcomes (e.g. increase of the interlayer spacing versus drugs arrangement and orientation. We also performed a molecular modeling study of these amine-containing drugs associated with different Na-MMT/drug complex models to analyze the orientation and arrangement of the drugs molecules in the complexes studied. Six amine-containing drugs (rivastigmine, doxazosin, 5-fluorouracil, chlorhexidine, dapsone, nystatin were found to successfully intercalate Na-MMT. These findings provide important insights on the interlayer aspect of the molecular systems formed and may contribute to produce more efficient drug delivery nanosystems.

Experimental studies of dynamic impact response with scale models of lead shielded radioactive material shipping containers

International Nuclear Information System (INIS)

Robinson, R.A.; Hadden, J.A.; Basham, S.J.

1978-01-01

Preliminary experimental studies of dynamic impact response of scale models of lead-shielded radioactive material shipping containers are presented. The objective of these studies is to provide DOE/ECT with a data base to allow the prediction of a rational margin of confidence in overviewing and assessing the adequacy of the safety and environmental control provided by these shipping containers. Replica scale modeling techniques were employed to predict full scale response with 1/8, 1/4, and 1/2 scale models of shipping containers that are used in the shipment of spent nuclear fuel and high level wastes. Free fall impact experiments are described for scale models of plain cylindrical stainless steel shells, stainless steel shells filled with lead, and replica scale models of radioactive material shipping containers. Dynamic induced strain and acceleration measurements were obtained at several critical locations on the models. The models were dropped from various heights, attitudes to the impact surface, with and without impact limiters and at uniform temperatures between -40 and 175 0 C. In addition, thermal expansion and thermal gradient induced strains were measured at -40 and 175 0 C. The frequency content of the strain signals and the effect of different drop pad compositions and stiffness were examined. Appropriate scale modeling laws were developed and scaling techniques were substantiated for predicting full scale response by comparison of dynamic strain data for 1/8, 1/4, and 1/2 scale models with stainless steel shells and lead shielding

In vitro antibacterial activity of a novel resin-based pulp capping material containing the quaternary ammonium salt MAE-DB and Portland cement.

Science.gov (United States)

Yang, Yanwei; Huang, Li; Dong, Yan; Zhang, Hongchen; Zhou, Wei; Ban, Jinghao; Wei, Jingjing; Liu, Yan; Gao, Jing; Chen, Jihua

2014-01-01

Vital pulp preservation in the treatment of deep caries is challenging due to bacterial infection. The objectives of this study were to synthesize a novel, light-cured composite material containing bioactive calcium-silicate (Portland cement, PC) and the antimicrobial quaternary ammonium salt monomer 2-methacryloxylethyl dodecyl methyl ammonium bromide (MAE-DB) and to evaluate its effects on Streptococcus mutans growth in vitro. The experimental material was prepared from a 2 : 1 ratio of PC mixed with a resin of 2-hydroxyethylmethacrylate, bisphenol glycerolate dimethacrylate, and triethylene glycol dimethacrylate (4 : 3 : 1) containing 5 wt% MAE-DB. Cured resin containing 5% MAE-DB without PC served as the positive control material, and resin without MAE-DB or PC served as the negative control material. Mineral trioxide aggregate (MTA) and calcium hydroxide (Dycal) served as commercial controls. S. mutans biofilm formation on material surfaces and growth in the culture medium were tested according to colony-forming units (CFUs) and metabolic activity after 24 h incubation over freshly prepared samples or samples aged in water for 6 months. Biofilm formation was also assessed by Live/Dead staining and scanning electron microscopy. S. mutans biofilm formation on the experimental material was significantly inhibited, with CFU counts, metabolic activity, viability staining, and morphology similar to those of biofilms on the positive control material. None of the materials affected bacterial growth in solution. Contact-inhibition of biofilm formation was retained by the aged experimental material. Significant biofilm formation was observed on MTA and Dycal. The synthesized material containing HEMA-BisGMA-TEGDMA resin with MAE-DB as the antimicrobial agent and PC to support mineralized tissue formation inhibited S. mutans biofilm formation even after aging in water for 6 months, but had no inhibitory effect on bacteria in solution. Therefore, this material shows

The ''nuclear car wash'': a scanner to detect illicit special nuclear material in cargo containers

International Nuclear Information System (INIS)

Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Dougan, A. D.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Pohl, B. A.; Prussin, S. G.; Walling, R. S.; Weirup, D. L.

2004-01-01

There is an urgent need to improve the reliability of screening cargo containers for illicit nuclear material that may be hidden there for terrorist purposes. A screening system is described for detection of fissionable material hidden in maritime cargo containers. The system makes use of a low intensity neutron beam for producing fission; and the detection of the abundant high-energy γ rays emitted in the β-decay of short-lived fission products and β-delayed neutrons. The abundance of the delayed γ rays is almost an order of magnitude larger than that of the delayed neutrons normally used to detect fission and they are emitted on about the same time scale as the delayed neutrons, i.e., ∼1 min. The energy and temporal distributions of the delayed γ rays provide a unique signature of fission. Because of their high energy, these delayed γ rays penetrate loW--Z cargoes much more readily than the delayed neutrons. Coupled with their higher abundance, the signal from the delayed γ rays escaping from the container is predicted to be as much as six decades more intense than the delayed neutron signal, depending upon the type and thickness of the intervening cargo. The γ rays are detected in a large array of scintillators located along the sides of the container as it is moved through them. Measurements have confirmed the signal strength in somewhat idealized experiments and have also identified one interference when 14.5 MeV neutrons from the D, T reaction are used for the interrogation. The interference can be removed easily by the appropriate choice of the neutron source

Stress corrosion cracking of candidate materials for nuclear waste containers

International Nuclear Information System (INIS)

Maiya, P.S.; Shack, W.J.; Kassner, T.F.

1989-09-01

Types 304L and 316L stainless steel (SS), Incoloy 825, Cu, Cu-30%Ni, and Cu-7%Al have been selected as candidate materials for the containment of high-level nuclear waste at the proposed Yucca Mountain Site in Nevada. The susceptibility of these materials to stress corrosion cracking has been investigated by slow-strain-rate tests (SSRTs) in water which simulates that from well J-13 (J-13 water) and is representative of the groundwater present at the Yucca Mountain site. The SSRTs were performed on specimens exposed to simulated J-13 water at 93 degree C and at a strain rate 10 -7 s -1 under crevice conditions and at a strain rate of 10 -8 s -1 under both crevice and noncrevice conditions. All the tests were interrupted after nominal elongation strains of 1--4%. Examination by scanning electron microscopy showed some crack initiation in virtually all specimens. Optical microscopy of metallographically prepared transverse sections of Type 304L SS suggests that the crack depths are small (<10 μm). Preliminary results suggest that a lower strain rate increases the severity of cracking of Types 304L and 316L SS, Incoloy 825, and Cu but has virtually no effect on Cu-30%Ni and Cu-7%Al. Differences in susceptibility to cracking were evaluated in terms of a stress ratio, which is defined as the ratio of the increase in stress after local yielding in the environment to the corresponding stress increase in an identical test in air, both computed at the same strain. On the basis of this stress ratio, the ranking of materials in order of increasing resistance to cracking is: Types 304L SS < 316L SS < Incoloy 825 congruent Cu-30%Ni < Cu congruent Cu-7%Al. 9 refs., 12 figs., 7 tabs

Detection of nuclear material by photon activation inside cargo containers

Science.gov (United States)

Gmar, Mehdi; Berthoumieux, Eric; Boyer, Sébastien; Carrel, Frédérick; Doré, Diane; Giacri, Marie-Laure; Lainé, Frédéric; Poumarède, Bénédicte; Ridikas, Danas; Van Lauwe, Aymeric

2006-05-01

Photons with energies above 6 MeV can be used to detect small amounts of nuclear material inside large cargo containers. The method consists in using an intense beam of high-energy photons (bremsstrahlung radiation) in order to induce reactions of photofission on actinides. The measurement of delayed neutrons and delayed gammas emitted by fission products brings specific information on localization and quantification of the nuclear material. A simultaneous measurement of both of these delayed signals can overcome some important limitations due to matrix effects like heavy shielding and/or the presence of light elements as hydrogen. We have a long experience in the field of nuclear waste package characterization by photon interrogation and we have demonstrated that presently the detection limit can be less than one gram of actinide per ton of package. Recently we tried to extend our knowledge to assess the performance of this method for the detection of special nuclear materials in sea and air freights. This paper presents our first results based on experimental measurements carried out in the SAPHIR facility, which houses a linear electron accelerator with the energy range from 15 MeV to 30 MeV. Our experiments were also modeled using the full scale Monte Carlo techniques. In addition, and in a more general frame, due to the lack of consistent data on photonuclear reactions, we have been working on the development of a new photonuclear activation file (PAF), which includes cross sections for more than 600 isotopes including photofission fragment distributions and delayed neutron tables for actinides. Therefore, this work includes also some experimental results obtained at the ELSA electron accelerator, which is more adapted for precise basic nuclear data measurements.

Development of internal dose assessment procedure for workers in industries using raw materials containing naturally occurring radioactive materials

International Nuclear Information System (INIS)

Choi, Cheol Kyu; KIm, Yong Geon; Ji, Seung Woo; Kim, Kwang Pyo; Koo, Bon Cheol; Chang, Byung Uck

2016-01-01

It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are 10 Bq·g-1 for 40K and 1 Bq·g-1 for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups (<0.1 mSv, 0.1-0.3 mSv, and >0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels (<0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and >1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries

Development of internal dose assessment procedure for workers in industries using raw materials containing naturally occurring radioactive materials

Energy Technology Data Exchange (ETDEWEB)

Choi, Cheol Kyu; KIm, Yong Geon; Ji, Seung Woo; Kim, Kwang Pyo [College of Engineering, Kyung Hee University, Yongin (Korea, Republic of); Koo, Bon Cheol; Chang, Byung Uck [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-09-15

It is necessary to assess radiation dose to workers due to inhalation of airborne particulates containing naturally occurring radioactive materials (NORM) to ensure radiological safety required by the Natural Radiation Safety Management Act. The objective of this study is to develop an internal dose assessment procedure for workers at industries using raw materials containing natural radionuclides. The dose assessment procedure was developed based on harmonization, accuracy, and proportionality. The procedure includes determination of dose assessment necessity, preliminary dose estimation, airborne particulate sampling and characterization, and detailed assessment of radiation dose. The developed dose assessment procedure is as follows. Radioactivity concentration criteria to determine dose assessment necessity are 10 Bq·g-1 for 40K and 1 Bq·g-1 for the other natural radionuclides. The preliminary dose estimation is performed using annual limit on intake (ALI). The estimated doses are classified into 3 groups (<0.1 mSv, 0.1-0.3 mSv, and >0.3 mSv). Air sampling methods are determined based on the dose estimates. Detailed dose assessment is performed using air sampling and particulate characterization. The final dose results are classified into 4 different levels (<0.1 mSv, 0.1-0.3 mSv, 0.3-1 mSv, and >1 mSv). Proper radiation protection measures are suggested according to the dose level. The developed dose assessment procedure was applied for NORM industries in Korea, including coal combustion, phosphate processing, and monazite handing facilities. The developed procedure provides consistent dose assessment results and contributes to the establishment of optimization of radiological protection in NORM industries.

The carbonaceous sorbent based on the secondary silica-containing material from oil extraction industry

Science.gov (United States)

Starostina, I. V.; Stolyarov, D. V.; Anichina, Ya N.; Porozhnyuk, E. V.

2018-01-01

The object of research in this work is the silica-containing waste of oil extraction industry - the waste kieselghur (diatomite) sludge from precoat filtering units, used for the purification of vegetable oils from organic impurities. As a result of the thermal modification of the sludge, which contains up to 70% of organic impurities, a finely-dispersed low-porous carbonaceous mineral sorption material is formed. The modification of the sludge particles surface causes the substantial alteration of its physical, chemical, adsorption and structural properties - the organic matter is charred, the particle size is reduced, and on the surface of diatomite particles a carbon layer is formed, which deposits in macropores and partially occludes them. The amount of mesopores is increased, along with the specific surface of the obtained product. The optimal temperature of sludge modification is 500°C. The synthesized carbonaceous material can be used as an adsorbing agent for the purification of wastewater from heavy metal ions. The sorption capacity of Cu2+ ions amounted to 14.2 mg·g-1 and for Ni2+ ions - 17.0 mg·g-1. The obtained values exceed the sorption capacity values of the initial kieselghur, used as a filtering charge, for the researched metal ions.

Morphological changes during enhanced carbonation of asbestos containing material and its comparison to magnesium silicate minerals

International Nuclear Information System (INIS)

Gadikota, Greeshma; Natali, Claudio; Boschi, Chiara; Park, Ah-Hyung Alissa

2014-01-01

The disintegration of asbestos containing materials (ACM) over time can result in the mobilization of toxic chrysotile ((Mg, Fe) 3 Si 2 O 5 (OH) 4 )) fibers. Therefore, carbonation of these materials can be used to alter the fibrous morphology of asbestos and help mitigate anthropogenic CO 2 emissions, depending on the amount of available alkaline metal in the materials. A series of high pressure carbonation experiments were performed in a batch reactor at P CO2 of 139 atm using solvents containing different ligands (i.e., oxalate and acetate). The results of ACM carbonation were compared to those of magnesium silicate minerals which have been proposed to permanently store CO 2 via mineral carbonation. The study revealed that oxalate even at a low concentration of 0.1 M was effective in enhancing the extent of ACM carbonation and higher reaction temperatures also resulted in increased ACM carbonation. Formation of phases such as dolomite ((Ca, Mg)(CO 3 ) 2 ), whewellite (CaC 2 O 4 ·H 2 O) and glushinskite (MgC 2 O 4 ·2H 2 O) and a reduction in the chrysotile content was noted. Significant changes in the particle size and surface morphologies of ACM and magnesium silicate minerals toward non-fibrous structures were observed after their carbonation

Annual energy analysis of concrete containing phase change materials for building envelopes

International Nuclear Information System (INIS)

Thiele, Alexander M.; Jamet, Astrid; Sant, Gaurav; Pilon, Laurent

2015-01-01

Highlights: • Adding PCM to concrete walls can significantly reduce the cooling needs of buildings. • Climate, season, and wall orientation strongly affect energy and cost savings. • The PCM melting temperature should be near the desired indoor temperature. • Benefits are maximum for outdoor temperature oscillating around set indoor temperature. • Adding PCM had little effect on heating energy needs and associated cost savings. - Abstract: This paper examines the annual energy and cost savings potential of adding microencapsulated phase change material to the exterior concrete walls of an average-sized single family home in California climate zones 3 (San Francisco, CA) and 9 (Los Angeles, CA). The annual energy and cost savings were larger for South- and West-facing walls than for other walls. They were also the largest when the phase change temperature was near the desired indoor temperature. The addition of microencapsulated phase change material to the building walls reduced the cooling load in summer substantially more than the heating load in winter. This was attributed to the cold winter temperatures resulting in nearly unidirectional heat flux on many days. The annual cooling load reduction in an average-sized single family home in San Francisco and in Los Angeles ranged from 85% to 100% and from 53% to 82%, respectively, for phase change material volume fraction ranging from 0.1 to 0.3. The corresponding annual electricity cost savings ranged from $36 to $42 in San Francisco and from $94 to $143 in Los Angeles. From an energy standpoint, the best climate for using building materials containing uniformly distributed microencapsulated phase change material would have outdoor temperature oscillations centered around the desired indoor temperature for the entire year

Mechanical and Electrical Properties of Sulfur-Containing Polymeric Materials Prepared via Inverse Vulcanization

Directory of Open Access Journals (Sweden)

Sergej Diez

2017-02-01

Full Text Available Recently, new methods have been developed for the utilization of elemental sulfur as a feedstock for novel polymeric materials. One promising method is the inverse vulcanization, which is used to prepare polymeric structures derived from sulfur and divinyl comonomers. However, the mechanical and electrical properties of the products are virtually unexplored. Hence, in the present study, we synthesized a 200 g scale of amorphous, hydrophobic as well as translucent, hyperbranched polymeric sulfur networks that provide a high thermal resistance (>220 °C. The polymeric material properties of these sulfur copolymers can be controlled significantly by varying the monomers as well as the feed content. The investigated comonomers are divinylbenzene (DVB and 1,3-diisopropenylbenzene (DIB. Plastomers with low elastic content and high shape retention containing 12.5%–30% DVB as well as low viscose waxy plastomers with a high flow behavior containing a high DVB content of 30%–35% were obtained. Copolymers with 15%–30% DIB act, on the one hand, as thermoplastics and, on the other hand, as vitreous thermosets with a DIB of 30%–35%. Results of the thermogravimetric analysis (TGA, the dynamic scanning calorimetry (DSC and mechanical characterization, such as stress–strain experiments and dynamic mechanical thermal analysis, are discussed with the outcome that they support the assumption of a polymeric cross-linked network structure in the form of hyper-branched polymers.

Gallium containing composites as a tunable material to understand neuronal behavior under variable stiffness and radiation conditions

Energy Technology Data Exchange (ETDEWEB)

Berg, Nora G.; Pearce, Brady L.; Rohrbaugh, Nathaniel [Department of Materials Science and Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Jiang, Lin [Materials Science and Engineering, University of New South Wales, Sydney (Australia); Nolan, Michael W. [Department of Clinical Sciences (College of Veterinary Medicine), Comparative Medicine Institute, North Carolina State University, Raleigh, NC 27606 (United States); Ivanisevic, Albena, E-mail: ivanisevic@ncsu.edu [Department of Materials Science and Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

2017-02-01

We report a composite biomaterial containing nanostructured GaOOH and Matrigel™ that can be modulated with respect to its stiffness and radiosensitization properties. A variety of concentrations of GaOOH were added to the composite to alter the mechanical properties of the material as well as to tune the radiosensitizing properties to the composite. PC-12 cells were used to study the combined effects of different stimuli on cell behavior. NGF was given to the cells to record their morphology as well as viability. An increase in the substrate stiffness caused an increase in neurite outgrowth but a decrease in cell viability. In addition, increasing the radiation dose decreased neurite outgrowth but increased cell viability when radiosensitizing particles were present. A subtractive effect between radiosensitizing and mechanical stimuli was observed when PC-12 cells were grown on the GaOOH containing composite. - Highlights: • A composite containing GaOOH and Matrigel can be used to study neurotypic cell behavior. • The composite material can be used to modulate multiple stimuli. • The neurotypic cell behavior can be altered during radiation based on the amount of GaOOH present.

The interaction between contacting barrier materials for containment of radioactive wastes

International Nuclear Information System (INIS)

Chang, Hao-Chun; Wang, Chun-Yao; Huang, Wei-Hsing

2012-01-01

Document available in extended abstract form only. The disposal of low-level radioactive wastes requires multi-barrier facilities to contain the wastes from contamination. Typically, the engineered barrier is composed of a concrete vault backfilled with sand/bentonite mixture. The backfill material is a mixture of bentonite and sand/gravel produced from crushing the rocks excavated at the site. With a great swelling potential, bentonite is expected to serve the sealing function, while the crushed sand/gravel improves the workability of the mixture. Due to the nature of radioactive wastes, the disposal site is designed for a service life of 300 years or more, which is much longer than typical engineering or earth works. With such a long service life, the site is subject to groundwater intrusion and geochemical evolution. The near-field environment evolution can be a complex problem in a disposal site. In the vicinity of the concrete vault in a disposal site, the high-alkali concrete environment can cause changes in the pore solution and alter the nature of backfill materials. Therefore, the interaction between the concrete and the backfill material needs to be assessed, such that the barriers serve the expected functions for a long time. Materials and Methods A locally available Zhishin clay and a bentonite originated from Black Hill, Wyoming, USA were used as raw clay materials in this study. Zhishin clay and Black Hill (BH) bentonite are mixed with Taitung area hard shale to produce the backfill material. An experimental program was conducted analysing the soil properties of these 2 bentonites. And an accelerated migration test was devised to understand the loss of calcium leaching of concrete on characteristics of backfill material. The 2 barrier materials (concrete and backfill) were placed in contact and then an electric gradient applied to accelerate the move of cations between the 2 barriers. Fig. 1 shows a schematic diagram of the accelerated migration test

The effect of dispersed materials on baro-membrane treatment of uranium-containing waters

International Nuclear Information System (INIS)

Kryvoruchko, Antonina P.; Atamanenkoa, Irina D.

2007-01-01

The paper investigated a treatment process of uranium-containing waters in a membrane reactor while using natural mineral kizelgur and synthetic sorbent SKN-1K with subsequent ultra- and nano-filtration separation of the mixture. The retention coefficient of U(VI) by membrane UPM-20 under conditions of quasi-stationary equilibrium reached the levels of 0.87-0.89 and 0.89-0.91, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. In the case of membrane OPMN-P and natural mineral kizelgur the retention coefficient of U(VI) was 0.990-0.991 and 0.993-0.996, respectively, while using natural mineral kizelgur and synthetic sorbent SKN-1K. Data regarding the state of water in membranes formed from natural mineral or synthetic sorbent on the surface of substrate membranes UPM-20 and OPMN-P made it possible to conclude that dispersed materials of different chemical nature affect the process of baro-membrane treatment of uranium-containing waters. (authors)

H1259 Container Foams: Performance Data on Aged Materials

International Nuclear Information System (INIS)

Linda Domeier

2002-01-01

Samples of the three cushioning foams used in the H1259 weapon storage container were obtained in 1997, 1998, 2000 and 2001 and tested for density, compression set and compressive strength using the same procedures specified for acceptance testing. Foams from six containers, all about 30 years old and located at Pantex, were evaluated. The bottom cushioning foam is a General Plastics polyurethane foam and the two side pads are rebonded polyurethane foams. All the tests were carried out at room temperature. When compared to the original acceptance requirements the foams were generally in-spec for density and compressive strength at 10% strain and were generally out-of-spec for compression set and compressive strength at 50% strain. Significant variability was noted in the performance of each foam sample and even more in the container-to-container foam performance. The container-to-container variability remains the major unknown in predicting the long-term suitability of these containers for continued use. The performance of the critical bottom cushion foams was generally more uniform and closer to the specified performance than that of the rebonded foams. It was judged that all the foams were adequate for continued use as storage container foams (not shipping) under controlled conditions to mitigate temperature extremes or high impact. This archived information is important in evaluations of the continued suitability for weapon storage use of the H1259 containers and other containers using the same foam cushions

Effect of different packaging materials containing poly-[2-(tert-butylamino) methylstyrene] on the growth of spoilage and pathogenic bacteria on fresh meat.

Science.gov (United States)

Dohlen, S; Braun, C; Brodkorb, F; Fischer, B; Ilg, Y; Kalbfleisch, K; Lorenz, R; Kreyenschmidt, M; Kreyenschmidt, J

2017-09-18

The objective of this study was to investigate the effect of novel antimicrobial packaging materials containing poly-[2-(tertbutylamino) methylstyrene] (poly(TBAMS)) on the growth of typical spoilage and pathogenic bacteria present on meat. The antimicrobial activity of materials containing different poly(TBAMS) concentrations was determined by comparing the bacterial counts on reference and sample materials at different temperatures and times and in the presence of meat components. Storage tests with poultry fillets and veal cutlets were conducted with samples vacuum packaged in the reference foil and foil containing 10% poly(TBAMS). After specific time intervals, typical spoilage microorganisms, total viable count (TVC), sensory changes and pH value were analysed. The results of the different poly(TBAMS) containing packaging materials showed an increase of the antimicrobial activity with an increasing amount of poly(TBAMS) in the base polymer. A high antimicrobial activity against inoculum of spoilage and pathogenic organisms typical for meat products was detected of a multilayer foil containing 10% poly(TBAMS) in the inner layer after 24h at 7°C. Gram positive-bacteria were more sensitive to poly(TBAMS) foil than gram-negative bacteria. In storage tests however, over the entire storage, a significant effect of this poly(TBAMS) foil on microbial growth on chicken breast fillets and veal cutlets could not be identified. Poly(TBAMS) packaging materials showed very good antimicrobial properties against a wide range of bacteria. However, for a significant inhibition of microbial growth on fresh meat, a higher amount of poly(TBAMS) was necessary to prolong the shelf life of meat. Copyright © 2017 Elsevier B.V. All rights reserved.

Concrete containments in Swedish nuclear power plants. A review of construction and material

International Nuclear Information System (INIS)

Roth, Thomas; Silfwerbrand, Johan; Sundquist, Haakan

2002-12-01

attention. Current investigation shows that the documentation on the concrete containment structures of the Swedish nuclear power stations is fairly complete after the authors have obtained new information through a survey during 2001 and included these data in the report. The target group of this report are structural engineers and other people interested in knowing how the prestressed concrete containments in the Swedish nuclear power stations are designed, detailed and constructed. Uprising questions regarding the structural behaviour of the containment structures ought to be evaluated by using present material properties and not the data describing the used building materials at the design stage. The aim of this research project is to gain new knowledge on life span questions regarding prestressing steel in concrete structures, partly generally and partly with focus on Swedish nuclear power stations and Swedish bridges. The project covers both bonded and un bonded prestressing steel. This report describes the containment structures for all Swedish nuclear power stations. The information is both given in Chapters 5 through 16 and assembled in tables in Appendix A. The intention is that the documentation shall grow and be supplemented as soon as new information, either new data describing the containment structures or new measuring results, will be obtained or produced within current research project. Design and detailing of prestressed concrete structures are among others based on the knowledge of time-dependent material changes regarding concrete (creep and shrinkage) and prestressing steel (relaxation). The intention is that the following items will treated: general evaluation; testing of prestressing steel and concrete properties; assessment of the risk of a time-dependent increase of brittleness of the prestressing steel; comparisons with codes; modelling of steel relaxation; unidimensional modelling of prestressing losses; regard to elevated temperatures

Shielded container

International Nuclear Information System (INIS)

Fries, B.A.

1978-01-01

A shielded container for transportation of radioactive materials is disclosed in which leakage from the container is minimized due to constructional features including, inter alia, forming the container of a series of telescoping members having sliding fits between adjacent side walls and having at least two of the members including machine sealed lids and at least two of the elements including hand-tightenable caps

Stress corrosion cracking tests on high-level-waste container materials in simulated tuff repository environments

International Nuclear Information System (INIS)

Abraham, T.; Jain, H.; Soo, P.

1986-06-01

Types 304L, 316L, and 321 austenitic stainless steel and Incoloy 825 are being considered as candidate container materials for emplacing high-level waste in a tuff repository. The stress corrosion cracking susceptibility of these materials under simulated tuff repository conditions was evaluated by using the notched C-ring method. The tests were conducted in boiling synthetic groundwater as well as in the steam/air phase above the boiling solutions. All specimens were in contact with crushed Topopah Spring tuff. The investigation showed that microcracks are frequently observed after testing as a result of stress corrosion cracking or intergranular attack. Results showing changes in water chemistry during test are also presented

The possibilities of the microwave utilization of wastes on the example of materials containing the asbestos

Directory of Open Access Journals (Sweden)

M. Pigiel

2010-04-01

Full Text Available The presented paper introduce some of the results of the investigations in the utilization of the materials containing asbestos in the existingin Wroclaw University of Technology Institute’s of Technology of Machines and the Automation Foundry and Automation Group themicrowave reactor. In the reactor’s heating chamber there is possible to recycle from 3 up to 5 kg of the batch at once. The temperaturewith which is possible to receive in it is approx. 1400 oC. The time of it’s achievement (in dependence from utilized material can take outfrom 25 up to 40 minutes.

Method of approximating the effects of blast mitigation materials on particulate-containing clouds formed by explosions

International Nuclear Information System (INIS)

Dyckes, G.W.

1983-09-01

A numerical model was developed for predicting the effect of blast mitigation materials on the rise and entrainment rate of explosively driven buoyant clouds containing radiotoxic particles. Model predictions for clouds from unmitigated explosions agree with published observations. More experimental data are needed to assess the validity of predictions for clouds from mitigated explosions

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

Energy Technology Data Exchange (ETDEWEB)

Strum, M.J.; Weiss, H.; Farmer, J.C. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

1988-06-01

This volume surveys the effects of welding on the degradation modes of three austenitic alloys: Types 304L and 316L stainless steels and Alloy 825. These materials are candidates for the fabrication of containers for the long-term storage of high-level nuclear waste. The metallurgical characteristics of fusion welds are reviewed here and related to potential degradation modes of the containers. Three specific areas are discussed in depth: (1) decreased resistance to corrosion in the forms of preferential corrosion, sensitization, and susceptibility to stress corrosion cracking, (2) hot cracking in the heat-affected zone and the weld zone, and (3) formation of intermetallic phases. The austenitic alloys are ranked as follows in terms of overall weldability: Alloy 825 (best) > Type 316L stainless steel > Type 304L stainless steel (worst). 108 refs., 31 figs., 7 tabs.

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Strum, M.J.; Weiss, H.; Farmer, J.C.; Bullen, D.B.

1988-06-01

This volume surveys the effects of welding on the degradation modes of three austenitic alloys: Types 304L and 316L stainless steels and Alloy 825. These materials are candidates for the fabrication of containers for the long-term storage of high-level nuclear waste. The metallurgical characteristics of fusion welds are reviewed here and related to potential degradation modes of the containers. Three specific areas are discussed in depth: (1) decreased resistance to corrosion in the forms of preferential corrosion, sensitization, and susceptibility to stress corrosion cracking, (2) hot cracking in the heat-affected zone and the weld zone, and (3) formation of intermetallic phases. The austenitic alloys are ranked as follows in terms of overall weldability: Alloy 825 (best) > Type 316L stainless steel > Type 304L stainless steel (worst). 108 refs., 31 figs., 7 tabs

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

Energy Technology Data Exchange (ETDEWEB)

Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D. (Lawrence Livermore National Lab., CA (USA)); Bullen, D.B. (Science and Engineering Associates, Inc., Pleasanton, CA (USA))

1988-04-01

Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs.

Survey of degradation modes of candidate materials for high-level radioactive-waste disposal containers

International Nuclear Information System (INIS)

Farmer, J.C.; Van Konynenburg, R.A.; McCright, R.D.; Bullen, D.B.

1988-04-01

Three iron- to nickel-based austenitic alloys (Types 304L and 316L stainless steels and Alloy 825) are being considered as candidate materials for the fabrication of high-level radioactive-waste containers. Waste will include fuel assemblies from reactors as well as high-level waste in borosilicate glass forms, and will be sent to the prospective repository at Yucca Mountain, Nevada. The decay of radionuclides in the repository will result in the generation of substantial heat and in fluences of gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including atmospheric oxidation; uniform aqueous phase corrosion; pitting; crevice corrosion; sensitization and intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This report is an analysis of data relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of the three austenitic candidate alloys. The candidates are compared in terms of their susceptibilities to these forms of corrosion. Although all three candidates have demonstrated pitting and crevice corrosion in chloride-containing environments, Alloy 825 has the greatest resistance to these types of localized corrosion (LC); such resistance is important because pits can penetrate the metal and serve as crack initiation sites. Both Types 304L and 316L stainless steels are susceptible to SCC in acidic chloride media. In contrast, SCC has not been documented in Alloy 825 under comparable conditions. Gamma radiation has been found to enhance SCC in Types 304 and 304L stainless steels, but it has no detectable effect on the resistance of Alloy 825 to SCC. Furthermore, while the effects of microbiologically induced corrosion have been observed for 300-series stainless steels, nickel-based alloys such as Alloy 825 seem to be immune to such problems. 211 refs., 49 figs., 10 tabs

Morphological changes during enhanced carbonation of asbestos containing material and its comparison to magnesium silicate minerals

Energy Technology Data Exchange (ETDEWEB)

Gadikota, Greeshma [Department of Chemical Engineering, Columbia University, 500 West 120th Street, New York, NY 10027 (United States); Natali, Claudio; Boschi, Chiara [Institute of Geosciences and Earth Resources – National Research Council, Pisa (Italy); Park, Ah-Hyung Alissa, E-mail: ap2622@columbia.edu [Department of Earth and Environmental Engineering, Columbia University, 500 West 120th Street, New York, NY 10027 (United States); Department of Chemical Engineering, Columbia University, 500 West 120th Street, New York, NY 10027 (United States); Lenfest Center for Sustainable Energy, Columbia University, 500 West 120th Street, New York, NY 10027 (United States)

2014-01-15

The disintegration of asbestos containing materials (ACM) over time can result in the mobilization of toxic chrysotile ((Mg, Fe){sub 3}Si{sub 2}O{sub 5}(OH){sub 4})) fibers. Therefore, carbonation of these materials can be used to alter the fibrous morphology of asbestos and help mitigate anthropogenic CO{sub 2} emissions, depending on the amount of available alkaline metal in the materials. A series of high pressure carbonation experiments were performed in a batch reactor at P{sub CO2} of 139 atm using solvents containing different ligands (i.e., oxalate and acetate). The results of ACM carbonation were compared to those of magnesium silicate minerals which have been proposed to permanently store CO{sub 2} via mineral carbonation. The study revealed that oxalate even at a low concentration of 0.1 M was effective in enhancing the extent of ACM carbonation and higher reaction temperatures also resulted in increased ACM carbonation. Formation of phases such as dolomite ((Ca, Mg)(CO{sub 3}){sub 2}), whewellite (CaC{sub 2}O{sub 4}·H{sub 2}O) and glushinskite (MgC{sub 2}O{sub 4}·2H{sub 2}O) and a reduction in the chrysotile content was noted. Significant changes in the particle size and surface morphologies of ACM and magnesium silicate minerals toward non-fibrous structures were observed after their carbonation.

Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants.

Science.gov (United States)

Adrados, A; De Marco, I; Lopez-Urionabarrenechea, A; Caballero, B M; Laresgoiti, M F

2013-01-01

In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm(3) reactor, swept with 1 L min(-1) N(2), at 500°C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg(-1)). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO(2); their HHV is in the range of 18-46 MJ kg(-1). The amount of COCO(2) increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char. Copyright © 2012 Elsevier Ltd. All rights reserved.

Plasma container

International Nuclear Information System (INIS)

Ebisawa, Katsuyuki.

1985-01-01

Purpose: To enable to easily detect that the thickness of material to be abraded is reduced to an allowable limit from the outerside of the plasma container even during usual operation in a plasma vessel for a thermonuclear device. Constitution: A labelled material is disposed to the inside or rear face of constituent members of a plasma container undergoing the irradiation of plasma particles. A limiter plate to be abraded in the plasma container is composed of an armour member and heat removing plate, in which the armour member is made of graphite and heat-removing plate is made of copper. If the armour member is continuously abraded under the effect of sputtering due to plasma particles, silicon nitride embedded so far in the graphite at last appears on the surface of the limiter plate to undergo the impact shocks of the plasma particles. Accordingly, abrasion of the limiter material can be detected by a detector comprising gas chromatography and it can easily be detected from the outside of the plasma content even during normal operation. (Horiuchi, T.)

Time, temperature, chemical and radiation exposure effects on the mechanical performance of polymeric materials used for the containment of radioactive waste. Abstract 56

International Nuclear Information System (INIS)

Brown, L.; Bui, V.T.; Bonin, H.W.

2004-01-01

'Full text:' The mechanical performance of materials used for the fabrication of materials used for the fabrication of a storage container for radioactive waste is dependent on the environment to which the container will be exposed over its lifetime. There exists a complex relationship between the many variables affecting the properties of the polymer and potentially decreasing the mechanical performance properties of the container. To further complicate the system, the degradation processes are often time dependant. Experimental results for Nylon 6,6, Semi-Aromatic Nylon, and Polycarbonate have been used as a basis for the development of a model, which represents the performance of a polymeric container used for the storage of radioactive waste over time. The experimental work aimed at providing information on the materials performance in a variety of environmental conditions, as well as a function of time. This included exposing the polymeric material samples to a mixed field of radiation in the SLOWPOKE-2 nuclear reactor. A series of dilution viscometry experiments have been used to relate the changes in mechanical performance to changes in the physical characteristics of the polymer molecules. This provided a valuable tool in the extrapolation of the model to other polymeric materials, and allowed for use of the model based on theoretical predictions of a polymer molecules reaction to various environmental conditions. (author)

Time, temperature, chemical and radiation exposure effects on the mechanical performance of polymeric materials used for the containment of radioactive waste. Abstract 56

Energy Technology Data Exchange (ETDEWEB)

Brown, L.; Bui, V.T.; Bonin, H.W. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)]. E-mail: Laura-lee.Brown@rmc.ca; bui-v@rmc.ca; bonin-h@rmc.ca

2004-07-01

'Full text:' The mechanical performance of materials used for the fabrication of materials used for the fabrication of a storage container for radioactive waste is dependent on the environment to which the container will be exposed over its lifetime. There exists a complex relationship between the many variables affecting the properties of the polymer and potentially decreasing the mechanical performance properties of the container. To further complicate the system, the degradation processes are often time dependant. Experimental results for Nylon 6,6, Semi-Aromatic Nylon, and Polycarbonate have been used as a basis for the development of a model, which represents the performance of a polymeric container used for the storage of radioactive waste over time. The experimental work aimed at providing information on the materials performance in a variety of environmental conditions, as well as a function of time. This included exposing the polymeric material samples to a mixed field of radiation in the SLOWPOKE-2 nuclear reactor. A series of dilution viscometry experiments have been used to relate the changes in mechanical performance to changes in the physical characteristics of the polymer molecules. This provided a valuable tool in the extrapolation of the model to other polymeric materials, and allowed for use of the model based on theoretical predictions of a polymer molecules reaction to various environmental conditions. (author)

Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

Energy Technology Data Exchange (ETDEWEB)

Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S

2004-07-01

It has previously been demonstrated that immobilization of Cs{sup +} and/or Sr{sup 2+} sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO{sub 3}){sub 3} solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO{sub 3}){sub 3} losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO{sub 3}){sub 3}. In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

Cesium release from ceramic waste form materials in simulated canister corrosion product containing solutions

International Nuclear Information System (INIS)

Vittorio, Luca; Drabarek, Elizabeth; Chronis, Harriet; Griffith, Christopher S.

2004-01-01

It has previously been demonstrated that immobilization of Cs + and/or Sr 2+ sorbed on hexagonal tungsten oxide bronze (HTB) adsorbent materials can be achieved by heating the materials in air at temperatures in the range 500 - 1300 deg C. Highly crystalline powdered HTB materials formed by heating at 800 deg C show leach characteristics comparable to Cs-containing hot-pressed hollandites in the pH range from 0 to 12. As a very harsh leaching test, and also to model in a basic manner, leaching in the presence of canister corrosion products in oxidising environments, leaching of the bronzoid phases has been undertaken in Fe(NO 3 ) 3 solutions of increasing concentration. This is done in comparison with Cs -hollandite materials in order to compare the leaching characteristics of these two materials under such conditions. Both the Cs-loaded bronze and hollandite materials leach severely in Fe(NO 3 ) 3 losing virtually all of the immobilized Cs in a period of four days at 150 deg C. Total release of Cs and conversion of hollandite to titanium and iron titanium oxides begins to be observed at relatively low concentrations and is virtually complete after four days reaction in 0.5 mol/L Fe(NO 3 ) 3 . In the case of the bronze, all of the Cs is also extracted but the HTB structure is preserved. The reaction presumably involves an ion-exchange mechanism and iron oxide with a spinel structure is also observed at high Fe concentrations. (authors)

Underground processing method for radiation-contaminated material and transferring method for buffer molding material

International Nuclear Information System (INIS)

Akasaka, Hidenari; Shimura, Satoshi; Asano, Eiichi; Yamagata, Junji; Ninomiya, Nobuo; Kawakami, Susumu.

1995-01-01

A bottomed molding material (buffer molding material) is formed into a bottomed cylindrical shape by solidifying, under pressure, powders such as of bentonite into a highly dense state by a cold isotropic pressing or the like, having a hole for accepting and containing a vessel for radiation-contaminated materials. The bottomed cylindrical molding material is loaded on a transferring vessel, and transferred to a position near the site for underground disposal. The bottomed cylindrical molding material having a upwarded containing hole is buried in the cave for disposal. The container for radiation-contaminated material is loaded and contained in the containing hole of the bottomed cylindrical molding material. A next container for radiation-contaminated materials is juxtaposed thereover. Then, a bottomed cylindrical molding material having a downwarded containing hole is covered to the container for the radiation-contaminated material in a state being protruded upwardly. The radiation-contaminated material is thus closed by a buffer material of the same material at the circumference thereof. (I.N.)

Presence of multiple sites containing polar material in spherical Escherichia coli cells that lack MreB.

Science.gov (United States)

Nilsen, Trine; Yan, Arthur W; Gale, Gregory; Goldberg, Marcia B

2005-09-01

In rod-shaped bacteria, certain proteins are specifically localized to the cell poles. The nature of the positional information that leads to the proper localization of these proteins is unclear. In a screen for factors required for the localization of the Shigella sp. actin assembly protein IcsA to the bacterial pole, a mutant carrying a transposon insertion in mreB displayed altered targeting of IcsA. The phenotype of cells containing a transposon insertion in mreB was indistinguishable from that of cells containing a nonpolar mutation in mreB or that of wild-type cells treated with the MreB inhibitor A22. In cells lacking MreB, a green fluorescent protein (GFP) fusion to a cytoplasmic derivative of IcsA localized to multiple sites. Secreted full-length native IcsA was present in multiple faint patches on the surfaces of these cells in a pattern similar to that seen for the cytoplasmic IcsA-GFP fusion. EpsM, the polar Vibrio cholerae inner membrane protein, also localized to multiple sites in mreB cells and colocalized with IcsA, indicating that localization to multiple sites is not unique to IcsA. Our results are consistent with the requirement, either direct or indirect, for MreB in the restriction of certain polar material to defined sites within the cell and, in the absence of MreB, with the formation of ectopic sites containing polar material.

Stress corrosion cracking of the tubing materials for nuclear steam generators in an environment containing lead

International Nuclear Information System (INIS)

Kim, Kyung Mo; Kim, Uh Chul; Lee, Eun Hee; Hwang, Seong Sik

2004-01-01

Steam generator tube materials show a high susceptibility to stress corrosion cracking (SCC) in an environment containing lead species and some nuclear power plants currently have degradation problems associated with lead-induced stress corrosion cracking in a caustic solution. Effects of an applied potential on SCC is tested for middle-annealed Alloy 600 specimens since their corrosion potential can be changed when lead oxide coexists with other oxidizing species like copper oxide in the sludge. In addition, all the steam generator tubing materials used for nuclear power plants being operated and currently under construction in Korea are tested in a caustic solution with lead oxide. (author)

Transport and storage container for radioactive materials

International Nuclear Information System (INIS)

Engelstaedter, R.; Henning, E.; Storch, S.

1987-01-01

The container consists of a reinforced concrete wall, which is lined on the outsinde with a steel sheet and on the inside by a steel sheet several times thicker than this steel sheet. A front container opening can be closed by a reinforced concrete lid, which consists of a concrete core. This is surrounded by a lining, a covering and a lost shell. The cover projects beyond the reinforced concrete lid, so that bolts can be introduced in holes equally distributed over the extent of the cover, which can be turned in the threaded holes of threaded anchors, which are fixed on the steel sheet on the side towards the reinforced conrete wall. (orig./HP) [de

Multifunctional materials and composites

Science.gov (United States)

Seo, Dong-Kyun; Jeon, Ki-Wan

2017-08-22

Forming multifunctional materials and composites thereof includes contacting a first material having a plurality of oxygen-containing functional groups with a chalcogenide compound, and initiating a chemical reaction between the first material and the chalcogenide compound, thereby replacing oxygen in some of the oxygen-containing functional groups with chalcogen from the chalcogen-containing compound to yield a second material having chalcogen-containing functional groups and oxygen-containing functional groups. The first material is a carbonaceous material or a macromolecular material. A product including the second material is collected and may be processed further to yield a modified product or a composite.

Luminescent hybrid materials functionalized with lanthanide ethylenodiaminotetraacetate complexes containing β-diketonate as antenna ligands

Energy Technology Data Exchange (ETDEWEB)

Aguiar, Franklin P.; Costa, Israel F.; Espínola, José Geraldo P.; Faustino, Wagner M.; Moura, Jandeilson L. [Departamento de Química-Universidade Federal da Paraíba, 58051-970 João Pessoa, PB (Brazil); Brito, Hermi F.; Paolini, Tiago B. [Departamento de Química Fundamental-Instituto de Química da Universidade de São Paulo, 05508-900 São Paulo, SP (Brazil); Felinto, Maria Cláudia F.C. [Instituto de Pesquisas energéticas e Nucleares-IPEN, 05508-900 São Paulo, SP (Brazil); Teotonio, Ercules E.S., E-mail: teotonioees@quimica.ufpb.br [Departamento de Química-Universidade Federal da Paraíba, 58051-970 João Pessoa, PB (Brazil)

2016-02-15

Three organic–inorganic hybrid materials based on silica gel functionalized with (3-aminopropyl)trimethoxysilane (APTS), [3-(2-aminoetilamino)-propil]-trimetoxissilano (DAPTS) and 3-[2-(2-aminoetilamino)etilamino] propiltrimetoxysilane (TAPTS) and subsequently modified with EDTA derivative were prepared by nonhomogeneous route and were then characterized. The resulting materials named SilXN-EDTA (X=1 for APTS, 2 for DAPTS and 3 for TAPTS) were used to obtain new lanthanide Ln{sup 3+}-β-diketonate (Ln{sup 3+}=Eu{sup 3+}, Gd{sup 3+} and Tb{sup 3+}) complexes covalently linked to the functionalized silica gel surfaces (named SilXN-EDTALn-dik, dik=tta, dbm, bzac and acac). The photophysical properties of the new luminescent materials were investigated and compared with those with similar system presenting water molecules coordinated to the lanthanide ions, SilXN-EDTALn-H{sub 2}O. The SilXN-EDTAEu-dik and SilXN-EDTATb-dik systems displayed characteristic red and green luminescence when excited by UV radiation. Furthermore, the quantitative results showed that the emission quantum efficiency (η), experimental intensity parameters Ω{sub 2} and Ω{sub 4}, and Einstein's emission coefficient (A{sub 0J}) of the SilXN-EDTAEu-dik materials were largely dependent on the ligands. Based on the luminescence data, the most efficient intramolecular energy transfer processes were found to the SilXN-EDTAEu-dik (dik: tta and dbm) and SilXN-EDTATb-acac materials, which exhibited more pure emission colors. These materials are promising red and green phosphors, respectively. - Highlights: • New highly luminescent hybrid materials containing lanthanide-EDTA complexes. • The effect of three silylanting agent on the adsorption and luminescent properties has been studied. • The luminescence sensitizing by different β-diketonate ligands have been investigated.

Neutron data error estimate of criticality calculations for lattice in shielding containers with metal fissionable materials

International Nuclear Information System (INIS)

Vasil'ev, A.P.; Krepkij, A.S.; Lukin, A.V.; Mikhal'kova, A.G.; Orlov, A.I.; Perezhogin, V.D.; Samojlova, L.Yu.; Sokolov, Yu.A.; Terekhin, V.A.; Chernukhin, Yu.I.

1991-01-01

Critical mass experiments were performed using assemblies which simulated one-dimensional lattice consisting of shielding containers with metal fissile materials. Calculations of the criticality of the above assemblies were carried out using the KLAN program with the BAS neutron constants. Errors in the calculations of the criticality for one-, two-, and three-dimensional lattices are estimated. 3 refs.; 1 tab

Type I Collagen and Strontium-Containing Mesoporous Glass Particles as Hybrid Material for 3D Printing of Bone-Like Materials.

Science.gov (United States)

Montalbano, Giorgia; Fiorilli, Sonia; Caneschi, Andrea; Vitale-Brovarone, Chiara

2018-04-28

Bone tissue engineering offers an alternative promising solution to treat a large number of bone injuries with special focus on pathological conditions, such as osteoporosis. In this scenario, the bone tissue regeneration may be promoted using bioactive and biomimetic materials able to direct cell response, while the desired scaffold architecture can be tailored by means of 3D printing technologies. In this context, our study aimed to develop a hybrid bioactive material suitable for 3D printing of scaffolds mimicking the natural composition and structure of healthy bone. Type I collagen and strontium-containing mesoporous bioactive glasses were combined to obtain suspensions able to perform a sol-gel transition under physiological conditions. Field emission scanning electron microscopy (FESEM) analyses confirmed the formation of fibrous nanostructures homogeneously embedding inorganic particles, whereas bioactivity studies demonstrated the large calcium phosphate deposition. The high-water content promoted the strontium ion release from the embedded glass particles, potentially enhancing the osteogenic behaviour of the composite. Furthermore, the suspension printability was assessed by means of rheological studies and preliminary extrusion tests, showing shear thinning and fast material recovery upon deposition. In conclusion, the reported results suggest that promising hybrid systems suitable for 3D printing of bioactive scaffolds for bone tissue engineering have been developed.

Fuel salt and container material studies for MOSART transforming system

Energy Technology Data Exchange (ETDEWEB)

Ignatiev, V.; Feynberg, O.; Merzlyakov, A.; Surenkov, A.; Zagnitko, A. [National Research Center, Kurchatov Institute, Moscow (Russian Federation); Afonichkin, V.; Bovet, A.; Khokhlov, V. [Institute of High Temperature Electrochemisty, Ekaterinburg (Russian Federation); Subbotin, V.; Gordeev, M.; Panov, A.; Toropov, A. [Institute of Technical Physics, Snezhinsk (Russian Federation)

2013-07-01

A study is under progress to examine the feasibility of single stream Molten Salt Actinide Recycling and Transmuting system without and with Th support (MOSART) fuelled with different compositions of actinide tri-fluorides (AnF{sub 3}) from used LWR fuel. New fast-spectrum design options with homogeneous core and fuel salts with high enough solubility for AnF{sub 3} are being examined because of new goals. The flexibility of single fluid MOSART concept with Th support is underlined, particularly, possibility of its operation in self-sustainable mode (Conversion Ratio: CR=1) using different loadings and make up. The paper summarizes the most current status of fuel salt and container material data for the MOSART concept received within ISTC-3749 and ROSATOM-MARS projects. Key physical and chemical properties of various fluoride fuel salts are reported. The issues like salt purification, the electroreduction of U(IV) to U(III) in LiF-ThF{sub 4} and the electroreduction of Yb(III) to Yb(II) in LiF-NaF are detailed.

Synthesis of one-dimensional metal-containing insulated molecular wire with versatile properties directed toward molecular electronics materials.

Science.gov (United States)

Masai, Hiroshi; Terao, Jun; Seki, Shu; Nakashima, Shigeto; Kiguchi, Manabu; Okoshi, Kento; Fujihara, Tetsuaki; Tsuji, Yasushi

2014-02-05

We report, herein, the design, synthesis, and properties of new materials directed toward molecular electronics. A transition metal-containing insulated molecular wire was synthesized through the coordination polymerization of a Ru(II) porphyrin with an insulated bridging ligand of well-defined structure. The wire displayed not only high linearity and rigidity, but also high intramolecular charge mobility. Owing to the unique properties of the coordination bond, the interconversion between the monomer and polymer states was realized under a carbon monoxide atmosphere or UV irradiation. The results demonstrated a high potential of the metal-containing insulated molecular wire for applications in molecular electronics.

Charging material into a container and closing the container

International Nuclear Information System (INIS)

Critchley, R.J.; Oldham, S.R.

1989-01-01

Radioactive waste material is passed from a cell through a port into a storage drum which has a mouth with a removable lid. A port door includes an electromagnet to hold the lid to the door and both are moved into the cell by a pneumatic cylinder to allow loading of the drum and after re-closure of door and lid swaging tools operate to deform the lid to form one or more lips which engage under the margin around mouth to resist separation of the lid from the drum. Seals are provided. The swaging tools, or a single swaging tool, can be rotatably mounted on the port door. (author)

Corrosion studies on selected metallic materials for application in nuclear waste disposal containers

International Nuclear Information System (INIS)

Smailos, E.; Fiehn, B.; Gago, J.A.; Azkarate, I.

1994-03-01

In previous corrosion studies, carbon steels and the alloy Ti 99.8-Pd were identified as promising materials for heat-generating nuclear waste containers acting as a radionuclide barrier in a rock-salt repository. To characterize the long-term corrosion behaviour of these materials in more detail, a research programme including laboratory-scale and in-situ corrosion studies has been undertaken jointly by KfK and ENRESA/INASMET. In the period under review, gamma irradiation corrosion studies of up to about 6 months at 10 Gy/h and stress corrosion cracking studies at slow strain rates (10 -4 -10 -7 s -1 ) were performed on three preselected carbon steels in disposal relevant brines (NaCl-rich, MgCl 2 -rich) at 90 C and 150 C (TStE 355, TStE 460, 15 MnNi 6.3). Moreover, results were obtained from long-term in-situ corrosion studies (maximum test duration 9 years) conducted on carbon steel, Ti 99.8-Pd, Hastelloy C4, Ni-resist D4, and Si-cast iron in boreholes in the Asse salt mine. (orig./MM) [de

Modeling of multi-species ion transport in cement-based materials for radioactive waste container

International Nuclear Information System (INIS)

Pang, X.Y.; Li, K.F.; Dangla, P.

2015-01-01

Through the conservations of heat and ions mass, a thermo-hydro-ionic model is established for radionuclide ions transport in cement-based porous barrier materials in radwaste disposal. This model is applied to the design and the safety assessment of a high-integrity container (HIC) used for near surface disposal of low- and intermediate-level radwaste. Five working cases are investigated in the safety assessment considering the internal nuclide ion release, internal heating and pressure accumulation, and external leaching. Comparative analysis shows that leaching increases concrete porosity from external side of container, internal heating of 10 K increase can considerably accelerate the nuclide transport process, and the internal pressure increases the transport rate to limited extent. It is shown that each increment of 10 mm in wall thickness will reduce the radioactivity release by 1.5 to 2 times. Together with the mechanical resistance of HIC under impact actions, the thickness of 100 mm is finally retained for design

49 CFR 173.9 - Transport vehicles or freight containers containing lading which has been fumigated.

Science.gov (United States)

2010-10-01

... 49 Transportation 2 2010-10-01 2010-10-01 false Transport vehicles or freight containers... MATERIALS REGULATIONS SHIPPERS-GENERAL REQUIREMENTS FOR SHIPMENTS AND PACKAGINGS General § 173.9 Transport... hazardous material. (b) No person may offer for transportation or transport a rail car, freight container...

Localized corrosion and stress corrosion cracking of candidate materials for high-level radioactive waste disposal containers in U.S

International Nuclear Information System (INIS)

Farmer, J.C.; McCright, R.D.

1989-01-01

Three ion-based to nickel-based austenitic alloys and three copper-based alloys are being considered in the United States as candidate materials for the fabrication of high-level radioactive waste containers. The austenitic alloys are Types 304L and 316L stainless steels as well as the high-nickel material Alloy 825. The copper-based alloys are CDA 102 (oxygen-free copper) CDA 613 (Cu7Al), and CDA 715 (Cu-30Ni). Waste in the forms of spent fuel assemblies from reactors and borosilicate glass will be sent to a proposed repository at Yucca Mountain, Nevada. The decay of radionuclides will result in the generation of substantial heat and in gamma radiation. Container materials may undergo any of several modes of degradation in this environment, including: undesirable phase transformations due to a lack of phase stability; atmospheric oxidation; general aqueous corrosion; pitting; crevice corrosion; intergranular stress corrosion cracking (IGSCC); and transgranular stress corrosion cracking (TGSCC). This paper is an analysis of data from the literature relevant to the pitting, crevice corrosion, and stress corrosion cracking (SCC) of these alloys

Pyrolysis behavior of different type of materials contained in the rejects of packaging waste sorting plants

Energy Technology Data Exchange (ETDEWEB)

Adrados, A., E-mail: aitziber.adrados@ehu.es [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain); De Marco, I.; Lopez-Urionabarrenechea, A.; Caballero, B.M.; Laresgoiti, M.F. [Chemical and Environmental Engineering Department, School of Engineering of Bilbao, Alameda. Urquijo s/n, 48013 Bilbao (Spain)

2013-01-15

Highlights: Black-Right-Pointing-Pointer Study of the influence of materials in the pyrolysis of real plastic waste samples. Black-Right-Pointing-Pointer Inorganic compounds remain unaltered. Black-Right-Pointing-Pointer Cellulosic components give rise to an increase in char formation. Black-Right-Pointing-Pointer Cellulosic components promote the production of aqueous phase. Black-Right-Pointing-Pointer Cellulosic components increase CO and CO{sub 2} contents in the gases. - Abstract: In this paper rejected streams coming from a waste packaging material recovery facility have been characterized and separated into families of products of similar nature in order to determine the influence of different types of ingredients in the products obtained in the pyrolysis process. The pyrolysis experiments have been carried out in a non-stirred batch 3.5 dm{sup 3} reactor, swept with 1 L min{sup -1} N{sub 2}, at 500 Degree-Sign C for 30 min. Pyrolysis liquids are composed of an organic phase and an aqueous phase. The aqueous phase is greater as higher is the cellulosic material content in the sample. The organic phase contains valuable chemicals as styrene, ethylbenzene and toluene, and has high heating value (HHV) (33-40 MJ kg{sup -1}). Therefore they could be used as alternative fuels for heat and power generation and as a source of valuable chemicals. Pyrolysis gases are mainly composed of hydrocarbons but contain high amounts of CO and CO{sub 2}; their HHV is in the range of 18-46 MJ kg{sup -1}. The amount of CO-CO{sub 2} increases, and consequently HHV decreases as higher is the cellulosic content of the waste. Pyrolysis solids are mainly composed of inorganics and char formed in the process. The cellulosic materials lower the quality of the pyrolysis liquids and gases, and increase the production of char.

Full scale impact testing for environmental and safety control of energy material shipping container systems

International Nuclear Information System (INIS)

Seagren, R.D.

1978-01-01

Heavily-shielded energy material shipping systems, similar in size and weight to those presently employed to transport irradiated reactor fuel elements, are being destructively tested under dynamic conditions. In these tests, the outer and inner steel shells interact in a complex manner with the massive biological shielding in the system. Results obtained from these tests provide needed information for new design concepts. Containment failure (and the resulting release of radioactive material to the environment which might occur in an extremely severe accident) is most likely through the seals and other ancillary features of the shipping systems. Analyses and experiments provide engineering data on the behavior of these shipping systems under severe accident conditions and information for predicting potential survivability and environmental control with a rational margin of safety

Development of a re-brazeable containment system for special nuclear material storage and transport

International Nuclear Information System (INIS)

Pierce, J.D.; Stephens, J.J.; Walker, C.A.; Hosking, F.M.; Curlee, R.M.

1995-01-01

This report describes a novel means of closing and sealing small type B radioactive material transport packages for surface or air transport as governed by 10CFR71 or NUREG-0360 has been developed at Sandia National Laboratories (SNL). This method is a controlled brazing process that may be used to attach and seal a closure lid to a containment vessel and then remove it at a later time. The process may be performed multiple times without the need for special preparations of the braze joint. A number of advantages for utilization of this technique have been determined. A brazed seal has integrity at high temperatures for better protection in accident or abnormal environments. A properly designed joint has essentially the same strength as the parent metal. A closure that is brazed, therefore, will no longer be the anticipated point of failure for a broad range of accident environments. This technique will allow the containment vessel design to be optimized with a lighter, more uniform wall thickness throughout. Finally, with a well defined process for sealing, mechanical inspection, leak testing, and then reopening at a later time, automation of the process is relatively straightforward and the overall system should be as easy to use as one that utilizes elastomeric seals for containment

Roadmapping the Resolution of Gas Generation Issues in Packages Containing Radioactive Waste/Materials - A Status Report

International Nuclear Information System (INIS)

Luke, D.E.; Hamp, S.

2002-01-01

Gas generation issues, particularly hydrogen, have been an area of concern for the transport and storage of radioactive materials and waste in the Department of Energy (DOE) Complex. Potentially combustible gases can be generated through a variety of reactions, including chemical reactions and radiolytic decomposition of hydrogen-containing material. Since transportation regulations prohibit shipment of explosives and radioactive materials together, it was decided that hydrogen generation was a problem that warranted the execution of a high-level roadmapping effort. This paper discusses the major gas generation issues within the DOE Complex and the research that has been and is being conducted by the transuranic (TRU) waste, nuclear materials, and spent nuclear fuels (SNF) programs within DOE's Environmental Management (EM) organizations to address gas generation concerns. This paper presents a ''program level'' roadmap that links technology development to program needs and identifies the probability of success in an effort to understand the programmatic risk associated with the issue of gas generation. This paper also presents the status of the roadmap and follow-up activities

Mechanism of NH{sub 3} desorption during the reaction of H{sub 2} with nitrogen containing carbonaceous materials

Energy Technology Data Exchange (ETDEWEB)

Juan F. Espinal; Thanh N. Truong; Fanor Mondragon [University of Antioquia, Medellin (Colombia). Institute of Chemistry

2005-07-01

The continued increase in demand for natural gas has stimulated the interest in coal conversion to methane as synthetic natural gas by hydropyrolysis of coal (pyrolysis in a H{sub 2} atmosphere). Because the produced raw gas contains considerable amounts of gaseous N-containing products that have to be removed before delivering to final users, the information on distribution of coal-N is important for designing purification processes. It has been reported in the literature that NH{sub 3} is the main nitrogen containing gas that is released during the hydropyrolysis process. Other gases such as HCN and N{sub 2} are also released but in a much smaller amount. To the best of our knowledge, the mechanism for NH{sub 3} desorption during hydrogen reaction with carbonaceous materials has not been studied. We carried out a molecular modeling study using Density Functional Theory in order to get an insight of the mechanism and thermodynamics for NH{sub 3} evolution using pyridinic nitrogen as a model of N-containing carbonaceous material. We propose a mechanism that involves consecutive hydrogenation steps that lead to C-N bond breakage and NH{sub 3} desorption to the gas phase. It was found that the first hydrogenation reaction is highly exothermic. However, further hydrogenations are endothermic. Several pathways for NH{sub 3} evolution were proposed and most of them show high exothermicity. 17 refs., 2 figs.

Characterization of Concrete Mixes Containing Phase Change Materials

Science.gov (United States)

Paksoy, H.; Kardas, G.; Konuklu, Y.; Cellat, K.; Tezcan, F.

2017-10-01

Phase change materials (PCM) can be used in passive building applications to achieve near zero energy building goals. For this purpose PCM can be added in building structures and materials in different forms. Direct incorporation, form stabilization and microencapsulation are different forms used for PCM integration in building materials. In addition to thermal properties of PCM itself, there are several other criteria that need to be fulfilled for the PCM enhanced building materials. Mechanical properties, corrosive effects, morphology and thermal buffering have to be determined for reliable and long-term applications in buildings. This paper aims to give an overview of characterization methods used to determine these properties in PCM added fresh concrete mixes. Thermal, compressive strength, corrosion, and microscopic test results for concrete mixes with PCM are discussed.

Fractionated-combustion analysis of carbonate-containing phases in composite materials of the hydroxyapatite-calcium carbonate system

Science.gov (United States)

Goldberg, M. A.; Shibaeva, T. V.; Smirnov, V. V.; Kutsev, S. V.; Barinov, S. M.; Grigorovich, K. V.

2012-12-01

Materials in the hydroxyapatite (HA)-calcium carbonate (CC) system were synthesized by a precipitation method from aqueous solutions. According to the data of X-ray phase analysis and IR spectroscopy, the powders consisted of CC and AB-type carbonate-substituted HA (CHA). In order to determine the content of carbonate-containing phases in materials, the temperature-temporal mode of fractionated-combustion analysis of carbon was developed. The quantitative phase ratios and the degree of substitution of carbonate groups in CHA were determined. It was shown that the degree of substitution of carbonate groups in CHA increased from 2.47 to 5.31 wt % as the CC content increased from 13.50 to 88.33 wt %.

Leaching studies of heavy concrete material for nuclear fuel waste immobilization containers

International Nuclear Information System (INIS)

Onofrei, M.; Raine, D.; Brown, L.; Hooton, R.D.

1989-08-01

The leaching behaviour of a high-density concrete was studied as part of a program to evaluate its potential use as a container material for nuclear fuel waste under conditions of deep geologic disposal. Samples of concrete material were leached in deionized distilled water, Standard Canadian Shield Saline Solution (SCSSS), SCSSS plus 20% Na-bentonite, and SCSSS plus granite and 20% Na-bentonite under static conditions at 100 degrees celsius for periods up to 365 days. The results of these leaching experiments suggest that the stability of concrete depends on the possible internal structural changes due to hydration reactions of unhydrated components, leading to the formation of C-S-H gel plus portlandite (Ca(OH) 2 ). The factors controlling the concrete leaching process were the composition of the leachant and the concentration of elements in solution capable of forming precipitates on the concrete surface, e.g., silicon, Mg 2+ and Ca 2+ . The main effect observed during leaching was an increase in groundwater pH (from 7 to 9). However, the addition of Na-bentonite suppressed the normal tendency of the pH of the groundwater in contact with concrete to rise rapidly. It was shown that the solution concentration of elements released from the concrete, particularly potassium, increased in the presence of Na-bentonite

Optimization of fly ash as sand replacement materials (SRM) in cement composites containing coconut fiber

Science.gov (United States)

Nadzri, N. I. M.; Jamaludin, S. B.; Mazlee, M. N.; Jamal, Z. A. Z.

2016-07-01

The need of utilizing industrial and agricultural wastes is very important to maintain sustainability. These wastes are often incorporated with cement composites to improve performances in term of physical and mechanical properties. This study presents the results of the investigation of the response of cement composites containing coconut fiber as reinforcement and fly ash use as substitution of sand at different hardening days. Hardening periods of time (7, 14 and 28 days) were selected to study the properties of cement composites. Optimization result showed that 20 wt. % of fly ash (FA) is a suitable material for sand replacement (SRM). Meanwhile 14 days of hardening period gave highest compressive strength (70.12 MPa) from the cement composite containing 9 wt. % of coconut fiber and fly ash. This strength was comparable with the cement without coconut fiber (74.19 MPa) after 28 days of curing.

Carbon black vs. black carbon and other airborne materials containing elemental carbon: Physical and chemical distinctions

International Nuclear Information System (INIS)

Long, Christopher M.; Nascarella, Marc A.; Valberg, Peter A.

2013-01-01

Airborne particles containing elemental carbon (EC) are currently at the forefront of scientific and regulatory scrutiny, including black carbon, carbon black, and engineered carbon-based nanomaterials, e.g., carbon nanotubes, fullerenes, and graphene. Scientists and regulators sometimes group these EC-containing particles together, for example, interchangeably using the terms carbon black and black carbon despite one being a manufactured product with well-controlled properties and the other being an undesired, incomplete-combustion byproduct with diverse properties. In this critical review, we synthesize information on the contrasting properties of EC-containing particles in order to highlight significant differences that can affect hazard potential. We demonstrate why carbon black should not be considered a model particle representative of either combustion soots or engineered carbon-based nanomaterials. Overall, scientific studies need to distinguish these highly different EC-containing particles with care and precision so as to forestall unwarranted extrapolation of properties, hazard potential, and study conclusions from one material to another. -- Highlights: •Major classes of elemental carbon-containing particles have distinct properties. •Despite similar names, carbon black should not be confused with black carbon. •Carbon black is distinguished by a high EC content and well-controlled properties. •Black carbon particles are characterized by their heterogenous properties. •Carbon black is not a model particle representative of engineered nanomaterials. -- This review demonstrates the significant physical and chemical distinctions between elemental carbon-containing particles e.g., carbon black, black carbon, and engineered nanomaterials

Assessment of surface acidity in mesoporous materials containing aluminum and titanium

Science.gov (United States)

Araújo, Rinaldo S.; Maia, Débora A. S.; Azevedo, Diana C. S.; Cavalcante, Célio L., Jr.; Rodríguez-Castellón, E.; Jimenez-Lopez, A.

2009-04-01

The surface acidity of mesoporous molecular sieves of aluminum and titanium was evaluated using four different techniques: n-butylamine volumetry, cyclohexylamine thermodesorption, temperature-programmed desorption of ammonia and adsorption of pyridine. The nature, strength and concentration of the acid sites were determined and correlated to the results of a probe reaction of anthracene oxidation to 9,10-anthraquinone (in liquid phase). In general, the surface acidity was highly influenced by the nature, location and coordination of the metal species (Al and Ti) in the mesoporous samples. Moderate to strong Brönsted acid sites were identified for the Al-MCM-41 sample in a large temperature range. For mesoporous materials containing Ti, the acidity was represented by a combination of weak to moderate Brönsted and Lewis acid sites. The Ti-HMS sample exhibits a higher acidity of moderate strength together with a well-balanced concentration of Brönsted and Lewis acid sites, which enhanced both conversion and selectivity in the oxidation reaction of anthracene.

Fiber reinforced concrete as a material for nuclear reactor containment buildings

International Nuclear Information System (INIS)

Mallikarjuna; Banthia, N.; Mindess, S.

1991-01-01

The fiber reinforced concrete as a constructional material for nuclear reactor containment buildings calls for an examination of its individual characteristics and potentialities due to its inherent superiority over normal plain and reinforced concrete. In the present investigation, first, to study the static behavior of straight, hooked-end and crimped fibers, recently developed nonlinear three-dimensional interface (contact) element has been used in conjunction with the eight nodded hexahedron and two nodded bar elements for concrete and steel fiber respectively. Then impact tests were carried out on fiber reinforced concrete beams with an instrumented drop weight impact machine. Two different concrete mixes were tested: normal strength and high strength concrete specimens. Fibers in the concrete mix found to significantly increase the ductility and the impact resistance of the composite. Deformed fibers increase peak pull-out load and pull-out distance, and perform better in the steel fiber reinforced concrete (SFRC) structures. (author)

Update on ASME rules for spent nuclear fuel and high level radioactive material and waste storage containments

International Nuclear Information System (INIS)

Ralph S. Hill III; Foster, G.M.

2005-01-01

In 2004, a new Code Case, N-717, of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) was published. The Code Case provides rules for construction of containments used for storage of spent nuclear fuel and high level radioactive material and waste. The Code Case has been incorporated into Section III of the Code as Division 3, Subsection WC, Class SC Storage Containments, and will be published in the 2005 Addenda. This paper provides an informative background and insight for these rules to provide Owners, regulators, designers, and fabricators with a more comprehensive understanding of the technical basis for these rules. (authors)

Development of evaluation methods for shipping and storage containers with an increased content of metallic residual materials - Further investigations (EBER II). Final report

International Nuclear Information System (INIS)

Zencker, U.; Qiao, L.; Droste, B.

2002-01-01

Further stress analyses were carried out in the second part of the process. Material damping was calculated on the basis of an analysis of shock wave propagation in cast iron. The container structure was modelled better than before. The focus was on the safety assessment concept. For detecting cracks in geometrically complex container sections, static and dynamic fracture-mechanical numeric calculations of a crack in a groove and in the mid of a container wall were carried out. The results were verified for specific applications for which solutions are already available. The results were integrated in simple diagrams for safety assessment. The safety assessment concept was verified in a case study on a fractured prototype container of defined material quality. It was shown that stress can be reduced by constructional improvements (deeper grooves, bottom ridge). (orig.) [de

Dictionary materials engineering, materials testing

International Nuclear Information System (INIS)

1994-01-01

This dictionary contains about 9,500 entries in each part of the following fields: 1) Materials using and selection; 2) Mechanical engineering materials -Metallic materials - Non-metallic inorganic materials - Plastics - Composites -Materials damage and protection; 3) Electrical and electronics materials -Conductor materials - Semiconductors - magnetic materials - Dielectric materials - non-conducting materials; 4) Materials testing - Mechanical methods - Analytical methods - Structure investigation - Complex methods - Measurement of physical properties - Non-destructive testing. (orig.) [de

High Efficient Enrichment and Activated Dissolution of Refractory Low Grade Rh-containing Material

Institute of Scientific and Technical Information of China (English)

WU Xiaofeng; DONG Haigang; TONG Weifeng; ZHAO Jiachun; ZENG Rui

2012-01-01

Aiming to the low-grade rhodium-containing waste materials,a new process was proposed to enrich and activate rhodium by smelting using iron oxide as a trapping agent and activator.A rhodium concentrate was obtained by the separation of base metals and precious metals.The concentrate was reacted with dilute aqua regia to obtain rhodium solution.The factors influencing the enrichment and activation effects were discussed in this paper.The results showed that the dissolution rate is greater than 99％ under the optimum conditions.In this process,the activation of rhodium was finished in the enrichment process.The iron oxide is both a trapping agent and activator,which simplifies the process and reduce the cost.

Residual Stresses In 3013 Containers

International Nuclear Information System (INIS)

Mickalonis, J.; Dunn, K.

2009-01-01

The DOE Complex is packaging plutonium-bearing materials for storage and eventual disposition or disposal. The materials are handled according to the DOE-STD-3013 which outlines general requirements for stabilization, packaging and long-term storage. The storage vessels for the plutonium-bearing materials are termed 3013 containers. Stress corrosion cracking has been identified as a potential container degradation mode and this work determined that the residual stresses in the containers are sufficient to support such cracking. Sections of the 3013 outer, inner, and convenience containers, in both the as-fabricated condition and the closure welded condition, were evaluated per ASTM standard G-36. The standard requires exposure to a boiling magnesium chloride solution, which is an aggressive testing solution. Tests in a less aggressive 40% calcium chloride solution were also conducted. These tests were used to reveal the relative stress corrosion cracking susceptibility of the as fabricated 3013 containers. Significant cracking was observed in all containers in areas near welds and transitions in the container diameter. Stress corrosion cracks developed in both the lid and the body of gas tungsten arc welded and laser closure welded containers. The development of stress corrosion cracks in the as-fabricated and in the closure welded container samples demonstrates that the residual stresses in the 3013 containers are sufficient to support stress corrosion cracking if the environmental conditions inside the containers do not preclude the cracking process.

Apparatus for the storage of transport- and storage-containers containing radioactive fuel elements

International Nuclear Information System (INIS)

Vox, A.

1983-01-01

The invention concerns an apparatus for the storage of transport and storage containers containing radioactive fuel elements. For each transport or storage container there is a separate silo-type container of steel, concrete, prestressed concrete or suchlike breakproof and fireproof material, to be placed in the open, that can be opened for removal and placing of the transport or storage container respectively. (orig.) [de

Evaluation of Nystatin Containing Chitosan Hydrogels as Potential Dual Action Bio-Active Restorative Materials: in Vitro Approach

Directory of Open Access Journals (Sweden)

V. Tamara Perchyonok

2014-11-01

Full Text Available Healing is a specific biological process related to the general phenomenon of growth and tissue regeneration and is a process generally affected by several systemic conditions or as detrimental side-effects of chemotherapy- and radiotherapy-induced inflammation of the oral mucosa. The objectives of this study is to evaluate the novel chitosan based functional drug delivery systems, which can be successfully incorporated into “dual action bioactive restorative materials”, capable of inducing in vitro improved wound healing prototype and containing an antibiotic, such as nystatin, krill oil as an antioxidant and hydroxyapatite as a molecular bone scaffold, which is naturally present in bone and is reported to be successfully used in promoting bone integration when implanted as well as promoting healing. The hydrogels were prepared using a protocol as previously reported by us. The physico-chemical features, including surface morphology (SEM, release behaviors, stability of the therapeutic agent-antioxidant-chitosan, were measured and compared to the earlier reported chitosan-antioxidant containing hydrogels. Structural investigations of the reactive surface of the hydrogel are reported. Release of nystatin was investigated for all newly prepared hydrogels. Bio-adhesive studies were performed in order to assess the suitability of these designer materials. Free radical defense capacity of the biomaterials was evaluated using established in vitro model. The bio-adhesive capacity of the materials in the in vitro system was tested and quantified. It was found that the favorable synergistic effect of free radical built-in defense mechanism of the new functional materials increased sustainable bio-adhesion and therefore acted as a functional multi-dimensional restorative material with potential application in wound healing in vitro.

Shear bond strengths of tooth coating materials including the experimental materials contained various amounts of multi-ion releasing fillers and their effects for preventing dentin demineralization.

Science.gov (United States)

Arita, Shoko; Suzuki, Masaya; Kazama-Koide, Miku; Shinkai, Koichi

2017-10-01

We examined shear bond strengths (SBSs) of various tooth-coating-materials including the experimental materials to dentin and demineralization resistance of a fractured adhesive surface after the SBS testing. Three resin-type tooth-coating-materials (BC, PRG Barrier Coat; HC, Hybrid Coat II; and SF, Shield force plus) and two glass-ionomer-type tooth-coating-materials (CV, Clinpro XT Varnish; and FJ, Fuji VII) were selected. The experimental PRG Barrier Coat containing 0, 17, and 33 wt% S-PRG filler (BC0, BC17, and BC33, respectively) were developed. Each tooth-coating-material was applied to flattened dentin surfaces of extracted human teeth for SBS testing. After storing in water for 32 days with 4000 thermal cycling, the specimens were subjected to the SBS test. Specimens after SBS testing were subjected to a pH cycling test, and then, demineralization depths were measured using a polarized-light microscope. ANOVA and Tukey's HSD test were used for statistical analysis. The SBS value of FJ and CV was significantly lower than those of other materials except for BC (p coating-materials demonstrated significantly higher SBS for dentin than the glass-ionomer-type tooth-coating-materials; however, they were inferior to the glass ionomer-type tooth-coating-materials in regards to the acid resistance of the fractured adhesion surface.

Performance of Sequoyah Containment Anchorage System

International Nuclear Information System (INIS)

Fanous, F.; Greimann, L.; Wassef, W.; Bluhm, D.

1993-01-01

Deformation of a steel containment anchorage system during a severe accident may result in a leakage path at the containment boundaries. Current design criteria are based on either ductile or brittle failure modes of headed bolts that do not account for factors such as cracking of the containment basemat or deformation of the anchor bolt that may affect the behavior of the containment anchorage system. The purpose of this study was to investigate the performance of a typical ice condenser containment's anchorage system. This was accomplished by analyzing the Sequoyah Containment Anchorage System. Based on a strength of materials approach and assuming that the anchor bolts are resisting the uplift caused by the internal pressure, one can estimate that the failure of the anchor bolts would occur at a containment pressure of 79 psig. To verify these results and to calibrate the strength of materials equation, the Sequoyah containment anchorage system was analyzed with the ABAQUS program using a three-dimensional, finite-element model. The model included portions of the steel containment building, shield building, anchor bolt assembly, reinforced concrete mat and soil foundation material

The use of repassivation potential in predicting the performance of high-level nuclear waste container materials

International Nuclear Information System (INIS)

Sridhar, N.; Dunn, D.; Cragnolino, G.

1995-01-01

Localized corrosion in aqueous environments forms an important bounding condition for the performance assessment of high-level waste (HLW) container materials. A predictive methodology using repassivation potential is examined in this paper. It is shown, based on long-term (continuing for over 11 months) testing of alloy 825, that repassivation potential of deep pits or crevices is a conservative and robust parameter for the prediction of localized corrosion. In contrast, initiation potentials measured by short-term tests are non-conservative and highly sensitive to several surface and environmental factors. Corrosion data from various field tests and plant equipment performance are analyzed in terms of the applicability of repassivation potential. The applicability of repassivation potential for predicting the occurrence of stress corrosion cracking (SCC) and intergranular corrosion in chloride containing environments is also examined

Asbestos-Containing Materials in Schools; Final Rule and Notice. Part III: Environmental Protection Agency. 40 CFR Part 763.

Science.gov (United States)

Federal Register, 1987

1987-01-01

The Environmental Protection Agency (EPA) has issued a final rule under the Toxic Substance Control Act (TSCA) to require all local education agencies (LEAs) to identify asbestos-containing materials in their school buildings and take appropriate action to control release of asbestos fibers. The LEAs are required to describe their activities in…

7 CFR 58.339 - Containers.

Science.gov (United States)

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Containers. 58.339 Section 58.339 Agriculture... Procedures § 58.339 Containers. (a) Containers used for the packaging of butter and related products shall be commercially acceptable containers or packaging material that will satisfactorily protect the quality of the...

21 CFR 113.60 - Containers.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 2 2010-04-01 2010-04-01 false Containers. 113.60 Section 113.60 Food and Drugs... CONSUMPTION THERMALLY PROCESSED LOW-ACID FOODS PACKAGED IN HERMETICALLY SEALED CONTAINERS Control of Components, Food Product Containers, Closures, and In-Process Materials § 113.60 Containers. (a) Closures...

Phosphorus containing sintered alloys (review)

International Nuclear Information System (INIS)

Muchnik, S.V.

1984-01-01

Phosphorus additives are considered for their effect on the properties of sintered alloys of different applications: structural, antifriction, friction, magnetic, hard, superhard, heavy etc. Data are presented on compositions and properties of phosphorus-containing materials produced by the powder metallurgy method. Phosphorus is shown to be an effective activator of sintering in some cases. When its concentration in the material is optimal it imparts the material such properties as strength, viscosity, hardness, wear resistance. Problems concerning powder metallurgy of amorphous phosphorus-containing alloys are reported

CONTAIN calculations; CONTAIN-Rechnungen

Energy Technology Data Exchange (ETDEWEB)

Scholtyssek, W.

1995-08-01

In the first phase of a benchmark comparison, the CONTAIN code was used to calculate an assumed EPR accident `medium-sized leak in the cold leg`, especially for the first two days after initiation of the accident. The results for global characteristics compare well with those of FIPLOC, MELCOR and WAVCO calculations, if the same materials data are used as input. However, significant differences show up for local quantities such as flows through leakages. (orig.)

Natural waste materials containing chitin as adsorbents for textile dyestuffs: batch and continuous studies.

Science.gov (United States)

Figueiredo, S A; Loureiro, J M; Boaventura, R A

2005-10-01

In this work three natural waste materials containing chitin were used as adsorbents for textile dyestuffs, namely the Anodonta (Anodonta cygnea) shell, the Sepia (Sepia officinalis) and the Squid (Loligo vulgaris) pens. The selected dyestuffs were the Cibacron green T3G-E (CI reactive green 12), and the Solophenyl green BLE 155% (CI direct green 26), both from CIBA, commonly used in cellulosic fibres dyeing, the most used fibres in the textile industry. Batch equilibrium studies showed that the materials' adsorption capacities increase after a simple and inexpensive chemical treatment, which increases their porosity and chitin relative content. Kinetic studies suggested the existence of a high internal resistance in both systems. Fixed bed column experiments performed showed an improvement in adsorbents' behaviour after chemical treatment. However, in the column experiments, the biodegradation was the main mechanism of dyestuff removal, allowing the materials' bioregeneration. The adsorption was strongly reduced by the pore clogging effect of the biomass. The deproteinised Squid pen (grain size 0.500-1.41 mm) is the adsorbent with highest adsorption capacity (0.27 and 0.037 g/g, respectively, for the reactive and direct dyestuffs, at 20 degrees C), followed by the demineralised Sepia pen and Anodonta shell, behaving like pure chitin in all experiments, but showing inferior performances than the granular activated carbon tested in the column experiments.

Bismuth silicate glass containing heavy metal oxide as a promising radiation shielding material

Science.gov (United States)

Elalaily, Nagia A.; Abou-Hussien, Eman M.; Saad, Ebtisam A.

2016-12-01

Optical and FTIR spectroscopic measurements and electron paramagnetic resonance (EPR) properties have been utilized to investigate and characterize the given compositions of binary bismuth silicate glasses. In this work, it is aimed to study the possibility of using the prepared bismuth silicate glasses as a good shielding material for γ-rays in which adding bismuth oxide to silicate glasses causes distinguish increase in its density by an order of magnitude ranging from one to two more than mono divalent oxides. The good thermal stability and high density of the bismuth-based silicate glass encourage many studies to be undertaken to understand its radiation shielding efficiency. For this purpose a glass containing 20% bismuth oxide and 80% SiO2 was prepared using the melting-annealing technique. In addition the effects of adding some alkali heavy metal oxides to this glass, such as PbO, BaO or SrO, were also studied. EPR measurements show that the prepared glasses have good stability when exposed to γ-irradiation. The changes in the FTIR spectra due to the presence of metal oxides were referred to the different housing positions and physical properties of the respective divalent Sr2+, Ba2+ and Pb2+ ions. Calculations of optical band gap energies were presented for some selected glasses from the UV data to support the probability of using these glasses as a gamma radiation shielding material. The results showed stability of both optical and magnetic spectra of the studied glasses toward gamma irradiation, which validates their irradiation shielding behavior and suitability as the radiation shielding candidate materials.

Container corporation: The financial, management, and operating implications of a universal beverage container recovery system: Executive summary

Energy Technology Data Exchange (ETDEWEB)

1991-01-01

This report contains an overview of the system. It discusses containers and container materials, elements of the system, and the container recovery cost structure. It includes a financial evaluation and addresses issues of implementation.

Sharps container

Science.gov (United States)

Lee, Angelene M. (Inventor)

1992-01-01

This invention relates to a system for use in disposing of potentially hazardous items and more particularly a Sharps receptacle for used hypodermic needles and the like. A Sharps container is constructed from lightweight alodined nonmagnetic metal material with a cup member having an elongated tapered shape and length greater than its transverse dimensions. A magnet in the cup member provides for metal retention in the container. A nonmagnetic lid member has an opening and spring biased closure flap member. The flap member is constructed from stainless steel. A Velcro patch on the container permits selective attachment at desired locations.

Container for hydrogen isotopes

Energy Technology Data Exchange (ETDEWEB)

1976-01-12

A container is described for storage, shipping and and dispensing of hydrogen isotopes such as hydrogen, deuterium, tritium, or mixtures of the same. The container is compact, safe against fracture or accident, and is reusable. It consists of an outer housing with suitable inlet and outlet openings and electrical feed elements, the housing containing an activated sorber material in the form, for example of titanium sponge or an activated zirconium aluminate cartridge. The gas to be stored is introduced into the chamber under conditions of heat and vacuum and is retained in the sorber material. Subsequently, it may be released by heating the unit to drive off the stored gas at desired rates.

Container for hydrogen isotopes

International Nuclear Information System (INIS)

1976-01-01

A container is described for storage, shipping and and dispensing of hydrogen isotopes such as hydrogen, deuterium, tritium, or mixtures of the same. The container is compact, safe against fracture or accident, and is reusable. It consists of an outer housing with suitable inlet and outlet openings and electrical feed elements, the housing containing an activated sorber material in the form, for example of titanium sponge or an activated zirconium aluminate cartridge. The gas to be stored is introduced into the chamber under conditions of heat and vacuum and is retained in the sorber material. Subsequently, it may be released by heating the unit to drive off the stored gas at desired rates

Container for hydrogen isotopes

International Nuclear Information System (INIS)

Solomon, D.E.

1977-01-01

A container for the storage, shipping and dispensing of hydrogen isotopes such as hydrogen, deuterium, tritium, or mixtures of the same which has compactness, which is safe against fracture or accident, and which is reusable is described. The container consists of an outer housing with suitable inlet and outlet openings and electrical feed elements, the housing containing an activated sorber material in the form, for example, of titanium sponge or an activated zirconium aluminate cartridge. The gas to be stored is introduced into the chamber under conditions of heat and vacuum and will be retained in the sorber material. Subsequently, it may be released by heating the unit to drive off the stored gas at desired rates

Development of neutron shielding concrete containing iron content materials

Science.gov (United States)

Sariyer, Demet; Küçer, Rahmi

2018-02-01

Concrete is one of the most important construction materials which widely used as a neutron shielding. Neutron shield is obtained of interaction with matter depends on neutron energy and the density of the shielding material. Shielding properties of concrete could be improved by changing its composition and density. High density materials such as iron or high atomic number elements are added to concrete to increase the radiation resistance property. In this study, shielding properties of concrete were investigated by adding iron, FeB, Fe2B, stainless - steel at different ratios into concrete. Neutron dose distributions and shield design was obtained by using FLUKA Monte Carlo code. The determined shield thicknesses vary depending on the densities of the mixture formed by the additional material and ratio. It is seen that a combination of iron rich materials is enhanced the neutron shielding of capabilities of concrete. Also, the thicknesses of shield are reduced.

Corrosion behaviour of container materials for the disposal of high-level wastes in rock salt formations

International Nuclear Information System (INIS)

Smailos, E.; Schwarzkopf, W.; Koester, R.

1986-01-01

In 1983-84 extensive laboratory-scale experiments (immersion tests) to evaluate the long-term corrosion behaviour of selected materials in salt brines and first in situ experiments were performed. In the laboratory experiments the materials Ti 99.8-Pd, Hastelloy C4 and hot-rolled low carbon steel (reference materials in the joint European corrosion programme) as well as cast steel, spheoroidal cast iron, Si-cast iron and the Ni-Resists type D2 and D4 were investigated. The investigated parameters were: temperature (90 0 C; 170 0 C, 200 0 C), gamma-radiation (10 5 rad/h) and different compositions of salt brines. The results obtained show that, in addition to Ti 99.8-Pd, also Hastelloy C4 and unalloyed steels are in principle suitable for being used for long-term stable HLW-containers if the gamma dose rate is reduced by suitable shielding. Furthermore, the susceptibility of Hastelloy C4 to crevice corrosion must be taken into account. Further studies will be necessary to provide final evidence of the suitability of the materials examined. These will mainly involve clarification of questions related to hydrogen embrittlement (Ti 99.8-Pd, unalloyed steels) and to the influence of pressure and saline impurities (e.g. antiJ, antiBr) on corrosion

Bilayered container << stone-concrete >> to store toxic materials and radioactive waste.; Dvukhslojnyj kontejner << kamen` - beton >> dlya khraneniya toksichnykh materialov i radioaktivnykh otkhodov.

Energy Technology Data Exchange (ETDEWEB)

Vagin, V V; Koltunov, B G; Kurylo, D A; Kosyak, A T; Izotov, Yu L [Dnepropetrovskij Inst. Chernoj Metallurgii, Dnepropetrovsk (Ukraine)

1994-12-31

A design of a universal container providing for the storage of toxic and radioactive waste with the hydrogen index from 2 to 12 pH has been developed. The construction is based on the lining of stone casting with high density and corrosion-resistance indices ensuring leak-proofness and operation reliability of the container under long terms of storage of agressive materials.

Arsenic removal using steel manufacturing byproducts as permeable reactive materials in mine tailing containment systems.

Science.gov (United States)

Ahn, Joo Sung; Chon, Chul-Min; Moon, Hi-Soo; Kim, Kyoung-Woong

2003-05-01

Steel manufacturing byproducts were tested as a means of treating mine tailing leachate with a high As concentration. Byproduct materials can be placed in situ as permeable reactive barriers to control the subsurface release of leachate from tailing containment systems. The tested materials had various compositions of elemental Fe, Fe oxides, Ca-Fe oxides and Ca hydroxides typical of different steel manufacturing processes. Among these materials, evaporation cooler dust (ECD), oxygen gas sludge (OGS), basic oxygen furnace slag (BOFS) and to a lesser degree, electrostatic precipitator dust (EPD) effectively removed both As(V) and As(III) during batch experiments. ECD, OGS and BOFS reduced As concentrations to <0.5mg/l from 25mg/l As(V) or As(III) solution in 72 h, exhibiting higher removal capacities than zero-valent iron. High Ca concentrations and alkaline conditions (pH ca. 12) provided by the dissolution of Ca hydroxides may promote the formation of stable, sparingly soluble Ca-As compounds. When initial pH conditions were adjusted to 4, As reduction was enhanced, probably by adsorption onto iron oxides. The elution rate of retained As from OGS and ECD decreased with treatment time, and increasing the residence time in a permeable barrier strategy would be beneficial for the immobilization of As. When applied to real tailing leachate, ECD was found to be the most efficient barrier material to increase pH and to remove As and dissolved metals.

Development of Separation Materials Containing Palladium for Hydrogen Isotopes Separation

International Nuclear Information System (INIS)

Deng Xiaojun; Luo Deli; Qian Xiaojing

2010-01-01

Displacement chromatography (DC) is a ascendant technique for hydrogen isotopes separation. The performance of separation materials is a key factor to determine the separation effect of DC. At present,kinds of materials are researched, including palladium materials and non-palladium materials. It is hardly replaceable because of its excellent separation performance, although palladium is expensive. The theory of hydrogen isotopes separation using DC was introduced at a brief manner, while several palladium separation materials were expatiated in detail(Pd/K, Pd-Al 2 O 3 , Pd-Pt alloy). Development direction of separation materials for DC was forecasted elementarily. (authors)

Closure Welding of Plutonium Bearing Storage Containers

International Nuclear Information System (INIS)

Cannell, G.R.

2002-01-01

A key element in the Department of Energy (DOE) strategy for the stabilization, packaging and storage of plutonium-bearing materials involves closure welding of DOE-STD-3013 Outer Containers (3013 container). The 3013 container provides the primary barrier and pressure boundary preventing release of plutonium-bearing materials to the environment. The final closure (closure weld) of the 3013 container must be leaktight, structurally sound and meet DOE STD 3013 specified criteria. This paper focuses on the development, qualification and demonstration of the welding process for the closure welding of Hanford PFP 3013 outer containers

Studies of Corrosion Resistant Materials Being Considered for High-Level Nuclear Waste Containment in Yucca Mountain Relevant Environments

International Nuclear Information System (INIS)

McCright, R.D.; Ilevbare, G.; Estill, J.; Rebak, R.

2001-01-01

Containment of spent nuclear fuel and vitrified forms of high level nuclear waste require use of materials that are highly corrosion resistant to all of the anticipated environmental scenarios that can occur in a geological repository. Ni-Cr-Mo Alloy 22 (UNS N60622) is proposed for the corrosion resistant outer barrier of a two-layer waste package container at the potential repository site at Yucca Mountain. A range of water compositions that may contact the outer barrier is under consideration, and a testing program is underway to characterize the forms of corrosion and to quantify the corrosion rates. Results from the testing support models for long term prediction of the performance of the container. Results obtained to date indicate a very low general corrosion rate for Alloy 22 and very high resistance to all forms of localized and environmentally assisted cracking in environments tested to date

Effect of heating oils and fats in containers of different materials on their trans fatty acid content.

Science.gov (United States)

Kala, A L Amrutha; Joshi, Vishal; Gurudutt, K N

2012-08-30

The nature of the container material and temperature employed for deep-frying can have an influence on the development of trans fatty acids (TFAs) in the fat used. The present study was undertaken to determine the effect of heating vegetable oils and partially hydrogenated vegetable fats with different initial TFA content in stainless steel, Hindalium (an aluminium alloy), cast iron and glass containers. Ground nut oil (oil 1), refined, bleached and deodorised (RBD) palmolein (oil 2) and two partially hydrogenated vegetable oils with low (fat 1) and high (fat 2) TFA content were uniformly heated at 175-185 °C over a period of 12 h. An increase in TFA content to 20 g kg⁻¹ was observed in oil 2 in the cast iron container, while a decrease in TFA content of 20-30 g kg⁻¹ was observed in fat 2 in all containers. The heating process of fats and oils also led to an increase in Butyro refractometer reading and colour values. This study showed that the TFA 18:1t content of oil 1, oil 2 and fat 1 increased with repeated or prolonged heating. The cast iron container showed the highest increase in TFA 18:1t for RBD palmolein (oil 2). The amount of linoleic acid trans isomers formed in the heating process was negligible. Fat 2 with high initial TFA content showed a decrease in TFA 18:1 and 18:2 on heating in all containers. Oils heated in glass and stainless steel containers showed less TFA 18:1t formation. Copyright © 2012 Society of Chemical Industry.

10 CFR 20.1904 - Labeling containers.

Science.gov (United States)

2010-01-01

... 10 Energy 1 2010-01-01 2010-01-01 false Labeling containers. 20.1904 Section 20.1904 Energy....1904 Labeling containers. (a) The licensee shall ensure that each container of licensed material bears... handling or using the containers, or working in the vicinity of the containers, to take precautions to...

Explicit Finite Element Modeling of Multilayer Composite Fabric for Gas Turbine Engine Containment Systems, Phase II. Part 3; Material Model Development and Simulation of Experiments

Science.gov (United States)

Simmons, J.; Erlich, D.; Shockey, D.

2009-01-01

A team consisting of Arizona State University, Honeywell Engines, Systems & Services, the National Aeronautics and Space Administration Glenn Research Center, and SRI International collaborated to develop computational models and verification testing for designing and evaluating turbine engine fan blade fabric containment structures. This research was conducted under the Federal Aviation Administration Airworthiness Assurance Center of Excellence and was sponsored by the Aircraft Catastrophic Failure Prevention Program. The research was directed toward improving the modeling of a turbine engine fabric containment structure for an engine blade-out containment demonstration test required for certification of aircraft engines. The research conducted in Phase II began a new level of capability to design and develop fan blade containment systems for turbine engines. Significant progress was made in three areas: (1) further development of the ballistic fabric model to increase confidence and robustness in the material models for the Kevlar(TradeName) and Zylon(TradeName) material models developed in Phase I, (2) the capability was improved for finite element modeling of multiple layers of fabric using multiple layers of shell elements, and (3) large-scale simulations were performed. This report concentrates on the material model development and simulations of the impact tests.

Human gliomas contain morphine

DEFF Research Database (Denmark)

Olsen, Peter; Rasmussen, Mads; Zhu, Wei

2005-01-01

BACKGROUND: Morphine has been found in cancer cell lines originating from human and animal cells. Thus, it became important to demonstrate whether or not actual tumours contain this opiate alkaloid. MATERIAL/METHODS: Human glioma tissues were biochemically treated to isolate and separate endogeno...... of the solutions used in the study nor was it present as a residual material in blank HPLC runs. CONCLUSIONS: Morphine is present in human gliomas, suggesting that it may exert an action that effects tumour physiology/pathology.......BACKGROUND: Morphine has been found in cancer cell lines originating from human and animal cells. Thus, it became important to demonstrate whether or not actual tumours contain this opiate alkaloid. MATERIAL/METHODS: Human glioma tissues were biochemically treated to isolate and separate endogenous...

Fire test of container for radioactive materials under the condition of transportation state

International Nuclear Information System (INIS)

Miyazaki, Sanae; Shimada, Hirohisa

1986-01-01

To secure the safe transportation of container for radioactive materials, furnace and open fire test for the thermal test of container are provided. Therefore, we have carried out furnace and open fire test using test model simulating a transportation state. Test model used in this test is made of stainless steel with diameter of 200 mm and length of 400 mm, and is set on the rest as in the case of transportation state. From the data on temperature measurement, some interesting results are obtained as follows. Near the surface of model, the temperature gradient in the direction perpendicular to the surface of model with the rest is greater than that without the rest. The temperature rise at the center of the model with the rest is less than that without the rest. In the experiment, temperature distributions are measured in the three radial directions. The temperature differences among three distributions in the model with rest are greater than that without rest. On the other hand, in the furnace test, the heat transfer coefficient on the surface of test model with the rest is 90 - 140 kcal/m 2 h · K for the range of furnace temperature from 700 to 950 deg C and this value is almost equal to the value without the rest. (author)

7 CFR 58.150 - Containers.

Science.gov (United States)

2010-01-01

... 7 Agriculture 3 2010-01-01 2010-01-01 false Containers. 58.150 Section 58.150 Agriculture... Identification § 58.150 Containers. (a) The size, style, and type of packaging used for dairy products shall be commercially acceptable containers and packaging materials which will satisfactorily cover and protect the...

Crosslinked polybenzimidazoles containing branching structure as membrane materials with excellent cell performance and durability for fuel cell applications

Science.gov (United States)

Hu, Meishao; Ni, Jiangpeng; Zhang, Boping; Neelakandan, Sivasubramaniyan; Wang, Lei

2018-06-01

Crosslinking is an effective method to improve the properties of high temperature proton exchange membranes based on polybenzimidazole. However, the compact structure of crosslinked polybenzimidazole hinders the phosphoric acid absorption of the membranes, resulting in a relatively poor fuel cell performance. Recently, we find that branched polymers can absorb more phosphoric acid with a larger free volume, but suffer from deteriorated mechanical strength. In this work, a new method is proposed to obtain excellent over-all properties of high temperature proton exchange membranes. A series of crosslinked polybenzimidazoles containing branching structure as membrane materials are successfully prepared for the first time. Compared with conventional crosslinked membranes, these crosslinked polybenzimidazole membranes containing branching structure exhibit a higher phosphoric acid doping level and proton conductivity, improved durability, lower swelling rate and comparable mechanical strength. In particular, the fuel cell base on the crosslinked and branched membrane with a 10% ratio of crosslinker in non-humidified hydrogen/air at 160 °C achieves a power density of 404 mW cm-2. The results indicate that the combination of crosslinking and branching is an effective approach to improve the properties of polybenzimidazole membrane materials.

Method of processing liquid wastes containing radioactive materials

International Nuclear Information System (INIS)

Matsumoto, Kaname; Shirai, Takamori; Nemoto, Kuniyoshi; Yoshikawa, Jun; Matsuda, Takeshi.

1983-01-01

Purpose: To reduce the number of solidification products by removing, particularly, Co-60 that is difficult to remove in a radioactive liquid wastes containing a water-soluble chelating agent, by adsorbing Co-60 to a specific chelating agent. Method: Liquid wastes containing radioactive cobalt and water-soluble chelating agent are passed through the layer of less water-soluble chelating agent that forms a complex compound with cobalt in an acidic pH region. Thus, the chelating compound of radioactive cobalt (particularly Co-60) is eliminated by adsorbing the same on a specific chelating agent layer. The chelating agent having Co-60 adsorbed thereon is discarded as it is through the cement- or asphalt-solidification process, whereby the number of solidification products to be generated can significantly be suppressed. (Moriyama, K.)

Mg-containing hydroxyapatite coatings on Ti-6Al-4V alloy for dental materials

Science.gov (United States)

Yu, Ji-Min; Choe, Han-Cheol

2018-02-01

In this study, Mg-containing hydroxyapatite coatings on Ti-6A1-4 V alloy for dental materials were researched using various experimental instruments. Plasma electrolytic oxidation (PEO) was performed in electrolytes containing Mg (symbols of specimens: CaP, 5M%, 10M%, and 20M%) at 280 V for 3 min. The electrolyte used for PEO was produced by mixing Ca(CH3COO)2·H2O, C3H7NaCaO6P, and MgCl2·6H2O. The phases and composition of the oxide films were evaluated by X-ray diffraction and field-emission scanning electron microscopy with energy dispersive X-ray spectrometry. The irregularity of the surface, pore size, and number of pores decreased as the Mg concentration increased. The ratio of the areas occupied and not occupied by pores decreased as the Mg concentration increased, with the numbers of both large and small pores decreasing with increasing Mg concentration. The number of particles on the internal surfaces of pores was increased as the Mg content increased. Mg content of all samples containing Mg ions showed higher in the pore outside than that of pore inside, whereas the Ca content was higher inside the pores. The P content of samples with the addition of Mg ions showed higher values inside the pores than outside. The Ca/P and [Mg + Ca]/P molar ratios in the PEO films decreased with Mg content. The crystallite size of anatase was increased with increasing Mg concentration in the solution.

Towards a Sustainable Wild Poliovirus Containment Strategy in ...

African Journals Online (AJOL)

Appropriate bio-safety containment and restricted access to stored materials containing WPV at the Virology Laboratory in Lusaka is a major step in the eradication of poliomyelitis in Zambia. Containment of these infectious materials will be particularly important in the post oral polio vaccination cessation era as there will be ...

Density functional theory study the effects of oxygen-containing functional groups on oxygen molecules and oxygen atoms adsorbed on carbonaceous materials.

Science.gov (United States)

Qi, Xuejun; Song, Wenwu; Shi, Jianwei

2017-01-01

Density functional theory was used to study the effects of different types of oxygen-containing functional groups on the adsorption of oxygen molecules and single active oxygen atoms on carbonaceous materials. During gasification or combustion reactions of carbonaceous materials, oxygen-containing functional groups such as hydroxyl(-OH), carbonyl(-CO), quinone(-O), and carboxyl(-COOH) are often present on the edge of graphite and can affect graphite's chemical properties. When oxygen-containing functional groups appear on a graphite surface, the oxygen molecules are strongly adsorbed onto the surface to form a four-member ring structure. At the same time, the O-O bond is greatly weakened and easily broken. The adsorption energy value indicates that the adsorption of oxygen molecules changes from physisorption to chemisorption for oxygen-containing functional groups on the edge of a graphite surface. In addition, our results indicate that the adsorption energy depends on the type of oxygen-containing functional group. When a single active oxygen atom is adsorbed on the bridge site of graphite, it gives rise to a stable epoxy structure. Epoxy can cause deformation of the graphite lattice due to the transition of graphite from sp2 to sp3 after the addition of an oxygen atom. For quinone group on the edge of graphite, oxygen atoms react with carbon atoms to form the precursor of CO2. Similarly, the single active oxygen atoms of carbonyl groups can interact with edge carbon atoms to form the precursor of CO2. The results show that oxygen-containing functional groups on graphite surfaces enhance the activity of graphite, which promotes adsorption on the graphite surface.

High security container

International Nuclear Information System (INIS)

Moreau, P.J.-M.; Monsterleet, G.A.

1979-01-01

This invention concerns containments, vessels or tanks for containing and protecting products or installations of various kinds, to be called by the general denomination 'containers'. Such products can be, inter alia, liquids such as natural gas, ammonia, vinyle chloride and hydrocarbons. Far from just forming simple means of storage, the containers used for this must now be capable of withstanding fire, sabotage for instance rocket fire, even impacts from aircraft, earthquakes and other aggressions of the same kind. The particular object of this invention is to create a container withstanding all these various agressions. It must also be considered that this container can not only be used for storing products or materials but also for enclosing particularly dangerous or delicate installations, such as nuclear or chemical reactors [fr

Fluorine analysis of human enamel around fluoride-containing materials under different pH-cycling by {mu}-PIGE/PIXE system

Energy Technology Data Exchange (ETDEWEB)

Komatsu, H., E-mail: kom@den.hokudai.ac.jp [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan); Yamamoto, H. [Graduate School of Dentistry, Osaka University, 1-8 Yamada-Oka, Suita 565-0871 (Japan); Matsuda, Y.; Kijimura, T.; Kinugawa, M.; Okuyama, K. [Graduate School of Dental Medicine, Hokkaido University, Kita-13, Nishi-7, Kita-ku, Sapporo 060-8586 (Japan); Nomachi, M. [Graduate School of Science, Osaka University, 1-1 Machikaneyama, Toyonaka 560-0043 (Japan); Yasuda, K. [Wakasa Wan Energy Research Center, 64-52-1 Hase, Tsuruga 914-0192 (Japan); Satoh, T. [Advanced Radiation Technology, TARRI, JAEA, 1233 Watanuki-Machi, Takasaki 370-1292 (Japan); Oikawa, S. [National Institute of Radiological Science, 4-9-1, Anagawa, Inage-ku, Chiba 263-8555 (Japan)

2011-10-15

The caries preventive effect of fluoride-containing materials (FCMs) might depend on the caries risk of the individuals. Two pairs of demineralizing and remineralizing solutions of pH-cycling were prepared for simulating low and high caries risk. The purpose of this study was to determine fluorine (F) uptake into human enamel around FCMs under different pH-cycling using the in-air {mu}-PIGE/PIXE system. Fluoride-containing glass ionomer cement (Fuji IX{sub GP} FAST CAPSULE (FN)), and composite resin (BEAUTIFIL II with FLUORO BOND SHAKE ONE (BS)) were used in this study. The pH-cycling (pH 6.8-4.5) was carried out for 5 weeks. After pH-cycling, the caries progression was analyzed using transverse micro-radiography (TMR). The fluorine and calcium distributions in the carious lesion in each specimen were evaluated using the PIGE/PIXE system. From TMR analysis, there was a difference in caries risk between the two kinds of pH-cycling. Although the caries preventive effect of BS and FN was confirmed at low risk, the effect at high risk was confirmed for FN only. From the analysis of the fluorine uptake in the outer 200 {mu}m of the lesion we concluded that there was no significant difference between the pH-cycling solutions. However, we found different fluorine concentrations in the enamel for the two FCMs. The decreased caries progression under high risk for FN indicated that an adequate amount of fluorine supplied from the material is required at higher caries risk. It was confirmed that the caries preventive effect of FCM depends on the caries risk. The fluorine analysis of teeth under various pH-cycling conditions gives information to evaluate the caries preventive effect of fluoride-containing materials according to the caries risk.

Fluorine analysis of human enamel around fluoride-containing materials under different pH-cycling by μ-PIGE/PIXE system

International Nuclear Information System (INIS)

Komatsu, H.; Yamamoto, H.; Matsuda, Y.; Kijimura, T.; Kinugawa, M.; Okuyama, K.; Nomachi, M.; Yasuda, K.; Satoh, T.; Oikawa, S.

2011-01-01

The caries preventive effect of fluoride-containing materials (FCMs) might depend on the caries risk of the individuals. Two pairs of demineralizing and remineralizing solutions of pH-cycling were prepared for simulating low and high caries risk. The purpose of this study was to determine fluorine (F) uptake into human enamel around FCMs under different pH-cycling using the in-air μ-PIGE/PIXE system. Fluoride-containing glass ionomer cement (Fuji IX GP FAST CAPSULE (FN)), and composite resin (BEAUTIFIL II with FLUORO BOND SHAKE ONE (BS)) were used in this study. The pH-cycling (pH 6.8-4.5) was carried out for 5 weeks. After pH-cycling, the caries progression was analyzed using transverse micro-radiography (TMR). The fluorine and calcium distributions in the carious lesion in each specimen were evaluated using the PIGE/PIXE system. From TMR analysis, there was a difference in caries risk between the two kinds of pH-cycling. Although the caries preventive effect of BS and FN was confirmed at low risk, the effect at high risk was confirmed for FN only. From the analysis of the fluorine uptake in the outer 200 μm of the lesion we concluded that there was no significant difference between the pH-cycling solutions. However, we found different fluorine concentrations in the enamel for the two FCMs. The decreased caries progression under high risk for FN indicated that an adequate amount of fluorine supplied from the material is required at higher caries risk. It was confirmed that the caries preventive effect of FCM depends on the caries risk. The fluorine analysis of teeth under various pH-cycling conditions gives information to evaluate the caries preventive effect of fluoride-containing materials according to the caries risk.

Containment Loads Analysis for CANDU6 Reactor using CONTAIN 2.0

International Nuclear Information System (INIS)

Kim, Tae H.; Yang, Chae Y.

2013-01-01

The containment plays an important role to limit the release of radioactive materials to the environment during design basis accidents (DBAs). Therefore, the containment has to maintain its integrity under DBA conditions. Generally, a containment functional DBA evaluation includes calculations of the key containment loads, i. e., pressure and temperature effects associated with a postulated large rupture of the primary or secondary coolant system piping. In this paper, the behavior of containment pressure and temperature was evaluated for loss of coolant accidents (LOCAs) of the Wolsong unit 1 in order to assess the applicability of CONTAIN 2.0 code for the containment loads analysis of the CANDU6 reactor. The containment pressure and temperature of the Wolsong unit 1 were evaluated using the CONTAIN 2.0 code and the results were compared with the CONTEMPT4 code. The peak pressure and temperature calculated by CONTAIN 2.0 agreed well with those of CONTEMPT4 calculation. The overall result of this analysis shows that the CONTAIN 2.0 code can apply to the containment loads analysis for the CANDU6 reactor

Radioactive wastes processing and disposing container

International Nuclear Information System (INIS)

Wada, Jiro; Kato, Hiroaki.

1987-01-01

Purpose: To obtain a processing and disposing container at low level radioactive wastes, excellent in corrosion and water resistance, as well as impact shock resistance for the retrieval storage over a long period of time. Constitution: The container is constituted with sands and pebbles as aggregates and glass fiber-added unsaturated polyester resins as binders. The container may entirely be formed with such material or only the entire inner surface may be formed with the material as liners. A container having excellent resistance to water, chemicals, freezing or melting, whether impact shock, etc. can be obtained, thereby enabling retrieval storage for radioactive wastes at the optimum low level. (Takahashi, M.)

Container Verification Using Optically Stimulated Luminescence

International Nuclear Information System (INIS)

Tanner, Jennifer E.; Miller, Steven D.; Conrady, Matthew M.; Simmons, Kevin L.; Tinker, Michael R.

2008-01-01

Containment verification is a high priority for safeguards containment and surveillance. Nuclear material containers, safeguards equipment cabinets, camera housings, and detector cable conduit are all vulnerable to tampering. Even with a high security seal on a lid or door, custom-built hinges and interfaces, and special colors and types of finishes, the surfaces of enclosures can be tampered with and any penetrations repaired and covered over. With today's technology, these repairs would not be detected during a simple visual inspection. Several suggested solutions have been to develop complicated networks of wires, fiber-optic cables, lasers or other sensors that line the inside of a container and alarm when the network is disturbed. This results in an active system with real time evidence of tampering but is probably not practical for most safeguards applications. A more practical solution would be to use a passive approach where an additional security feature was added to surfaces which would consist of a special coating or paint applied to the container or enclosure. One type of coating would incorporate optically stimulated luminescent (OSL) material. OSL materials are phosphors that luminesce in proportion to the ionizing radiation dose when stimulated with the appropriate optical wavelengths. The OSL fluoresces at a very specific wavelength when illuminated at another, very specific wavelength. The presence of the pre-irradiated OSL material in the coating is confirmed using a device that interrogates the surface of the enclosure using the appropriate optical wavelength and then reads the resulting luminescence. The presence of the OSL indicates that the integrity of the surface is intact. The coating itself could be transparent which would allow the appearance of the container to remain unchanged or the OSL material could be incorporated into certain paints or epoxies used on various types of containers. The coating could be applied during manufacturing

Improvements in containers

International Nuclear Information System (INIS)

Guy, R.

1977-01-01

An improved container is described for transporting radioactive materials, such as irradiated Magnox fuel elements. It has a lid fixed to the container body and at the corners of the lid has shock absorbers that project from the corners and have part-spheroidal shape. The centre of curvature of the surface of the spheroid is positioned within the lid, so that impact loads on a shock absorber tend to hold it to the container rather than dislodge it. The shock absorbers may be Al-Si alloy castings. (U.K.)

Improvements in or relating to the production of metal-containing material in particulate form

International Nuclear Information System (INIS)

Woodhead, J.L.; Scott, K.T.B.; Ball, P.W.

1977-01-01

The process described refers mainly to production of the material in the form of very small spheres. It comprises forming a metal compound-containing gel precipitate by mixing a solution or sol of the metal compound with a soluble organic polymer and contacting the mixture with a precipitating reagent to precipitate the metal as an insoluble compound bound with the polymer. The precipitate is then subjected in the liquid phase to a breaking down and dispersing process to produce an intermediate product suitable for spray drying, and the intermediate product is spray dried to form the particulate product. The breaking down and dispersing process may be performed by means of a colloid mill or vibratory stirrer. Examples of application of the process are described. (U.K.)

Corrosion behaviour of container materials for geological disposal of high-level waste. Joint annual progress report 1983

International Nuclear Information System (INIS)

1985-01-01

Within the framework of the Community R and D programme on management and storage of radioactive waste (shared-cost action), a research activity is aiming at the assessment of corrosion behaviour of potential container materials for geological disposal of vitrified high-level wastes. In this report, the results obtained during the year 1983 are described. Research performed at the Studiecentrum voor Kernenergie/Centre d'Etudes de l'Energie Nucleaire (SCK/CEN) at Mol (B), concerns the corrosion behaviour in clay environments. The behaviour in salt is tested by the Kernforschungszentrum (KfK) at Karlsruhe (D). Corrosion behaviour in granitic environments is being examined by the Commissariat a l'Energie Atomique (CEA) at Fontenay-aux-Roses (F) and the Atomic Energy Research Establishment (AERE) at Harwell (UK); the first is concentrating on corrosion-resistant materials and the latter on corrosion-allowance materials. Finally, the Centre National de la Recherche Scientifique (CNRS) at Vitry (F) is examining the formation and behaviour of passive layers on the metal alloys in the various environments

Shipping container for nuclear fuels

International Nuclear Information System (INIS)

Housholder, W.R.; Greer, N.L.

1976-01-01

A container for nuclear materials is described wherein a specially and uniquely constructed pressure vessel and gamma shield assembly for holding the nuclear materials is provided in a housing, and wherein a positioning means extends between the housing and the assembly for spacing the same, insulation in the housing essentially filling the space between the assembly and housing, the insulation comprising beads, globules or the like of water encapsulated in plastic and which, in one important embodiment, contains neutron absorbing matter

19 CFR 10.601 - Retail packaging materials and containers.

Science.gov (United States)

2010-04-01

...-Central America-United States Free Trade Agreement Rules of Origin § 10.601 Retail packaging materials and...), HTSUS. (b) Effect on regional value content calculation. If the good is subject to a regional value... originating or non-originating materials, as the case may be, in calculating the regional value content of the...

[The effects of topical fluoridation of Ketac Molar Aplicap glass-ionomer material on the growth of cariogenic bacteria contained in the dental plaque].

Science.gov (United States)

Płuciennik-Stronias, Małgorzata; Zarzycka, Beata; Bołtacz-Rzepkowska, Elzbieta

2013-01-01

Dental caries is a bacterial disease. The most important element used in caries prevention is fluoride, which is derived from the air, diet or fluoride-containing preparations and materials, e.g. glass-ionomer restorations. Modern fluoride-containing restorative materials are capable of releasing fluoride to the environment. Fluoride can be also accumulated in glass-ionomer cements, thus an attempt was made to saturate these materials with fluoride. The aim of the study was to evaluate the effect of topical fluoridation of Ketac Molar Aplicap glass-ionomer cement on the growth of Lactobacillus spp. in the dental plaque. The study was carried out in 15 patients with good oral hygiene, in whom 35 fillings with conventional glass-ionomer material, Ketac Molar Aplicap, were performed. After 6 months, three-day dental plaque from these fillings was examined. Next, fluoride was rubbed on the glass-ionomer surface and the examination of three-day dental plaque was repeated. No statistically significant differences (p = 0.143) in the amounts of Lactobacillus spp. in the plaque collected prior to and after topical fluoridation were revealed. Fluoride rubbed in the conventional glass-ionomer cement, Ketac Molar Aplicap, did not affect the amount of Lactobacillus spp. in the dental plaque growing on this material.

Processing method and device for radioactive waste containing surfactant

International Nuclear Information System (INIS)

Yukita, Atsushi; Yoshikawa, Ryozo; Izumida, Tatsuo; Nishi, Takashi; Hattori, Yasuo.

1997-01-01

Washing liquid wastes generated in washing facilities in a nuclear power plant are collected in a liquid waste collecting tank. A suspension containing a powdery active carbon is supplied to the liquid waste collecting tank. Organic ingredients such as of a surfactant, oil ingredients and radioactive materials in the form of ions contained in the washing liquid wastes are adsorbed to the powdery active carbon. The washing liquid wastes containing the powdery active carbon and granular radioactive materials are led into an active carbon separating and drying device. The powdery active carbon and granular radioactive materials contained in the washing liquid wastes are filtered and separated by a filtering plate, and accumulated as filtered materials on the surface of the filtering plate. The purified washing liquid wastes are discharged to the outside. The filtered materials are dried by hot steams (or hot water) and dried air. The filtered materials are peeled from the filtering plate. The filtered materials, in other word, dried powdery active carbon and granular radioactive materials are transported to and burnt in an incinerator. (I.N.)

Species identification of processed animal proteins (PAPs) in animal feed containing feed materials from animal origin.

Science.gov (United States)

Axmann, Sonja; Adler, Andreas; Brandstettner, Agnes Josephine; Spadinger, Gabriela; Weiss, Roland; Strnad, Irmengard

2015-01-01

Since June 2013 the total feed ban of processed animal proteins (PAPs) was partially lifted. Now it is possible to mix fish feed with PAPs from non-ruminants (pig and poultry). To guarantee that fish feed, which contains non-ruminant PAPs, is free of ruminant PAPs, it has to be analysed with a ruminant PCR assay to comply with the total ban of feeding PAPs from ruminants. However, PCR analysis cannot distinguish between ruminant DNA, which originates from proteins such as muscle and bones, and ruminant DNA, which comes from feed materials of animal origin such as milk products or fat. Thus, there is the risk of obtaining positive ruminant PCR signals based on these materials. The paper describes the development of the combination of two analysis methods, micro-dissection and PCR, to eliminate the problem of 'false-positive' PCR signals. With micro-dissection, single particles can be isolated and subsequently analysed with PCR.

Experience in usage of T-108 titrimetric laboratory unit for precision analysis of uranium-containing materials

International Nuclear Information System (INIS)

Ryzhinskij, M.V.; Bronzov, P.A.

1989-01-01

Possibilities of the T-108 device of potentiometric titration for precise determination of uranium in various uranium-containing materials are studied, the results being presented. Principle flowsheet of the device and the sequence of analytic procedure of uranium potentiometric titration are considered. U 3 O 8 , UO 2 and UF 4 were used as materials to be analyzed, state standard samples of K 2 Cr 2 O 7 -SSS 2215-81 and U 3 O 8 SSS 2396-83P- as standard samples. It is shown that relative standard deviation during titration using the T-108 device is mainly determined by the error of determination of the final titration point potention and it must not exceed 0.002 for uranium titration considered. The conclusion is made that the variant of potentiometric titration of uranium with the use of the T-108 device is not inferior in its accuracy to gravimetry and surpasses it in productivity and possibility of automation. 4 refs.; 2 figs.; 2 tabs

Morphology control between microspheres and nanofibers by solvent-induced approach based on crosslinked phosphazene-containing materials

International Nuclear Information System (INIS)

Zhu Yan; Huang Xiaobin; Fu Jianwei; Wang Gang; Tang Xiaozhen

2008-01-01

Multi-morphology control between monodisperse microspheres and uniform nanofibers was successfully achieved by adjusting the ratio of solvent composition. Through the condensation polymerization between hexachlorocyclotriphosphazene and 4,4'-sulfonyldiphenol, the corresponding hybrid inorganic-organic materials appeared. The morphology of both microspheres and nanofibers contained excellent size and shape: the monodisperse microspheres with 0.7-0.9 μm in diameter and the uniform nanofibers with 60 nm in outer diameter. We applied the concept of three-dimensional Hansen solubility parameters for the initial explanation. The activity of the primary colloid particles and the solubility of triethylamine-hydrogen chloride crystal were considered as two factors for the mechanism explanation. This interesting research shows that the nano- and micro-materials with high crosslinked molecule structure and prepared by condensation polymerization can also achieve the morphology transition. It fills the blank in nano-morphology transition research and will provide great information for the research about the control of different morphology preparations based on polymer nanomaterials

Analysis of failures in concrete containments

International Nuclear Information System (INIS)

Moreno-Gonzalez, A.

1989-09-01

The function of Containment, in an accident event, is to avoid the release of radioactive substances into the surroundings. Containment failure, therefore, is defined as the appearance of leak paths to the external environment. These leak paths may appear either as a result of loss of leaktightness due to degradation of design conditions or structural failure with containment material break. This document is a survey of the state of the art of Containment Failure Analysis. It gives a detailed description of all failure mechanisms, indicating all the possible failure modes and their causes, right from failure resulting from degradation of the materials to structural failure and linear breake failure. Following the description of failure modes, possible failure criteria are identified, with special emphasis on structural failure criteria. These criteria have been obtained not only from existing codes but also from the latest experimental results. A chapter has been dedicated exclusively to failure criteria in conventional structures, for the purpose of evaluating the possibility of application to the case of containment. As the structural behaviour of the containment building is very complex, it is not possible to define failure through a single parameter. It is therefore advisable to define a methodology for containment failure analysis which could be applied to a particular containment. This methodology should include prevailing load and material conditions together with the behaviour of complex conditions such as the liner-anchorage-cracked concrete interaction

Process for the production of superconductor containing filaments

Energy Technology Data Exchange (ETDEWEB)

Tuominen, Olli P. (Candler, NC); Hoyt, Matthew B. (Arden, NC); Mitchell, David F. (Asheville, NC); Morgan, Carol W. (Asheville, NC); Roberts, Clyde Gordon (Asheville, NC); Tyler, Robert A. (Canton, NC)

2002-01-01

Superconductor containing filaments having embedments of superconducting material surrounded by a rayon matrix are formed by preparing a liquid suspension which contains at least 10 weight percent superconducting material; forming a multicomponent filament having a core of the suspension and a viscose sheath which contains cellulose xanthate; and thereafter, regenerating cellulose from the cellulose xanthate to form a rayon matrix.

Testing, verification and application of CONTAIN for severe accident analysis of LMFBR-containments

International Nuclear Information System (INIS)

Langhans, J.

1991-01-01

Severe accident analysis for LMFBR-containments has to consider various phenomena influencing the development of containment loads as pressure and temperatures as well as generation, transport, depletion and release of aerosols and radioactive materials. As most of the different phenomena are linked together their feedback has to be taken into account within the calculation of severe accident consequences. Otherwise no best-estimate results can be assured. Under the sponsorship of the German BMFT the US code CONTAIN is being developed, verified and applied in GRS for future fast breeder reactor concepts. In the first step of verification, the basic calculation models of a containment code have been proven: (i) flow calculation for different flow situations, (ii) heat transfer from and to structures, (iii) coolant evaporation, boiling and condensation, (iv) material properties. In the second step the proof of the interaction of coupled phenomena has been checked. The calculation of integrated containment experiments relating natural convection flow, structure heating and coolant condensation as well as parallel calculation of results obtained with an other code give detailed information on the applicability of CONTAIN. The actual verification status allows the following conclusion: a caucious analyst experienced in containment accident modelling using the proven parts of CONTAIN will obtain results which have the same accuracy as other well optimized and detailed lumped parameter containment codes can achieve. Further code development, additional verification and international exchange of experience and results will assure an adequate code for the application in safety analyses for LMFBRs. (orig.)

Rapid Generation of Superheated Steam Using a Water-containing Porous Material

Science.gov (United States)

Mori, Shoji; Okuyama, Kunito

Heat treatment by superheated steam has been utilized in several industrial fields including sterilization, desiccation, and cooking. In particular, cooking by superheated steam is receiving increased attention because it has advantages of reducing the salt and fat contents in foods as well as suppressing the oxidation of vitamin C and fat. In this application, quick startup and cut-off responses are required. Most electrically energized steam generators require a relatively long time to generate superheated steam due to the large heat capacities of the water in container and of the heater. Zhao and Liao (2002) introduced a novel process for rapid vaporization of subcooled liquid, in which a low-thermal-conductivity porous wick containing water is heated by a downward-facing grooved heating block in contact with the upper surface of the wick structure. They showed that saturated steam is generated within approximately 30 seconds from room-temperature water at a heat flux 41.2 kW⁄m2. In order to quickly generate superheated steam of approximately 300°C, which is required for cooking, the heat capacity of the heater should be as small as possible and the imposed heat flux should be so high enough that the porous wick is able to dry out in the vicinity of the contact with the heater and that the resulting heater temperature becomes much higher than the saturation temperature. The present paper proposes a simple structured generator to quickly produce superheated steam. Only a fine wire heater is contacted spirally on the inside wall in a hollow porous material. The start-up, cut-off responses and the rate of energy conversion for input power are investigated experimentally. Superheated steam of 300°C is produced in approximately 19 seconds from room-temperature water for an input power of 300 W. The maximum rate of energy conversion in the steady state is approximately 0.9.

Container for liquid metal

International Nuclear Information System (INIS)

Abe, Yoshihito; Imazu, Takayuki; Ueda, Sabuo; Ueya, Katsumi.

1980-01-01

Purpose: To arrange a vapor trapping member of a specific structure at the inlet part of a cylindrical gap formed by the inner peripheral surface of the circular opening of a container and the outer peripheral surface of a rotary plug thereby to prevent ingress of vapor in the upper part of the cylindrical gap for a long period of time. Constitution: A sealing material receiving tray is fitted to the container side of the inlet part of a cylindrical gap, and a partition plate is fitted to the rotary plug side. The tray is filled with a sealing material consisting of a large number of steel balls, mesh wire gages and the like, and the partition plate is placed in the tray thereby to carry out sealing of the container. Liquid metal vapor evaporating from the liquid level of the liquid metal adheres to the sealing material to fill the gap, and therefore ingress of vapor to the upper part of the cylindrical gap is prevented, and there is no possibility of causing seal cutting due to the use for a long period. (Sekiya, K.)

Liquid-liquid phase separation in particles containing secondary organic material free of inorganic salts

Science.gov (United States)

Song, Mijung; Liu, Pengfei; Martin, Scot T.; Bertram, Allan K.

2017-09-01

Particles containing secondary organic material (SOM) are ubiquitous in the atmosphere and play a role in climate and air quality. Recently, research has shown that liquid-liquid phase separation (LLPS) occurs at high relative humidity (RH) (greater than ˜ 95 %) in α-pinene-derived SOM particles free of inorganic salts, while LLPS does not occur in isoprene-derived SOM particles free of inorganic salts. We expand on these findings by investigating LLPS at 290 ± 1 K in SOM particles free of inorganic salts produced from ozonolysis of β-caryophyllene, ozonolysis of limonene, and photo-oxidation of toluene. LLPS was observed at greater than ˜ 95 % RH in the biogenic SOM particles derived from β-caryophyllene and limonene while LLPS was not observed in the anthropogenic SOM particles derived from toluene. This work combined with the earlier work on LLPS in SOM particles free of inorganic salts suggests that the occurrence of LLPS in SOM particles free of inorganic salts is related to the oxygen-to-carbon elemental ratio (O : C) of the organic material. These results help explain the difference between the hygroscopic parameter κ of SOM particles measured above and below water saturation in the laboratory and field, and have implications for predicting the cloud condensation nucleation properties of SOM particles.

46 CFR 160.028-6 - Container.

Science.gov (United States)

2010-10-01

... 46 Shipping 6 2010-10-01 2010-10-01 false Container. 160.028-6 Section 160.028-6 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) EQUIPMENT, CONSTRUCTION, AND MATERIALS... Container. (a) General. Containers for the stowage of signal pistols and pistol projected parachute red...

Depleted uranium concrete container feasibility study

International Nuclear Information System (INIS)

Haelsig, R.T.

1994-09-01

The purpose of this report is to consider the feasibility of using containers constructed of depleted uranium aggregate concrete (DUCRETE) to store and transport radioactive materials. The method for this study was to review the advantages and disadvantages of DUCRETE containers considering design requirements for potential applications. The author found that DUCRETE is a promising material for onsite storage containers, provided DUCRETE vessels can be certified for one-way transport to disposal sites. The author also found that DUCRETE multipurpose spent nuclear fuel storage/transport packages are technically viable, provided altered temperature acceptance limits can be developed for DUCRETE

Analysis of stress-strain relationship in materials containing voids by means of plastic finite element method

International Nuclear Information System (INIS)

Shiraishi, Haruki; Tabuchi, Masaaki

2000-01-01

Applying the finite element method in two dimensions, an analysis is performed to derive the stress-strain relationship of material containing voids in matrix, and which is subjected to large deformation. The conditions assumed for the analysis are applicability of continuum body mechanics, Mises yield criterion, J2 flow theory, power work-hardening, plane stress in two-dimensional system and uniform cyclically recurring void distribution. Taking as example a case of material presenting 0.3 work-hardening, it is indicated from the analysis that: With voids arrayed in square lattice, total elongation would be little affected by change in void size; With a void spacing in lattice of 10 μ m, a uniform elongation 12-14% should be obtained in a wide range of void sizes from 0.01 to 8.0 μm; Tensile strength should start to lower at a void areal fraction of around 1%; A sharply lowered uniform elongation of a level far below 1% should be presented by material of low work-hardening exponent. The severe decline of ductility seen with 316 stainless steel upon neutron irradiation at temperatures around 600 K is interpreted as resulting from a combination of low work-hardening and the presence of voids in matrix. (author)

Immobilisation of radwastes in glass containers and products formed thereby

International Nuclear Information System (INIS)

Macedo, P.B.; Simmons, C.J.; Lagakos, N.; Simmons, J.H.; Tran, D.C.

1980-01-01

A method of preventing the dissemination of toxic material to the environment comprises forming an admixture of toxic material and glass packing in a hollow doped glass container of high silica content, or forming the admixture in a first container and then depositing at least a portion of the admixture in a hollow doped glass container of high silica content. The glass container is then heated to collapse its walls and to seal the container so that the toxic material is entrapped and sealed within the collapsed doped glass container. The thermal expansion coefficient of the container may be decreased prior to use by exchanging hydrogen ion in pores thereof with other cations followed by collapsing the pores. (author)

Screening of heavy metal containing waste types for use as raw material in Arctic clay-based bricks

DEFF Research Database (Denmark)

Belmonte, Louise Josefine; Ottosen, Lisbeth M.; Kirkelund, Gunvor Marie

2016-01-01

In the vulnerable Arctic environment, the impact of especially hazardous wastes can have severe consequences and the reduction and safe handling of these waste types are therefore an important issue. In this study, two groups of heavy metal containing particulate waste materials, municipal solid...... waste incineration (MSWI) fly and bottom ashes and mine tailings (i.e., residues from the mineral resource industry) from Greenland were screened in order to determine their suitability as secondary resources in clay-based brick production. Small clay discs, containing 20 or 40% of the different...... brick discs obtained satisfactory densities (1669-2007 kg/m3) and open porosities (27.9-39.9%). In contrast, the fly ash brick discs had low densities (1313-1578 kg/m3) and high open porosities (42.1-51. %). However, leaching tests on crushed brick discs revealed that heavy metals generally became more...

Properties of concrete containing different type of waste materials as aggregate replacement exposed to elevated temperature – A review

Science.gov (United States)

Ghadzali, N. S.; Ibrahim, M. H. W.; Sani, M. S. H. Mohd; Jamaludin, N.; Desa, M. S. M.; Misri, Z.

2018-04-01

Concrete is the chief material of construction and it is non-combustible in nature. However, the exposure to the high temperature such as fire can lead to change in the concrete properties. Due to the higher temperature, several changes in terms of mechanical properties were observed in concrete such as compressive strength, modulus of elasticity, tensile strength and durability of concrete will decrease significantly at high temperature. The exceptional fire-proof achievement of concrete is might be due to the constituent materials of concrete such as its aggregates. The extensive use of aggregate in concrete will leads to depletion of natural resources. Hence, the use of waste and other recycled and by-product material as aggregates replacements becomes a leading research. This review has been made on the utilization of waste materials in concrete and critically evaluates its effects on the concrete performances during the fire exposure. Therefore, the objective of this paper is to review the previous search work regarding the concrete containing waste material as aggregates replacement when exposed to elevated temperature and come up with different design recommendations to improve the fire resistance of structures.

Polyethers containing 4-(carbazol-2-yl)-7-arylbenzo[c]-1,2,5-thiadiazole chromophores as solution processed materials for hole transporting layers of OLEDs

Science.gov (United States)

Krucaite, G.; Tavgeniene, D.; Xie, Z.; Lin, X.; Zhang, B.; Grigalevicius, S.

2018-02-01

Two polyethers containing electroactive pendent 4-(carbazol-2-yl)-7-arylbenzo[c]-1,2,5-thiadiazole moieties have been synthesized by the multi-step synthetic route. Full characterization of their structures is presented. The polymers represent derivatives of very high thermal stability with initial thermal degradation temperatures of 425 °C and 431 °C. Glass transition temperatures of the amorphous materials were also very high and reached values of 154 °C and 163 °C. The electron photoemission spectra of thin layers of the polymers showed ionization potentials of 5.84 eV and 5.93 eV. Hole-transporting properties of the polymeric materials were tested in the structures of organic light emitting diodes with Alq3 as the green emitter and electron transporting material. An electroluminescent device containing hole-transporting layer (HTL) of the polymer with electroactive 4-carbazolyl-7-phenylbenzo[c]-1,2,5-thiadiazole moieties exhibited turn on voltage of 6.2 V, maximum photometric efficiency of 2.5 cd/A and maximum brightness exceeding 300 cd/m2. The device containing HTL of the polymer with 4-carbazolyl-7-(1-naphtyl)benzo[c]-1,2,5-thiadiazole moieties demonstrated turn on voltage of 5.2 V, maximum photometric efficiency of 1.6 cd/A and maximum brightness exceeding 1500 cd/m2. The efficiencies were about 30-90% higher than that of the device containing widely used hole transporting layers of poly(9-vinylcarbazole).

Safety system for reactor container

International Nuclear Information System (INIS)

Shimizu, Miwako; Seki, Osamu; Mano, Takio.

1995-01-01

A slanted structure is formed below a reactor core where there is a possibility that molten reactor core materials are dropped, and above a water level of a pool which is formed by coolants flown from a reactor recycling system and accumulated on the inner bottom of the reactor container, to prevent molten fuels from dropping at once in the form of a large amount of lump. The molten materials are provisionally received on the structure, gradually formed into small pieces and then dropped. Further, the molten materials are dropped and received provisionally on a group of coolant-flowing pipelines below the structure, to lower the temperature of the molten materials, and then the reactor core molten materials are gradually formed into small pieces and dropped into the pool water. Since they are not dropped directly into the pool water but dropped gradually into the pool water as small droplets, occurrence of steam explosion can be reduced. The occurrence of steam explosion due to dropped molten reactor core material and pool water is suppressed, and the molten materials are kept in the pool water, thereby enabling to maintain the integrity of the reactor container more effectively. (N.H.)

Dual Neutral Particle Beam Interrogation of Intermodal Shipping Containers for Special Nuclear Material

Science.gov (United States)

Keith, Rodney Lyman

Intermodal shipping containers entering the United States provide an avenue to smuggle unsecured or stolen special nuclear material (SNM). The only direct method fielded to indicate the presence of SNM is by passive photon/neutron radiation detection. Active interrogation using neutral particle beams to induce fission in SNM is a method under consideration. One by-product of fission is the creation of fragments that undergo radioactive decay over a time period on the order of tens of seconds after the initial event. The "delayed" gamma-rays emitted from these fragments over this period are considered a hallmark for the presence of SNM. A fundamental model is developed using homogenized cargos with a SNM target embedded at the center and computationally interrogated using simultaneous neutron and photon beams. Findings from analysis of the delayed gamma emissions from these experiments are intended to mitigate the effects of poor quality information about the composition and disposition of suspect cargo before examination in an active interrogation portal.

Natural material-decorated mesoporous silica nanoparticle container for multifunctional membrane-controlled targeted drug delivery

Directory of Open Access Journals (Sweden)

Hu Y

2017-11-01

Full Text Available Yan Hu,1 Lei Ke,2 Hao Chen,1 Ma Zhuo,1 Xinzhou Yang,1 Dan Zhao,1 Suying Zeng,1 Xincai Xiao1 1Department of Pharmaceutics, School of Pharmaceutical Science, South-Central University for Nationalities, 2Department of Medicinal Chemistry, School of Pharmacy, Tongji Medical College, Huazhong University of Science and Technology, Wuhan, People’s Republic of China Abstract: To avoid the side effects caused by nonspecific targeting, premature release, weak selectivity, and poor therapeutic efficacy of current nanoparticle-based systems used for drug delivery, we fabricated natural material-decorated nanoparticles as a multifunctional, membrane-controlled targeted drug delivery system. The nanocomposite material coated with a membrane was biocompatible and integrated both specific tumor targeting and responsiveness to stimulation, which improved transmission efficacy and controlled drug release. Mesoporous silica nanoparticles (MSNs, which are known for their biocompatibility and high drug-loading capacity, were selected as a model drug container and carrier. The membrane was established by the polyelectrolyte composite method from chitosan (CS which was sensitive to the acidic tumor microenvironment, folic acid-modified CS which recognizes the folate receptor expressed on the tumor cell surface, and a CD44 receptor-targeted polysaccharide hyaluronic acid. We characterized the structure of the nanocomposite as well as the drug release behavior under the control of the pH-sensitive membrane switch and evaluated the antitumor efficacy of the system in vitro. Our results provide a basis for the design and fabrication of novel membrane-controlled nanoparticles with improved tumor-targeting therapy. Keywords: multifunctional, membrane-controlled, natural materials, mesoporous silica nanoparticles, targeted drug delivery

Transportation of Hazardous Evidentiary Material.

Energy Technology Data Exchange (ETDEWEB)

Osborn, Douglas.

2005-06-01

This document describes the specimen and transportation containers currently available for use with hazardous and infectious materials. A detailed comparison of advantages, disadvantages, and costs of the different technologies is included. Short- and long-term recommendations are also provided.3 DraftDraftDraftExecutive SummaryThe Federal Bureau of Investigation's Hazardous Materials Response Unit currently has hazardous material transport containers for shipping 1-quart paint cans and small amounts of contaminated forensic evidence, but the containers may not be able to maintain their integrity under accident conditions or for some types of hazardous materials. This report provides guidance and recommendations on the availability of packages for the safe and secure transport of evidence consisting of or contaminated with hazardous chemicals or infectious materials. Only non-bulk containers were considered because these are appropriate for transport on small aircraft. This report will addresses packaging and transportation concerns for Hazardous Classes 3, 4, 5, 6, 8, and 9 materials. If the evidence is known or suspected of belonging to one of these Hazardous Classes, it must be packaged in accordance with the provisions of 49 CFR Part 173. The anthrax scare of several years ago, and less well publicized incidents involving unknown and uncharacterized substances, has required that suspicious substances be sent to appropriate analytical laboratories for analysis and characterization. Transportation of potentially hazardous or infectious material to an appropriate analytical laboratory requires transport containers that maintain both the biological and chemical integrity of the substance in question. As a rule, only relatively small quantities will be available for analysis. Appropriate transportation packaging is needed that will maintain the integrity of the substance, will not allow biological alteration, will not react chemically with the substance being

27 CFR 20.145 - Encased containers.

Science.gov (United States)

2010-04-01

... 27 Alcohol, Tobacco Products and Firearms 1 2010-04-01 2010-04-01 false Encased containers. 20.145... Denatured Alcohol § 20.145 Encased containers. Completely denatured alcohol may be packaged by distributors in unlabeled containers which are completely encased in wood, fiberboard, or similar material so that...

Tritium containment in fusion facilities

International Nuclear Information System (INIS)

Anderson, J.L.

1978-01-01

The key environmental control systems that have been identified and are being developed are listed. A brief description of each of the following systems is given: primary process materials, permeation barriers, secondary containment, tritium waste treatment, emergency tritium cleanup, maintenance procedures, and tertiary containment

Port Security: Container Cargo Control

Directory of Open Access Journals (Sweden)

Vladivoj Vlaković

2006-05-01

Full Text Available illicittrafficking of threat materials, especially explosives, chemicalsubstances and radioactive or nuclear material. The transportof the threat materials by using sea routes is an advantageto te"orists especially because of the possible use of ship containers.The container is the basis of world trade. It is assumed thatthe world total movement in containers is about 200 millionTEUs ("20-foot equivalent units" per year. The list of materialstransported by containers which should be subject to inspectionwith the aim of reducing the acts of te"orism includes explosives,narcotics, chemical weapons, hazardous chemicalsand radioactive materials.Of special interest is nuclear te"orism. The risk of nuclearte"orism carried out by sub-national groups should be considerednot only in the construction and/or use of nuclear device,but also in possible radioactive contamination of large urbanareas.The system of ship containers control is an essential componentof «smart border» concept. Modem personnel, parcel,vehicle and cargo inspection systems are non-invasive imagingtechniques based on the use of nuclear analytical techniques.The inspection systems use penetrating radiations: hard x-rays(300 keV or more or gamma-rays from radioactive sources(137Cs and 60Co with energies from 600 to 1300 keV that producea high resolution radiograph of the load. Unfortunately,this information is "non-specific" in that it gives no informationon the nature of objects that do not match_ the travel documentsand are not recognized by a visual analysis of the radiographicpicture. Moreover, there are regions of the containerwherex and gamma-ray systems are "blind" due to the high averageatomic number of the objects i"adiated that appear asblack spots in the radiographic image.The systems being developed are based on the use of fast, 14Me V, neutrons with detection of associated a-particle from nuclearreactionbywhichneutrons are produced (d+t>a+n.Jnsuch a way the possibility to

Nuclear reactor container

International Nuclear Information System (INIS)

Hosaka, Seiichi.

1988-01-01

Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

Development of Zr-containing advanced reduced-activation alloy (ARAA) as structural material for fusion reactors

Energy Technology Data Exchange (ETDEWEB)

Chun, Y.B., E-mail: borobang@gmail.com [Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Kang, S.H. [Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Lee, D.W. [Nuclear Fusion Engineering Development Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, S. [National Fusion Research Institute, Daejeon (Korea, Republic of); Jeong, Y.H. [Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Żywczak, A. [AGH University of Science and Technology, Academic Centre of Materials and Nanotechnology, Kraków (Poland); Rhee, C.K. [Nuclear Materials Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-11-01

Highlights: • Creep and impact resistances of reduced activation ferritic–martensitic steel are enhanced by the addition of Zr. • A 5 ton scale heat of Zr containing RAFM steel, ARAA, has been produced for material property evaluation. • The physical, thermal, magnetic and mechanical properties of ARAA are quite similar to those of Eurofer 97. - Abstract: Korea has developed an advanced reduced-activation alloy (ARAA) as a structural material for helium-cooled ceramic reflector test blanket module (HCCR-TBM) applications. The present paper describes the history of alloy development and the properties of ARAA, which has been produced at a 5 t scale using vacuum induction melting and electro-slag re-melting methods. ARAA is a 9Cr–1.2W based ferritic–martensitic steel with 0.01 wt.% Zr. The mechanical properties, thermal properties and physical and magnetic properties of ARAA show similar temperature dependencies to those observed for Eurofer 97. However, ARAA exhibits a much longer creep–rupture time than conventional RAFM steel, which suggests a positive effect on Zr addition. The enhanced creep strength of ARAA by the addition of Zr is attributed to the reduced temperature-dependence of the yield strength.

Elaboration of vitreous and vitrocrystalline basalt materials containing simulated radioactive ash wastes. Study of some physical properties and leaching behaviour

International Nuclear Information System (INIS)

Lebeau, M.J.

1988-01-01

This work is a preliminary study of a matrix for containment of incinerator ashes. Basalt presents a good capacity for ash incorporation (up to 50%), glasses obtained are homogeneous and the low viscosity at 1300 0 C allows an industrial preparation. Vitrocrystalline products have a dendritic texture which can be controlled by cooling rate and are composed of magnesioferrite, pyroxene, plagioclase and vitreous fraction depending on the filling material ratio. Uranium and thorium, for actinide simulation in ashes, are localized in the glass. Glass leaching decrease with ash content and the alteration film presents a better retention of some elements, such as uranium and thorium. Vitrocrystalline materials are less leachable than glasses. Interesting possibilities are shown for use of basalt as filling material in underground storage, since basalt decrease glass deterioration [fr

Experimental analysis of plastic materials containing radionuclides for decontamination viability

International Nuclear Information System (INIS)

Tazaki, Kazue; Nakano, Mikio; Takehara, Teruaki; Ishigaki, Yasuhito; Nakagawa, Hideaki

2015-01-01

After the Fukushima Daiichi Nuclear Power Plant (FDNPP) accident on 11 March, 2011, the high radioactive dosage was found in polluted water for agriculture use at Baba, Haramachi, Minami-Soma, Fukushima Prefecture, Japan. Field experiment for decontamination of water had been studied by using commercial plastic materials. The agricultural water comes from Tetsuzan dam is full of radioactive-contaminated water. Experimental analysis showed that the plastic materials can take up radioactive elements for several months soaked in the polluted agricultural water. The quantitative analyses using X-ray fluorescence analysis, Ge semiconductor and scanning electron microscopy equipped with energy dispersive X-ray spectrometer (SEM-EDS), revealed the detection of the radionuclides on the plastic materials with diatom and clays. The results suggest the adsorption of radionuclides on the surface of plastic materials due to FDNPP accident. The plastic materials associated with clays and diatoms could be stronger carriers of radionuclides in the polluted water. Adherence of diatoms to the plastic fiber in the water for 7 months suggested that some plastic materials were taking up heavy metals (Zn, Ba, Pb, Sb) with radioactive elements (Cs etc.). Mechanisms by which radioactive pollutants and microorganisms are adsorbed onto and desorbed from clays at aqueous interface can be understood by combining chemical analysis with electron microscopy observation. (author)

Irradiation Effects in Fortiweld Steel Containing Different Boron Isotopes

International Nuclear Information System (INIS)

Grounes, M.

1967-07-01

Tensile specimens and miniature impact specimens of the low alloyed pressure vessel steel Fortiweld have been irradiated at 265 deg C in R2 to two neutron doses, 6.5 x 10 18 n/cm 2 (> 1 MeV) and 4 x 10 19 n/cm 2 (thermal) and also 9.0 x 10 18 n/cm 2 (> 1 MeV) and 6 x 10 19 n/cm 2 (thermal). Material from three laboratory melts, in which the boron consisted of 10 B, 11 B and natural boron respectively, were investigated. The results both of tensile tests and impact tests with miniature impact specimens show that the 10 B-alloyed material was changed more and the 11 B-alloyed material was changed less than the material containing natural boron. At the higher neutron dose the increase in yield strength (0.2 % offset yield strength) was 11 kg/mm in the 10 B containing material compared to 5 kg/mm in the 11 B-containing material. The decrease in total elongation was 5 and 0 percentage units respectively. The transition temperature was increased 190 deg C at the higher neutron dose in the 10 B-alloyed material, 40 deg C in the 11 B-alloyed material and 80 deg C in the material containing natural boron